ML20045D931
| ML20045D931 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/30/1993 |
| From: | Cavanaugh G, Lowe A, Redmond K ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20045D927 | List: |
| References | |
| BAW-2160, NUDOCS 9306300239 | |
| Download: ML20045D931 (97) | |
Text
-r 4 BAW-2160 June 1993 ANALYSIS OF CAPSULE 97*
BALTIM0RE GAS & ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO, 1
-- Reactor Vessel Material Surveillance Program'-- ,
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BAW-2160 June 1993 ANALYSIS OF CAPSULE 97" BALTIM0RE GAS & ELECTRIC COMPANY LALViki CLIFFS iiUCLEAR POEn PLANT UNIT NO.1
-- Reactor Vessel Material Surveillance Program --
by A. L. Lowe, Jr., PE K. R. Redmond W. R. Stagg B&W Nuclear Service Company and G. P. Cavanaugh ABB-Combustion Engineering (Neutron Dosimetry Evaluations - Section 6)
B&W Document No. 77-2160-00 I (See Section 10 for document signatures)
B&W NUCLEAR SERVICE COMPANY Engineering and Plant Services Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 BW!!na?i%r
SUMMARY
This report describes the results of the examination of the second capsule (Capsule 97*) removed from the Baltimore Gas and Electric Co. Calvert Cliffs Nuclear Power Plant Unit No. I as part of the reactor vessel surveillance program. The program monitors the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor vessel materiais by the Lesling I and evaluation of tension and Charpy impact specimens. The program was designed in accordance with the requirements of ASTM Specification E185-66. The program i meets the requirements of ASTM Specification E185-73.
The capsule received an average fast fluence of 2.64 x 10'8 n/cm2(E > 1.0 MeV) and the predicted fast fluence for the reactor vessel T/4 location at the end of the tenth cycle is 1.07 x 10 n/cm 2(E > 1 MeV). Based on the calculated fast i flux at the vessel wall, an 80% load factor, and the planned fuel management, the projected fast fluence that the Calvert Cliffs Nuclear Power Plant Unit No. I reactor pressure vessel inside surface will receive in 40 calendar years of operation is 3.46 x 10 n/cm 2(E > 1 MeV) and the corresponding T/4 fluence is 2
calculated to be 1.89 x 10 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 used for predicting the change in both the increase in the RT NDT and the decrease in upper-shelf properties due to irradiation are conservative.
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CONTENTS Page
- 1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1
- 2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
- 3. SURVEILLANCE PROGRAM DESCRIPTION . . . . . . . . . . . . . . . . . 3-1
- 4. PRE-IRRADIA110N IL515 . .................. . . 4-1
- 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 . . . . . . . . . . . . . . . . . . . . 5-2 5.4. Charpy V-Notch Impact Test Results . . . . . . . . . . . . . 5-2
- 6. NEUTRON FLUENCE AND DOSIMETRY ANALYSIS . . . . . . . . . . . . . . 6-1 6.1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2. Fluence Evaluation Methodology. . . . . . . . . . . . . . . . 6-2 6.3. Vessel Fluence . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.4. Projected Fluence Distribution ...............6-7 6.5. Capsul e Fl uence . . . . . . . . . . . . . . . . . . . . . . . 6-8
- 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. Pressurized Thermal Shock (PTS) Evaluation .........7-2 7.4. Neutron Fluence Analysis . . . . . . . . . . . . . . . . . . 7-3
- 8.
SUMMARY
OF RESULTS ........................8-1
- 9. SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . . . . . . . . . . . . . . 9-1
- 10. CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1
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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. Tension Test Stress-Strain Curves . . . . . . . . . . . . . . . . . . D-1 E. References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-1 List of Tables Table 3-1. Specimens in Surveillance Capsule 97 . . . . . . . . . . . . . . 3-2 1-2. Chemical Composition and Heat Treatment of Surveillance Materials . 3-3 5-1. Conditions of Thermal Monitors in Capsule 97* . . . . . . . . . . 5-4 5-2. Tensile Properties of Capsule 97* Base Metal and Weld Metal 2
Irradiated to 2.64 x 10' n/cm (E > 1 MeV) . . . . . . . . . . . 5-5 5-3. Charpy Impact Results for Capsule 97" Base Metal Long'itudinal (LT) Orientation, Heat No. C-4441-1, 2.64 x 10 n/cm . . . . . . 5-6 5-4. Charpy Impact Results for Capsule 97" Base Metal Transverse (TL) Orientation, Heat No. C-4441-1, 2.64 x 10'8 n/cm2 . . . . . . 5-6 5-5. Charpy impact Results for Capsule 97* Base Metal Heat-Affected Zone Material, Heat No. C-4441-1, 2.64 x 10 n/cm2 . . . . . . . . 5-7 5-6. Charpy Imp'act Results for Capsule 97* Weld Metal, 33A277/3922, 2.64 x 10 n/cm 2 . . . . ... . . . . . . . . . . . . . . . . . 5-7 6-1. Neutron Flux Dosimeters . . . . . . . . . . . . . . . . . . . . 6-12 6-2. Peak Fast Neutron (E > 1.0 MeV) Flux and fluence Summary Beginning of Life to End of Cycle 10 . . . . . . . . . . . . . . 6-12 6-3. Peak Neutron (E > 0.1 MeV) Flux and Fluence Summary Beginning of Life to End of Cycle 10 . . . . . . . . . . . . . . 6-13 6-4. Peak Thermal Neutron Flux and Fluence Summary Beginning of Life to End of Cycle 10 . . . . . . . . . . . . . . 6-13 6-5. Peak Iron DPA Summary: Beginning of Life to End of Cycle 10 . . . 6-14 6-6. Normalized Flux and DPA Radial Distribution Through the Reactor Vessel Beginning of Life to End of Cycle 10 . . . . . . . 6-14 6-7. Extrapolated Peak Fast, E > 0.1, and Thermal Neutron Plus DPA at the Vessel / Clad Interface . . . . . . . . . . . . . . . . . . 6-14 6-8. Calvert Cliffs Unit 1 Surveillance Capsule 97*
Measured Activities . . . . . . . . . . . . . . . . . . . . . . . 6-15 6-9. Calvert Cliffs Unit 1 Core Power History . . . . . . . . . . . . 6-16 6-10. Normalized Neutron Energy Spectrum at the Surveillance Capsule 97* . . . . . . . . . . . . . . . . . . . . . 6-17 6-11. Surveillance Capsule 97 Interpreted Fast Neutron Fluxes . . . . . 6-18 6-12. Surveillance Capsule Lead Factors for Fast Neutron (E > 1.0 MeV) Fluence . . . . . . . . . . . . . . . . . . . . . . 6-1B
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Tables (Cont'd)
Table Page 6-13. Surveillance Capsule 97* Interpreted Thermal Neutron Fluxes . . . 6-19 7-1. Comparison of Calvert Cliffs Unit 1, Capsule 97" Tension Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 7-2. Summary of Calvert Cliffs Unit 1 Reactor Vessel Surveillance Capsule Tensile Test Results .. . . . . . . . . . . . . . . . . . . . . . 7-5 7-3. Observed Vs. Predicted Changes for Capsule 97" Irradiated Charpy impact Properties - 2.64 x 10 n/cm' (E > 1 MeV) . . . . . . . . 7-6 7-4. Evaluation of Reactor Vessel End-of-Life Pressurized Thermal Shock Criterion - Calvert Cliffs Unit 1 . . . . . . . . . . . . . 7-7 A-1. Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials - Calvert Cliffs Unit 1 . . . . . . . A-3 A-2. Type and Quantity of Specimens Contained in Each Irradiation Capsule Assembly . . . . . . . . . . . . . . . . . . . . . . . . . A-4 B-1. Tensile Properties of Unirradiated Shell Plate Material, Heat No. C-4441-1, Longitudinal . . . . . . . . . . . . . . . . . . B-2 B-2. Tensile Properties of Unirradiated Shell Plate Material, Heat No. C-4441-1, Transverse . . . . . . . . . . . . . . . . . . . B-2 B-3. Tensile Properties of Unirradiated Shell Plate HAZ Material, Heat No. C-4441-1, Transverse . . . . . . . . . . . . . . . . . . B-3 B-4. Tensile Properties of Unirradiated Weld Metal 33A277/3922 . . , . . B-3 C-1. Charpy Impact Data from Unirradiated Base Material, Longitudinal Orientation, Heat No. C-4441-1 . . . . . . . . . . . . . . . . . . C-2 C-2. Charpy impact Data from Unirradiated Base Material, Transverse Orientation, Heat No. C-4441-1 . . . . . . . . . . . . . . . . . . C-2 C-3. Charpy Impact Data from Unirradiated Base Metal, Heat-Affected Zone, Heat No. C-4441-1 . . . . . . . . . . . . . . . . . . . . . . C-3 C-4. Charpy Impact Data from Unirradiated Weld Metal, 33A277/3922 . . . C-3 List of Fioures Figure 3-1. Reactor Vessel Cross Section Showing Location of Capsule 97" in Cal vert Cl if f s Uni t 1 . . . . . . . . . . . . . . . . . . . . . . 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 Compartment Assembly.
(Four per Capsule) ... . . . . . . . . . . . . . . . . . . . . . 3-7 5-1. Photographs of Thermal Monitor Melt Wire Capsules as Removed From Surveill ance Capsule . . . . . . . . . . . . . . . . . . . . 5-8 i 5-2. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal, Transverse Orientation . . . . . 5-9 I
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Figure Page 5-3. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal Heat-Affected Zone . . . . . . . 5-10
- 4. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Weld Metal 33A277/3922 . . . . . . . . . . . .
5-11 5-5. Charpy Impact Data for Irradiated Base Metal, Longitudinal Orientation,. Heat No.'C-4441-1 . . . . .. . . . . . . . . . . ... 5-12 5-6. Charpy Impact Data-for Irradiated Base Metal, Transverse Orientation, Heat No. C-4441-1. . . . . . . . ... . . . . . . . . 5-13' 5-7. Charpy Impact Data for Irradiated Base Metal, Heat-Affected .
Zone, Heat.No. C-4441-1 . . . . . . . . . . . ... . . . . . . . . . 5-14 .
5-8. Charpy Impact Data for Irradiated ~ Weld Metal, 33A277/3922 . . . . . 5-15 5-9. Photographs of Charpy Impact Specimen Fracture Surfaces -
Base Metal, Longitudinal . . . . . .
. . . . . . . . . . . . . . . 5-16 5-10. Photographs of Charpy Impact Specimen. Fracture Surfaces -
Base Metal, Transverse . . . . . . . . . . . ... . . . . . . . . 5-17; 5-11. Photographs of Charpy Impact Specimen Fracture Surfaces -
Base Metal, Heat-Affected Zone . . . . . . . . . . . . . . . . . 5-18 5-12. Photographs of Charpy Impact Specimen Fracture Surfaces -
Weld Metal 33A277/3922. . . . . . . . . . . . . . . . . . . . . . . 5 .
6-1. DDT Planar Geometry . . . . . . . . . . . . . . . . . . . . . .-. 6-20' 6-2. Surveillance Capsule 97" . . . . . . . . . . . . . . . . . . . .. 6 6-3. Cycle I to 10 Planar Average Core Power Distribution . . . . . . 6-22' 6-4. Cycle 1 to 10 Average Axial Power Distribution . . .
. . . . . . .l6-23 6-5. Cycle 1 to 10 Fast Neutron Fluence Distribution at the Vessel /Cl ad Interf ace . . . . . . . . . . .. . . . . . . . . . . . .
6-24 '
6-6. Cycle 1 to 10 Average Fast Neutron Axial Flux Distribution . .. . 6-25; 6-7. Cycle 11 Planar Average Core Power Distribution . . . . . . . . . 6-26 6-8. Cycle 11 Average Axial Power Distribution . . . ._ . . . . . . . . 6-27 '
6-9. Extrapolated fast Neutron Fluence. Distribution at the-Vessel /
Clad Interface 32-Effective Full Power (2700LMwT) Years _ . . . . . 6-28 7-1. Comparison of Unirradiated and~ Irradiated Charpy Impact Data Curves for Plate Material Longitudinal Orientation, . Heat No. C-4441-1 . . 7-8 7-2. Comparison of Unirradiated and Irradir.ted Charpy Impact Data Curves-for Plate Material Transverse Orientation',' Heat.No. C-4441-1 . . . 7-9 7-3. Comparison of Unirradiated and Irradiated Charpy-Impact Data Curves for Base Metal, Heat-Affected-Zone, Heat No. C-4441-1 . . . . . . 7-10 7-4. Comparison of Unirradiated.and-Irradiated Charpy Impact Data Curves for Weld Metal 33A277/3922 . . . . . . . . . . . . ~- .-.
