ML20127M273
| ML20127M273 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/30/1992 |
| From: | Low A, Napolitano R, Spaar D BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20127M267 | List: |
| References | |
| BAW-2177, NUDOCS 9211300108 | |
| Download: ML20127M273 (78) | |
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BAW-2177 November 1992 ANALYSIS OF CAPSULE W-97 ENTERGY OPERATIONS, INC.
WATERFORD GENERATING STATION, UNIT NO. 3
-- Reactor Vessel Material Surveillance Program --
by A. L. Lowe, Jr., PE R. E. Napolitano D. M. Spaar W..R. Stagg B&W Document No. 77-2177 (See Section 10 for document signatures)
'B&W NUCLEAR SERVICE COMPANY Engineering and Plant Services Division P. O. Box 10935
-Lynchburg,; Virginia 24506-0935' BWlik"fc%r
SUMMARY
This report describes the results of the examination of the first capsule (Capsule W-97) of the Entergy Operations, Inc., Waterford Generating Station, Unit No. 3 reactor vessel surveillance program. The objective of the program is to monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reacter vessel materials by the testing and evaluation of tension and Charpy impact specimens. The program was designed in accordance with the requirements of ASTM Specification E185 73.
The capsule received an average fast fluence of 6.47 x 10 18 n/cm' (E > 1.0 MeV) and the predicted fast fluence for the reactor vessel T/4 incation at the end of the fourth cycle is 2.74 x 10 18 n/cm' (E > 1 MeV) . 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 Waterford Generating Station, Unit No. 3 reactor pressure vessel inside surface will receive in 40 calendar years of operation is 3.69 x 10 I9 n/cm2 (E > 1 MeV) and tha corresponding T/4 fluence is calculated to be 1.97 x 10 l9 n/cm' (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 ft-lb transition temperature and a decrease in upper-shelf energy. These results demonstrated that the current techniques ased for predicting the change in both the increase in the RT NDT and the decrease in upper-shelf properties due to irradiation are conservative.
- ii -
S Q Nn5 E Eo N m
CONTENTS Page
- 1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1
- 2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
- 3. SURVEILLANCE PROGRAM DESCRIPTION ... . . . . . . . . . . . . . 3-1
- 4. PRE-lRRADIATION TESTS . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Tension Impact Tests Tests . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2. . . . . . . . . . . . . . . . . . . . . . . . . 4-2
- 5. POST-IRRADIATION TESTING . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Visual Examination and Inventory . . . . . . . . . . . . . . 5-1 5.2. Thermal Monitors . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3. Tension Test Results . . . . . . . . . . . . . . . . . . . . 5-1 5.4. Charpy V-Notch Impact Test Results . . . . . . . . . . . . . 5-2 C. NEUTRON FLUENCE . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2. Vessel Fluence . . . . . . . . . , . . . . . . . . . . . . . 6-4 6.3. Capsule Fluence . . . . . . . . . . , . . . . . . . . . . 6-5 6.4. Fluence Uncertainties . . . . . . . . . . . . . . . . . . . 6-6
- 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 7.4. Operating Limitations . . . . . . . . . . . . . . . . . . . . 7-5 7.5. Pressurized Thermal Shock (PTS) Evaluation . . . . . . . . . 7-5 7.6. Neutron Fluence Analysis . . . . . . . . . . . . . . . . . . 7-5
- 8.
SUMMARY
OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . . 8-1
- 9. SURVEILLANCE CAPSULE REMOVAL SCHEDULE , . . . . . . . . . . . . . . 9-1
- 10. CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1
- iii -
nmasw &:= Lean GUSSERVICE COMPANY
Contents (Cont'd)
APPENDIXES Page A. Reactor Vessel Surveillance Program Background Data and Information . A-1 B. Pre-Irradiation Tensile Data . . . . . . . . . . . . . . . . . . . . B-1 C. Pre-Irradiation Charpy impact Data . . . . . . . . . . . . . . . . . C-1 D. Fluence Analysis Methodology . . . . . . . . . . . . . . . . . . . . D-1 E. Capsule Dosimetry Data . . . . . . . . . , . . . . . . . . . . . . E-1 F. Tension Test Stress-Strain Curver . . . . . . . . . . . . . . . . . . F-1 G. References ......... . . . . . . . . . . . . . . . . . . . G-1 List of Tables Table 3-1. Specimens in Surveillance Capsule W-97 . . . . . . . . . . . . . 3-2 3-2. Chemical Composition and Heat Treatment of Surveillance Materials . 3-3 5-1. Conditions of Thermal Monitors in Capsule W-97 . . . . . . . . . 5 5-2. Tensile Properties af Capsule W-97 Base Metal and Weld Metal Irradiated to 6.47 x 10' n/cm' (E > 1 MeV) . . . . . . . . . . . 5-4 5-3. CharpyImpactResultsforCapsuleW-97BaseMetalLongitudinal (LT) Orientation, Heat No. M-1004-2, 6.47 x 10'" n/cm . . . . . . 5-5 5-4. Charpy Impact Results for Capsule W-97 Base Metal Transverse (TL) Orientation, Heat No. M-1004-2, 6.47 x 10 n/cm' . . . . . . 5-5 5-5. Charpy impact Results for Capsule W-97 Base Metal Heat Affected Zone Material, Heat No, M-1004-2, 6.47 x 10 n/cm' ... . . . 5-6 5-6. Charpy Imp"act Results for Capsule W-97 Weld Metal, 8811 A/t.45, 6.47 x 10 n/cm' . . . . . . . . . . ... . . . . . . . . . . . . 5-6 6 1. Surveillance Capsule Dosimeters . . . . . . . . . . . . . . . . . 6-6 6-2. Waterford Unit 3 Reactor Vessel f ast Flux . . . . . . . . . . . . 6-7 6-3. Calculated Waterford Unit 3 Reactor Vessel Fluence . . . . . . . 6-8 6-4. Calculated Waterford Unit 3 Reactor Vessel DPA . . . . . . . . . 6-9 6-5. Fluence, Flux, and DPA for 97* Surveillance Capsule . . . . . . . 6 6-6. Surveillance Capsule Measurements . . , . . . . . . . . . . . . 6-10 6-7. Axial Power Data Affecting Flux.. . . . . . . , , . . . , . . . . 6-11 7-1, Comparison of Waterford Unit 3, Capsule W-97 Tension Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . 7-6 7-2. Summary of Waterford Unit 3 Reactor Vessel Surveillance Capsules Tensile Test Resuits . . . . . . . . . . . . . . . . . . . . . . 7-7 7-3. Observed Vs. Predicted Changes for Capsule W-97 Irradiated Charpy:
Impact Properties - 6.47 x 10 18 n/cm2 (E > 1 MeV) . . . . .-. . . 7-8 7-4. Evaluation of Reactor Vessel End-of-Life (32 EFPY) Fracture Toughness - Waterford Unit 3 . . . . . . . . . . . . . . . . . . . 7-9 7-5. Evaluation of Reactor Vessel End-of-Life p2_EFPY) Upper-Shelf Energy - Waterford Unit 3 . . . . . . . . . . . . . . . . . . . . 7-10 7-6. Evaluation of Reactor Vessel End-of-Life Pressurized-Thermal Shock Criterion - Waterford Unit 3 . . . . . . . . . . . . , . . . 7-11 iv -
SW#.LYdME-
4 Tables (Cont'd) i Table Page A-1. Unirradiated Impact Properties and Residual Element Content Data - I of Beltline Region Materials Used for Selection of Surveillance P: mram Materials - Waterford Unit No. 3 . . . . . . . . . . . . A A-2. L a and Quantity of Specimens Contained.in Each Irradiation Capsule Assembly . . . . . . . .=. . . . . . . . . . . . . . . . . A-4 B-1. Tensile Properties of Unirradiated Shell Plate Material, Heat No. M-1004-2, Longi tudinal . . . . . . . . . . . . . . . . . . B-2 B-2. Tensile Properties of Unirradiated Shell Plate Material, ;
Heat No. M-1004-2, Transverse . . . . . . . . . . . . . . . . . . . B-2 1 B-3. Tensile Properties of Unirradiated Shell Plate HAZ Material, '
Heat No. M-1004-2, Transverse . . . . . . . . . . . . . . . . . . . B-3 B-4. Tensile Properties of Unirradiated Weld Metal 88114/0145 . . . . . B-3 C-1. Charpy impact Data from Unirradiated Base Material, Longitudinal Orientation, Heat No. M-1004-2 . . . . . . . . . . . . . . . . . C-2 C-2. Charpy impact Data from Unirradiated. Base Material, Transverse Orientation, Heat No. M-1004-2 . . . . . . . . . . . . . . . . . . C-2 C-3. Charpy impact Data from Unirradiated Base Metal, Heat-Affected-Zone , He at No. M- 1004-2 . . . . _ . . . . . . . . . . . . . . . . . C-3 C-4. Charpy Impact Data from Unirradiated Weld Metal, 88114/0145 . . . . C-3 D-1. Flux Normalization Factor for 97* Capsule . . . . . . . . . . . . . D-8 E-1. Detector Composition and Shielding . . . . . . . . . . . . . . . . E-2 E-2. Measured Specific Activities (Unadjusted) for Dosimeters in 97* Capsule . . . . . . . . . . . . . .. . . . . . . . . . . . . E-3 E-3. Dosimeter Activation Cross Sections, b/ atom . . . . . . . . . . . . E-4 List of Fiaures Figure 3-1, Reactor Vessel Cross Section Showing Location of Capsule W-97 in Waterford Unit 3 . . . . . . . . . . . . . . . .-. .-. . . . . . 3-4 3-2. Typical Surveillance Capsule Assembly Showing Location of Specimens and Monitors . . . . . . . . . . . . . . . . ... . . . 3-5 3-3. Typical Surveillance Capsule Tensile - Monitor Compartment-Assembly (Three per Capsule) . . . . . . . . . . . . . . . . . . 3-6 3-4. Typical Surveillance Capsule Charpy Impact Com- . ment Assembly (Fouc per Capsule) ............. ._. . . . . . . . 3-7 5-1. Photographs of Thermal Monitor. Helt Wire Caps ..es as-Removed From Surveill ance Capsule . - . . . . . . . . . . . . . . . . . . . . 5-7 5-2. Photographs of Tested Tension Test Specimens and Corresponding.
Fractured Surfaces - Base Metal, Transverse Orientation , . . . . .-5-8 5-3. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal Meat-Affected Zone- . . . . . . . 5 -v-SWM!Mh
- . _ = .
Fiaures (Cont'd)
Figure Page 5-4. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Weld Metal 88114/0145 . ... . . . . . . . . 5-10 5-5. Charpy impact Data for Irradiated Base Metal, longitudinal Orientation, Heat No. M-1004-2 . . . . . . . . . . . . . . . . . 5-11 5-6. Charpy Impact Data for Irradiated Base Metal, Transverse Orientation, Heat No, M-1004-2 ..................b12 5-7. Charpy impact Data for Irradiated Base Metal, Heat-Af'ected Zone, Heat No. M- 1004-2 . . . . . . . . . . . . . . . . . . .. . . 5-13 5-8. Charpy Impact Data for Irradiated Weld Metal, 88114/0145. . . . . 5-14 5-9. Photographs of Charpy Impact Specimen Fracture Surfaces -
Base Metal, Longitudinal . . . . . . . . . . . . . . . . . . . . 5-15 5-10. Photographs. of Charpy Impact Specimen Fracture Surfaces -
Base Metal, Transverse . . . . . . . . . . . . . . . . . . . . . 5-16 5-11. Photographs of Charpy Impact Specimen Fracture Surfaces -
Base Metal, Heat-Affected Zone . . . . . . . . . . . . . . . . . 5-17 5-12. Photographs of Charpy Impact Specimen Fracture Surfaces -
Weld Metal 88114/0145 . . . . . . . . . . . . . . . . . . . . . . 5-18 6-1. General Fluence Determination Methodology . . . . . . . . . . . . 6-2 6-2. Fast Flux, Fluence and DPA Distribution Through Reactor Vessel Wall ........................... 6-12 6-3. Azimuthal M ux and Fluence Distributions at Reactor Vessel Inside Surtace . . . . . . . . . , . . . . . . . . . . . . . . . 6-13 6-4. Relative Axial Variation of E > 1 MeV Flux / Fluence . . . . . . . 6-14 6-5. Radial Dimensions Used in Modeling Capsule and Pressure Vessel Regions .........................6-3 7-1. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Plete Materiti Longitudinal Orientation, Heat No. M-1004-2 . 7-12 7-2. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Plate Material Transverse Orientation, Heat No. M-1004-2 . . 7-13 7-3. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal, Heat-Affected-Zone, Heat No. M-1004-2 . . . . . . 7-14 7-4. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Weld Metal 88114/0145 . . . . . . . . . . . . . . . . . . . . 7-15 A-1. Location and Identification of Materials Used in the Fabrication of Waterford Unit 3 Reactor Pressure Vessel , , . . . . . . . . . A-5 A-2. Locatica of Beltline Region Materials in Relationship to the Reactor Vessel Core . . . . . . . . . . . . . . . . . . . . . . . . A-6 A-3. Location of Longitudinal Welds in Waterford Unit 3 Upper and Lower Shell Courses . . . . . . . . . . . . . . . . . . . . . . . . A-7 A-4. Location of Surveillance Capsule Irradiation Sites in Waterford Unit 3 . . . . . . . . . . . . . . . . . . . . . . . . . A-8 C-1. Charpy Impact Data froc Unirradiated Base Metal (Plate),
Longitudinal Orientation, Heat No, M-1004-2 . . . . . . . . . . . . C-4 C-2. Charpy Impact Data from Unirradiated Base Metal (Plate),
Transverse Orientation, Heat No. M-1004-2 . . . . . . . . . . . . . C-5 C-3, Charpy Impact Data from Unirradiated Heat-Affected-Zone Base Metal , Heat No. M-1004-2 . . . . . . . . . . . . . . . . . . . C-6
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Fiaures (Cont'd)
Figure Page C-4. Charpy Impact Data for Unirradiated Weld Metal, 88114/0145 . . . . C-7 C-5. Charpy impact Data for Unirradiated Correlation Monitor Material . C-8 D 1. Rationale for the Calculation of Dosimeter Activities and Neutron Fl ux in the Capsul e . . . . . . . . . . . . . . . . . . . . . . . . D-9 D-2. Rationale for the Calculation of Neutron Flux in the Reacter Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . D-10 0-3. Plan View Through Reactor Core Midplane (Reference R-e Calculation Model) . . . . . . . . . . , . .-. . . . . . . . . . 0-11 F-1. Ten: ion Test Stress-Strain Curve for Base Metal Plate Heat H-1004-2, Specimen No. 2L6, Tested at 70F . . . . . . . . . . . . . F-2 F-2. Tension Test Stress-Strain Curve for Base-Metal Plate Heat M-1004-2, Specimen No. 2K5, Tested at 250F . . . . . . . . . . . . F-2 F-3. Tension Test Stress-Strain Curve for Base Metal Plate Heat M-1004-2, Specimen No. 2K2, TestedHat 550F . . . . . . . . . . . . F-3 F-4. Tension Test Stress-Strain Curve for Base Metal Heat Affected Zone, Heat H-1004-2, Specimen No. 4K3, Tested at 70F . . . . . . . F-3 F 5. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone, Heat M-1004-2, Specimen No. 4KK, Tested at 250F , . . . . . . F-4 F-6. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone, Heat M-1004-2, Specimen No. 4J4, Tested at 550F . . . . . . . F-4 F-7. Tension Test Stress-Strain _ Curve for Weld Metal 88114/0145, Specimen No. 3JM, Tested at 70F . . . . . . . . . . . . . . . . . F-5 F-8. Tension Test Stress-Strain Curve for Weld Metal 88114/0145, Specimen No. 3KK, Tested at 250F . . . . . . . . . . . . . . . . . F-5 F-9. Tension Test Stress-Strain Curve for Weld Metil 88114/0145,
-Specimen No. 3KY, Tested at 550F . . . . . . . . . . . . . . . . . F-6
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- l. INTRODUCTION This report describes the results of the examination of the first capsule (Capsule W-97) of the Entergy Operations, Inc., Waterford Generating Station, Unit N9 3 (Waterford Unit 3) reactor vessel material surveillance program (RVSP). The capsule was removed and evaluated after being irradiated in the Waterford Unit 3 reactor as part of the reactor vessel materials surveillance program (Combustion Engineering (C-E) Report C-NLM-003') . The capsule 18 experienced a fluence of 6.47 x 10 n/cm' (E > 1 MeV), which is the equivalent of approximately six effective full power years' (EFPY) operation of the Baterford Unit 3 reactor vessel inside surface.
The objective of the program is to moritor the effects- of neutron irradiation on the tensile and impact properties or reactor _ pressure vessel materials under actual operating conditions. The surveillance program for Waterford Generating Station Unit No. 3 was designed and furnished by Combustion Engineering, Incorporated (C-E) as described in TR-C-MCS-0012 and conducted in accordance with 10CFP50, Appendix H , 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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- 2. BACKGROUND i
l The ability of the reactor pressure vessel to resist fracture is the primary l 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 region !
f of the vessel because it is exposed to neutron irradiation. The general effects f
j of fast neutron irradiation a the mechanical properties of low alloy ferritic j sted s cuch as SA533, Grade B, used in the fabrication of the Waterford Unit 3 j 9 actor vessel, are well characterized and documented in-the literature. The h low alloy ferritic steels used in the beltline region of reactor vessels exhibit l
an increase in ultimate and yield strength properties with a corresponding i decrease in ductility after i rradiation. The most significant mechanical property change in reactor pressure vessel steels is the increase in temperature l
i for the transition from brittle to ductile fracture accompanied by a reduction l in the Charpy upper shelf energy value. :
Appendix G to 10CFR50, " Fracture Toughness Requirements,"' specifies minimum i l
fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of eater cooled power reactors, and provides specific guidelines for determining the ,
pressure-temperature limitations for operation of the RCPB. .The toughness and worational requirements are specified to provide adequate safety margins during_
1
! any condition of normal operation, including anticipated operational occurrences l and system hydrostatic tests, to which the pressere boundary may be subjected j over its. service lifetime. Although the requirements-of Appendix G to 10CFR50 l became effective on August 16, 1973, the requirements are applicable to all l boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effective date.
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l Appendix H to 10CFR50, " Reactor Vessel Materials Surveillance Program Requirements,"' defines the material turveillance program required to monitor 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 thermal environment. Fracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel. These data will permit determination of tFe 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 (B&PV) Code, Section 111. "Nulear Power Plant Components." This method utilizes fracture mechanics cancepts and the reference nil ductility temperature, RTt4DT, which is defined as the greater of the drop weight nil-ductility ttansition temperature (per ASTM E 208') or the temperature that is 60F below that at which the matcrial exhibits 50 ft lbs and 35 mils lateral expansion. The RT n' a given material liDT is used to index that material to a reference stress intensity factor curve (K IR curve), which appears in Appendix G of ASME B&PV Code Section 111. The K curve IR is a lower bound of dynamic, static, and crack arrest fracture toughness results obtained from several heats of presrure vessel steel. When a given material is indexed to the K IR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. A"owable operating limits can then be determined using these allowable stress insensity factors.
The RT and, in turn, the operating limits of a nuclear power plant, can be t4DT 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 containing 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 t4DT for radiation embrittlement. This adjusted RT is used to index the material t4DT to the K IR cut ve which, in turn, is used to set operating limits for the nuclear 2-2 13 W itsefEi h r
, power plant. These new limits take into account the effects of irradiation on the reactor vessel materials. !
Appendix G,10CFR50, also requires a minimum Charpy V-notch upper-shelf energy l l of 75 ft lbs for all beltline region materials unless it is demonstrated that l 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.
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- 3. SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for Waterford Unit 3 comprises six surveillance capsules designed to monitor the effects of neutron 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 near the inside wall of the reactor vessel at the locations shown in Figure 3-1.
The six capsules, designed to be placed in holders attached to the reactor vessel wall are positioned near the peak axial and azimuthal neutron flux. During the four cycles of operation, Capsule W 97 was irradiated in the 97' position .
adjacent to the reactor vessel wall as shown in Figure 3-1.
Capsule W-97 was removed during the fourth refueling shutdown of Waterford Unit
- 3. The capsule cont':ned Charpy V-notch impact test specimens fabricated from the one base metal (SA533, Grade B1) both longitudinal and transverse orienta-tion, oae heat affected-zone, and a weld metal. Tension test specimens were fabricated from the base metal, heat-affected-zone, and weld metal. The number of specimens of each material contained in the capsule are described in Table 3-1, and the location of the individual specimens within the capsule are described ir. Figures 3-2 through 3-4. The chemical composition and heat treatment of the surveillance material in Capsule W-97 are described in Table 3-2.
All plate and heat-affected zone specimens were machined from the 1/4-thickness 3 (1/4T) location of the plate material. Weld metal specimens were machined throughout the thickness of the weldment. Charpy V-notch and tension test specimens were cut from the surveillance material such that they were oriented tsith their longitudinal axes either parallel or perpendicular to the principal working direction.
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_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ l
The neutron dosimeters contained in Capsult ..-97 are as follows:
Threshold Material Shieldina Reaction Enerov (Mev) Half-Life _
Uranium None/Cd O' (n, f) Sr" 0.7 28.0 years Sulfur None S8 ' (n.p) p ar 2.9 14.3 days Iron None Fe" (n.p) Mn" 4.0 312.5 days Nickel Cd Ni" (n.p) Co" 5.0 70.9 days Copper Cd Cu (n.a) Co" 7.0 5.27 years Titanium None Ti (n r) Sc 8.0 83.8 days Cobalt None/Cd Co" (n,y) Co" Thermal 5.27 years Four thermal monitors of low melting alloys were included in the W 97 capsule.
The eutectic alloys and their melting points are as follows:
Melting Allov Composition. wt% Point. F 80.0 Au, 20.0 Sn 536
{ 90.0 Pb, 5.0 Sn, 5.0 Ag 558 l 97.5 Pb, 2.5 Ag 580 l 97.5 Pb, 0.75 Sn, 1.75 Ag 590 i
t lable 3-1. Specimens in Surveillance Capsule W-91
)
! Number of l Test Soecimens i LQts_r rial Description Tension CVN ImDBCt l Base Metal (M 1004-2) l- Longitudinal -
12 l Transverse . 3 12 4
Heat-Affected Zane 3 12 1
- Weld Metal (68114/0145) 1 12
- j. Total Per Capsule 9 48 --
r l 32
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Table 3-2. Chemical Composition and Heat Treatment of Surveillance Malerials Chemical Comonsitiot.. w/o Heat No.w Weld Metal Element H-1004-2 88114/0145" C 0.23 0.23 Mn 1.38 1.35 P 0.005 0.008 S 0.005 0.005 Si 0.23 0.16 Ni 0.58 0.22 Cr 0.01 0.05 Mo 0.57 0.57 Cu 0.03 0.04 Heat Treatment Heat No. Temp. F lime. h Coolina _,
Plate 1575150 4 Water Quenched (M 1004-2) 1220125 4 Furnace Cooled 1150125 40 furnace Cooled to 600F Weld Metal 1100-1175 40 1/2 Furnace Cooied to 600f (88114/0145)
(a) Chemical analysis by Combustion Engineering of surveillance program test plate.'
(b)ChemicalanalysigbyCombustionEngineeringofsurveillanceprog.am test weld metal.
3-3 BWURE?a%r l
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Figure 3-1. Reactor Vessel Cross section showing Location of Camule W 97 in Waterford Unit 3 18(P 4
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l 4. PRE IRRADIATION TESTS 1
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 material properties to the extent practical from available j material, as required for compliance with Appendixes G and H to 10CFR50.
The pre-irradiated specimens were tested by Combustion Engineering as part of the development of the Waterford Unit 3 surveillance program. The details of the testing procedures are described in C-E Repsrt TR-C-HCS-002' and are summarized here to provide continuity.
4.1. Tension Tests Tensiot. test specimens were fabricated from the reactor vessel shell plate, HAZ 4 metal, and weld metal. The specimens were 3.00 inches long with a reduced section 1.50 inches long by 0.250 inch in diameter. The tensile terts were performed using a Riehle universal screw testing machine with a maximum capacity of 30,000 lb and separate scale ranges between 50 lb and 30,000 lb. The machine -
is capable of constant cross head rate or constant strain rate operation.
Elevated temperature tests were performed in a 21/2" 10 x 18" long high temperature tensile testing furnace with a temperature limit of 1800"F. A Riehle high temperature, dual range extensometer was used for monitoring specimen elongation.
Tensile testing was conducted in accordance with ASTM E 8, " Tension Tests of Metallic Materials:"' and/or Recommended Practice E-21, "Short-Time Elevated Teeperature Tension Tests of Materials, except as modified by Section 6.1 of Recommended Practice E-184, " Effects of High-Energy Radiation on the_ Mechanical Proporties of Metallic Matereials."'
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i For each materia' type and/or condition, nine specimens in groups of three were i
tested at room temperature, 250 and 550F. All test data for the pre-irradiation l tensile specimens are given in Appendix B.
4.2. Impact Tests Charpy V-notch impact tests were conducted in accordance with the requirements ;
of ASTM E23-72" on a Model SI-l BLH Sonntag Universal Impact Machine certified i to meet Watertown standards." Test specimens were of the Charpy V-notch type, which were nominally ').394 inch square and 2.165 inches lorig.
Impact test data for the unirradiated baseline reference matarials are presented t in Appendix C. Tables C-1 through C 4 contain the basis data that. are plotted in Figures C-1 through C-4. These data were replotted and re-evaluated to be consistent with the irradiated Charpy curves and evaluations.
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i 1
- 5. POST-IRRADIATION TESTING 5.1. Visual Examination and Inventory The capsule was inspected and photographed upon receipt and confirmed that the 1 markings as those of Capsule W-P7. The contents of the capsule were inventoried and found to be consir, tent with the surveillance program report inventory. All specimens were visually examined and no signs of abnormalities were found. There was no evidence of rust or of the penetration of reactor coolant into the capsule.
5.2. Thermal Monitors Surveillance Capsule W-97 contained three temperature monitor holder blocks each containing four fusible alloys with different melting points. Each of the thermal mnitors was inspected and the results are tabulated in Table 5-1.
Photographs of the monitors are shown in Figure 5-1.
From these data, it can be concluded that the irradiated specimens had been l- exposed to a maximum temperature no greater than 580F during the reretor vessel operating period. This is not significantly greater than the nominal inlet ,
temperature of 550F, and is considered acceptable. However, the partly melted or slumped appearance -of the 558F monitor is probably due to an irradiation induced creep mechanism and not the result of actual melting. This being the case, then the maximum temperature was no greater than 558F which is the most likely case. This behavior has been seen in other surveillance capsules. There appeared to be no signs of a significant temperature gradient along the capsule length.
5.3. Tension Test Results The results of the post-irradiation tension tests are presented in Table 5-2.
Tests were performed on specimens at room temperature, 250, and 550F. They were tested- on a 55,000-1b load capacity MTS servohydraulic - computer-controlled universal test machine. All tests were run using stroke control with an initial 5-1 BWltKML
actuator travel rate of 0.005 inch pier minute through yield point. Past specimen yielding an actuator speed of 0.040 inch per minute was used. A 4 pole exteasion device with a strain gaged extensometir was used to determine the 0.2% yield point. Test conditions were in accordance with the applicable requirements of ASTM A370 77." For each material type and/or condition, specimens were tested at room temperature, 250 and 550f. The data for both the heat affect zone specimen and the weld metal specimen, tested at 250F, were lost because of a test machine malfunction. The tension-compression load cell used had a certified accuracy of better than +0.5% of full scale (25,000 lb). Photographs of the tension test specimen fractured surfaces are prosented in figures 5-2 through 5-4.
In general, the ultimate strength and yield strength of the material increasec with a corresponding slight decrease in ductility as compared to the unirradiated values; both effects were the result of neutron radiation damage. The type cf behavior observed and the degree to which the' material properties changed is within the range of changes to be expected for the radiation environment to which the specimens were expcstd.
The results of the pre-irradiation tension tests are presented in Appendix B.
5 4. 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 5-3 through 5-6 and figures 5-5 through 5 8. Photographs of the Charpy specimen fractere surfaces are presented in figures 5-9 through 5-12. The Charpy V-notch impact tests were conducted in accordance with the requirements of ASTM E23-88" on a Satec SI-lK impact tester certified to meet Watertown standards'.'
The data show that the materials exhibited a sensitivity to irradiation within the values to be evpected based on their chemical composition and the fluence to whi:.h they were exposed. Detailed discussion 'of the results are provided in Section 7.
The results of the pre-irradiation Charpy V-notch impact tests are given in Appendix C.
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lable 5-1. Conditions of Thermal Monitors in Caosule W-97 Capsule Melt Post-Irradiation Seament Jrnstrature Condition Al 536F Melted (Top) 558F Melted (slumped?)
580F Unmelted 590F Unmelted A4 536F Melted (Middle) 558F Palted (slumped?)
580F Unmelted 590F Unmelted A7 536F Helted (Bottom) 550F Melted (slumpeJ?) :
580F Unmelted 590F Unmelted
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_ _ _ _ - . . . . - . . . _ . _ _ _ _ . ~ . . . . . _ . _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . - - . . _ . _ _ _ . - . _ _ . _ . . _ _ . _
i i Table 5-2. Tensile Properties of Capsule g-97 I Base Metal and Weld l Metal Irradiated to 6.47 x 10 n/cm' (E > 1 MeV)** l f
i !
Strenath, psi Fracture Elonaation. % Reduction Specimen Test Temp, Load, Stress, Strength, in Area, No. __ F Yield Ultimate lbs psi osi Uniforg Total %
i i i Base Metal. N-1004-2. Transverse i
- 2L6 70 70,400- 92,600 3,097 173,000 63,100 11.7 26.2 63.5 i
! 2K5 250 65,500 85,800 2,834 175,3C3 57,700 10.2 23.1 67.1 i
l 2K2 '550 63,500 90,000 2,994 162,900 61,000 10.2 23.0 62.5 t
4 Base Metal Heat-Affected Zone. M-1004-2 l V' i ** 4K3 70 69,500 93,500 2,844 184,700 57,500 7.0 70.3 68.9 i
i i 4KK ---. ---- ---- --- ---- --- ---- ---- ----
i i
! 4J4 550 69,600 91,000 2,913 194,700 59,300 6.4 18.5 E9.5
', Weld Metal 88114/0145-i 3JM 70 84,500 95,900 3,351 187,100 68,300 7.3 7.9* 63.5 i
i Q3 3KK --- ---- ---- ---- ---- ---- ---- ---- ----
EE 4
^
3KY 550 74,000 93,200 2,766 187,700 56,400 7.9 22.6 70.0 i
- i
- Results not valid . specimen necked and fractured outside extensometer gage length.
~** Stress-strain curves are presented in Appendix F.
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Table 5 3. Charpy Impact Results for Capsule W 9*/ Base Metal Longitudinal (LT) Orientation, Heat No. M 1004-2, 6.47 x 10 n/cm i
Test Impact lateral Shear Specimen Temperature, Energy, Expansion, Fracture, ID F ft lbs inch % _
14Y 50 7.5 0.006 0 llD 25 24.0 0.021 0 13D 0 22.5 0.025 10 133 20 36.0 0.032 10 i ISC 30 84.5 0.061 40 12P 35 76.0 0.054 40 14C 50 72.5 0.057 40 15K 70 90.0 0.069 80 132 100 113.0 0.075 70 llE 150 156.0* 0.093 100 14K 200 152.0* 0.093 100 llY 550 157.0 0.082 100
- Values used to determino upper-shelf energy valt i per ASTM E-185.