. ... . . s . 7-11 A-1. Location:and Identification of Materials Used in the Fabrication of Calvert Cliffs Unit-I Reactor Pressure Vessel .: . . . . . . . . A C-1. Charpy Impact Data from Unirradiated Base Metal (Plate), .
Longitudinal Orientation, Heat No.~ C-4441-1. _ . . . . . . . . . . . C-4' C-2. Charpy Impact- Data from Unirradiated Base Metal (Plate), .
Transverse.. Orientation,-Heat No. C-4441-1 . . . . . . . . ... . . ..C-5 C-3. Charpy Impact Data from Unirradiated. Heat'-Affected-Zone Base Metal , Heat No. C-4441-1 . . . . . . . . . . . . . . .' . . . . C .
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Fiaures'(Cont'di Figure ~Page 10 -4. Charpy Impact Data for Unirradiated Weld Metal,: 33A277/3922 . . .. . C-7 D-1. Tension Test Stress-Strain Curve for Base Metal. Plate Heat '
C-4441-1,' Specimen No. IKA,-Tested at 70F . . . . . . . . . . . . .'D-2 i D-2. Tension Test Stress-Strain Curve for Base Metal Plate Heat- .
C-4441-1, Specimen No. IKM, Tested at 250F . . . . . . . . . . . . D-2.
-D-3. Tension Test Stress-Strain Curve for Base Metal Plate' Heat >
C-4441-1, Specimen'No. 1K7,= Tested:at 550F .. . . . . . . . . . . D-3 D-4. Tension Test Stress-Strain Curve for: Base Metal Heat-Affected
- Zone, Heat C-4441-1, Specimen No. 4J5, Tested at 70F .. . .. . . D-3 ,
D-5. Ter* 19r Test Stress-Strain Curve for Base Metal Heat-Affected . '
Zone, Heat C-4441-1, Specimen No. 4J4, Tested at 250F . . . . . . . D-4: -
D-6. Tension Test. Stress-Strain Curve for Base Metal Heat-Affected Zone, Heat C-4441-1, Specimen No. 4JP, Tested at 550F . . . . . . . D-4 '
D-7. Tension Test Stress-Strain Curve for Weld Metal 33A277/3922, Specimen No. 3JD, Tested at 70F . . . . . . . . . . . . . . . . . . D-5 .
D-8. Tension Test Stress-Strain Curve for Weld Metal 33A277/3922, -
Specimen No. 3JB, Tested at 250F . . . . . . . . . . . . . .. . . D-5 D-9. Tension Test Stress-Strain. Curve for Weld Metal.33A277/3922, Specimen No. 3JJ, Tested at 550F . . . . . . . . . . . . . . . . . D-61 :
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- 1. INTRODUCTION This report describes the results of the examination of the second capsule (Capsule 97") of the Baltimore Gas & Electric Co., Calvert Cliffs Nuclear Power Plant Unit No.1 (Calvert Cliffs Unit 1) reactor vessel material surveillance-program (RVSP). The capsule was removed and the contents evaluated after being irradiated in the Calvert Cliffs Unit I reactor as part of the reactor vessel materials surveillance program (Combustion Engineering (C-E) Report B-NLM-010').
The capsule experienced a fluence of 2.64 x 10 n/cm' (E > 1 MeV), which is the equivalent of approximately 21 effective full power years' (EFPY) operation of the Calvert Cliffs Unit I reactor vessel inside surface based on Cycle 11 fuel schemes to E0L. The first capsule (Capsule 263*) from this program was removed and evaluated after the third fuel cycle and the results are reported in Battelle report" dated December 15, 1980.2 The objective of the program is_ to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions. The surveillance program for Calvert Cliffs Unit I was designed and furnished by Combustion Engineering, Incorporated (C-E) as described in CENPD-34' and conducted in accordance with 10CFR50, 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 The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water-cooled reactors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to the highest levels of neutron irradiation.
The general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels such as SA533, Grade B, used in the fabrication of the Calvert Cliffs Unit I reactor vessel, are well characterized and documented in the literature. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. The most significant mechanical property change in reactor pressure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the Charpy upper-shelf energy value.
Appendix G to 10CFR50, " Fracture Toughness Requirements,"5 specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitations for operation of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10CFR50 became effective on August 16, 1973, the requirements are applicable to ' all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effective date.
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Appendix H to 10CFR50, " Reactor Vessel Materials Surveillance Program l Requirements,"4 defines the material surveillance 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 the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life.
A method for guarding against brittle fracture in reactor press'ure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel (B&PV) Code,Section III, " Nuclear Power Plant Components."' This method utilizes fracture mechanics concepts and the reference nil-ductility temperature, RTNDT, which is defined as the greater of the drop weight nil-ductility transition temperature 7
(per ASTM E208 ) or the temperature that is 60F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion. The RT NDT f a given material is used to index that material to a reference stress intensity factor curve (KIR' curve), which appears in Appendix G of ASME B&PV Code Section III. The K curve IR is a lower bound of dynamic and crack arrest fracture toughness results obtained from several heats of pressure 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. The operating limits can then be determined using these allowable stress intensity factors.
The RT and, in turn, the operating limits of a nuclear power plant, can be NDT 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 NDT to adjust it for radiation embrittlement. This adjusted RT is used to index the material NDT to the K IR curve which, in turn, is used to set operating limits for the nuclear BWitnMa % r
power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.
Appendix G, IrrFR50, also requires a minimum Charpy V-notch upper-shelf energy of 75 ft-lbs for all beltline region materials unless it is demonstrated that lower values of upper-shelf fracture energy will provide an adequate margin for deterioration as the result of neutron radiation. No action is required for a material that does not meet the 75 ft-lb requirement provided the irradiation deterioration does not cause the upper-shelf energy to drop below 50 ft-lbs. The regulations specify that if the upper-shelf energy drops below 50 ft-lbs, it must be demonstrated in a manner approved by the Office of Nuclear Regulation that the lower values will provide adequate margins of safety.
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- 3. SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for Calvert Cliffs Unit I comprises six surveillance capsules designed to monitor the effects of the neutron and thermal environments on the materials of the reactor pressure vessel core region. The c'apsules, which were inserted into the reactor vessel before initial plant startup, are posi-tioned near the inside wall of the reactor vessel at the locations shown in Figure 3-1. The capsule positions were designed to be near the peak axial and azimuthal neutron flux. During the ten cycles. of operation, Capsule 97" was irradiated in the 97" position adjacent to the reactor vessel wall as shown in Figure 3-1.
Capsule 97* was removed during the tenth refueling shutdown of Calvert Cliffs Unit 1. The capsule contained Charpy V-notch impact test specimens fabricated from the one base metal (SA533, Grade B1), heat-affected-zone material, and a weld metal . Base metal Charpy V-notch impact specimens were fabricated from material in both the transverse and the longitudinal orientations. Tension test specimens were fabricated from the base metal, heat-affected-zone material, 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 in Figures 3-2 through 3-4. The chemical composition and heat treatment of the surveillance material in Capsule 97* are described in Table 3-2.
All plate and heat-affected-zone specimens were machined from the 1/4-thickness (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 with their longitudinal axes either parallel or perpendicular to the principal working direction.
3-1 BGbV NUCLEAR TECHNOLOGIES
There are three ' sets of nine dosimeters per surveillance capsule; one each located in the top, middle, and bottom of the capsule. The neutron dosimeters contained in each set in Capsule 97* are as follows:
Threshold Material Shieldina Reaction Enerav (Mev) Half-Life Uranium None/Cd Ursa (n,f) Cs7 1.2 30.2 years 2
Sulfur None S (n,p) p 2 2.9 14.3 days 64 Iron None Fe (n,p) Mn 64 4.0 312.5 days Nickel Cd Ni'8 (n,p) Co 68 5.0 70.9 days Copper Cd Cu' (n,a) Co* 7.0 5.27 years Titanium None Ti (n,p) Sc'8 8.0 83.8 days Cobalt None/Cd Co** (n,y) Co* Thermal 5.27 years There are three sets of four thermal monitors of low-melting alloys; one each located at the top, middle, and bottom of Capsule 97 . The eutectic alloys and their melting points are as follows:
Melting Alloy ComDosition. wt% Point. F 92.5 Pb, 5.0 Sn, 2.5 Ag 536 90.0 Pb, 5.0 Sn, 5.0 Ag 558 97.5 Pb, 2.5 Ag 580 97.5 Pb, 0.75 Sn, 1.75 Ag 590 Table 3-1. Specimens in Surveillance Caosule 97*
Number of Test ~Soecimens Material Description Tension CVN 1moact Base Metal, C-4441-1 Longitudinal 3 12 Transverse -
12 Heat-Affected Zone 3 12 Weld Metal (33A277/3922)* 1 12 Total Per Capsule 9 48
- Weld wire heat and flux lot identifiers.
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Table 3-2. Chemical Composition and Heat Treatment of Surveillance Materials Chemical Composition, w/o Heat No. Weld Metal C-4441-l) 33A277/3922" Element (D-7206-3) (9-203)
C 0.26 0.15 Mn 1.29 1.05 P 0.011 0.014 S. 0.016 0.013 Si 0.24 0.20 Ni 0.64 0.18 Cr 0.08 0.06 Mo 0.69 0.55 Cu 0.12 0.24 Heat Treatment Heat No. Temp. F Time, h Coolina Plate 1600 25 4 Water Quenched (C-4441-1) 1225 25 4 Air Cooled 1150 25 40 Furnace Cooled to 600F Weld Metal 1150 25 40 Furnac#e Cooled to 600F (33A277/3922)
(a) Chemical analysis by Combustion Engineering of surveillance program test plate."
(b) Chemical analysis by Combustion Engineering of surveillance program test weld metal."
CRCBGW NUCLEAR WiTECHNOLOGIES J
i Figure 3-1. Reactor Vessel Cross Section-Showing Location of Caosule 97* in Calvert Cliffs Unit 1 U '
180 c------ 0
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104 h-~ --
,~~,n f ld y \ inlet i , \ Nozzle I
( gCore Shroud )
g Core Support Barrel . L /
/
\ ' Vessel Vessel 0
970 %
y Reactor Vessel , 263 Vessel,/ E '
~ -
a .
Vessel 83
' 2770 .
,2 s /
\ .
I I /
%e # ' *== ==.8 ..____- % , smo sus /
i
// ,
A /
N, n
/
O i
3-4 BWs?##fsinLv q l
7 Figure 3-2. Typical Surveill'ance: Capsule Assembly Showing location of Specimens and Monitors 1 ;
Lock Assembly Y5 7
/- >
Wedge Coupling; Assembly Tensile - Monitor- m Compartment s , -
NW
/ ::
'" > Charpy impact Compartments Tensile-Monitor '3 -
Compartment N sp
- 4 ss ,
sv Nsa > Charpy impact Compartments sg
/P i s a Tensile -Monitor.
Compartment -
- sa kI i a
- ~
as,ypygyS?J.y -- 1
- I
'l Figure 3-3. Typical surveillance Capsule Tensile'- Monitor. .
Compartment Assembly iihree per Caosuler ';
Wedge Coupling - End Cap
. Flux Spectrum Menitor Cadmium Shielded Flux Monitor Housing .. . :
Stainless Steel Tubing- ,
Stainless. Steel Tubing Cadmium' Shield-Threshold Detector s
/ Threshold Detector-N Flux Spectrum Monitor y .
l J L i- Quartz Tubing Temperature Monitor- N[ :
Temperature Monitor-L M lting Alloy-Housing - 1 J
Tens.le i Specimen t-Split Spacer Y; '
k Tensile Specimen Housing -
-Rectangular Tubing l i
~
-Wedge Coupling - End Cap .
N 3-6 BWsWv'aVS"amim -
m
Figure 3-4. Typical Surveillance Capsule Charpy Impact !
Compartment Assembly (Four- per Caosule) 1
.)
[ -Wedge Coupling - End Cap
\[
Charpy impac; Specimens ;
f % '
Spacer = a
% s
%/
N -
/\
" Rectangular Tubing Wedge Coupling - End Cap N :
3-7 BWs"fsefEi h r
- 4. PRE-IRRADIATION TESTS Unirradiated material was evaluated for two purposes: (1) to establish a baseline of data to which irradiated properties data could be referenced; and (2) to determine those material properties to the extent practical from available 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 Calvert Cliffs Unit 1 surveillance program. The details of 8
the testing procedures are described in C-E Report TR-ESS-001 and the results are summarized in Appendices B and C. The Charpy data was re-evaluated to be consistent with the irradiated data.