Table 5 4. Charpy Impact Results for Capsule W-97 Base Metal Transverse (TL)
Orientation, Heat No. M 1004-2, 6.47 x 10 n/cm' 1 Test impact lateral Shear i
Specimen Temperature, Energy, Expansion, Fracture, i 10 F ft-lbs inch %
216 -50 7.0 0.006 0 214 -25 11.0 0.011 5 25A 0 15.0 0.017 5 210 10 36.0 0.031 10 22P 20 53.0 0.04( 20 26K 35 54.0 0.047 40 25K 50 73.0 0.057 50 217 70 71.5 0.057 100 22L 100 88.5 0.070 70 245 150 121.5* 0.086 100 244 200 125.0* 0.086 100=
22M 550- 124.0 0.081 100
- Values used to detemine upper-shelf energy value per ASTM E-185.
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Table 5 5. Charpy impact Results for Capsule W-97 Base Metal Heat-Affected Zone Material, Heat No. M-1004-2, 6.47 x 10 n/cm' Test impact Lt.teral Shear-Specimen Temperature, Energy, Expansion, Fracture, 10 F ft-lbs inch %
415 -100 21.0 0.014 5 46Y - 85 35.0 0.026 25 455 - 65 53.5 0.036 15 45J - 50 90.0 0.060 60 42C 0 115.5 0.071 80 ,
43D 10 101.0 0.007 70 44U 20 121.0 0.073 85 45K 50 119.5 0.077 85 45Y 70 155.0* 0.083 100 41M 100 163.5* 0.090 100 474 150 150.0* 0.071 100 -
414 550 >240.0 --. --- i
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Table 5-6. Charpy Impact Results for Capsule W-97 Weld Met?l, 88114/0145, 6.47 x 10 n/cm' Test Impact lateral Shear Specimen Temperature, Energy, Expansion, Fracture, ID F ft-lbs inch- %
35J -50 7.0 0.006 0 32T -25 11.0 0.011 5 31M -0 15.0 0.01/ 5 32K 10 36.0 0.031 10 362 20 53.0 0.044 20 34L 35 54.0 0.047 40 31T 50 73.0 0.057 50 326 70 71.5 0.057 100 35K 100 88.5 0.070 70 33C 150 121.5* 0.086 100 374 200 125.0* 0.086~ 100 33B 550 124.0 0.081 100
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- 6. NEUTRON FLUENCE 6.1. Introduction The neutron fluence (time integral of flux) is a quantitative way of expressing the cumulative exposure of a ma.erial to a pervading neutron flux over a specific period of time. Fast neutron fluence, defined as the fluence of neutrons having energies greater than 1 MeV, is the parameter that is presently 'ised 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 _ 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 tne fluence, the fluence gradf ent 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) t iey provide only rudimentary information abcut 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, l measurements and calculations are necessary complements of each other. and together thoy 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 described in the following paragraphs. The details of the procedures and methods are presented in general terms in Appendix D-and in BAW-1485P." 6-1 i BWHEEVa L r
i 4 l Fioure 6-1. General Fluence Determination Methodoloov s
- Measurements of Neutron Analytical Determination of !
! Dosimeter Actitivies Dosimeter Activities and Neutron flux-( Adjusted Energy Dependent Neutron l Flux l Reactor Operating ! Neutron History and Pre-i Fluence dicted Future
- operation i-l 4
i i i Analytical Determination of Dosimeter Activities and Neutron Flux l The analytical calculation of the space and energy dependent neutron flux in the '
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{ quadrature of 48 sectors (S8), a third order LeGendre polynomial scattering approximation (P3), the BUGLE cross section set" with 47 neutron energy groups . ! and a fixed distributed source corresponding to the time weighted average power distribution for the applicable irradiation period. i j- In addition to the flux in the test specimens, the D0TIV calculation determines j the saturated specific activity of the various neutron' dosimeters located in the surveillance capsule using.the ENDF/B5 dosimeter reaction cross sections." The saturated activity of each dosimeter is then adjusted by a factor which corrects for the fraction of saturation attained during the dosimeter's actual:-(finite)- 4 4 j 6-2 l BWMWe%
irradiation history. Additional corrections are made to account for the following effects:
- Photon . induced fissions in V dosimeters (without this correction the results underestimate the measured activity).
- Short half-life of isotopes produced in nickel, iron and titanium dosimeters (71 day 00-58, 312 day Mn-54 and 84 day Sc-46 respectively).
(Without this correction, the results could be biased high or low depending on-the long-term versus short term power histories.) Measurement of Neutron Dosimeter Activities The accuracy of neutron fluence predictions is improved if the calculated neutron flux is comoared with neutron dosimeter measurements arijusted for the effects noted above. The neutron dosimeters located in the surveillance capsules are listed it Table 6-1. Both activation type and fission type dosimeters were used. The ratio of measured dosimeter activity to calculated dosimeter activity (M/C) is determined for each dosimeter, as discussed in Appendix D. These M/C ratios are evaluated on a case-by-case basis to assess the dependability or veracity of cach individual dosimeter response. After carefully evaluating all 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 DOT-calculated flux at all points of interest. Neutron Fluence The determination of the neutron fluence from the time averaged flux requires only a simple multipl_ication by the time in EFPS (effective full-power seconds) over which the flux was averaged, i.e. fij ( A T) =[,43,,hT where f (AT) - Flucce at (i,j) accumulated over time T (n/cm'), g g - Energy group index, 6-3 SWR &Wc% c _ - - _ - - _ - _ u
= Time-average flux at (1,j) in energy group g, (n/cm'-sec),
us AT - Irradia+1on time, EFPS. Neutron fluence was calculated in this analysis for the following components over the indicated operating time: 4 Test Specimens: Capsule irradiation time in EFPS Fluence Monitors: 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 atom (DPA) of iron. The iron DPA is an exposure index giving the fraction of iron atoms in an iron specimen which would be displaced during an irradiation. 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).* A DFA cross section for iron is given in the ASTM Standard in 641 energy groups. DPA per second is determined by multiplying 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 47 neutron group structure of BUGLE; the DPA was then determined by summing the group flux times the DPA cross section over the 47 energy groups 4 and multiplying by the time of the irradiation. 5.2. Vessel Fluence The maximum fluence (E > 1 MeV) exposure of the Waterford Unit 3 reactor ves:el during Cycles 1 to 4 was determined to be 5.13 x 1038 n/cm' based on a maximum neutron flux of 3.66 x 10' n/cm 2-s. The maximum fluence occurred at the clad-ding / vessel interface at an azimuthal location of approximately 1 degree from a major horizontal axis of the core (Figure 6-3). Cumulative DPA results were calculated at the quarter T positions and are presented in Table 6-4. 6-4 BWiinMafeLe
l [ Fluence data were extrapolated to 32 EFPY of operation based on two assumptions: 1 (1) the future fuel cycle operations do not differ significantly from the cycles-l 1 to 4 design, and (2) the latest calcalated (or extrapolated) flux remains l constant from E004 through 32 EFPY. The ertrapolation was carried out.from EOC l 4 to 32 EFPY. The cycle averaged fluxes for future cycles are assumed to be the ! average flux experienced during cycles 1 to 4. l Fast fluence and DPA (displacements per atom) gradients relative to the-inside f surface of the vessel wall are shown in Figure 6-2. Reactor vessel neutron j fluence lead factors, which are the ratio of the neutron flux at the clad i interface to that in the vessel wall at the T/4, T/2 and 3T/4 locations, are l 1.87, 4.03, and 9.17, respectively. DPA lead factors at the same locations are , 1.62, 2.79, and 5.00, respectively. The relative fluence as a function of azimuthal angle is shown in Figure 6-3. The peak average flux from cycles 1 to 6 occurred at about I degree with a corresponding value of 3.66 x 10' n/cm'- s . 4 l The flux and fluence results were corrected using the final measured to l calculated activity ratio (M/C) derived from the capsule (0.958) and were also { corrected to account for an axial power peak (1,08). The M/C ratio is detailed l in Appendix D. The axial fluencs, which was normalized-over the height of the ! core and assumed to be proportional to the axial power distributions in the-f peripheral assemblies, was averaged over cycles 1 to 4. Table 6 7 shows the l nodal values used to obtain the axial factors. These values were based on time-l averaged nodal values obtained from the customer. Figure 6-4 shows the axial i flux variation, overlaid by an image of the capsule showing the axial factors in } each dosimeter compartment. l. 6.3. Caosule Fluenca i The 97* capsule was irradiated in Waterford Unit 3 for Cycles 1 to 4, 4.44 EFPY, ?
- at a location 7. degrees off a major horizontal axis. The cumulative fast fluence i
5 at the center of the surveillance capsule was calculated to be 6.47 x 10 n/cm'.
- j. lhis fluence valua represents an average-value for the-various locations in the capsule. -It includes an axial peaking factor of 1.08 and a normalization factor
; of 0.958. The fluence is .approximately 6% higher at the : center of the charpy 1
j; specimens closest to the-core and approximately 6%-lower at the center of the 4
!- 6-5 f3EEErEs"cdEmv I -
charpy specimens away from the core. Figure 6-5 shows a sketch of__the' capsule and pressure vessel, which includes the radial dimer.sions from the core center supplied by the customer, although the dimensions-have been converted from the inches in which the informatica was supplied to the~ centimeters which were used in the modelling. 6.4. Fluence Uncertainties Surveillance capsules provide neutron dosimetry information as well as materials - data at various points during the lifetime of power reactors. -The dosimetry results, measured-to-calculated ratios, obtained from numerous analyses utilizing the same methodology provide a measure of confidence in the analytical techniques and a benchmarking for the methodology used to determine vessel fluence. Table = 6-6 presents a comparison of the results of fourteen surveillance capsule analyses which utilized B&W's methodology. I Table 6-1. Surveillance Caos'ule Dosimeters Lower Energy 3 Limit for- Isotope Dosimeter Reactions Rgg+ ion. MeV Hal f-L i fe
- Ni (n . p )Co 2.3 70.8 days 5'Fe (n , p)Mn 2.5 312.5 days
Cu (n ,a)"Co 3.7 5.27 years
T i (n , p) **Sc .1.9 83.8) days assV(n , f)'87Cs 1.1 30.0 years
Co (n , y)"Co thermal 5.27 years-
'*% action activities measured for capsule flux evaluation.
I 6 SW##efhr
. - - - = - - - - . - - . . - . . .
Table 6-2. Waterford Unit 3 Reactor Vessel Fast-Flux flux n/cm'-s Fast Flux (E > 1 MeV). n/cm'- s (E > 0.1 MeV) Inside Surface ** Inside Surface Cycle (Max location) T/4 3T/4 (Max location) Cycles 1 to 4 3.66E+10 1.96E+10 3.99E+9 7.91E+10 5 EFPY 3.66E+10 1.96E+10* 3.99E+9* 6 EFPY 3.66E+10 1.96E+10* 3.99E+9* 7 EFPY 3.66E+10 1.96E+10* 3.99E*9* 8 EFPY 3.66E+10 1.96E+10* 3.99E+9* 16 EFPY 3.66E+10 1.96E+10* 3.99E+9*- 24 EFPY 3.66E+10 1.96E+10* 3.99E+9* 32 EFPY 3.66E+10 1.96E+10* 3.99E+9*
- Divide flux at inside surface by the appropriat, lead f actors on page 6-5 to obtain these T/4 and 3T/4 fast flux values.
** Clad / Base metal interface at 221.54 cm from core center.
6-7 G W .*f a r ci! L v
i l Table ( 3. Calculated Waterford Unit 3 Reactor Vessel Fluence-Fast Fluence. n/cm' (E > l- MeV) Cummul ative Inside-Surface
- Irradiation Time (Max location) T/4' T/2 _,3T/4 j
- End of Cycle 4 5.13E+18 2.74E+18 1.27E+18 5.59E+17 5 EFPY 5.76E+18 3.08E+18* 1.43E+18* 6.29E+17*
6 EFP) 6.92E+18 3.70E+18* 1.72EtlS* 7.54E+17* 7 EFPY 8.07E+18 4.32E+18* 2.00E+18* 8.80E+17* l 8 EFPY 9.22E+18 4.93E+18* 2.29E418* 1.01E+18* 16 EFPY 1.84E+19 9.86E+18* 4.58E+18* 2. die +18* 24 EFPY 2.77E+19 1.48E+19* 6.87E+18* 3.02E+18* 32 EFPY 3.69E+19 1.97E+19* 9.15E+18* -4.02E+18*
*Calcalated using 1.00 1.67- 4.03 9.17 these lead factors.
" Clad / Base metal interface at 221.54 ci,1 -: core center.
Conversion Factors Fluence (E > 1 MeV) 1.50E-21** 1.72E-21** 2.16E-21** 2.73E-21** to DPA.
** Multiply fast fluence values (E > 1 MeV) in units of n/cm2 by these factors to obtain the corresponding DPA values.
a 6-8 G Wit M i k r
l l Table 6-4. Calculated Waterford Unit 3 Reactor Vessel DPA DPA. Displacements / Atom (Total) _ Cummulative inside-Surface"> Irradistion Time (Max location) T/4 T/2 3T/4 End of Cycle 1 7.67E-3* 4.73E-3* 2.76E-3* 1.63E-3* 5 EFPY 8.62E-3* 5.31E-3* 3.10E-3* -1.72L-3* ! 6 EFPY 1.03E-2* 6.38E-3* 3.72E-3* 2.06E-3* 1 EFPY 1.21E-2* 7.44E-3* 4.33E-3* 2.41E-3* 8 EFPY 1.38E-2* 8.50E-3* 4.95E-3* 2.75E-3* 16 EFPY 2.76E-2* 1.70E-2* 9.91E-3* 5.50E-3* 24 EFPY 4.14E-2* 2.55E-2* 1.49E-2* 8.25E-3* < 32 EFPY 5.51E-2* 3.40E-2* 1.98E-2* 1.10E-2*
- Clad / Base metal interface at 221.54 cm from core center.
- Calculated using these 1.00 1.62 2.79 5.00 lead factors Conversion Factors Fluence (E > 1 MeV) 1.50E-21** 1.72E-21** 2.15E-21**- 2.73E-21**
to DPA. ( ** Fast fluence values (E > 1 MeV) in units of n/cm2 were multiplied by these factors to obtain the corresponding DPA values. Table 6-5. Fluence. Flux. and DPA for 97* Surveillance Capsule E > 0.1 E > 1.0 MeV -MeV-Flux Fluence, Flux, Caosule Irradiation Time n/cm 3 n/cm 2 DPA n/cm 2 W-97 Cycles 1 to 4 4.62E+10 6.47E+18 9.25E-3 8.63E+10 (4.44 EFPY) l l 6-9 BWMAL%"%
Table 6-6. Surveillance Caosule Measureme011 Heasured/ , Plant Caosule Calculated Arkansas One, Unit 1 AN1-C 1.04 Rancho Seco RSI-F 1.03 Crystal River-3 CR3-F 0.99 Oconee Unit 1 OCl-C 1.04 Oconee Unit 2 .CC2-E 0.98 Davis-Besse DB1-LGI- 1.08 Crystal River-3 CR3-LG1 1.06 Oconee Unit 3 OC3-D -1.00 Davis-Besse TEl-D 1.03-St. Lucie W-83 1.08 Shearon Harris 0 0.88 Zion Unit 1 Y. 1.11 Millstone Unit 2 W-104 0.99 Millstone Unit 2 W-97 C.94 Average M/C for 14 surveillance data points - 1.02 1 Sigma standard deviation of data base = 0.06 10 S W A' M /SiiG5 v.
i i 4 4- ] Table 6-7. Axial Power Data Affectina Flux Hgde Exposure
- Relative Exo.