\
l l
4-1 BW!!sMEiM =r
- 5. POST-IRRADIATION TESTING 5.1. Visual Examination and Inventory The capsule was inspected and photographed upon receipt. The markings were confirmed as those of Capsule 97*. The contents of the capsule were inventoried and found to be consistent 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 97 contained three temperature monitor holder blocks each containing four fusible alloys with different melting points. Each of the thermal monitors 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 exposed to a maximum temperature no greater than 580F during the reactor vessel operating period. This is not significantly greater than the nominal inlet temperature of 548F, and is considered acceptable. These observations are similar to those reported for Capsule 263*. There was complete melting of the 536F melting point alloy and partial melting of the 558F melting point alloy.
This indicates that the temperature was probably at 558F for a period of several seconds during some portion of the irradiation of Capsule 263*. The partly melted or slumped appearance of the 580F monitor in Capsule 97 is probably due to an irradiation induced creep mechanism and is not the result of actual melting. This being the case, then the maximum temperature was no greater than 580F and most probably not greater than 558F. This type of behavior has been seen in other surveillance capsules." There appeared to be no signs of a I temperature gradient along the capsule length.
~
+
BGW NUCLEAR TECHNOLOGIES
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-lb load capacity MTS servohydraulic computer-controlled universal test machine. All tests were run using stroke control with an initial actuator travel rate of 0.005 inch per minute through yield point. Past specimen yielding an actuator speed of 0.040 inch per minute was used. A 4-pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point. The tension-compression load cell used had a certified accuracy of better than +0.5% of full scale (25,000 lb). Test conditions were in accordance with the applicable requirements of ASTM A370." The data for the base metal specimen, tested at 250F, was lost because of a test machine malfunction.
Photographs of the tension test specimen fractured surfaces are presented in l
l Figures 5-2 through 5-4.
In general, the ultimate strength and yield strength of each material increased with a corresponding slight decrease in ductility as compared to the unirradiated values;- both effects were the result of neutron radiation damage. The type of behavior observed and the degree to which the material properties changed is within the range of changes to be expected for the radiation environment to which - -
the specimens were exposed.
The results of the pre-irradiation tension tests are presented in Appendix B.
5,4, Charoy V-Notch Impact Test Resul_t_s 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 fracture 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" on a Satec SI-lK impact tester certified to meet NIST ("Watertown") standards." .
l The data show that the materials exhibited a sensitivity to irradiation within the values to be expected based on their chemical composition and the fluence to which they were exposed. Detailed discussion of the results are provided in Section 7.
l 5-2 BW!!n2ELr
The results of the pre-irradiation Charpy V-notch impact tests _ are given in-
-Appendix C.
5-3 BWilnE?a % r
Table 5-1. Conditions of Thermal Monitors in Caosule 97*
Capsule Melt Post-Irradiation Seament Temperature Condition Al 536F Melted (Top) 558F Melted 580F Unmelted 590F Unmelted A4 536F Melted (Middle) 558F Melted -
580F Unmelted 590F Unmelted A7 536F Melted (Bottom) 558F Melted 580F Unmelted 590F Unmelted 5-4 BWitnMiiMim i i
1 1
Table 5-2. Tensile' Properties of Capsule 97* Base Meta'l and Weld '
t Metal Irradiated to 2.64 x 10" n/cm' (E > 1 MeV)***
Strenath, psi Fracture Elonaation. % Reduction Specimen Test Temp, load, Stress, Strength, in Area, No. F Yield Ultimate -lbs ksi ksi Uniform Total %
Base Metal (Lonaitudinal): C-4441-1 IKA 70 93,300 119,800 3,925 202.7 80.0 11.1 22.3 60.6<
IJM** 250 ---- ----- --- ----- ---- ---- ---- ----
IK7 550 82,600 110,600 3,830- 192.9 78.0 8.4 19.0 ~59.6 Base Metal Heat-Affected Zone. C-4441-1
^ }"
4J5 70 91,300 98,300 2,946 199.8 60.0 4.0* 17.8* 70.0
'4J4 250 84,300 91,000 3,141 185.1 64.0 2.5* 15.4* 65.4 4JP 550 82,700 104,500 3,664. -155.9 74.6 8.9 16.41 52.1 4
l Weld Metal 33A27//3922' 3JD 70 -94,500 109,600 3,454 190.3 7 10 . 4 ' 11.2 25.7 63.0-53 3JB. 250- 87,200- 101,200 -3,293 196.7 67.1 10.3 23.1 65.9 EE g3 3JJ 550- 86,900 -107,000 4,638 262.4 94.5- 11.6 23.2 64.0 sE R .
Results not valid Especimen' necked and~ fractured outside extensometer gage length..
- Data for specimen' lost because of equipment ' malfunction.
I>
m
- Stress-strain curves areLpresented in Appendix ~D.
s 9
- [
u _ . _ _ _ _ _ _ _ _ . _ . . . _ _ . .. _ - _ . _ . _ . . - . - . . . . . . - . _ . _ . . - . . - . . .: ; _ . _ _ _ _ . _ _ _ _ _ . _ . _ _ _
. ~ . - .
a
-l P
Table 5-3. - Charpy . Impact Results for Capsule 97* Base > Metal Longitudinal,
- (LT) . Orientation, Heat No. C-4441-1, '2.64 x 10 n/cm 2.
Test -Impact ~ Lateral . / Shear Specimen Temperature, Energy, : Expansion,- _
Fracture,.
ID F ft-lbs inch %
'120 70 13.5 16 10- '
-12L 90 -12.5 12 15' 12P 100 .26.5- 26 :20 ,
12T 110L 22.5 23 20- ' -
12X 120 34.5 31~ 40 124 130 44.0 39 40' 123 150 52.0 42 ~- :50:
12A 170' .51.0 46 55 12H 200' 71.5 57: .: 70 L 125 240 98.5* 80 ~
100' 1 l127 280 105.0* 86' 100 126 550 99.0 81 100
- Values used to determine upper-shelf energy value per ASTM E .185."
. Table 5-4. Charpy.'. Impact- Results for Capsule 97* Base Metal Transverse -(TL)L Orientation, Heat- No'. C-4441-1, 2.64 x 10'." n/cm8 ,
a Test. Impact . Lateral' . Shear:
Specimen. Temperature,- Energy, . Expansion, Fracture, ID- F ft-lbs inch -%
e 23J 70 16.5 -16 10 246 100 23.0 :23 20 23K- 130. 26.5 28 :40 234L 140 33.5 32 -40 225 150 37.0 36; '45 23M 160 43.0 41' 50 -!
23L 180 47.5 45 50' 22L 200 58.0 49 60 '
244 220 ' 67.0 . . 61 - 80 ,
245 240 83.0* 70 95 -
226 280 85.0* - 72 100-243 550 90.0 77. -100:
- Values used tol determine upper-shelf energy value per. ASTM E-185."'
~5-6
.SWeb""ca.".-
q
I r Table-5-5. Charpy. Impact Results for Capsule 97* Base _ Metal Heat-Affected?
Zone -Material, Heat' No.' C-4441-1, 2.64 x 10 n/cm'
? Test' Impact . . Lateral- Shear Specimen Temperature, ' Energy, : Expansion, Fracture, ID F ft-lbs inch ' '%-
L
~
42E -40 -23.0- 16 -- 20 42T' -20 23.0 16 :30 42C 0 51.0. -33 -40 41Y 20- 43.5 32 50' -
427 40 61.5 -36 65- "
420 =70 43.5 ..33- 40 3 428 80 56.0 41- -70 1 426 100 42.0: 34- 50 42P 130 82.0 60- 65~
42A 160 87.0* 67 100 425 200. 75.0* 51- 100 7j 410 -550 129.5 80 100 .;
- Values used to determine upper.-shelf energy value per. ASTM E-185."' j L Table 5-6. Charpy; Impact Results for? Capsule 97*- ,
^ Weld Metal, 33A277/3922, 2.64 x -10_ n/cm 2~ '
Test Impact lateral Shear. !
Specimen Temperature,' Energy, Expansion, . Fracture, 10 F ft-lbs ~ inch % ;
32B 'O 8.0. 15- 5 32E 20 24.0 - 23 .30' >
32T 40 28.0 25 30 i 326 50 23.5- 24 35-32C 60 37.0 ' 34 40-32U 70 65.5 51 55: ~
32J 82.5 100 64. .70 327 130 91.5 74 -85 32D 160 89.0 75 90 32A 200 104.0* - 82 100 324 200 107.0* 82- 100' 325 550 ~113.0 . 88J 100'
- )
9
- Values lused to determine upper-shelf energy value per ASTM E-185."
q l
5-7 sg,ype=, thy -
. , ,- + r,--r r
Figure 5-1. Photographs of Thermal Monitor Helt Wire ,
Caosules as Removed From Surveillance Caosule l
__.,7, .
, ^
590F
%- - ;. , q - a r .:ro p l
.- ~n.
- g. mas: s .s64 w nu -
- s. _ ,.
~
l l
.. y .. =
580F Too Group
_g g &
T k \
.. , << 1 558F u 536F
~
'd9 A
'g 590F 580F Middle Group
- L aan a m. .,, , ,
. 558F
-v - .-
.b ;;
.. _. z y .
l y %-
- ~
536F l
590F
. . , . . . . +
_s i
4t.
v.
e,:,
.4-580F Bottom Group ' ,
~
558F
_. _ ,7 _ _ , ._. ..
l r
'Nk I ( h, hr-536F 5-8 Q
adWTECHNOLOGIES BGW NUCLEAR
Figure 5-2. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal. Lonaitudinal Orientation J
Specimen 1KA (70F) !
l i
No Photograph Specimen IJM (250F) i Specimen 1K7 (550F) ,
, ,(' ~-
No Photograph 4
' $v Specimen 1KA (70F) Specimen IJM (250F) Specimen IK7 (550F) l 5-9 B Wi!EE?it!A m ,
i
_* _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - c - - - - - -. - - - - -
- -. .. - . . . _ . . .- _... . ..~. - . .. .. -. .. .
l Figure 5-3. Photographs of Tested Tension Test Specimens and Corresponding ;
Fractured Surfaces - Base Metal Heat-Affected Zone t i
- , . . . . . . . , . , . . . , , . , .n i
l
._m._.;-u..~.-_._,.2.___a-Specimen 4J5 (70F) t Specimen 4J4 (250F)
L...,_,,,-..-.4..*,- . s.-- ' ' ' --
-2- l Specimen 4JP (550F) '
1 3 + l s . .,. -
14
- lW Specimen 4J5 (70F) g Specimen 4J4 (250F)
Specimen 4JP (550F) 1
~
5-10 l BWs*fEElaMim l
l
-l
i Figure 5-4. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces. - Weld Metal 33A277/3922 ;
- . _ , . _ . - _ _ . , _ . , _ _ _ . ~ . - - _ -
~
l
, , ~ o.w s--- e +m su o <- n : ~. s aa.s u.+.a mawM n+ n:~4 s 7
Specimen 3JD (70F) l l
Specimen 3JB (250F)
Specimen 3JJ (550F) !
t I
i r., , ;
i , ,, t
> !: t l
, {. .
o, t ;
i
.*'t
- r-w ,; I
,,w .
< l5 !
,,s.. . . , -
.t, ;
, Specimen 3JD (70F) Specimen 3JB (250F) Specimen 3JJ (550F) l 5-11 aswsuct.can
- OWscavice COMPANY 1
l L_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . . _ . __ .
Figure 5-5. Charpy Impact Data:for Irradiated Base Metal, 4 tonoitudinal Orientation. Heat No. C-4441 100 j M
- 75 '-
gw __ _ _ _ _ _ _ _ _.
m e E 25 _
- I I I I I I I I I I I I I O ,
,D.10 Z
~
e g'O.00 -
~
g0.05 ,
w i-g 0.04
_ _ _ _ _ _ _ _ _ _e _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
e k O.02 -
e CC W
e
)'o- I' I I I I I I I I I I I ..
zm PLANTS CALVERT CLIFTS, LNIT I m, W-97 IEC _ TCv (35 M.E) +132F T
- IT '
CV (50 FT-LB)
- T CV (30 FT-LB) *IIF 160 C 1024-l h e V-USE ( Av0) c ,40 - a 5 ,
120 -
i 100 - .
?
$So e
.k ,o
- ___ _ __ _ _ _ _ _ _ a______________._____.
e
. 40 - .
_"i g4 m ggt SA-533. Dr. .. 81
20 -
- a
'm 2.64vton/cma C4441-I t WT._ W.