1 3.272- . 0.750 2 4.360 1.000-1 4.666' l.070~ 3 l 4 4.731 1.085 5 4.726 1.084 l 6 4.704 1.079 l 7 4.679 1.073 l 8 4.649 1.066 i 9 4.602 1.055 l' 10 4.491 1.030 1 11 4.179 0.958 l 12 L2.6.2 0.750 l Avg 4.351 . t l OThese exposure values are based upon the nodal-values for assemblies [9,1] and [9,2] supplied by the customer and time-averaged for. cycles 1 through 4. I 4 4 L l i i i 1 I I i i 4 l 6-11 l SWsTn%"cWmv
* -r- + . - ~ , - ,-s . <- - - , , , m ._,,,,.,,,np ,m , , - , r,--.-,e . . . - - - - , ,,. .ve v- -,w.ne,-, r- g.- -.- --
1 Figure 6-2. Fast Flux, Fluence and OPA Distribution
, Throuah Reactor Vessel Wall 1
h .
, '-, . _ . LF(D) = 1.62
'A '
.4 y -
_ . . _ _ . . _ _ _ _ . _ . '.a,_.._ __ . . _ _LF (D) =._2.7 9
'A 0.5
'O -.- ---
J21.54 cm ---_ b ' i Inside Surface 'A,A ,
-.Claa. . Base M.. eta. l ...-.--y, y
A ' 0.3 - ----- -- - - - - - ---
'b;- -- - - -- ---- ---
fg 227.02 cm T/4 N. N LF(F) = 1.87 .U
--- - --4, e
l 0.2
-- --- -= - - - - - -
5
'A ,
g - / 2 ,L. w ' T/2 i f o,1 - _. . . . . . - _ . - 232.50 cm
. LF(F) .= 4.03 . . _ - .. g _ . _._. - . 2.
'3, i
~
L 3T/4 0.05 - Rux/ Fluence DPA- --___ . _ .23Mem_ _ _. u ..6... LF(F) = 9.17
. .. . . .- . _ . . . . . .. g._.
LF(F)is Lead Factor for Fluence 243.45 cm h LF(D) is Lead Factor for DPA . Outside Surface 0.03 ' I I 220 225 230 235 240 245
, Radial Position (cm)-
6-12 B Wi!?% ct L r
Fiqure 6-3. Azimuthal Flux and Fluence Distributions at Reactor Vesse! Inside Surface 1.1 i 1ifrq 0.9 - t ,
- r ,
0.8 - 5 - en E
.L
? 0.7 -
i T:s . _ E 0.6 - Q ri t,
; 0.5 -
Iw3 0.4 -
~3 Elt
~
i SE : I t i , n 0.3
. 0 10 20 30 40 50 Azwmthat Position (degnmos) 3 4
I
Fiaure 6-4. Relative Axial Variation of E > 1 MeV FluxlFluence 1.2 -- r j < i i t
! t -*
- t +-
l t } 1 l.
, 3 survemance capsule
! ! i 1.1 -
i -- i
/
r' - S. 2 I ;
%-+N i
1 1% , t o ; 1 ; A 1 ! E 1 r ! - I e i !k jli i !
.)k o 1 4 1
C i i-e . t 3 ! . 1 , m E i 1 ! O 3 i { a- t . i i e o.g ; ! ,
, i . . I . i j j Bottom Dos;. Comp. Middle Dos. Comp. Top Dosimeled Comp;;
To a: Axial Factor'= 1.06 Axial Factor d 1.07 ! Axial Factor ni 1.04 g
- i 8 I t g j y 4)
+i i
0.8 . - l *- e ! -,- - I . 1 i I i 1 . i N (1 - - - I J - 4 E I
'. i-mI g i. t i SE I b I '
I 0.7 t
- Og 0 20 40 60 80 100 120 140 100 Distance from Bottom of core (in.)
4
'I
i l
- Figure 6-5. Radial Dimensions Used in Modeling i Caosule and Pressure Vessel Reaions l
l NOTE: Distances are llam core conter, MATERIALS: E SS004 WATER PV STEEL
].
.].
0..j.1
=4 -
I i I ABCD EFG HI JK LM N POSITION DISTANCE (cm) POSITION DISTANCE (crn) POSITION DISTANCE (cm) } A 215.31 F 21727 K 21922 B 215.67 G 411.50 L 220.98 l C 216.32 H 217.95 M 221.54 0 210.59 1 218.22 N 243.45 E 217.04 J 218.87 6-15 BWs*fas?i % r
Page Intentionally left Blank B Wsisenia:ANY
- 7. DISCUSSION OF CAPSULE RESULTS L3. Pre-Irradiation Property Data The weld metal and base metals were selected for inclusion in the surveillance program in accordance with the criteria in effect at the time the program was designed for Waterford Unit 3. The applicable selection critarion was based on the unirradiated properties only. A review of the original unirradiated properties of the reactor vessel core beltline region materials indicated no significant deviation f r ..* expected properties. Based on the design end-of-service peak neutron fluence value at the 1/4T vessel wall location and the copper content of the base metals, it was predicted that the end-of-service Charpy uppee-shelf energy (USE) will not be below 50 ft-lb.
7.2. Irradiated Property Data 7.2.1. Tensile Properties Tables 7-1 and 7-2 compare irradiated properties from Capsule W-97 with the unirradiated tensile properties. At both room temperature and elevated temperature, the ultimate and yield strength changes in the base metal as a l 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. The changes observed in the base metal are such as to be considered tsithin acceptable limits. The changes, at both room temperature and 550f, in the properties of the base metal are equivalent to those observed for the weld l 1 metal, indicating a similar sensitivity of both the base metal and the weld metal to irradiation damage. In either case, the changes in tensile properties are l insignificant relative to the analysis of the reactor vessel materials at this time period in the reactor vessel service life. 7-1 SW!!=nihr
- ~ . . - , , - , , ..a- >-.,s n w.-.u.. .w., w- ..u.u- x- - . - + a t
The general behavior of the tensile properties as a function of neutron - irraciation is an increase in both ultimate and yield strength and a decrease in-j ductility as measured by both total elongation and reduction of area. The most-l significant observation from these data is that the weld metal exhibited slightly gieater sensitivity to neutron radiation than the base metal. j l.2.2. Imoact Procarties 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 compares the observed change: in irradiated Charpy impact properties with the predicted l changes. A comparison of the Charpy data curves are presented in Figures 7-1 I ! through 7-4. The 30 ft-lb transition temperature shift for the base metal in the longitudinal orientation is conservative compared to the value predicted using Regulatory. l Guide 1.99, Rev. 2" and when the margin is added the predicted value is very conservative. However, the 30 ft-lb transition temperature shift . in the l transverse orientation is not in good agreement with. the predicted using j Regulatory Guide 1.99, Rev. 2, and when the margin is added the preoicted value ! equals the measured value without any margin- for conservatism. It would be expected that these values for the longitudinal orientation would exhibit good agreement when it is considered that the data used to develop Regulatory Guide 1.99, Rev. 2, was taken from data obtained from longitudinal oriented specimens. The transition temperature measurements at 30 ft-lbs for the weld metal is in relatively good agreement with the predicteo' shift using Regulatory Guide 1.99, Revision 2 but the predicted value is not conservative. The predicted shift , being slightly under estimated indicates that the estimating technique based on the Regulatory Guide 1.99, Rev. 2, is not overly conservative for predicting the 30 ft-lb transition temperature shift. Since-the method requires that a margin l be added to the calculated value tc provide a conservative value, the final shift value using Regulatory Guide 1.99, Revision 2, is conservative, and future evaluations shouid be based on Position 2 when additional data are available , which will help to account for some of the over-conservatism in the application of Regulatory Guide 1.99, Position 1. 7-2 BW.HaWsLa
The data for the decrease in Charpy USE due to irradiation showed relatively good agreement with predicted values for the base metal. The weld metal decrease in Charpy USE was over predicted by 200 percent. However, the poor comparison of the measured weld metal data with the predicted value is to be expected in view of the lack of date. for low copper-content materials at medium fluence values that were used to develop the estimating curves. Results from other surveillance capsules also indicate that RT estimating NOT curves have greater inaccuracies than originally thought. These inaccuracies are a function of a number of parameters related to the basic data available at the time the estimating curves were established. These parameters may include inaccurate fluence values, inaccurate 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 errors that resulted from using the 30 ft-lb data base to predict the shift 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 NDT shift as given in Regulatory Guide 1.99, Revision 2, are considered-adequate for predicting the RT shift of those materials for which data are not NDT , available. These curves will be used to establish the pressure-temperature operational limitations for the irradiated portions of the reactor vessel until ' the time that improved prediction curves are developed and approved. The relatively poor agreement of the change in Charpy upper-shelf energy for the weld metal does support the conservatism of the prediction curves for low copper-content materials. However, for low copper-content base materials the predicted values are not conservative. Although the prediction curves are conservative for the weld metal in that they generally predict a larger decrease in upper-shelf l energy than is observed for a given fluence and copper content, the conservatism can unduly restrict the operational limitations. These data support the contention that the upper-shelf energy drop curves wiil have -to be revised as more reliable data become available; until that time the design curves used to predict the decrease in upper-shelf energy of the controlling materials are considered conservative. 7-3 BW#EEVSLv
7.3. Reactor Vessel Fracturef e 'c an An evaluation of the reactor vessel end-of-life fracture toughness was made and the results are presented in Table 7-4. The fracture toughness evaluation shows that the controlling base metal will have a T/4 wall location end-of-life RTwo7 of 69F based on Regulatory Guide 1.99, Revision 2, including a margin of 24F. The controlling weld metal will have a T/4 wall location end-of-life RTuo7 of 34F based on Regulatory Guide 1.09. Revision 2, including a margin of 52F. These predicted shifts may be excessive since data from the first surveillance capsule exhibited measured RTuoy values that are comparable to the Regulatory Guide mean values. It is estimated that the end-of-life RTuor shift for both the controlling base metal and weld metal will be significantly less than the value predicted using Regulatory Guide 1.99, Revision 2 because the 9 of future surveillance data will permit a reduction in the applied margin. This reduced shift will permit the calculation of less restrictive pressure-temperature operating limitations than if Regulatory-Guide 1.99, Revision 2, was used. An evaluation of the reactor vessel end of-life upper-shelf energy for each of the materials used in the reactor vessel fabrication was made and the results are presented in Table 7-5. This evaluation was made because the base metals used to fabricate the reactor vessel are ch'aracterized by upper-shelf energies measured only in the longitudinal orientation. Consequently, when adjusted for the transverse orientation are expected to be sensitive to neutron radiation damages'and exhibit values significantly lower than the longitudinal value. The method used to evaluate the radiation induced decrease in upper-shelf energy is the method defined in' Regulatory Guide 1.99, Revision 2, ' which is the same procedure used in Revision 1. The method of Regulatory Guide 'I.99, Revision 2, shows that the base metals used in the fabrication of the beltline region 'of the reactor vessel will have -an upper-shelf energy greater than 50 ft-lbs through the 32 EFPY design life based - on the T/4 wall' location. Regulatory Guide 1.99 method also predicts an upper-s:1 elf energy above 50 ft-lbs for th~e controlling base metal at the vessel inside wall. The weld metal upper-shelf energies unirradiated values are so high as to 7-4 SW#M&%v
i preclude any change of the values decreasing below 50 ft-lbs during the 32 EFPY ! design life. Based on the first surveillance capsule data, it is estimated that ; ! the controlling vessel base metal upper-shelf energy will remain above the i required 50 ft-lbs during the vessel design life. L4. Operati no limitations The current normal pressure-temperature operating limitations are designed for - l operation through ' 8 EFPY. Based on the fluence calculations performed for
! Capsule W-97 and the results of the Charpy impact test results, the current operating limitations may be extended to 10.5 EFPY. However, any changes must l; be verified by confirmatory calculations and, in addition, any changes in the l fuel cycle designs will require t. review and possible verification for extension j from the original 8 EFPY limit.
- 7.5. Pressurized Thermal Shock (PTS) Evaluation i
The pressurized thermal shock evaluation shown in Table 7-6 demonstrates that the Baterford Unit 3 reactor pressure vessel is well below the screening criterion limits and, therefore, need not take any additional corrective action as required l by the regulation. 7.6. Ne v.'_Is.f Fluence Analysis 2 These new an ases calculated an end-of-life fluence value of 3.69 x 10 n/cm ] (E > 1 MeV) at the reactor vessel inside surface peak location. The correspond-l ing value for the vessel wall T/4 location is calculated to be 1.97 x 10 n/cm2 l (E > 1 MeV). These values do.not represent a reduction compared to the values calculated based on the design basis fluence values. 1 a i m i i i l 7-5 i SWTanh
s
} Table 7-1. Comparison of Waterford Unit 3. Capsule W-97 Tension Test Results. ; Room Temo Test Elevated Temo Test
- l Unirr** Irrad Unirr** Irrad Base Metal -- M-1004-2. Transverse Fluence, 10 18 n/cm' (E > 1 MeV) 0 6.47 0 6.47 i
! Ultimate tensile strength, ksi 89.0- 92.6 87.0 90.0 0.2% yield-strength, ksi 68.1 70.4 64.5 63.5 Uniform elongation, % 11.0 11.7 9.9 10.2 Total elongation, % 27.3 26.2 22.3 23.0 Reduction of area, % 68.2 63.5 65.3 62.5 l Base Metal -- Heat-Affected Zone l
Fluence, 10 18 n/cm' (E > 1 MeV) 0 6.47 0 6.47 , Ultimate tensile strength, ksi 91.3 93.5 86.7 91.0 0.2% yield strength, ksi 68.2 69.5 60.2 69.6 Uniform elongation, % 6.8 7.0 6.6 6.4 Total elongation, % - 21.3 20.3 20.3 18.5 Reduction of area, % 69.4 68.9- 66.6 69.5 Weld Metal -- 88114/0145 Fluence, 10 I0 n/cm2 (E > 1 MeV) 0 6.47- 0 6.47 Ultimate tensile strength, ksi 92.2. 95.9 88.3 93.2 0.2% yield strength, ksi 81.0 84.5 - 72.2 74.0 Uniform elongation, % 9.6 7.3 9.2 7.9 Total elongatica, % 27.7 - *** 23.7 22.6 I Reduction of area, % 70.7 63.5 69.4 70.0
- Test temperature is 550F.
l ** Average of the 1 ewer yield strength data in Appendix 8. ! ***See footnote Table 5-2. l 7-6 GW##enen =r -
Table 7-2. Summary of Waterford Unit 3 Reactor Vessel Surveillance Caosule Tensile Test Results Ductility. % ; Strenath. ksi Cap. F Test Total Reduction - Material I.D. 10}gence,2 Temp, F Ultimate o%I *) Yield o%(a) Elon. o%I ") of Area o%I *) n/cm ; i Base metal -- 0.00 71 89.0' -- 68.1 -- 27.3 -- 68.2 -- Transverse 550 87.0 -- 64.5 -- 22.3 -- 65.3 -- (M-1004-2) W-97 6.47 70 92.6 +4 70.4 +3 26.2 -4 63.5 -7 550 90.0 +3 63.5 -2 23.0 +3 62.5 4 Base metal -- 0.00 71 91.3 -- 68.2 -- 21.3 -- 69.4 -- Heat-affected 550 86.7 -- 60.2 -- 20.3 -- 66.6 -- L zone _(M-1004-2) W-97 6.47 70 93.5 +2 69.5 +2 20.3 -5 68.9 -1 7 w 550 91,0 +5 69.6 +16 18.5 -9 69.5 +4 Weld metal -- 0.00 71 92.2 -- 81.0 -- 27.7 -- 70.7 -- (88114/0145) 550 88.3 -- 72.2 -- 23.7 -- 6S' . 4 -- W-97 6.47 70 95.9 +4 84.5 +4 --" -- 63.5 -10 550 93.2 +6 74.0 +2 22.6 -5 70.0 +1 ' i t
Change relative to unirradiated. ;
"See footnote' Table 5-2.