1 t t l' I t t t t t t t I g
-100 -50 0 50 100 150 200 250 300 350 400 450 500. 550- ' 60C ,
l TEST TEMPERATURE. F-5-12 aswnuCLEAR BWSERIMCE CONOANY - ,
.I
g- .
- r. J u
'l Figure 5-6. Charpy Impact Data:for Irradiated Base Metal,.
Transverse Orientation. Heat'No. C-4441-1 .
i 100 i
x .
o 75 -
4 2 .
l
!E 2 - I 0
' ' ' ' I ' ' I ' I I 'I~
0.10 l 3 '
g0.00 -
O,06 -
.i g0.04 0.02 -
.J O
' ' ' ' ' I ' '. ' ' ' I I r
?
200 m, CALVERT CLIFFS. (NIT I m, V-97 (90 - T CV (35 H.E) +849F T
cv (so Fr-Ls) *' N <
160 - I Li(30 FT-LS)
' * ' 3Y m C 84 j V-USE (Ave) t-840 -
5 .;
{ 120 t 100 -
- I 8
go -
U n 4 BO '"'
k 40 -
mygygg '5A-533. Gr.- Bf;
- - ORID(TATION TRANSVERSE 20 -
m 2. 64 = t o n/es,'
my ,c; C444s-t
.g -t t t t t 1 i 1 1 t t t t-l
-800' -50 0 50- 100 150 200 250 300 350 400 450 500 .550 :Soo: j TEST TEPPERATURE, F 5-13 aswmucwan '
SMCE COfWPAAfr
.. . . . - . . . ~ . . .
. a b
i W
i I
Figure 5-7. Charpy' Impact. Data for-Irradia'ted Base Metai,- .
Heat-Affected-Zone.-Heat No.-C-4441 i 100 -
M
. ys -. ,
e e e .
-?
m _____
__w_____________________
- k. e k 25 -
i O
' ' ' ' ' ' ' ' ' ' I I ' '
,0.lo Z
m 58
~
-l ,
9 . ..
o.06 -
- W .
0.04 -
- g e , ,_e_______________________
o -
5 02 6-
. 3, i i i i , , , , i- i i -i ,
zoo PLANTe CALVERT CLIFFS, LNIT 1 m, V-97 190 - T Cv (35 M) *4tF 1
T *#
CV (50 PT-LS)
T ~'Y iso CV (30 FT-LB)
- h C ya pyaj elft-tb. 3 t , ,,
_ 4 5
y i2o -
100 -
b 8 e y ,o _ e W e ,
b O 5 4 e so -
j k
- _______a _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
e e ,
.o
___ _ _ ____ _ __ _ __ ____ _ _ __ _ ____ _ _ MTERIAL ~ SA-533. Gr. BI ,
- ORIENTATION HAZ ,
2o. -
m 2. 64 s l o i' n/cm8 j W_AT No. C4441-I j o' I ' ' ' ' ' I ' ' ' ' '
+
-too -50 o . 50 100 iso 200 250 300 350 -400 450 500 550 , soo
[
TEST TE,4'ERATURE. F 5-14 SW#Ai!MMl"=nv : l l
i
4 i
r Fiaure 5-8. Charny-Impact Data for Irradiated Weld Metal. 33A277/3922 -
800 f
M
- 75 -
j
. gw _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
u
. .e 5 25 -
' ' ' ' I I I I I I '
0 !
,0.10 .
z m -
i
[0.00 0.06 0 - -
g.04 - - - _ _ _ _ _ _ _
i
'd0.02 -
5 t-
' I I I ' ' ' ' ' I '
0 2=
PLANTS CAL M CLI M . UNIT CAPSLLE: W-97 t
100 - TCV (35 MLE) + 5Y
+6IF TCV (50 FT-LB)
F iso _ T v(30 C FT-La) 3 C v.uSE ( AVG) 106H-the t 140 -
5 m
420 -
e goo -.
8 .
W . so -
W l N eo -
k 40 -
MA g gag WELD-LIDOE 0001 CRIENTATION
. .~ 2,64w80* n/em3 20 -
m
& EAT No. 33A277/3922 I ' I I I I '~ I I I I I '
O
-100 -50 0 50 t00 150 - 200 : 250 300- 350 400 450 500 550 -soo TEST TEMPERATURE. F 5-15 sswmucuan BWSERWCE COMRnnW s
1 1
j Figure 5-9. Photographs of Charpy Impact Specimen Fracture i Surfaces - Base Metal. Lonaitudinal l E ]!! E E 5FE CfMEN NO 12v TEST TEMPERATuq . 73r E
SPECIMEN NO.12L TEST TEMPEP ATURE + 90F SPECIMEN NO 12P T E ST TEMPEF ATURE + 100F SPE CIME tt NO 127, TEST TEMPE AATURE + 110F t p g , 77r-- ., y..,._-
-3 .. ;. .
yew sr ;z
___- t
&$ gg mg" i .
, .p _
SPE C MEN NO 12 A TE ST TE MPE R ATUAE
- 120F S5'EQMEN NO 124 TEST TEMPE RATUhE + 130F 2 Ns g . 3 Q .u ss. ' " wf --
- f yg T a
h .- g .. . :
I gs
} =
g ,
- .. {'. g*
., e , . . .
] SPECIMEN NO 123. TEST TEMPERATURE + 150F SPECIMEN NO 12 A ItST TEMPERATURE + 170F
- ~ - C \h .
? '
j j -
g.r.,. g 4
- '- ( .
4 4 ,,
9 h_ . .
SPECIMEN NO 12M. TEST TEMPERATURE + 200F SPECIMEN NO 125. TEST TEMi tRATURE + 240F i .W-r 5'
s,2, E N ,m , a uS1,Mee.1ue . mf ssC-N NO 1x usr uMnFJat~ sse '
5-16 l BWs%Mfahr l
\
- -. . -_... -. l
1 Figure 5-10. Photographs of Charpy Impact Specimen Fracture Surfaces - Base Metal. Transverse
^
of ' K gy Q , . _ , .w
~
. .? ,\ * * '
- g. Q Q "*. g
$$ h'. -
Li't y.4py
'.'t !. N J ;JJ TEdi Tt'.*t'f4AlVPE +7DE r, ++-
fdg hh LFECtME N NG 246 T Lf,I T E MPE RA T UP.E
. . ., tgppy g.,gp--
100F
- , .- ,, 9 I
v', . ** P s, .$d 'D , e i,
I
.k. . 9 a %K '
f.phT. +y 6 .r lf:4e/N.
- W'**q,
- ~ * ^** -
L
, g {
mm -
GPE 2 W'4 NC 22* TEST TEMPE RATUPE + 13')F SPEC' MEN NO 234 TEST TEMPE AATURE + 1425 i
EEER
.{ P! O ME N !.O .;;b TE ST T E Mrg SAT vag .1505
-v,9 py ~. ~
' ** k cn-j en trec:MENfaQ DM Ti ST MVPf RATUPE - 160F w, -
((-
sm d
, , ~
o'
, 3*
~
S m .
1 SPE C: MEN NO 23 TE ST TEMPERATURE + 1BOF SPE C:VEN hO 22L. TEST TEMPERATURE + 2OOF gg ~ven mn ew sw- p4 .s m p: -
, ~f ,
- f gs p ,j' i,
- u. 4 i . --
t l
SPEC! MEN NO 244. TEST TEMFERATURE + 220f SPEC! MEN NO 245. TEST TEMPERATURE + 240F
- q~ p ,. .: n- :n-
.- .4* - % ..
7.,
'y . , *a
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/ '
- T
i ,
h I - ~ .
i vt !.ME N NO 22(. TE ST TEMrEH ATURE . 2BOF SPEONEN NC 243 TEST TEMPf RATURE + $50F l 5-17 B WsYafffE0!=v
r l
l Figure 5-11. Photographs of Charpy Impact Specimen Fracture Surfaces - Base Metal. Heat-Affected Zone 4N,V,1 h _-
O d SPECtME N NO .12I TEST TEMPERATURE 40F asj.gg SPECIMEN hD 42T. TEST TEMPERATURE 20F 5 , ) $. . b a" &.g Sc'ECIMEN NO 41Y. TEST TEMPERAT URE ,20F
'FE C'ME% NO 420. TEST TE MPERAT URE OF
-**&# % '*P-SPECtMt% hO 427 TEST TEMPERA TUAf + 40F SPECIMEO NO 42D TEST TEMPERATURE + 70F
.;., , , pp yg.w g v,. - - - - -
SPECIMEN NO 428. TEST TEMPERATURE + 8CF SPECtMEN NO 426, NST TEMPERATURE + 100F l
d(l' , i, ' ( '- #
- =*
SPECtMEN NO 42P. TEST TEMPERATURE + 130F SPE CIMEN NO 42A. TEST TEMPERATURE + 16W SFEC.VEN NO 425 TE ST TE MPE RATURE e'2OOF SPECIMEN NO 41U. TEST TEMPERATURE + %OF
- SW#sefSihr
(-
Figure 5-12. Photographs of Charpy Impact Specimen Fracture
! Surfaces - Weld Metal 33A277/3922
, _ _ _ ex -
> ~- .
> e .
' g'g
- p. ) .y $
l W ,0. -
a bi t t.WLN t4U J 4 ti i t h i i t Met h A l umb UF SPECIMEN NO. 32E. TEST 1EMPERATURE + 20F 4 i, 1
- s.
- ~-
.;- .. q .. . - .
l - a
... ,-X % ~ , *
. ,<;},; >< : ' .-
$_ j i -
Li I me
- l !?d i-r ' . . ,._,.s.+ ? vas SPECIMEN NO 327. TEST TEMPERATURE +40F SPECIMEN NO. 326. TEST TEMPERATURE + 50F
- ryme- , ~.-nnnnww , - s~
, SPECAMEt rJO 32C. TEST TEMPERATURE + 60f SPEC 1 MEN NO 32u. TEST TEMPERATURE + 70F y sw- .
8 E{
is.
ht .
( i l
SPECIMEN NO. 32J. TE ST TEMPERATURE + 100F SPECIMEN NO. 327. TEST TEMPERATURE + 130F
(, , . --
a D
,h .
. %'. ,p
[ -"[ ' ,
i 1
=
eu.wm 44.ia. '5. m ,
, SPECIMEtt NO. 32D. TEST TEMPERATURE + 160F SPECIMEN NO 32A. TEST TEMPERATURE + 200F
,/
eag *
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- 6. NEUTWON FLUENCE AND DOSIMETRY ANALYSIS 1
6.1. Introduction The objective of the, analysis of the surveillance capsule dosimetry measurements s to provide a measure of the fast neutron (E > 1.0 MeV) fluenc'e distribution. -
, in the reactor vessel and in the surve'llance capsule. The fast neutron fluence t . .
distribution in the reactor vessel is used to determine the degree of embrittle . .
ment of the reactor vessel steel by means of an empirical correlation,: such as - ,
the correlation given in USNRC Regulatory. Guide 1.99 Revision 2.'" This. {
correlation is based on a statistical fit between measured shifts in the 30 ft~-lb Charpy V-notch energy level and the fast neutron fluence with the chemistry of -
the steel as a parameter- .
In this work the fluence distribution in the vessel is presented relative to the. i l
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absolute value at the peak location. This provides a more general description-of the fluence distribution throughout the vessel. The key results of this
- analysis are the fast neutron fluence at the surveillance capsule and at the peak !
location on the vessel. The remainder of the results provided in this work, such ;
as the neutron fluence for E > 0.1 MeV, iron displacements-per-atom (DPA), and l the thermal neutron fluence are provided more for research applications;than for.
l the direct determination- of the degree of embrittlement of the reactor vessel. j The fluence determined at the capsule can be applied in a limited way. to.the ,
verification of the fluence / shift correlation, by comparing the measured shifts !
- obtained for the test samples in the surveillance capsule and .those. predicted ,
using the existing correlation, such as the correlation given in USNRC Regulatory -
. Guide 1.99 Revision-2. .The limitation is due to the small number of material
-samples ( that. represent only- a few of the possible combinations of. material chemistry compositions, and' the small range in the neutron . fluence that the samples experience.
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The fluence is not measured directly by the dosimeters but is derived from an interpretation of the measured dosimeter activities. The dosimeter activities are interpreted using a calculational model that incorporates the following components:
- A methodology for calculating the transport of neutrons from the source region to the surveillance capsule and the reactor vessel.
- A representation of the spatial position and dimensions-of the reactor fuel, vessel internal structures and the reactor vessel.
- The dosimeter activation cross sections as a function of neutron energy.
- The variation in the neutron flux at the dosimeter position as a function of the reactor power level and irradiation time.
- The neutron source distribution in the reactor fuel as a function of energy, position and time.
The major portion of the fast neutron fluence evaluation is based on a calculational model. The dosimeter activity measurements represent samples over a portion of the energy, spatial and time components of the fluence distribution experienced by the reactor vessel. Table 6-1 shows the lower energy limits for the dosimeter reactions and the half lives of the products whose activities are measured. The principal use of the activity measurements is to obtain estimates of the bias and uncertainties associated with the calculational model that is applied to the analysis. The estimates of the biases and uncertainties become more accurate as the number of model-to-activity-measurement comparisons is increased. The range of application of these biases and uncertainties is determined on the basis of the degree of difference between the measurement configurations that were used to derive them and the actual conditions that apply to the calculation of the reactor vessel fluence distribution. The future application of these biases and uncertainties is limited by the degree of difference between the current calculational models and models applied to future analyses.
6.2. Fluence Evaluation Methodoloav The methodology for determining the capsule and vessel fluence is based on the discrete ordinates method as used in the DOT IV" computer code. The DOT IV i
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methodology uses a multigroup approximation to the energy variable in the transport equation solution, a Legendre expansion of the flux distribution and the scattering terms, an S,, approximation to the direction variables and i scattering quadrature, and a finite difference approximation to the spatial variables. A model of the Calvert Cliffs Unit I core-to-vessel geometry was constructed based on engineering drawings and approximations to the actual three dimensional geometry required by the two dimensional DOT IV methodology.
The source distribution in the reactor fuel was derived from three-dimensional reactor-core diffusion-theory depletion calculations using the ABB-CE ROCSw207 DIT*"/MC*2 computer codes. The MC computer code provided power distribution data at the pin-by-pin level of detail for each fuel assembly. This distribution was then adapted to the DOT IV geometry and input format requirements.
The two dimensional nature of the calculation required that the three dimensional source be represented by some synthesis of two dimensional distributions. In this analysis the radial source distribution used in the DOT IV planar model was referenced to a core midplane distribution. The axial component of the source distribution was treated using a combination of axial geometry DDT IV models.
A combination of RZ DOT models was applied to the calculation of the synthesized 3-D fluence distribution at the capsule and reactor vessel positions. In addition to the axial distribution of the source in the reactor core, the axial variation in the water density in the core bypass region and the axial variation in the core shroud dimensions were considered.
The DOT IV models used a3 P expansion of the cross section transfer matrix and the flux moments, an S8 r-theta and r-z quadrature, and the 40 energy group (22 neutron and 18 gamma ray) DLC-23 22 (CASK) transport cross section set.
The effective dosimeter reaction cross sections were calculated using the SAND 23 2
computer code with the DLC-97 (DOSDAM ') ENDFB/V based library. The neutron energy spectrum used by SAND was based on the spectrum calculated with the D0T IV model of the surveillance capsele. The input data to SAND also included the saturated activities calculated from the measured activities for each of the dosimeters included in the analysis.
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The saturation factor is introduced in order to convert the measured activities, which are the result of a variety of source distributions at various power levels during the time period of irradiation, into an equivalent activity that would have resulted had the source distribution been constant and at the 100% power level over the entire irradiation. The following two equations are applied to this calculation.
i A"'= N,A"*' x A" x CF (6-1) xIx5 where:
A., - saturated activity (dps/a)
A , - measured activity (dpm/mg-dosimeter material)
Aw - atomic weight (g/ mole)
N, - Avogadro's Number (atoms / mole)
I - isotopic abundance of target isotope S - saturation factor.(see Equation 6-2)
CF = units conversion factor 11, p S = [ -i x f, x e " 7' x ( 1 - e 'd " '-) (6-2) i=1 f,,7 where:
th gjme jnggrvag P; - average core power for the i P,,, - core full power (2700 MwT)
T; -
time from end of interval "i" to reactor shutdown
.t i - length of time interval "i" f4 = radial and axial core power distribution factor at full power A = radioactive decay constant BW!!nnfaa%v
The SAND code calculates an effective cross section using the DOT IV neutron flux spectrum at the surveillance capsule as given in the .following equation:
- A,,,
8* .
(6-3) 3 i.o $ (E)dE q 1
The fast neutron flux for neutrons with energies above 1.0 MeV can then be calculated for each dosimeter material as follows: l
't o (E > 1.0 NeV ) . = ."
A' .(6-4)
'a .;
Table 6-1 lists the dosimeters in surveillance' capsule 97". There are three sets j of dosimeters in three spatial locations (top, middle, bottom). The Cobalt-59 f dosimeters, shown as Cobalt-Aluminum material in Table 6-1, provide a measure of ~ I the thermal neutron fl ux. The rest of the dosimeter mate' rials represent-' ;
reactions that measure the neutron flux spectrum above -1.0 MeV. l Figure 6-1 shows a sketch of the D0T IV planar geometry. Figure 6-2-shows a more detailed sketch of the surveillance capsule model. The average planar and axial' ] '
power distributions for the Calvert Cliffs Unit 1_ Cycle 1 to Cycle'10. irradiation period are shown in Figures 6-3 and 6-4. The planar power ' distribution was actually used at a much")reater level of detail but:is shown at the assembly by. l assembly level for the purpose of display. )!
Once the fluence at the critical weld location .is-determined via dosimetry and- !
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transport analysis, then the ARTuay can be: calculated using the methodology of. j Regulatory Guide 1.99, Revision 2. The key components to this calculation are; [
the neutron fluence (E > 1.0 MeV) and the nickel and copper content to obtain the~ !
chemistry factor. The ARTuor. for the surveillance capsule test materials. can .
also be calculated in the same manner using the appropriate fast neutron fluence ;
for the respsctive sample locations.
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i Extrapolation of the vessel fluence distribution to future times is accomplished using the same calculational model except that the source distribution is changed to the expected average distribution for the future time interval. If 100 percent power level or coolant density conditions are also expected to be different then the extrapolation model can be appropriately adjusted. If too large a deviation is made from the original calculational model then some adjustment should be made to account for changes in the applicable biases and uncertainties that have been established for the model. Once the extrapolated fluence distribution is obtained then the ARTuor can be calculated for the critical location of the vessel for the extrapolated time interval.
6.3. Vessel Fluence The total energy output of Calvert Cliffs Unit 1 from the initial startup to the end of Cycle 10, when the surveillance capsule was removed, is equivalent to 3.5 8
x 10 Effective Full Power (2700 MwT) Seconds (11.07 Effective Full Power Years).
After this period of irradiation the peak fast neutron (E > 1.0 MeV) fluence on 2
the reactor vessel is 1.96 x 10 n/cm . The corresponding peak fast neutron flux is 5.63 x 10' n/cm*-S .
The analysis by Perrin et al 2 quoted a peak fast neutron flux on the reactor 2
vessel of 4.66 x 10' (n/cm -s) . This result applies to the Cycle 1 to 3 time period and is based on the 263* capsule dosimetry. A subsequent analysis done by Nair and Williams 2s quoted a peak fast neutron flux on the reactor vessel of 4.88 x 10' ( n/cm'- s ) . This result was quoted for the Cycle 1 to 4 time interval although it was implied that the same flux applied to the Cycle 1 to 3 period of irradiation. Applying the calculational model employed in this current analysis yields a peak fast neutron flux of 5.2 x 10' (n/cm2 -s) at the reactor vessel over the Cycle I to 3 time interval. Extending the time interval to Cycles 1 to 4 gives a peak flux of 5.4 x 10' (n/cm2 -s) . Comparison of this current calculational model with the two previous models shows a difference of +11.6%
relative to the Perrin results and 410.6% relative to the Nair and Williams results.
Table 6-2 provides a summary of the peak fast neutron flux and fluence values in addition to the gradient through the reactor vessel. Tables 6-3 through 6-5 show B W !!n 2 a!!& m i
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the corresponding values for the neutron fluence E > 0.1 MeV, thermal neutron fluence, and iron dpa. The iron dpa is based on the ASTM Standard E693-79 28 (updated 1985). Table 6-6 shows the relative attenuation of these quantities in the radial direction through the vessel.
The azimuthal distribution for the fast neutron fluence at the vessel / clad interface for the time interval from initial startup to the end of Cycle 10 is shown in Figure 6-5. The distribution is normalized to the peak fluence occurring at the azimuthal location opposite the core-flat center-line (0 - 0").
This azimuthal distribution applies to the region of the vessel opposite the mid-region of the active core. Opposite the top and bottom regions of the core the distribution becomes somewhat flatter. The corresponding distributions for the E > 0.1 MeV neutron fluence, the thermal neutron fluence, and the irnn dpa at the vessel / clad interface, are very much the same as that given for the fast neutron fluence distribution. The axial distribution for these quantities is given in Figure 6-6. The axial distribution is normalized to the Cycle 1 to 10 average planar power distribution at full power (2700 MwT).
6.4, Proiected Fluence Distribution The extrapolation of the vessel fast neutron fluence distribution to 32 Effective Full Power Years used the Calvert Cliffs Unit 1 Cycle 11 source distribution as an estimate of the average that would be experienced over the period of extrapolation. Figures 6-7 and 6-8 show the planar and axial power distributions that correspond to this source distribution. Table 6-7 shows the extrapolated values at the peak vessel / clad interface location for the fast, E > 0.1, and thermal neutron fluence plus the Iron DPA. The attenuation through the vessel is similar to that given in Table 6-6.
The azimuthal distribution of the extrapolated fast neutron fluence (to 32 EFPY) is given in Figure 6-9. The axial distribution is similar to that shown in Figure 6-6 except that it is reduced uniformly by a factor of 1% due to the lower axial peaking of Cycle 11. As before, the E > 0.1 and thermal neutron fluence plus the iron dpa distributions closely follow the fast neutron distribution.
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6.5. Caosule Fluence The specific activities of the reactor vessel surveillance capsule dosimeters were measured by Babcock & Wilcox Nuclear Environmental Services.27 The results of these measurements are shown in Table 6-8. Reference 27 indicates that the copper dosimeter was difficult to measure and the measurement for the middle dosimeter location was relatively unreliable with a large measurement uncertain-ty. As a result the copper dosimeter activity measurements were not included in the fluence analysis.
The power history of the Calvert Cliffs Unit I core from the initial startup to the end of Cycle 10 is given in Table 6-9. This data is used in Equation 6-2 to calculate the saturat. ion factor for the measured activities. The saturation factor is applied to the measured activities so that a comparison can be made between the calculated fast neutron flux and the flux interpreted from the activity measurements.
The activity data for the cadmium shielded and unshielded uranium-238 dosimeters were used to determine the degree of equivalent U-235 contamination of the U-238 assuming that the cadmium shields remained effective. The measured fission rates in the shielded and unshielded U-238 dosimeters are related to the fraction of fissions due to fissionable impurities, such as U-235, through the effective fission cross sections for the shielded and unshielded dosimeters. These I 23 effective cross sections are calculated using the SAND code and then applied in combination with the measured activities for the shielded and unshielded U-238 dosimeters to determine the level of equivalent U-235 combination.
The U-238 dosimeters in the top and middle compartments showed an equivalent U-235 contamination at the 614 ppm and 692 ppm levels, respectively. The U-238 dosimeter in the bottom compartment showed a lower level of equivalent contamination, 396 ppm. Corrections to account for the U-235' fission fraction in the measured activities for the shielded and unshielded U-238 dosimeters are based on these levels of equivalent U-235 contamination. No corrections were made to account for photofission in the U-238 because it is estimated to be on the order of a few percent and omitting that correction would yield a higher estimate of the fast neutron flux.
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The corrections for the photofission in U-238 are not applied to the "best estimate" fast neutron fluxes derived from the U-238 dosimeters because there is little or no measurement data currently available to confirm the gamma ray flux levels calculated at the dosimeter position. In addition the effect on the interpreted surveillance capsule fast neutron fluence is small in comparison to other contributors to the uncertainty. When the overall uncertainties in the complete analysis approach the level of a few percent and measurements of the gamma ray component of the radiation field become available, then this particular correction should be included and treated in more detail. .
The normalized neutron energy spectrum calculated with DOT 4 and used in SAND to determine the effective dosimeter reaction cross sections is shown in Table 6-10.
The spectrum, as shown, is normalized such that the integral over the entire energy range will be 1.0.