U3 kE tu , R$
. - . .= .-
[ i; i 6 1 1 I I j: e j Table 7-3. ObservedVs.PredictedChangesforCapsuleW-97 Irradiated i Charpy Impact Properties - 6.47 x 10' n/cm' (E > 1 MeV) : I l Difference Predicttd 4 Per R.G. 1.99/2 > Observed Withog) With Material' Unirrad. Irrad. Diff. Margin flargin(b)
- Increase in 30 ft-lb Trais. Temo.. F Base Material (M-1004-2)
Longitudinal 0 +3 +3 18 35 i Transverse - 29 +7 +36 18 35-t Haat-Affected Zone (M-1004-2) -106 -90 +16 13 35 , '?ca Weld Metal (88114/0145) - 80 -44 +36 39 78 j Decrease in Charry USE. ft-lb 4 Basa Material (M-1004-2) I Longitudinal 170 154 -16 N.A. 19 " i
' Transverse 141 123 -18 N.A. 16" Heat-Affected Zone (M-1004-2) 17.7 156 -14 N.A. 19 "
- - Weld Metal (881 W DI45) '156 143 -13 N.A. 26" F
- f I (*I Mean value per Regulatory Guide 1.99, .~4evision 2, May 1988.
i (b)Mean value per Regulatory Guide 1.99, Revision 2, May 1988; plus margin. (c) Bounding value per Regulatory Guide 1.99, Revision 2, May 1983 (includes margin). I
.N.A. - Not applicable.
Table 7-4. Evaluation of Reactor Ve55e. End-of-Life (32 EFPY) Fracture Touchne55 - Waterford Unit 3 Material
- Chemical _1311m4tdGL[1 genie
- End-of - L i fe RT,,,,,, T**
M4tgIla M igrj.311gn t'empesttion. Inside I/4 Wall f ab. Mat' t. R! actor vessel Heat w/o" Surface location Initial Inside T/4 Wall Code Beit1Ine Locat1on N+raber" 7ype Copper *1cheI n/ca' n/cm? R T,,,," Surfate Location M-1003-1 Intermed. Shell' 56488-1 SA533, Gr. 8 0.02 0.71 3.64E+19 1.95E+19 -30 +23 +17 M-1003-2 intermed. Shell 56512-1 5A533, Gr. B 0.02 0.67 3.64E+19 1.95E+19 -50 +3 -3 M 1003-3 Intermed. Shell 56484-1 SA533, Gr. B 0 02 0.70 3.64E+19 1.95E+19 .42 +11 +5 M-1004-1 Lower Shell 57326-1 SA533, Gr. B 0.0' O.62 3.69E*19 I.97E+19 -15 +39 +32 M-1C04-2 Lowr Shell 57286-1 5A533 Cr. 8 0.03 0.58 3.69E+19 1.97E*19 +72 +76 +49 tower Shell 57357-1 5A533, Gr. 8 0.03 0.57 3.f9E+19 I.97E*19 -10 +44 +37 M-1004-3 101-171 Mid. Ctrcum. Weld WW88114/ ASA Weid/ 0.05 0.16 3.64E*19 1.95E+I9 -70 +45 +34 Ft0145 Linde 0091 101-124-A.-6.-C Intenmed. Longtt. Weld CE Lots PTim Wald/ 0.02 0.96 3.64E+19" 1.95E+19* -60 +12 +a BOLA, H004 3ype 8018 y 101-142-A.-B.-C Lower longit. Weld .WW83653/ ASA Weld / 0.03 0.20 3.69E+19'* 1.
- 9* -80 +14 +3 to IL3536 Linde 0091 "Per Regulatory Guide 1.99, Revision 2. May 1988.
"Per section 6 of this report using n-stron transps t calenlation methods. " Materials chemical compositions per response to Generic Letter 92:-01. ** Fluence valve for longitudinal weld wit' maximum n value. "Per res'.nmse to Generic Letter 72-01.
E
i! !iItL lili;j ;i[I[jl , ;ljf F{ j [:! L TlL tf ,h + :tt;
. o l n t l o
- a it Y P s Wa 2 2 2 2 2 2 2 3 3 3 2 3 2 3 F b4/Loc 3 3 3 3
- EI > > > 2 > > > > >
- T dt
- 3 t ef a0e t t t i m 5 d [a i r 2 2 2 2 2 2 2 3 3 3 3 3 3 3 2 2 3
~
t s > > > > > > > :> > i s n }g n E I U d E S" U 2 l l a ti 1 g
*0 "
r W 4 8 0 0 0 2 0 - o L- / i 1 s 9 8 8 2 5 5 f O E 9 9 4 e 8 8 7 1 > >
- r 1
/L T
e d e G x t t t u - a a R s "0 *0 l W m er diei lr 9 1 6 7 8 8 5 f its P sg 7 8 7 1 5 7 7 1 1 5 5
> l d
_ E I n$ e y w g 1 _ 1 9 . r l a E b 0 -- e i 1
*4 *7 *4 *6 .
0 n t i U t "0 "6 "4 A. A. e - . E n f 9 9 9 0 9 9 5 d I 1 1 4 W n f i L l e l n *3 "9 d 9 9 9 9 9 9 9 n a o 'a l h i 1 1 1 1 1 1 1 1 1 a S W ta/c + + + + * + + + +
. - " 3 cn E E E E E E E E 5
E 7 er r de ec /o 5 5 5 9 9 9 9 -9 9 9 7 7 7 5 9 9 i e p t n T L w ae 1 1 l 1 1 1 1 1 1 U p 'sur l "9 "9 i d _. e - e 9 9 9 9 9 9 9 w Od e c 's EL 1 1 1 1 1 1 1 1 1
) + + * + + + + + +
Y e Lis ac f/ t E E E E E E E E m P n rn 4 4 6 6 6 6 6 6 6 4 9 9 9 4 4 6 9 5 a F I u s E S 3 3 3 3 3 3 3 3 3 e h
- 2 3 .
l e 1
- 0 2 8 2 6 0 t
h n h 7 6 7 6 5 6 2 ( ll o i c 1
%. t i
aai R 0 0 0 0 0 0 0 0 0 w e iriic t "o .
- f s d i ems / r d e
_. L t eow ah p t 2 2 2 3 3 3 5 2 3 o t a p h . _ - MC m n 0 0 0 0 0 0 0 0 0 t 1 c f o g e 0 i C 0 0 0 0 0 0 0 0 0 m - r o- E 2 b n 9 a - d B 8 B 8 8 8 o f i r n . . . . . . 1 t e l - E r r r r r r /9 /8 /9 1 a t a e l t t o G G G G G G d0 d1 d0 u e e ~ l 0 l 0 l 0 e r . . . . . .le e8 e l c L m - s T 3 3 3 3 3 3 W e W W e a et . d _ 3 3 3 3 3 3 d e d c e l s 5 5 5 5 5 5 A n A p An Si M y Si u a ew e A A A A A A r -s. a t l t 5 S S S S S AL MT AL a V o _ A p C v d ec r / D O / . s s y n o sH 3 8 n o w g a _ t t ae "r 1 1 1 8 2 4 6 6 9 1 1 1 4 1 5 t 1 4 o . 63 56 8 a t 9 r t o r i e l c eb 8 1 8 2 8 5 31 lA 35 1 t e n s a n e ev l a Hm R e n e M u 4 6 6 6 7 5 4 3 2 3 8O 5 5 5 5 5 5 7 7 WF C8 WF L E0 t 83 L P r a eo m r p h n f l r u i t
. u e i s t h
e s _ f t d 2 e r w . s e1 o s l n - h i e n r dl 0 r- t - r n W o g e e r n cl o iei d d i n p e 2 p o tst l t
. l e i s
- i. s w9 p o u f i 2 sa l l l e i W v v n l r t J. e c l l l W g n e o a e l n a 1V o e e e 6 o R t i n t a e
_ u lr l h h h S S S l l l o L t i r t i t
. o i d e n v i i l Loe tt n . . . e e e t l l l w .
g n 9 p s u L d g a lci d d d h h h r d o 9 e o t u r p i c e v wwn e t aal a S S S l l 1 m g i i u E M et C m s o n r g l Rl e r r r r r r e e e e e e . r e r e e i c o el n a _ il t t t w w w d t w d i h t l n o v e l
. _ n nI n o Lo toL iM n o u a r G n 5 _ I I I t G f c of f o o t o o
_ 7 C C y r 6 e e o h e u e t
'6 sa d
e e . 8 8 t n c l s 1 a o a n 0 . b l . l i s v o b 't . at p n e ct a e s o l a e A A g t a Md g 1 2 3 1 2 3 4 2 e e i c e a T .C 3 3 3 4 4 4 7 1 2 4 R S r en er d m e i b a 0 0 0 0 0 0 0 0 0 0 0 0 1 1 1
- r r t v r s t F 1 1 1 1 1 1 1 i 1 e e a i e a s
- - - - - - 0 J 0 P P M F P B E M h M M M M 1 I 1 " " " ** " "
- y$ aE .
5 , _, si
.It j!l 4 }' f j ifj ! .j ;5 .! i
I I V 7 Table 7-6. Evaluation of Reactor Ve5$el End-of-Life Pre 55urized Thermal Shock Criterion - Waterford Unit 3 i Material Estloated ' , Cheetcal inside PTS Evtiga11pe. T" Maltda; thscrictlpn Composition. Surface Inside Reactor Vessel w/o" Heat EOL Fluence initial Surface Screening I weltline tocrtion Mumber" Type Copper Nickel n/cs' R T. ." Rim Criteria ! Intermed. Shell 56488-1 SA533. Gr. 8 0.02 0.71 3.64E+19 30 +31 213 l Intermed. Shell 56512-1 SA533. Gr. 8 0.02 0.67 3.64E*19 -50 *11 270 ! . Intermed. Shell 56884-1 54533.'Gr. 8 0.02 0.70 3.64E+19 -42 +19 270 Lover Shell 57326-1 SA533. Gr. B 0.03 0.62 3.69E+19 -15 +46 270 f 4 tower Shell 57286-1 SA533. Cr. B 0.03 0.58 3.69E+15 +22 +83 270 ! tower Shell 57359-1 5A533. Gr. B 0.03 0.62 3.69E+19 -10 +11 270 i Mid. Ctrtwo. Weld WW88114/ ASA Weld / 0.05 0.16 3.64E+19 -70 +45 30G t , Ft0145 Linde Co11 i Intermed. Longit. Weld CE tots MM4 Weld / 0.02 0.96 3.64E+19" 44 +37 210 [' 80LA/H004 Type 8018 N' tower te.nett. Weld WW83653/ ASA Weld / 0.03 0.20 .69E+I9" -E0 +23 270 0 TL3536 Linde 0091 l l' "Per 10CFR50. Section 50.61. Fracture Toughness Regut.eeents for Protection Agatast Pressorial Thermal Shock Ewnts."
"Per Section 6 of this. report using neutron transport calculation methods.
i { " Materials the=1 cal compositions per response to Generic Letter 92-01. ' i
" Fluence value for longitudinal weld with manteve value.
"I"r ' esponse r to Generte Letter 92-01.
G . k i [
- _ ~ __ _ - _ _ . _ _ _ . _
figure 7-1. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Plate Material : Lonoitudinal Orientation. Heat No. M-1004-2 , 100 ; , , i i , j ( 75 - - 5
$ $0 Unirradiated _
t \ E 25 - = Fluente: 6.47 x 10" n!cm2 -
' ' ' ' I 0
0.10 , g- , , , i
' 0 08 - ^ -
3 _ Unittadiated C
'e 0.06
- d -
Fluente: 6.47 x 10 n/cm?
-$ 0.04 -
; ,4p 3
$ 0.02 - -
0 220 -
; ; j g g ,
200 - A = 16ft lbs _ 180 o 160 1
- Unitradiated g 140 - -
\
@ \
j 120 - - 100 5 I; - - 80 1 E 60 3F 40
- ; 3F 20 MATERIAL SA 533.CIBitt) -
HEAT NO, M 1004 2 , I f I I I I 0 100 0- 100 -200 300 400 500- 600 Test Temperature, F l l L l 7-10 BWltKWfWna
figurc 7-2. Comparison of Unirradiated and Irradiated Charpy impact Data Curves for Plate Material Transverse Orientation. Iteat No. M-10Q4-2
'00 i ; i i ; ; 75 -
5 4 (
'o _
Unitradiated g 25 - Fluence: 6 47 x 10" n/cm' i i t 1 I ' ~ 0 0 10 g g g ; j j E
~%
e 0 OB .- Unittadiated - 5 j 0 06
+- Fluence: 6.47 x 10" n/cm2
~
-{0.04 l-27F
$ 0 02 -
/
3 0
/ I I I I '
I 220 ; ; ; i j i 200 180 l - I g 160 I e - l e 14n - Unitradiated 4 E a = 18tt Ibs 1 e \ I - j 120 - l @ 100 5
- Fluence: 6.47 x 10" n/cm2 -
1 5 80
-E -
60 32F 40
= --- 3 6 F 20
- MATERIAL SA 533.Cid1(T)
HE AT NO. M 1004 2 l l 1 I I I l
-100 0 100 200 300 400 500 600 Test Temperature, F 7-13 BWuna?%%r
Figure 7 3. Comparison of Unirradiated and Irradiated Charpy impact Data Curves for Base Metal. Heat-Affected-Zone. Heat No. M 1004-2 100 i 3 i i i
'9
'. 75 - -
; = Unitradiated 5 50 -
3 fluence: 6.47 x 10 n/cmr 3 25 0 O 10 g g ; ; ; ; diated a 0 06 - 2
@ Fluence: 6.47 x 10'8 n/cm j 0 04 -
18F a
. $ 0.02 - -
! E . ! 0
- 220 i i j i i i 200 i
l 180 f A = 14ft lbs
~
.[
~ '
Unitra0iated t . ~ '
/
i g 140 - - i ! 5 120 - -' - t 2-b 100 - - i e l E 80
- - Fluence: 6.47 x 10 n/cm' i a e
i 60
= 18F 40
- 16F
- - 20
- MATERIAL SA 533.CIB1(HA2)-
- . ' HEAT NO. M 1004 2 i l i I I I I O
L 200 100' O 100 200- 300 400 - 500 i Test Temperature, F ! 7-14 L S W # M SIA b u
- _ _ _ _ _ _ _ _ - - . . . - . ~ . . - , - . - - - _ . , . - - - ~ . . _ . -w.s _ .--.....m.,.-.,,,,-.....--.-.n,,_,,--,v-. ._-_,,__4 .<,._,.y.,--+., .--,--.,,..,,.,,,~,,,,,-,,,r,,
4 i , Figure 7-4. Comparison of Unirradiated and Irradiat0d Charpy . j 1mpact Data Curves for Weld Metal 88114/0145 I i 100 ; i i i i 1 . 75 -
~
- 2 -
Unitradiated ! $0 - Tluence: 6.47 x 10" n/cm? [
~
25 h* i , I I I I l I i, O 10 j j i i i i s ~, _ l g 0 08 Unittadiated . E ~ j 0.06 = Fluence: 0.47 x 10" n/cm2
- j. jg 0 04 -
{c 31F 0
$ 0 02 -
-E I I I I I I O
220 ; i i g i ; i 200 180 - - A = 13ft lbs - 160 Unitradiated 6 I i 8 120 -
\ -
l l 2,
- 4 -
l L' 100 Fluence:6.47 x10 n/cm' ! B - u 80 60 32F
-f f 40 36F i
20 1 MATERIAL - W Ei.' 431.il HEAT NO. 88114/0145 , i i 1 I I I g 100 0 100 200 300 400- 500 200 Test Temperature, F 7-15 S WifAMT4 & v
i i .i 4 I I i Page Intentionally left Blank 15EliB&W NUCLEAR Id WSERVICE COMPANY _
I
- 8.