Table 6-11 shows the interpreted fast neutron flux for each of the analyzed dosimeters. The calculated axial variation of the fast neutron fluence at the surveillance capsule covers a range of about 8%. The variation in the top, middle, and bottom compartment fast neutron fluxes based on the dosimetry measurements ranges from about 4% to 15%, but some of this spread is caused by uncertainties in the individual dosimeter measurements. Based on the 00T4 calculations the variation in fast neutron flux is about 1.5% azimuthally within the capsule, and radially about 16%. The test samples would experience most of the variation in their fluence level from the radial and axial components of the distribution.
Using four of the fast neutron dosimeters (Iron, Nickel, U-238 and Titanium) the Beginning-of-Life to End-of-Cycle 10 average value of the fast neutron (E > 1.0 MeV) flux at the top, middle and bottom dosimeter compartments in the W-97 surveillance capsule were 6.92, 6.45 and 6.50 (10' n/cm'- s ) . The location of the peak fast neutron flux was determined to be between the middle and top -
dosimeter compartments. The value of the peak fast neutron flux normalized to the E0C 10 measurements was calculated to be 4% higher' than that of the top dosimeter compartment (i.e. 7.20 x 10' n/cm'-s). The equivalent calculated peak flux without the E0C 10 normalization, but including a calculational bias of 3%
is 7.55 x 10' n/cm'-s or within 5%.
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The relative value of the fast neutron fluence at the peak location on the reactor vessel to peak value for a particular surveillance capsule is expressed as a " lead factor." The concept of the lead factor derives from the idea that the irradiation of the specimens in the surveillance capsule would be accumulated at a higher rate than the limiting location on the reactor vessel and would thus
" lead" the vessel material as a damage indicator. The lead factor in terms of the peak fast neuti on (E > 1.0 MeV) is defined as the peak fast neutron fluence for the surveillance capsule divided by the peak fast neutron fluence at the vessel / clad interface for a specified time of irradiation. The lead factor is mainly determined by two items; the location of the capsule relative to the reactor vessel, and the source distribution for the irradiation interval. The source distribution is mostly a function of the fuel management employed during the irradiation interval.
For a given surveillance capsule each fuel cycle would typically show a different lead factor. Table 6-12 shows the lead factors calculated for the 97* and 263" surveillance capsules over three intervals of irradiation: Cycles 1-3, Cycles 1-10, and Cycle 11. The lead factors differ slightly between the 97" and 263" capsules because of differences in the positioning of the capsules relative to the reactor vessel . The lead factors differ among the irradiation intervals because of the variation in the source distribution due to changes in the fuel management.
l The Cobalt / Aluminum dosimeters in the surveillance capsule provide a measure of the thermal neutron flux / fluence in the capsule. The interpretation of the measurements is complicated by the variation in the boron level in the reactor coolant and by the difficulty of creating an adequate mesh for the DOT calculation. As a result of these two modelling problems extrapolation of the l thermal neutron fluence distribution from the surveillance capsule location and even the interpreted value at the capsule include much higher uncertainties than those estimated for the fast neutron fluences. Typically they should be used to provide an indicator of the general level of the thermal neutron fluence at the l capsule. Table 6-13 shows the results of the dosimetry analysis of the thermal neutron dosimeters. The most interesting result is the difference in the thermal neutron fluence for the bottom dosimeter versus the top and middle dosimeters.
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The equivalent U-235 fission contamination in the U-238 dosimeters show a similar pattern. The primary cause of this result is believed to be a 0.95 centimeter thick bracket of inconel that surrounds the bottom set of dosimeters and does not exist around the top or middle set of dosimeters. The bracket is part of the fixture that attaches the surveillance capsule assembly to the reactor vessel.
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Table 6-1. Neutron Flux Dosimeters Threshold Monitor Material Reaction Enerav (MeV) Hal f-Li fe 1 Cobalt /A1 Co68(n,y)Co' Thermal 5.27 years (Cadmium Shielded) 2 Uranium
- U2 '8(n, f)Cs' 1.2 30.2 years 3 Titanium Ti'8(n,p)Scd " 8.0 83.8 days 4 Iron Fe'd(n,p)Mn6 ' 4.0 312.5 days 5 Cobalt /Al Co( n , y) Co* Thermal 5.27 years 6 Uranium
- U23s(n,f)Cs 1.2 30.2 years (Cadmium Shielded) 7 Nickel Ni'8(n p)Co 68 3.5 70.8 days (Cadmium Shielded) 8 Copper Cu" (n,a)Co* 7.0 5.27 years (Cadmium Shielded) 9 Sulfur S 2(n,p)P 32 2.9 14.3 days 238
- U depleted in U23s to 500 ppm as estimated by the supplier.
Table 6-2. Peak Fast Neutron (E > 1.0 MeV) Flux and Fluence Summary Beainnino of Life to End of Cycle 10 Fast Neutron Fast Neutron Flux Fluence 10' n/cm2 -s (10'8 n/cm')
location BOL to EOC 10 BOL to E0C 10 Surveillance Capsule 7.56 2.64 (Calculated)
Vessel / Clad Interface 5.63 1.96 1/4 T Thru Vessel 3.08 1.07 1/2 T Thru Vessel 1.44 0.51 3/4 T Thru Vessel 0.64 0.22 SW#sefanLuv ,
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I Table 6-3. Peak Neutron (E > 0.1 MeV) Flux and Fluence I Summary Becinnina of Life to E.nd of Cycle 10 l Peak Neutron Peak Neutron Flux Fluence E > 0.1 MeV E > 0.1 MeV 10" n/cm'-s (10 n/cm2)
BOL to EOC 10 BOL to E0C 10 location (3.49 x 10 8EFPS) (3.49 x 10e EFPS)
Surveillance Capsule 1.32 4.61 (Calculated)
Vessel / Clad Interface 1.13 3.95 1/4 T Thru Vessel 0.93 3.26 1/2 T Thru Vessel 0.64 2.23 3/4 T Thru Vessel 0.37 1.28 Table 6-4. Peak Thermal Neutron Flux and Fluence Summary Beainnina of Life to End of Cycle 10 Peak Thermal Peak Thermal Neutron Flux
- Neutron Fluence *
(10" n/cm2 -si (10 n/cm')
BOL to E0C 10 BOL to EOC 10 8
location (3.49 x 10 8EFPS) 13.49 x 10 EFPS)
Surveillance Capsule 1.43 5.00 (Calculated)
Vessel / Clad Interface 0.53 1.83 1/4 T Thru Vessel 0.011 0.04 1/2 T Thru Vessel 0.001 0.004 3/4 T Thru Vessel 0.0001 0.002
- Note the model does not include the component of the thermal neutron flux due to sources such as neutrons thermalized outside the reactor vessel.
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" .g Table 6-5. Peak Iron DPA Summary: Bcainnina of-Life to End of Cycle 10-Peak' Iron DPA 4 per EFPS Peak Iron DPA (10 ' doa/s) - ( 10~2 doa) l BOL to EOC 10 BOL to EOC 10.
Location (3.49 x 10 8EFPS) (3.49 ' x '10' EFPS)
Surveillance Capsule 1.10 3.83 m (Calculated) 'l Vessel / Clad Interface 0.85 2.95' 1/4 T Thru Vessel 0.53 1.86 1/2 T Thru Vessel 0.31- 1.07 3/4 T Thru Vessel 0.16- 0.56 Table 6-6. Normalized Flux and DPA Radial Distribution Through the ,
Reactor Vessel Beainnina of Life to'End of Cycle 10 i Normalized to Value at Vessel / Clad Interface Fast Neutron E > 0.1 MeV Thermal .
Location Flux ' Neutron Flux , Neutron Flux Iron'DPA Vessel / Clad Interface 1.0 1.0 1.0 1.0' .;
1/4 T Thru Vessel 0.547 0.825 0.021 0.6'29 0.258 0.564 0.002' 1/2 T Thru Vessel 0.362 3/4 T Thru Vessel 0.114- 0.325 0.0002 0.189 ;
Table 6-7. Extrapolated Peak Fast, E > 0.1, and Thermal . Neutron :
Plus DPA at the Vessel / Clad Interface E > 0,1 MeV Thermal ,
Irradiation Fast Neutron Neutron- Neutron Interval 8
Fluence Fluence Fluence . . Iron DPA (10 EFPS) (10 n/cm') (10 n/cm2) (10 n/cm') (10-2 doa)" .
BOL to 3.49- 1.96 3.95 1.83. -2.95 '
, E0C 10 B0C 11 to 6.60 1.50' 2.96 1.30l : 2.' 24 '
32 EFPY. 1 32 EFPY. 10.09 3.46 6.91 3.13~ 5.19 6-'14 BW!!KWIMKm
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Table 6-8. Calvert Cliffs Unit 1 Surveillance Capsule 97* Measured Activities CD Target Measured Specific Activity Dosimeter Shield Nuclide Product Nuclide (uCi/a-Taraet)
CCl-97*,1 2G Ti (top) No Ti-46 Sc-46 3.913 x 10 2 CCI-97 ,4 2G Ti (middle) No Ti-46 Sc-46 3.876 x 10 2 C01-97*,7 2G Ti (bottom) No Ti-46 Sc-46 3.886 x 10 2 C01-97",1 7G Sh Cu Yes Cu-63 Co-60 No Sample C01-97*,4 7G Sh Cu Yes Cu-63 Co-60 2.518 x 10' CCl-97",7 7G Sh Cu Yes Cu-63 Co-60 No Sample CCl-97",1 6G Sh Ni Yes Ni-58 Co-58 1.973 x 10 8 CCl-97",4 6G Sh Ni Yes Ni-58 Co-58 1.70a x 10' CCl-97",7 6G Sh Ni Yes Ni-58 Co-58 1.886 x 10 CCl-97",1 3G Fe No Fe-54 Mn-54 1.293 x 10 CC1-97",4 3G Fe No Fe-54 Mn-54 1.123 x 10 CCl-97*,7 3G Fe No Fe-54 Mn-54 1.174 x 10' CCl-97*,1 OG Sh Co/Al Yes Co-59 Co-60 7.595 x 10' CCl-97",4 OG Sh Co/Al Yes Co-59 Co-60 7.106 x 10' CCl-97",5 OG Sh Co/Al Yes Co-59 Co-60 6.887 x 10' CCl-97",1 4G Co/Al No Co-59 Co-60 5.565 x 10 5 CCl-97 ,4 4G Co/Al No Co-59 Co-60 5.548 x 10 5 CCl-97*,5 4G Co/Al No Co-59 Co-60 3.734 x 10 5 CCl-97",1 IG U-238 No U-238 Cs-137 5.434 x 10' C01-97*,1 SG Sh U-238 Yes U-238 Cs-137 2.596 x 10' C01-97",4 IG U-238 No U-238 Cs-137 5.4'44 x 10' CCl-97*,4 5G Sh U-238 Yes U-238 Cs-137 2.450 x 10' CCl-97",7 IG U-238 No U-238 Cs-137 3.955 x 10' CCl-97",7 SG Sh U-238 Yes U-238 Cs-137 2.294 x 10' BW!!nNinom
Table 6-9. Calvert Cliffs Unit 1 Core Power History
% Full Power (2700 MwT)
Year Jan. Feb. Mar. Aor. May June July Auo. Sept. Oct. Nov. Req.
1974 1975 16.9 30.5 43.1 41.1 57.4 68.5 80.9 41.0 70.0 92.4 85.6 90.6 1976 88.3 87.2 92.2 52.0 93.3 91.7 92.1 92.9 86.2 90.5 79.2 61.4 1977 0.0 0.0 0.0 67.5 74.4 86.6 92.3 92.7 97.1 85.6 95.8 87.0 1978 56.3 0.0 0.0 58 1 62.5 89.8 87.5 90.0 90.8 2 90.9 15.5 48.5 1979 33.7 94.1 95.4 66.6 0.0 0.0 1112 87.8 95.0 90.0 63.1 57.3 1980 47.1 41.4 89.4 84.9 82.0 94.5 96.0 96.0 96.2 46.3 0.0 0.0 1981 54.1 98.4 96.4 78.4 89.5 81.4 43.2 92.0 94.6 72.5 88.2 99.2 1982 98.5 99.4 98.8 5221 0.0 0.0 2221 75.3 65.8 99.0 96.5 92.6 1983 93.5 88.6 97.4 80.4 97.8 91.0 98.5 91.1 89.2 0.0 0.0 53.3 1984 94.7 95.8 80.1 99.7 17.7 98.3 99.3 88.4 99.5 96.5 78.4 50.2 1985 88.8 96.8 99.2 15.1 0.0 0.0 0.0 12.1 91.5 76.5 98.1 98.0 1986 91.9 98.3 76.0 97.7 97.1 95.3 90.8 96.3 99.1 69.6 0.0 0.0 1987 52* 95* 90* 0.0* 11* 99* 34* 85* 97* 100* 63* 100*
1988 95* 94* 97* 23* 0.0 0.0 71.7 91.4 98.5 84.8 60.7 99.1 1989 74.7 96.2 7.6 26.5 15.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 1990 0.0 0.0 0.0 13.8 0.0 0.0 0.0 0.0 0.0 73.9 98.8 33.2 1991 90.3 37.5 98.8 93.8 54.4 0.0 42.2 98.9 98.8 89.7 99.2 99.0 1992 99.3 99.0 97** N/A N/A N/A N/A N/A N/A h/A N/A N/A
- Cycle 9 estimated average percent full power from hourly plant computer logs.
- March 1-19, 1992 Note: Underlined cells indicate the beginning or end of each fuel cycle.
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Table 6-10. Normalized Neutron Enerav Spectrum at the Surveillance Caosule 97" Top Energy Energy of Group Normalized Spectrum Normalized Spectrum Group (MeV) (MeV) (by Grouo. Not Per MeV) 1 14.92 9.79 x 10 O 00026 2 12.21 4.08 x 10 O.00090 3 10.00 1.20 x 10' O.00217 4 8.187 2.80 x 10' O.00512 5 6.360 6.11 x 10-8 0.00852 6 4.966 8.76 x 10 O.00788 7 4.066 1.17 x 10 2 0.01233 8 3.012 2.45 x 10-2 0.01337 9 2.466 3.65 x 10'* 0.00423 10 2.350 3.69 x 10 O.01928-11 1.827 6.02 x 10 2 0.04327 12 1.108 1.02 x 10 O.05686 13 0.5502 1.95 x 10 O.08583 14 0.111 9.53 x 10 O.10263 15 3.355 x 10' 1.65 x 10+' O.04581 16 5.83 x 10 l.02 x 10+2 0.04909 17 1.01 x 10 5.17 x 10+' O.03735 18 2.90 x 10 1.51 x 10+ 0.02761 19 1.07 -.x 10~6 4.53 x 10+' O.03452 i i
20 3.06 x 10 1.11 x 10+' O.02815 21 5.32 x 10 4 2.32 x 10+6 0.02727 4
22 4.14 x 10 1.06 x 10+' O.38752 135#atf5 h v
Table 6-11. Surveillance Capsule 97" Interpreted Fast Neutron Fluxes Reaction Effective Interpreted Saturation Cross Section Fast Neutron Dosimeter Factor (10'#4 cm2 ) Flux (n/cm2 3)
CCl-97",1 2G Ti 0.677 0.0243 6.73 x 10' CCl-97*,4 2G Ti 0.650 0.0243 6.94 x 10' CCl-97*,7 2G Ti 0.677 0.0243 6.68 x 10' CCl-97",1 7G Sh Cu N/A N/A No Sample CCl-97*,4 7G Sh Cu N/A N/A -
N/A CCl-97*,7 7G Sh Cu N/A N/A No Sample .
CCl-97*,1 6G Sh Ni 0.695 0.155 6.53 x 10' C01-97",4 6G Sh Ni 0.664 0.155 5.90 x 10' CCl-97*,7 6G Sh Ni 0.695 0.155 6.25 x 10' '
CCl-97*,1 3G Fe 0.485 0.121 7.31 x 10' CCl-97*,4 3G Fe 0.485 0.121 6.35 x 10' CCl-97",7 3G Fe 0.485 0.121 6.63 x 10' CCl-97",1 1G U-238 0.195 0.428 7.22 x 10' C01-97*,1 SG Sh U-238 0.195 0.428 7.14 x 10' CCl-97",4 1G U-238 0.197 0.428 6.70 x 10' CCl-97",4 SG Sh U-238 0.197 0.428 6.62 x 10' CCl-97 ,7 IG U-238 0.195 0.428 6.52 x 10' CCI-97 ,7 SG Sh U-238 0.195 0.428 6.44 x 10' Table 6-12. Surveillance Capsule Lead Factors for i Fast Neutron (E > 1.0 MeV) Fluence ;
Irradiation Interval Surveillance Capsule 97* Surveillance Capsule 263*
i Cycle-1-3 1.36 1.39 Cycle 1-10 1.34 1.37 Cycle 11 1.28 1.30 l 6-18 BW!!&%?5%r
Table 6-13. Surveillance Caosule 97* Interoreted Thermal Neutron Flux _es Reaction Effective Interpreted Thermal Saturation Cross Section Neutron Flux Dosimeter Factor (102' cm2 ) (n/cm2 -s)
CCI-97*,1 OG Sh Co/Al 0.507 21.6 1.59 x 10"*
CCl-97*,4 OG Sh Co/Al 0.512 21.6 1.59 x 10"*
CCl-97*,5 OG Sh Co/Al 0.507 21.6 1.00 x 10"*
CCl-97",1 4G Co/Al 0.507 21.6 1.59 x 10" CCl-97 ,4 4G Co/Al 0.512 21.6 1.59 x 10" C01-97*,5 4G Co/Al 0.507 21.6 1.00 x 10"
- Thermal neutrons as defined by the cadmium cutoff energy.
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Fioure 6-2. Surveillance Cansule 97*
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5 0.7 '
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e 0.5 -
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-200 -100 0.0 100 200 Axial Distance from Core Midplane (cm)
Bottom of active core Top of active core i
6-23 B W it?vsW W e m
Figure 6-5. Cycle 1 to 10 Fast Neutron Fluence Distribution at the Vessel / Clad Interface 1.0 -
7m .
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-100 0 100 Axial Distance from Core Midplane (cm)
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cla p core support barrel
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- Figure 6-9. : Extrapolated Fast Neutron Fluence: Distribution at the . Vessel /- 1 Clad Interface 32-Effective Full Power (2700 MwT) Years 'i 1.0" g ,
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6-28 S W.Ya n1E! L v c- ..%+ a y( ' 7y'
. _ _ . . _ , ~. . _ _ _ _
- 7. ~ DISCUSSION OF CAPSULE RESULTS q
r 7.1. 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 Calvert Cliffs Unit 1. The applicable selection criterion was based on the unirradiated properties only. A review of -the original- unirradiated 'l properties of the reactor. vessel . core beltline region materials. indicated no ,
significant deviation from 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 upper-shelf energy (USE) will not be_ below 50 ft-lb.
.t 7.2. Irradiated'Pronerty Data 7.2.1. Tensile Properties Tables 7-1 and 7-2 comptre irradiated properties- from Capsule 97* with the unirradiated tensile _ properties. At both room temperatureLand elevated J temperature, the ultimate' and yield strength changes in the base metal as. a result of irradiation and the corresponding changes -in ductility 'are within the. ,
limits observed for similar materials. There is ~ some ' strengthening, as-indicated by increases in ultimate and yield' strengths 'and decreases in ductility .;
properties. The changes observed in the base metal are such as to be-considered within 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 -
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 +
insignificant relative to the analysis of ~ the reactor vessel' materials -at this time period in the reactor vessel service life.
i l
j
R&W NUCLEAR WTECHNOLOGIEB .
I The general behavior of the tensile properties as a function of neutron irradiation is an increase in both ultimate and yield strength and a decrease in ductility as measured by both total elongation and reduction of area.
7.2.2. Imoact Properties The behavior of the Charpy V-notch impact data is most significant for the calculation of the reactor system's operating limitations. Table 7-3 compares the observed changes in irradiated Charpy impact properties with the predicted changes. Comparisons of the unirradiated and irradiated Charpy data curves are presented in Figures 7-1 through 7-4.
The 30 ft-1b transition temperature shift for the base metal in the longitudinal orientation is slightly non-conservative compared to the value predicted using Regulatory Guide 1.99, Rev. 2,18 but when the margin is added the predicted value is conservative. Also, the 30 ft-lb transition temperature shift in the transverse orientation is slightly non-conservative compared to the value predicted using Regulatory Guide 1.99, Rev. 2, but again, when the margin is added the predicted value is conservative.
The transition temperature measurement at 30 ft-lbs for the weld metal is not in good agreement with the predicted shift using Regulatory Guide 1.99, Revision 2 and the predicted value is very conservative. The large disparity between the measured and predicted weld metal data is not unexpected since little weld metal data at medium fluence values was available when the prediction method of Regulatory Guide 1.99 Revision 2 was developed.
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 40 percent.
7.3. Pressurized Thermal Shock (PTS) Evaluation The pressurized thermal shock evaluation shown in Table 7-4 demonstrates that the Calvert Cliffs Unit 1 intermediate longitudinal weld (20291/12008) will exceed the PTS screening criteria before current end-of-license date. The weld will reach the screening criteria at a fluence of 2.61 x 10 n/cm2 . Based on the
~
MI'fBGW NUCLEAR U2> TECHNOLOGIES
4 calculated fast flux at the vessel wall, an 80% load factor and the continuation of the cycle 11 fuel management, this fluence will be reached in February 2004.
7.4 Neutron Fluence Analysis These new analyses calculated an end-of-life fluence value of 3.46 x 10'8 n/cm2 (E > 1 MeV) at the reactor vessel inside surface peak location. The correspond-ing value for the vessel wall T/4 location is calculated to be 1.89 x 10 n/cm 2 (E > 1 MeV). These values represent a small increase compared to the values calculated based on the fluence values from the first surveillance capsule.
~
j'
- r? v8[il?BGW NUCLEAR TECHNOLOGIES
Table 7-1. Comparison of Calvert Cliffs Unit 1, Caosule 97* Tension Test Results Room Temo Test Elevated Temn Test
- Unirr** 1rrad Unirr** Irrad Base Metal - lonaitudinal: C-4441-1 Fluence,10 n/cm 2(E > 1 MeV) 0 2.64 0 2.64 Ultimate tensile strength, ksi 93.9 119.8 90.6 110.6 0.2% yield strength, ksi 70.6 93.3 65.7 82.6 Uniform elongation, % 9.2 11.1 9.3 8.4.
Total elongation, % 26.7 22.3 23.3 19.0 Reduction of area, % 71.4 60.6 66.6 59.6 Base Metal - Heat-Affected Zone fluence, 10'8 n/cm2(E > 1 MeV) 0 2.64 0 2.64 Ultimate tensile strength, ksi 85.1 98.3 84.6 104.5 0.2% yield strength, ksi 66.7 91.3 66.0 82.7.
Uniform elongation, % 6.8 *** 6.7 8.9 Total elongation, % 26.0 *** 22.0 16.4 Reduction of area, % 73.0 70.0 67.3 52.1 Weld Metal -- 33A277/3922 Fluence,10 n/cm 2(E > 1 MeV) 0 2.64 0 2.64 Ultimate tensile strength, ksi 88.5 109.6 84.7 107.0 0.2% yield strength, ksi 74.7 94.5 66.9 86.9 Uniform elongation, % 9.8 11.2 9.6 11.6 Total elongation, % 29.0 25.3 26.3 23.2 Reduction of area, % 72.7 63.0 70.4 64.0
- Test temperature is 550F.
- Average of the lower yield strength data in Appendix B. -
- See footnote Table 5-2.
7-4 l BWitnE?ti%r l l
Table 7-2. Summary of Calvert Cliffs Unit 1 Reactor Vessel Surveillance Capsule Tensile Test Results Ductility. %
Cap. Fluence, Test Strenoth. ksi Material I.D. 10 n/cm 2
Temp, F Ultimate M(a) Yield M I") Total Elon. M(a) Reduction of Area M(a)
Base metal --
0.00 RT 93.9 --
70.6 --
26.7 --
71.4 --
Longitudinal 550 90.6 --
65.7 --
23.3 --
66.6 --
(C-4441-1) 263* 0.62 RT 102.0 + 8.6 79.1 +12.0 25.8 - 3.4 67.5 - 5.5 550 97.6 + 7.7 71.6
- 9.0 22.7 - 2.6 60.6 - 9.0 97" 2.64 RT 119.8 +27.6 93.3 +32.2 22.3 -16.5 60.6 -15.1 550 110.6 +22.1 82.6 +25.7 19.0 -18.5 59.6 -10.5 Base metal --
0.00 RT 85.1 --
66.7 --
26.0 --
73.0 --
Heat-affected 550 84.6 --
66.0 --
22.0 --
67.3 --
y zone m (C-4441-1) 263" 0.62 RT 99.0 +16.3 79.1 +18.6 20.3 -21.9 69.0 - 5.5 550 95.5 +12.9 72.1 + 9.2 22.4 + 1.8 60.9 - 9.5 97" 2.64 RT 98.3 +15.5 91.3 +36.9 --'" ---
70.0 - 4.1 550 104.5 +23.5 82.7 +25.3 16.4 -25.5 52.1 -22.6 Weld metal --
0.00 RT 88.5 --
74.7 --
29.0 --
72.7 --
(33A277/3922) 550 84.7 --
66.9 --
26.3 --
70.4 --
263* 0.62 RT 98.2 +11.0 84.0 +12.4 27.5 - 5.2 66.3 - 8.8 550 97.0 +14.5 79.1 +18.2 22.5 -14.4 64.6 - 8.2 6 97* 2.64 RT 109.6 +23.8 94.5 +26.5 25.7 -11.4 63.0 -13.3 hR 550 107.0 +26.3 86.9 +29.9 23.2 -11.8 64.0 - 9.1 SE N!