SUMMARY
Of RESULTS l The analysis of the reactor vessel material contained in the first surveillance capsule (Capsule W-97) removed for evaluation as pa t of the Waterford Generating Station Unit No. 3 Reactor Vessel Surveillance Program, led to the following conclusions:
- 1. The capsule received an average fast fluence of 6.47 x 10 n/cm' (E >
1.0 MeV) . The. predicted fast fluence for the reactor vessel T/4 location at the end of the fourth fuel cycle is 2.74 x 10 n/cm' (E > 1 MeV).
- 2. The fast fluence of 6.47 x 10 n/cm' (E > 1 MeV) increased the RTuor of ~
the capsule reactor vessel core region shell materials by a maximum of 40F.
- 3. Based on the calculated fast flux c+ the vessel wall, an 80% load i
factor and the planned fuel management, the projected fast fluence that the Waterford Generating Station Unit No. 3 reactor pressure vessel inside surface will receive in 40 calendar year's operation is 3.69 x 10 n/cm' (E > 1 M'9
- 4. The i v rease in the RTuoy for the transverse oriented shell plate material wu in poor agreement with that predicted by the currently used design curves of RT versus fluence (i.e., Regulatory Guide NDT 1.99, Revision 2).
- 5. The increase in the RTuoi for the weld metal was in good agreement with that predicted.
- 6. Neither the base metal nor the weld metal upper shelf energies at the i T/4 location, based on surveillance capsule results, are predicted to decrease below 50 ft-lbs prior to 32 EFPY.
- 7. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in the base metal and the weld metal RTuor proper-ties due to irradiation are conservative except for the basa metal transverse properties.
- 8. The current techniques (i.e... Regulatory Guide 1.99, Revision 2) used to predict the change in the base metal and the. weld metal-Charpy l
1
~
BWllMfL
upper-shelf properties due to irradiation are in good agreement with the base metal and conservative for the weld metal.
- 9. The analysis of the neutron dosimeters demonstrated that the analytical techniques used to predict the neutron flux and fluence were accurate.
I l i I 8-2 S W !!?v n ef a u,y
- 9. SURVEILLANCE CAPSLLE REMOVAL SCHEDULE Based on the post irradiation test results of Capsule W-97 and the recommended eithdrawal schedule of Table 1 of E185 the following schedule i recommended for the examination of the remaining capsules in the Waterford Generating Station, Unit No. 3 RVSP:
i Evaluation Schedule Capsule Location of Lead Removal Expected Capsule Identification Capsul es Factor'd Time Fluence (n/cm')'" . W-83 83* 1.26 15 EFPY 2.19 x 10"' W-104 104" 0.81 Spare"' (2.99 x 10) W-263 263* 1.26 26 EFPY 3.69 x 10 W-277 277* 1.26 Spare"' (4.65 x 10) W-284 284* 0.81 Spare"' (2.99 x 10)
Reference reactor vessel irradiation locations, Figure 3-1.
'"The factor by which the capsule fluence leads tha vessels maximum inner wall fluence.
'd Estimated fluence values based on current fuel-cycle designs.
" Spare capsule to be irradiated and available for an intermediate evaluation, if data needed, to support licensing requirements or provide data for license ?
renewal. Capsule withdrawal at 32 EFPY will have estimated fluence as defined in brackets (). 9 BWAUM!Mby
i i i ! 10. CERTIFICATION [ l : l l The specimens were tested, and the data obtained from Entergy Operations, Inc., Baterford Generating Station, Unit No. 3, reactor vessel surveillance Capsule W-97 were evaluated using accepted techniques and established standard methods and l procedures in accordance with the requirements of 10CFR50, Ap),endixes G and H. 4 I I ff _
, M. ll//6/72 A. L. Lows, Jr. , P.f(/ Date i Project Technical Manager i
i . This report has been reviewed for technical content and accuracy. I i O?9O bd M. J. dea n (Material Analysis) iiini.m Date
- M&SA Unit l > 0
,\ .Y I$i;h v ~~ u 1 9.1 '! L. Petrusha (Fluence Analysis) / Date
- Perforr-e Analysis Unit Verific.ation of independent review.
} qFSimpd h KE Kwy nlghu i X. E. Moore, Mant jer Date i M&SA Unit i This report is approved for release.
/YY T. L. Baldwin, P.E.-
l UfM/92-Date l Program Manager j i i ~ 10 1 SW.W A1H&nt
. _ . _ _ _._._ _ _ . .__1 . _ _ _ . _ . _ . _ _ _ . _ _
Page Intentionally l_ eft Blank s
, SWit?v'nni!:sar i
APPENDIX A Reactor Vessal Surveillance Program Background Data and Information 1 l l 1 1 l i A-1 as asw eicuuta
.I.3 WSERVICE COMPANY
- 1. Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with E185-73, are shown in Table A-1. The locationr, of these materials within the reactor vessel are shown in Figure: A-1 throisgh A 4,
- 2. Defini, tion ?f Beltline Re11on The beltline region of Waterford Unit 3 was defined in accordance with the definition given in ASTM ele,5 73.
- 3. Cac*,ule Identification lhe .apsules used in the Waterford Unit 3 surveillance program are identified below by identification, location, and original target fluence.'
Capsule Capsule Capsule Approximate Target Removal Identification Location
- Refueling Fluence, n/cm#
1 W-97 97* 7 6.0 x 10" 2 W-104 104* 19 1.6 x 10" 3 W-284 284* 30 2.5 x 10" 4 W-263 263* Standby --- 5 W-277 277* Standby --- 6 W-83 83* Standby ---
- 4. Specimens Per Surveillance Caosule The type and quantity of each material contained in each surveillance capsule is shown in Table A-2.
A-2 SW!stefahr
i 4
- r j Table A-1. Unirradiated Impact Properties and Residual Element Content !
i Data of Beltline Region Materials Used for Selection of l j Surveillance Program Materials - Waterford Unit No. 3 2###' i i[ Charpy Impact Cata,
- i. Lonettudinti ;
Fabricator viaterial 30 50 35 i 1 Material Ident., Belillne Drop wt it-Ib, ft-1b, MLE, USE, R T ,o, th*elstry. wt% I Code Heat No. Material Type Region Location T.,o,, F F F F ff-lb i __Cu *l P 5 ,i M-1003-1 56488-1 SA533, Gr. B .'atermed. Shell -30 -30 -10 -10 144 -30 0.02 0.71 0.004 0.010 I t M-1003-2 56512-1 5A533, Gr. B Intermed. Shell -50 -55 -12 -15 149 -50 0.02 0.67 0.006 0.007
- j. M-1003-3 56484-1 SA533, Gr. B intermed. Shell -50 -22 -2 -10 138 -42 0.02 (, . 70 0.007 0.009 ;
e M-1004-1 57326-1 SA'233, Gr. B Lower Shell -50 +10 +25 +20 163 -15 0.03 0.62 0.006 0.008 i I M-1004-2 57286 8 5A533, Gr. B tower Shell -20 +37 +62 +55 144 22 0.03 0.58 0.005 0.005 < r
- M-1004-3 57359-I SA533, Gr. B Lewer Shell -50 +12 +30 +25 145 -10 0/s3 0.62 0.007 0.007 t 6' l
{ 101-171 88114/0145 ASA Weld /Linde 0091 MIME Circus. -70 --- --- -- ---
-70 0.05 0.16 0.068 0.008 t
101-124-A.-B -C BOLA /H004 PWIA Weld / Type 8018 letermed. Lcagit. -50 --- -- --- ---
-60 0.02 0.% 0.010 0.016
- j IOI-142.A.-B.-C 83653/3536 ASA Weld /Llade 0091 Lower tongit. -80 --- --- --- ---
-80 0.03 0.20 0.007 0.00s i I l
, i I h I i n E-i i ) ! 1 i M i d l
~ _ _ - - - _ _ _ _ _ _ - _ - _ _ - _ .
i Table A-2. Tvoe and Quantity of Specimens Contained in Each Irradiation Capsule Assembly Base Metal Weld Metal Correl. Target (Heat No M-1004-2) (88114/0145)'" HAZ (Heat Material" Capsule F!uence" Impact _ No. M-1004-2) Total Specimens 2 location (n/cm ) L T Tensile impact Tensile Impact Tensile Impact Impact Tensile Vessel 97* 6.0 x 10 12 12 3 12 3 12 3 -- 48 9 Vessel 104* 1.6 x 10 -- 12 3 12 3 12 3 12 48 9 Vessel 284* 2.5 y 10 12 12 3 12 3 12 3 -- 48 9 Vessel 263* Standby -- 12 3 12 3 12 3 12 48 9 , Vessel 277* Standby 12 12 3 12 3 12 3 -- 48 9 k Vessel 83* Standby 12 12 3 12 3 12 3 -- 48 9 TOTALS 48 72 18 72 18 72 18 24 288 54
" Adjusted to nearest value attainable during scheduled refueling. " Reference material correlation monitors. 'd Weld wire / weld flux lot combination.
g L = Longitudinal 6 g T - Transverse EE 55 BR h
,5
- _ _ _ . . _ . _ . _ _ _ _ . _ _ _ _ _ _ . . _ . _ _ _ _ _ _ . _ _ _ _ . . _ _ _ . _ _ . ~ . _ _ . .
Figure A-1. l.ocation and Identification of Materials Used in the fabrication of Waterford Unit 3 Reactor Pressure Vessel EiffAITOR VESSEL BELTLINE MATERIAL $ MOT SHOWN INTERMEDIATE SHELL t WELD SE AM No. 101 124B l WELD dEAM No. 101 1240 PLATE No. M 1003 2 LOWER SHELL _ , WELD SE AM No. 101 142B Y ' ' N "" 1.J WELD SE AM No. 101 142C "- PLATE No. M 1004 2 g@ %g' 5/c \A-c\ya i E l. w I (f 42"ID f 30'ID OUTLET l INLET NOZZLE L NOZZLE 7 UPPER TOl' TERMEDIA INTERMEDIATE SHE L SHELL GIRTH SEAM _ =j / LONGITUDINAL WELL WELD No. SEAM No. 101 124 A INTERMEDIATE SHELL ~ INTERMEDIATE SHELL PLATE No. M 1003 3 f # PLATE No. M 10031
- j INTERMEDIATE TO !.OWER 6 SHELL GIRTH SEAM WELD No. 101 171 4 LOWER SHELL
- PLATE No. M 10041 LOWER SHELL PLATE -
No, M 1004 3 LOWT:R SHELL _
} gT
~' ~
LONGITUDINAL WELD SEAM No. 101 142A -
. k _ ,
f= s
'l!i!!!i!!!!!!!!!!!!!!!!!!!Ef A-5 SWifAINhv
_ . _ . _ _ . . _ . . _ . _. _ _ . _ _ _ _ _ _ . _ _ . . - _ _ . _ _ _ _ _ _ _ _ _ . . _ _ _ . _ . - - _ _ ~ . , Figure A 2. Location of Beltline Region Materials in Relationship to the Reactor Vessel Core i A J , X . N h.. 1
'\h
%f h~'o V CL' h tw N ji n i
.. .. 7 z
s N s 3 b \
-r . b S i N
o s s s i __ m o i 8
' s o.
k N N 28 N m" u N N Centerlinev b 11 U CL' j ' ' ~ C ofcore 06 em a h s i N >g
% l N Ba N I N <
*s
- s 4-- 172 s s s
\ \
N p N _ . . . , N N y y CL' ( ) l I
\
= Centerline of Weld A-6 RWit&5Hhr
Figure A-3. Location of Longitudinal Welds in Waterford Unit 3 Upper and Lower Shell Courses 0 101 124 C ,- 7- M 1003 2
/-
101-124 A 270- t 90 Middle Snell M 1003 3 101 -124 B M 10031 30* 180 0 101 -142C [ M 1004 3 101 -142 A 270 y ~ 90 Lower Shell M 1004 2 x 101 142B - M 10041 l 180 A-7 B W itn EV W e a__
l Fiaure A-4. LocA11on of Surveillance Caosule Irradiation Sites in Waterford Unit 3 l 180 l ( 0utlet Nozzle < ,i
- --- - ,l (7 \
I 's ! Vessel
/
0 l 104 \ s~~~A f r -~ f "~ % IA I \ Nozzle
/ ,
i / gCore Shroud
')
\ _
-Core Support Barrel
/
/ .
N '
' Vessel Vessel 0 l
970 ' I - y Reactor Vessei g 263 Vessel # 8
~
s / Vessel
/ - l,' 271*
83* I 3 2h ( / s ' 1 I /
\ -
%~~s ~J , ./
3
//
\ /
\ /
\i /
i s-------) 0 0 A-8 SW##Ne%, c
APPENDIX B Pre-lrradiation Tensile Data B-1 53WS?n"v5?Sii$$Avr
Table B-1. Tensile Properties of Unirradiated Shell Plate _ Material . Ht aL1h_ti 1004-2. lonaitudinal Test Reduction Elongation, Specimen Temp, Strenath ksi___ fracture Fracture, ksi of Area, Total /Unif. No. F Yield
- Ultimate load, lb Strenath Stress % % __
IJ2 71 68.6/66.7 88.5 2640 53.9 189 71.4 29/11.3 IJ1 71 70.4/67.4 88.4 2700 55.1 180 69.4 27/11.3 IK2 71 70.0/68.6 90.1 2700 55.1 193 71.4 30/11.7 IJA 250 63.7/63.1 82.5 2640 53.9 176 69.4 24/ 9.2 IK3 250 66.1/64.9 84.1 2640 53.9 176 69.4 24/ 9.3 IJL 250 63.7/63.1 83.3 2700 55.1 180 69.4 26/ 9.3 IJ6 550 63.7/---- 85.5 2700 55.1 193 71.4 23/ 9.3 IJC 550 63.1/---- 85.9 2700 55.1 208 73.4 26/ 9.8 IJ3 550 62.5/ -- 85.6 2760 56.3 173 67.3 25/10.2
- Lower and upper yield strengths.
Table B-2. Tensile Properties of Unirradiated Shell Plate Material. Heat No M-1004-2. Transverse Test Reduction Elongation, Specimen Temp. Strenath ksi fracture Fracture, ksi of Area, Total /Unif. No, f Yield
- Vltimate load, lb Strenoth Strq11 % %
2KC 71 69.2/68.6 89.7 2880 58.8 192 69.4 27/10.8 2KT 71 68.4/67.2 88.4 2820 58.8 188 70.0 29/11.3 2KB 71 69.8/68.6 89.0 2940 60.0 196 65.3 26/10.8 2KD 250 65.5/64.3 83.7 2700 55.1 180 69.4 23/ 9.7 2JE 250 64.6/64.6 83.9 2820 57.6 188 r9.4 21/ 9.3 2L2 250 64.9/64.3 82.3 2940 60.0 163 63.3 23/ 9.3 2J7 550 64.9/---- 37.2 2880 58.8 169 65.3 23/10.2 2KP 550 63.7j---- 86.9 3000 61.2 188 67.3 22/ 9.8 2KU 550 64.9/---- 87.0 2880 58.8 160 63.3 22/ 9.8
- Lower and upper yield strengths.
B-2 IBWitnMaPa m
Table B-3. Tensile Properties of Unirradiated Shell Plate HAZ Material. Heat No, M 1004-2. Trantverse Test Reduction Elongation, Specimen Temp, Strenath, ksi Fracture Frnture, ksi of Area, Total /Unif. No. F Yicid* Ultimate Load. lb Sirenath Stress % % 4KT 71 71.0/68.6 91.0 2820 57.6 188 69.4 22/ 7.3 4JJ 71 68.6/68.0 90.9 2820 57.6 183 69.4 21/ 7.0 4K4 71 69 8/6L'.0 91.9 2820 57.6 188 69.4 21/ 6.2 4KP 250 64 3/63.7 84.3 2640 53.9 176 69.4 21/ 5.4 4J5 250 63.1/63.1 84.0 2640 53.9 165 67.3 19/ 5.8 4KE 250 63.7/63.7 84.8 2640 53.9 176 69.4 21/ 5.4 4JT 550 60.6/60.0 86.3 2820 57.6 166 65.3 21/ 6.7 4JE 550 63.1/61.8 86.9 2820 57.6 176 67.3 20/ 6.8 4J4 550 60.0/58.7 86.9 2820 57.6 176 67.3 20/ 6.3
- Lower and upper yield strengths.
Table B-4. Tensile Properties of Unirradiated Weld Metal 88114/0145 Test Reduction Elongation, Specimen Temp, Strenath. ksi Fracture fracture ksi of Area, Total /Unif. No. F Yield
- Ultimate LQad. lb Strenath Stress % %
3KE 71 85.7/M.0 92.9 2760 56.3 184 69.4 27/ 9.3 3J3 71 84.5/80.2 91.6 2760 56.3 197 71.4 27/ 9.3 3K1 71 84.5/80.8 92.0 2700 55.1 193 71.4 29/10.2 3L2 250 79.6/73.9 86.9 2700 55.1 180 69.4 22/ 7.7 3J5 250 80.2/75.9 87.2 2760 56.3 197 71.4 21/ 7.8 , 3JC 250 79.6/73.5 85.4 2530 52.6 184 71.4 23/ 7.3 3K4 550 72.2/---- 88.8 2580 52.6 172 69.4 24/ 9.5 3KA 550 72.2/---- 88.2 2640 53.9 189 71.4 24/ 9.3 3KU 550 72.2/---- 87.9 2820 57.6 176 67.3 23/ 8.8
- Lower and upper yield strengths.
B-3 B W!!nani % v
Page Intentionally left Blank c I B W !!K i?Sh's% m
APPENDIX C Pre-Irradiation Charpy impact Data C-1 SWi!?vnViisiar
Table C-1. Charpy Impact Data From Unirradiated Base Material, Lonoitudinal Orientation. Heat No. M-1004-2 Absorbed lateral Shear Specimen Test Temp, Energy, Expagston, Fracture, ID F ft-lb 10 in. % 13T -80 7.0 3 0 156 -40 12.0 10 10 152 -40 11.5 10 10 15J 0 14.5 14 15 123 0 48.5 41 20 112 40 86.0 65 40 llP 40 105.0 78 65 153 80 107.0 72 75 11A 80 130.0 90 80 12D 120 131.0 89 85 14P 120 147.0 92 90 147 140 158.0 91 100 llM 160 169.5 94 100 137 160 177.5 90 100 127 210 168.5 95 100 126 210 175.0 90 100 Table 0-2. Charp, impact Data From Unirradiated Base Material, Transverse Orientation. Heat No M 1004-2 Absorbed lateral Shear Specimen Test Temp, Energy, Expagsion, Fracture, ID F ft-lb 10 in. % 264 -80 9.0 7 0 21L -60 10.0 6 0 22E -40 20.5 18 10 , 238 -40 28.5 23 10 25P 0 44.0 37 20 26B 0 65.5 50 25 21K 40 68.5 55 40 23L 40 73.0 57 60 26A 80 118.0 82 75 23A 80 130.0 77 75 212 120 123.5 81 90 255 120 141.0 90 100 21P 160 136.0 88 100 254 160 138.5 92 100 25T 210 143.5 88 100 23C 210 145.0 90 100 C? BW!!na?5%r
Table C-3. Charpy impact Data from Unirradiated Rase Metal, Heat-Affected Zone. Heat No. M-1004 , , Absorbed lateral Shear Specimen Test Temp, Energy, Fracture, Expags' ori, 10 F ft-lb 10 in. % 417 -150 6.5 9 0 44C -135 10.5 8 0 451 -120 21.5 17 10 4" 42J - 80 44.5 33 30 1 67 - 80 76.5 50 45 45r - 40 113.5 73 70 443 - 40 116.5 77 ts 466 0 118.0 77 75
.. Ag1 q }38,Q gg gQ d 126.0 84 75 40 162.0 88 100 9
80 163.5 91 100 80 177.0 84 100
% s.' 120 152.5 90 100 4 '
W 120 183.5 89 100 160 164.5 87 106 s 160 183.5 89 100 Ishle C-4. Charov Imoact Data from Unirradiated Weld Metal. 88114/0145 Absorbed !.ateral Shear Specimen Test Temp, Energy, Fracture, Expagsfon, ID F ft-lb 10 in. % ! 354 -180 3.5 4 0 316 -lM 5.5 3 0 36A 8.0 6 10 36D 13.5 12 20 310 - 80 45.5 36 30 32A - 40 83.5 61 50 313 40 96.5 69 75 357 0 122.5 80 80 I 341 0 130.5 95 85 Slh 40 142.0 95 90 344 40 149.0 97 100 33T 80 146.0 94 100 32M 80 158.0 96 100 37L 120 153.5 96 100 k 3EG 120 162.5 97 106 3C 160 148.5 94 100
) 36J 160 171.0 94 100 C-3 B Ws95 NEIAEnv
"' U
Figure C . Charpy Impact Data From Unirradiated Base Metal (Plate). lonaitudinal Orientation. Heat No. M-1004-2 00 , ; i: , ; , , :. 7s - 3 E 50 - m a j 25 - 4 0 ; I l- 1 I I l 0.10 j g , ;, ; g ; e e e e e' O.08 -
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. . - - - . - . . . -. -. - .. .- ._ - -. - . . . - ~ _ . - . . -
i 1 Figure C-2.- Charpy Impact Data From Unirradiated Base Metal i (Plate). Transverse Orientation. Heat No. M-1004-2 100 ; ,0- : g: ; ; ; o
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j Figure C-3. Charpy Impact Data From Unirradiated Heat-Affected-Zone Base Metal. Heat No, M-1004-2 100 3 i ;; ; ; ;
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- 200 100 0 100 200 300 400 Sus nest Temperature. F
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Fiaure C-4. Charoy Imoact Data From Unirradiated Weld Metal. 88114/0145
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Page Intentionally left Blank i l { i i l i I l l f l l h B&W NUCLEAR I2 WSLTVirE COMPANY
/
- APPENDIX D Fluence Analysis Methodology D-1 SW," feen %v
1 i i
- l I
j 1. Analvtical Method q b A semi-empirical method is used to calculate the capsule and vessel flux. The l j method employs explicit modeling of the reactor vessel and internals and uses an j . average core power distribution in the discrete ordinates transport code D0TIV, l j version 4.3. DOTlv calculates the energy and space dependent neutron flux for
~
I the specific reactor under consideratior. This semi-empirical method is conven-iently outlined in Figures D-1 (capsule flux) and D-2 (vessel flux). { l The two-dimensional transport code 00TIV was used to calculate the enargy- and l space-dependent neutron flux at all points of interest in the reactor system. ! DOTIV uses the discrete ordinates method of solution of the Boltzmann transport: equation and has multi-group .<nd asymmetric scattering capability. The reference l calculational model is an R-O geometric representation of a plan viev through th'e ! reactor core midplane which includes the core, core liner. coolant, core barrel, t
- thermal shield, pressure vessel, and concrete. The material and geometry model, f represented in Figure D-3, uses one-eighth core symmetry. In order to include l reasonable geometric detail within the computer memory liinitations, the code l parameters are specified as P3 order of scattering, Se quadrature, and 47 energy
- groups. The P 3 order of scattering adequately describes the' predominately forward scattering of neutrons observed in the deep penetration of steel and l water media, as demonstrated by the close agreement between measured and calculated dosimeter activities. The S, symmetric quadrature has generally j produced accurate results in discrete ordinates solutions for similar problems,
( and is used routinely in the B&W R-O D0T analyses. l Flux generation in the core was represented by a fixed distributed source which the code derived based on a combined 23s0 and 2asPu fission spectrum, the input l relative power distribution, and a normalization factor to adjust flux level to ! the desired power density, i l Geometrical Confiourati u l [ For modeling- purposes, the actual geometrical configuration was divided into
- three parts,-as shown in Figure D-3. The first part, Model "A," was used to i generate the energy-dependent angular flux at'the inner boundary of Model "B,"
which began at the inner surface of the core barrel. Model A included a detailed ! D-2 l- BUl!!?AYcWr
i.
)
j representation of the core baffle (or liner) in R-O geometry that has been I checked for both metal thickness and total metal volume to ensure that the DOT l approximation to the actual geometry was as accurate as possible for these two } very important parameters. The second, Model B, contained an explicit represen-j tation of both surveillance capsules and associated components for the applicable j time periods. The B&W Owners firoup's Flux Perturbation Experiment verified ! that the surveillance capsule must be explicitly included in the DOT models used l for capsule and vessel flux cticulations in order to obtain the desired accuracy. I Detailed explicit modeling of the capsule, capsule holder tube, and internal
- components were therefore incorporced into the DOT calculational models. The third, Model "C," was similar to Model B except that no capsule was included.
{ Model C was used in determining the vessel flux in quadrants that did not contain i a surveillance capsule; typically these quadrants contain the azimuthal flux peak E on the inside surface of the reactor vessel. ! An overlap region of approximately 52.95 cm was specified between Model A and j Models B or C. The width of this overlap region, which was fixed by the l placement of the Model A vacuum boundary and the Model B boundary source, was
- determined by an iterative process that resulted in close agreement between the
- overlap region flux as predicted by Models A and 8 or C. The outer boundary was j placed sufficiently far into the concrete shield (cavity wall) that the use of I
a " vacuum" boundary condition did not cause a perturbation in the flux at the points of interest. Macroscooic Cross Sections
- Macroscopic cross sections, required for transport analyses, were.obtained with l- the mixing code GIP. Nominal compositions were used for the structural metals.
Coolant compositions were determined using the average boron concentration over [. a fuel cycle and the bulk temperature of the region. The core region was a l l homogeneous mixture of fuel, fuel cladding, structure,. and coolant. l The cross-section library presently used is the (47-neutron group and 20-gamma group) BUGl.E coupled set. The dosimeter reaction cross sections'are based on the !. ENDF/85 library, and are listed in - Table E-3. The measured and - calculated dosimeters activities are compared in Table D-1. 1 D-3 SW##4nhv
- - . - ._ - - - _ . - ~_. . .
Distributed Source The neutron population in the core during full power operation is a function _of neutron energy, space, and time _ The time dependence was accounted for in the analysis by calculating the time-weighted average neutron source, i.e. the neutron source corresponding to the time-weighted average power distribution. The effects of the other two independent variables, energy and space,- were - 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 was assumed to be invariant and independant 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. Sv3,g= l Pf) x \RPD,jX,] magnitude spatial where: Sv%
= Energy-and space-dependent neutron source, n/cc-sec,.
v/X - Fission neutron production rate, n/w-sec, Po - Average power density in core, w/cc, RPD, - Relative power density at interval (1,j), unitless, X, - Fission spectrum, fraction of fission neutrons having energy in group "g," i - Radial coordinate index, j - Azimuthal ;oordinate index, . 1 g -- Energy group'index. The spatial dependence of the flux is directly related to the RPD. Even though j the entire (eighth-core symmetric) RPD was modeleo in the analysis, only the peripheral fuel assemblies contributed significantly to the ex-core _ flux._ The axial average RPD distribution is calculated on a quarter-core symmetric' basis for the entire capsule irradiation period. The time-weighted average RPD D-4 BWitk**ia%
distribution is used to generate the normalized space ano energy dependency of the neutron source. Calculations for the energy and space dependent, time-averaged flux were performed for the midpoint of each DOI interval throughout the model. Since the reference model calculation produced fluxes in the R-e plane that averaged over the core height, an axial correction factor was required to adjust tnese fluxos to the capsule elevation. This factor was calculated to be 1.08. A pin-by-pin RPD was provided by the customer, and was subsequently used to produce the source for use in the DOTIV code. 1.1. Caosule Flux and Fluence Calculation As discussed above, the DOTIV code was _ used to explicitly model the capsule assemblies and to calculate the neutron flux as a function of energy within the capsules. 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 D,, in ( Ci/gm) is: D= j ( 3 . 7 x10') A, E {e o (E) n $ (B) { jF - j(1-e '*'# ) e '*' *M 1 where: N = Avogadro's number, i An - Atomic weight of target material n, f i - Either weight fraction of target isotope in n-th material or the fission yield cf the desired. isotope, an (E) = Group-averaged cross sections for material n (listed in Table E-3)'
- d(E) = Group averaged fluxes calculated by DOTIV analysis, Fj = Fraction of full power during j-th _ time interval, tj Ar = Decay constant of the ith isewpe, T - Sum of total irracioiion time, i.e., residual . time in reactor, and the wait time between reactor shutdown and counting times, D-5 BWR&%"c b - . - . - _ . --- - --. - _ _ . _ - . .. . = - -
r,- Cumulative time from reactor startup to end of j-tb time period, t = Length of the j-th time period 3 Adjustments were made to the calculated dosimeter activities to correct for the effects listed below: 23W Photofission adjustments to U dosimeter activities Axial correction factor to adjust for axial power distribution After making these adjustments the calculated dosimeter activities were used with the corresponding measured activities to obtain the measured to calculated i activity ratios or flux normalizaticn factors: D, (measured) C; - , D,(calculated) These normalization factors wm: evaluated, averaged, and then used to adjust the calculated test specimen flux and fluence for aach capsule to be consistent with the dosimeter measurements. The flu.c normalization factors are given in Table D-1. Note that the Co-60 dosimeters are typically not used in the determination of the final normalization factor to be applied to the calculated flux due to the fact that they do not respond in the regions of interest, E > 1.0 MeV and E > 0.1 HeV, and the thermal region is not accurately calculated-in the DOT analysis.
- 2. Vessel Fluence Extracolatiqn For past core cycles, fluence values in the pressure vessel were calculated as described above. Extrapolation to future cycles was required to predict the useful vessel life. Two time periods were considered in the extrapolation: 1) operation to date for which vessel fluence has been calculated. 2) future fuel cycles which no analyses exist, j For the Waterford Unit 3 analysis, time period I was through cycle 4, and time ,
period 2 covered cycles from the end of cycle 4 through 32 EFPY. The flux and fluence for time perit. 4 2 was estimated by assuming that the flux at the inside surface of the pressure sessel (PVIS) for future cycles was the same as that calculated for cycles 1 through 4. This was a reasonable assumption because the D-6 BW!!?venb%v
i first four cycles were similar in that fresh fuel was loaded in the peripheral _ j locations in each of the cycles. l It was found in the Waterford Unit 3 a1alysis and is shown in figu"e 6-3 that the l peak fluence at the PVIS occurred at approximately 1 degree off the major axis for cycles 1 to 4. For this reason, the flux used to extrapolate from E0C 4 to l 32 EFPY was the flux calculated at I degree. Future analyses will ascertain the I actual effects. i i i I-- i l i ] i i i i a I e . i i e i i. i i , -1 i i k i i j D-7 i 4 GWWvMfx"=%
.r w e. ,,w..,-w..- ., ..%. - - - . - . - .m.,w..-e-..m,..-.--,-.-.,.-e*,-...._ sew,em.ww.,,,,3w--w.er 3- o w-e . g e , ,-w., -.
- _ 1
! 1 l' Table D-1. Flu'x Normalizr. tion Factor for 97'C_gpglg i 4 -< j Measured Calculated Flux i Dosimeter Ac t i v i ty ,- Activity,*' Normalization
- l Repetion _.uC',a uCi/c Factor l iii ( n , p)Co 1835.0 1900.7 0.965
- **T i ( n . p )Sc 365.9 342.6 1.067
- "Fe (n , p)Mn 1427.0 1442.0 0.990
Cu (n ,o)"Co 8.377 8.847 0.947 '
23sU(n , f) *C s 8.157'* 9.957 0.819 i Co ( n , y)"Co'*' 4.360E+5 3.409E+5 I.286 l
Co ( n , g )"Co* 4.860E+4 5.643E+4 0.862
- -Averaged
- 0;958a! t
Average of three dosimeter wires, except- for U powder capsules.
*'Each listed activity was determined as the average of thrae calculated activities.
" Average of, asured to calculated activity ratios for each dosimeter-type.
- The V dosimeters werE powder capsules, three shielded and three bare. Since the U-238 dosimeter response in the thermal range is negligible,- the shielding should-have virtually-no effect on the respanse of-the dosimeters. The three shielded U. samples showed good agtaement with the calculations. However. of the three b're U-238 dosimeters, one was unrecoverable and one jave unreasonabk esults. _ Therefore, the U values in this table are for four dosimeters, three shielded and one bare, averaged together.-
- Bare dosimeters. i
'"Cd-shielded dosimeters.
Average of.all dosimeters except_Co.
4 __ F I D-8 __ SWFAT"c h v l
. .. . . . - . - . - . . . . . - - .- ~. - -. . .
figure D-1. Rationale for the Calculhtion of Dorimeter Activities ano Neutron Flux in the Caosule ENDF/B4 Cross Sections Geometry & Quadraturs Power Distn-ENDF/B5 Doslmeter Reae. for Modst A DOT butions Since . tion Cross Sections _ Capsule Inser. l tion l If GIP If if If
~ it Cross Sections l-- DOT 4.3 Radial Power Model A l , Shape Appfled Gwmetry &
Quadrature Model B 1I If u DOT 4.3 Angular Flux Model B 4 At Barrel V Dosimeter Power History ActWes of Capsule (PRHIST Code) Fint
- Calcu:ated c ActMtles n Measured Dostmeter Axlal ActMiles Correction Factor Y Normalization Factor -
y ' Capsute Flux w M/C Ratio D-9 S Ws?seni h r
1 Figure D-2. Rationale for tne Calculation of Neutron Flux in the Reactor Vessel ENDF/B4 Cross Sections Geometry & Quadrature Power Distri. ENDF/BS Dosimeter Reae. for Model A DOT butions Since tJon Cross Sections Capsule Inser. tion 1I GIP_. Y y if U Cross Sections Radial Power DOT 4.3
- Model A Shape ApplIGd Geometry &
Quadrature Model B II y Jf DOT 4.3 Angular FicI Model C j< At Banel Normalization Factor from Capsule Fluence Maylls (from the Diagram on the Previous Page) Mal Correction Factor Y Time Averaged Vessel Flux
> Maximum Vessel Location , 4 (E,R,0)
D-10 T3EsY$5fEi$$$my
- .. . . _ = _ . -
)
l = e d M o % \5
?3 n
o i t l a 9 %*% C u B c t s l a o1 e C O fO4*. o5 s3H d M o R 5 e k .a $ j 1*L~% c n e r e f e R ( t?" %g e = n a l o
=
d i M 91 e r o Q*g+ C r g o _ t c a e u R p
=
h g u
/ c A l
d e o I o r \
\
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\
i V e \ n a l P 3 W 0 0 e M e = i F a W k e
]
Rg aRg
? "ln~ k1s
- I
- 4! :! ;
4 i APPENDIX E Capsule Dosimetry Data E-1 B W !! K i?fi % y
._ ~ _ _ _ . _ _ . _ . _ __ _ _ __ .. _ _ _ . _ ._.. _ _ . _- _ _ _ _ _ . _ . . _ _ _ Tabic E-1 lists the characteristics of-the neutron' dosimeters. Tables E-2 and , E-3 show the measured activity per gram of target material (i.e . per gram of uranium, nickel, etc.) for each capsule's dosimeters. Activation cross sections for the various materials were flux-weighted with the 235 U fission spectrum shown in Table E-4. Table E-1. Detector Composition and Shieldhg Detector Ptitfr_tal Shieldina Reaction Ni Wire Cd N 1 (n . p)Co Co Wire Bare Co(n, y)* Co Co Wire Cd Co(n, y)* Co Fe Wire Bare 5*Fe(n p)Mn Cu Wire Bare Cu(n.a)* Co 238 U0 3 Cd U(n , f)'87Cs U30s Bare aaV(n,f)'8'Cs Ti Wire Cd Ti (r., p)Sc - E BWN,.w
- -. - ._.. - .. - ..- - - - ..~. ..- - - . .-. - . .
I-i l l Table E-2. Measured Specific Activities (Unadjusted) j for Dosimeters in 97* CaDsule j Dosimeter Activity, j- (uCi/am of Taraet) j Detector Material Dosimeter Reaction udder Center lower Ni Wire Ni ( n , p) 6'Co 1808 1768 -1929 Co Wire (b) **Co(n,y)" Co 4.357E+5 4.895E+5 3.887E+5 l Co Wire (sh) 6'Co(n,y)'Co 5.163E+4 4.442fd A.?76E+4 {. Fe Wire Fe ( n , p)"Mn 1462 1374 1445-l j Cu Wire Cu(n,a)* Co 8.519 7.907 8.705 i 238 U(n , f)7Cs 7.946 7.941 8.424 U Powder (sh) U Powder (b)- 23sU ( n , f)70s 16.00 -- 8.317 ! -s l Ti Wire **T i ( n , p )Sc 332.3 354.9 410.5 i . i f I i i I I i i i 4 i i i E-3 S Wi E t?cti k v
Table E-3. Dosimeter Activation Cre:s Sections, b/ atom'd Group Upper No. Energy-(eV) **Ti(n,p) ' 'U(n, f) Fe(n.p) "Ni(n,p) Cu(n,a) Co(n,y) 1 1.733+7 2.407-1 1.215+0 2.803-1. 3.215-1 3.641-2' 7.968-4 2 1.419+7 2.667-1 1.033+0 4.260-1 4.980-1 4.535-2 8.380-4 s 3 1.221+7 2.600-1 9.851-1 4.728-1 5.734-1 5.360-2 7.033-4 4 1.000+7 2.356-1 9.933-1 4.769-1 5.971-1 3.842-2 6.978-4 5 8.607+6 2.043-1 9.898-1 4.759-1 5.988-1 1.926-2 9.431-4 6 7.408+6 1.555-1 8.240-1 4.687-1 5.845-1 9.389-3 2.214-3 7 6.065+6 9.645-2 5.588-1 4.266-1 5.141-1 2.956-3 2.455-3 8 4.966+6 3.766-2 5.452-1 3.041-1 3.347-1 4.568-4. 2.871-3 9 3.679+5 5.573-3 5.292-1 1.998-1 2.424-1 3.600-5 3.269-3 10 3.012+6 4.747-4 5.282-1 1.371-1 1.674-1 5.844-6 3.523-3 11 2.723+6 6.816-6 5.365-1 8.061-2 1.232-1 1.692-6 3.772-3 , 12 2.466+6 1.100-6 5.398-1 5.715-2 9.340-2 6.645-7 3.938-3 13 2.365+6 3.770-7 5.404-1 5.134-2 8.278-2 4.712-7 4.006-3 14 2.346+6 3.427-7 5.410-1 4.564-2 7.227-2 3.305-7 4.090-3 15 2.23)+6 2.326-7 5.358-1 2.892-2 4.600-2 1.1B1-7 4.337-3 6 16 1.921+6 8.518-8 4.799-1 8.181-3 2.440-J 1.500-8 4.931-3 3 17 1.653+6 0.000-0 3.154-1 2.933-3 1.206-2 0.000-0 6.222-3 18 1.353+6 0.000-0 4.480 2 6.824-4 3.758-3 0.000-0 8.205-3 19 1.003+6 0.000-0 1.296-2 5.308-5 1.362-3 0.000-0 -7.473-3 20 8.209+5 0.000-0 3.820-3 4.367-6 1.156-3 0.000-0 6.519-3 21 1.427+5 0.000 0 1.553-3 6.842-7 9.891-4 0.000-0 6.905-3 22 6.081+5 0.000-0 6.233-4 =1.097-7 7.958-4 0.000-0 7.598-3 23 4.393+5 0.000-0 2.846-4 8.051 6.086-4 0.000-0 9.233-3 24 2.688+5 0.000-0 1.635-4 5.515-8 4.483-4 0.000-0 8.724-3 25 2.972+5 0.000-0 1.001-4 3.448-8 3.058-4 0.000-0 1.058-2 26 1.835+5 0.000-0 7.720-5 1.197 B 1.577-4 0.000-0 1.322-2 27 1.111+5 0.000-0 6.115-5 0.000-0 6.464-5 0.000-0 1.780-2 28 6.738+4 0.000-0 6.174-5 0.000-0 7.780-6 0.000-0 -3.155-2 29 4.087+4 0.000-0 6.984-5 0.000-0 0.000-0 0.000 3.211-2 30 3.183+4 0.000-0 7.894-5 0.000-0 0.000-0 0.000-0 3.692-2 E-4 BW#nniinkv l
._ _ __ __ . _ _ _ _ __ __ _ . . _ . _ _ _ _ _ _ _ ~ _ . - _
4 i Table E-3. Dosimeter Activation Cross Sections, b/ atom"' lCont'd) Group _ Upper No. Energy (eV) **Ti(n.p) 238 V(n,f) fe(n,p) 68 Ni(n.p) es cu(na) 6'Co(n,y) 4 - 31 2.606+4 0.000-0 8.361-5 0.000-0 0.000-0 0.000-0 9.668-2 ! 0',000-0 32 2.418+4 '0.000-0 8.624-5 0,000-0 0.000-0 3.587-2 l 33 2.188+4 0.000-0 9.269-5 0.000-0 0.000-0 0.000-0 5.816-2 34 1.505+4- 0.000-0 9.681-5 0.000-0 0.000-0 0.000-0 9.916-2 , 35 7.108+3 0.000-0 3.211-5 0.000-0 0.000-0 0.000-0 1.906-1 36 3.355+3 0.000-0 3.380-9 0.030-0 0.000-0 0.000-0 4.447-2 37 1.585+3 0.000-0 8.094-4 0.000-0 0.000-0 0.000-0 2.462-2 38 4.540+2 0.000-0 1.279-5 0.000-0 0.000-0 0.000-0 2.424-1 39 2.145+2 0.000-0 1.857-3 0.000-0 0.000-0' O.000-0 7.332+1 40 1.013+2 0.000-0 2.814 0.000-0 0.000 0.000-0 2.782+0-41 3.727+1 0.000-0 1.518-4 0.000-0 0.000-0 0.000-0. 1.730+0 42 1.068+1 0.000-0 7.968-5 0.000-0 0,000-0 0.000-0 2.361+0 43 5.044+0 0.000-0 5.481-7 0.000-0 0.000-0 0.000-0 3.533+0 44 .l.855+0 0.00J-0 5.600-7 0.000 0.000-0 0.000-0 5.344+0 45 8.764-1 0.000-0 1.100-6 0.000-0 0.000-0 0.000-0 7.722+0 46 4.140-1 0.000-0 2.000-6 0.000-0 -0.000-0 0.000-0 1.464+1 I 47 1.000-1 0.000-0 4.300-6 0.000-0 0.000-0 0.000-0 2.922+1
*ENDF/85 values that have been flux weighted (over BUGLE energy groups) based on a 23s u fission. spectrum in the fast energy range plus a 1/E-shape in the intermediate energy range.
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- Figure F-2. Tension Test Stress-Strain Curve for Base Metal Plate Heat M-1004-2. Specimen No. 2K5. Tested at 250F Specimen: 2KS Test Temp.: 250 F( 121 C)
Strength Yield: 65484. - UTS: 85778. _ eme, e2. . 05 - 530. X - 2 d e d g em i _ , , , , g b b M M O & .5 _ ama. c h 40 'g C 0 .E G - _ 200. & c c W W
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- 89. _
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- l. Program for Irradiation Surveillance of Waterford Unit Three Reactor Vessel Materials, C-NLM-003. Revision 1, October 30, 1974.
- 2. Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Waterford Unit 3 Reactor Vessel Materials, TR-C-MCS-0ql, Combustion Engineering, Inc., Windsor, Connecticut.
- 3. Code of _ Federal Regulation, Title - 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix H, Reactor Vessel Material Surveillance Program Requirements.
- 4. Code of Federal Regulation, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix G, Fracture Toughness Requirements.
- 5. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Appendix G, Protection Against-Nonductile Failure (G-2000).
- 6. ASTM Standard E208, " Method for Conducting Drop-Weight Test to Determine Nil Ductility Transition Temperature of Ferritic Steels," in ASTM Stanoards, American Society for Testing and Materials, Philadelphia, PA.
- 7. A. G. Ragl, et al., Louisiana Power and Light Waterford Steam Electric Station Unit No. 3, Evaluation of Baseline Specimens, Reactor Vessel Materials Irradiation Surveillance Program, TR-C-MCS-002, Combustion Engineering, Inc., Windsor, Connecticut, August 1977.
- 8. ASTM Standard E8, " Standard Methods of Tension Testing of Metallic Materi-als," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 9. ASTM Standard E21, " Standard Recommended Practice for Elevated Temperature Tension Tests of Metallic Materials," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 10. ASTM Standard E184, " Standard Practice for Effects of High-Energy Neutron -
Radiation on the Mechanical Properties of Metallic Materials," in - ASTM Standards, American: Society for Testing and Materials, Philadelphia, PA. G-2 SW##aniE&w '
- 11. ASTM Designation U?-72, " Method for Notched Bar Impia:t Testing of Metallic Materials," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 12. Standardized Specimens for Certification of Charpy Impact Specimens from the Army Materials and Mechanics Research Center, Watertown, Mass. 02172, Attn:
DRXMR-MQ.
- 13. ASTM Designation A370-77, " Methods and Definitions for Mechanical Testing of Steel Products," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 14. ASTM Designation E23-86, " Methods for Notched Bar Impact Testing of Metallic Materials," in ASTM Standards, american Society for Testing and Materials, Philadelphia, PA.
- 15. ASTM Designation E185-XX (to be released), Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA. ,
- 16. S. Q. King, Pressure Vessel Fluence Analysis for 177-FA Peactors, BAW-1485P.
Revision 1, Babcock & Wilcox, Lynchburg, VA, March 1988.
- 17. B&W's VersNn of DOTIV Version 4.3, Filepoint 2A4, "One- and Two-Dimensional Transpon Code System," Oak Ridge National Laboratory, Distributed by the Radicion Shielding Information Center as CC-429, November 1, 1983.
- 18. " Bugle - 80 Coupled 47 Neutron, 20 Gamma-Ray,3 P , Cross Section Library for LWR Shielding Calculations," Radiation Information Shielding Center, DLC-75.
- 19. Dosimeter File ENDF/B5 Tape 531, distributed March 1984, National Neutron Data Center, Brookhaven National Laboratory, Upton, Long Island, NY.
- 20. American Society of Testing Materials, Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (DPA), E693-79, (Reapprov-ed 1985).
- 21. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel Material, Reaulatory Guide 1.99. Revision 2, May 1988.
G-3 S W # n U "c h 22, Code of Federal Regulations, Title 10, Part 50,61, fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, e G-4 13WitsafE h r _ _- - - _ - - - - -}}