"' Change relative to unirradiated.
ll 'N See footnote Table 5-2.
~~
.. ~
Table 7-3.
Observed Vs. Predicted Ch'anges for Capsule2 97* Irradiated Charpy Impact Properties 2.64 x 10' n/cm (E > 1 MeV)
Difference Predicted Per R.G. 1.99/2'd Observed Chemistry Without With Material Unirrad. Irrad. Diff. Factor Marain Marain' Marain increase-in 30 ft-lb Trans. Temo.. F Base Material (C-4441-1) longitudinal +8 +116 108 84 106 34 140 Transverse- +22 +133 111 84 106 34 140
- Heat-Affected Zone (C-4441-1) -56 - 41- 84 106 34 140 Weld Metali(33A277/3922) -50 + 43 93- 119 150 56 206 Decrease in Charoy USE. ft-lb Base Material-(C-4441-1) longitudinal 138 102 36 N.A. N.A. JN.A. 37*'
Transverse 106- 84 22 N.A. N.A. N.A. 28*'
Heat-Affected Zone (C-4441-1) 130 81 49 N.A. N.A. ~ N.A. 3 4 *'
Weld Metal _(33A277/3922) 160 106- 54'- N.A. N.A. -N.A. 78*'
E g Calculated per Regulatory. Guide 1-99,_ Revision:2,~May 1988.
- ' Bounding value per Regulatory Guide 1.99, Revision 2,- May 1988 (includes margin).
h kt N.A.:_ Not: applicable.
ll s
._____m_., nu-__ _, ##wn , a
a
..Y Table 7-4. ' Evaluation of Reactor Vessel End-of-Life Pressurized Thermal Shock Criterion - Calvert Cliffs Unit 1 Material Estimated Chemical Inside PTS Evaluation. F" ,
Material Description Composition, Surface .
Inside Reactor Vessel Heat w/o'd - EOL Fluence. Inittal Chemistry Surface Screening.
Beltline Location Number" . Type Copper Nickel- n/cm" R Tuo, ,
- Factor Margin.- RTns- . Criteria <
intertned. Shell ' C4351-2 SA533, Gr.-B 0.11 - 0.55 3.41E+19 +20 74 34- + 152. . 270 Intemed.' Shell- C4441-2 SA533, Gr. B 0.12. 0.64 3.41E+19 -30 84 34- +115 270 Intenned. Shell C4441-1 SA533, Gr..B 0.12 0.64 3.41E+19 +10 84 .34 +155- 270-Lower Shell C4420-1 . SA533, Gr.'B - 0.13 0.54 3.46E+19. +10. 90 34 +163~ 270 Lower Shell B8489-2 SA533, Gr. B 0.11' .:0.56 . 3.46E+19 -10 74 34- '
+122 270 Lower Shell B8489-1 SA533, Gr. B 0.11 0.53 3.46E+19 -20 74 34 +112- 270 ~
y Mid. Circum. Weld 33A277/ - ASA Weld /. 0.23' O.23- .3.41E+19 120 56 ' - +134 300 g 3922 Linde 0091 Intermed. Longit. Weld 20291+12008/ ASA Weld.
~
0.21 ' O.88 3.41E+19"' -50 210 56 A283 270 3833 Linde 1092-
' lower Longit. Weld 21935/_ ASA Weld / 0.21 0.69 3.46E+19" [- 56 ]'_ _179 66 +247- 270 - ~
3922 Linde-1092
Per 10CFR50, Section 50.61 Fracttre Toughness Requirements for' Protection 'Against Pressurized Thermal Shock Events."'
"Per Section 6 of.this report using neutron -transport calculation methods.
" Materials chemical compositions per response to Generic letter 92-01."
" Fluence value for longitudinal weld with maximum value.
h ' *Per -response to Generic ' Letter 92-01."
"' Generic-value per 10CFR50, Section 50.61.
E :
a:
laI .
[-
6.._.u_am_ 2 _ _ . _ .2.__L. m___l_____ - _ - - _ - im. w' N-a- u-- # sM". I-iD u WNg Ep-W-*L-grg/ g ^W.*P"- *7 e-N*1 v't v1 '-
- t"j'8-u ?'GI'4- =P 4 JM d4 wbi *- es e, ,y 1- tr * .-s_.- m
c Figure 7-1. Comparison of Unirradiated and Irradiated _ .
Charpy Impact Data-Curves'for Plate Material :
Loncitudinal Orientation; Heat No, C-4441-1 im
~
LN1mADIATED PLUENCE: 2.64a l0n/ce' n So _______ ___ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
25 -
y 0
,0.10 b
2' O.00 - -
,0 UN! MADIATED g 0.C6 ' * "' '
W 6I luf* en-O.04 kO.02 -
5 0
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v 100 -
5 iso ,
g L g __ (NIMADIATED it
_ 6 USEa36ft-Lbe FUJEM:Es 2.64ml0 n/cm' 4
U k lo0 _ _
90 -
W
>=
0 A110P _.m. ,5 60 -
010EF +
20 -
ORIENTATION LONGI M INAl.
E T NO. C4441-1 0
-100 -50 0 50 800 150 200 250 300 350 400' 450 500 550- 500-TEST TDOERATURE, F 7-8 aswaruCs.saa BWSERVICE COMPenfY
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.l Figure 7-2. _ Comparison of Unirradiated and Irradiated !
Charpy Impact Data Curves for. Plate Material i Transverse Orientation. Heat No.-C-4441-l' ioo ,
u :!
M LNIFTLADI ATED
.3 .
FL.LENCEe 2.54x IO n/cma b .i gSo ________ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
w 25 -
h
?
' ' ' ' ' I ' ' ' ' ' ' '
O
,0.10 ,
Z
~
O.00 -
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a g g g gp , - FLtfDE s 2.64a lO r./me 0.04 -
- Q.02 -
) 0 ---
' ' ' ' ' ' ' ' ' ' ' I '
zoo 100 -
t
)
- 2 b, 160 >
140 -
- 6 USEs22ft-tbe .l 120 -
LNIRRADIATED 100 -
I' BO -
6124F
- PLUENCEa 2.64sl0n/cma.
g e0 40 -
OlllP ;
ERgg SA-533, Gr. BI y 20 -
Crt!ENTAbON TRAMSVEREE C4446-1 FEAT NO,
]
g 0 1 f' i t t f t f t f f f t f 'j
-100 -so 0 50 100 150 200 250 300 350 400 -450 .500 550 600-l TEST TEMPERATURE, F i
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Figure.7-3. Comparison of Unirradiated and Irradiated Charpy Impact Data ,
Curves for Base Metal. Heat-Affected-Zone. Heat No. C-4441-1 100 y LNIRRADIATID '
+ 75 -
FLLENCE s 2. 64= 10 '* n/ce' y 50 .____
I E
2s -
t t ' f f -f 1 f t f f f f '
o
,O.to 2 -l
. LNIFv1A0! ATED gO.00 l
h.06 g0 FLUENCE: 2.64ml0'*n/ ems.
W 8122F !
g0.04 g
1 g
~
i 0.02 -
f 5
)
t f t f t i f f ' f ! f f O
i 200 l 100 -
f60 d
C 140 -
4 _.
h Ji
- 120 -
(N!RRADIATED A USE=49ftalbe 4 too -
@ ?
g= - nmcE. 2. se, i o ,, .m=
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40 -
- A41F
__j______________ _ ,, ,, ,,, ,,
ORIENTATION W 20 -
G T No.' C4441 ' ' ' ' ' ' ' ' ' ' ' ' '
0
-iso -t00 -50 0 50 100 150 200 250 300 350 400' :450 500.' ' elso TEST TD4N:RATURE.. F 7-10
^ CSUCKBGWNUCLEAR Vw'JM# TECHNOLOGIES
ez Figure _7-4. Comparison _of Unirradiated and Irradiated:Charpy' Imoact Data Curves for Weld Metal 33A277/3922 l, R
ioo
~ (NIMMo! ATE l I
'M 75 -
FLtENCEe 2,64mion/cm' g ,,
1 s, _
o ! ' ' ' ' ' ' ' ' ' ' '
0.lo I
[0.00
- UNIRRAotATED
- o.06 -
-897 + FLUENCEi 2.54x 10 n/cma
_h0.04 -
g ___
o.02 -
a '
a o ' ' ' ' ' ' ' ' ' '
200 100 -
i Il
- ,] ISO UNIMtAo!ATED 9
t ~
. A USEs54ft-Ibe 5
m r 120 -
If 100 -
FLUENCE s 2.64x t o '8n/e=8 so -
A B7F y oo -
): /
L _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _
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40 -
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_ _ _ _ _ f._ _ _= L _ _d ., _ _ _ _ _ _ _ _ _ _ _
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20- -
ORIENTAT!oN W.AT No. N/3922 t i 1 t i f f i f I o ! I
-t50 -t00 -50 0- 50 t00 150 200 250 300 -350 400 450 500 550 ~>
TEST TEMPERATURE, F 7-11 i B&W NUCLEAR BWSARWM COASPAAfY L_' j
- 8.
SUMMARY
OF RESULTS ,
The analysis of the reactor vessel material contained in the second surveillance capsule (Capsule 97") removed for evaluation as part of the Calvert Cliffs Nuclear Power Plant Unit No.1 Reactor Vessel Surveillance Program, led to the following conclusions:
2
- 1. The capsule received a peak fast fluence of 2.64 x 10 n/cm (E > 1.0 MeV). The predicted fast fluence for the reactor vessel T/4 location at the end of the tenth fuel cycle is 1.07 x 10 n/cm* (E > 1 MeV).
- 2. Based on the calculated fast flux at the vessel wall, an 80% load factor and the continuation of the Cycle 11 fuel management, the projected fast fluence that the Calvert Cliffs Nuclear Power Plant Unit No. I reactor pressure vessel inside surface will receive in 40 calendar year's operation is 3.46 x 10 n/c' m' (E > 1 MeV).
- 3. Neither the base metal nor the weld metal upper-shelf energies at the T/4 location, based on surveillance capsule results, are predicted to -
decrease below 50 ft-lbs prior to 32 EFPY.
- 4. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change was conservative for the base metal after the margin was added. The prediction for the weld metal was overly conservative.
- 5. 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 upper-shelf properties due to irradiation are in good agreement with the base metal and conservative for the weld metal.
- 6. The analysis of the neutron dosimeters demonstrated that the analytical techniques used to predict the neutron flux and fluence were accurate.
~
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- 9. SURVEILLANCE CAPSULE REMOVAL SCHEDULE Based on the post-irradiation test results of Capsule 97 and the recommended withdrawal schedule of Table 1 of E185" the following schedule is recommended for the examination of the remaining capsules in the Calvert Cliffs Nuclear Power Plant Unit No. 1 RVSP:
Evaluation Schedule i
Location of Lead Removal Time Expected Capsule C a p s ul e s Factor" E0C/ Year 2
Fluence (n/cm )ici 83" 1.28 17/2007 3.46 x 10
104" 0.96 16/2005'd' (3.29 x 10)
277" 1.28 IB/2005'd' (4.43 x 10)
284* 0.96 16/2005'd' (3.29 x 10'8) i Reference reactor vessel irradiation locations, figure 3-1.
"The factor by which the capsule fluence leads the vessels maximum inner wall fluence.
'*' Estimated fluence values based on current fuel cycle designs.
'd' Spare capsule to be irradiated and available for an intermediate evaluation, if data needed, to support licensing requirements or provide data for license renewal. Recommended time for re-evaluation of withdrawal schedule is indicated. Capsule withdrawal at 32 EFPY will have estimated fluence as defined in brackets ().
B W!!EEEsiPu%mr
.\
- 10. CERTIFICATION The specimens were tested, and the data obtained from Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Power Plant Unit No.1, reactor vessel surveil-lance Capsule 97* were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10CFR50, Appendixes G and H.
7 qfB
/ x /C
/Y A t!kE. Ntbite}lf/3 avl! tbw'e', Jr. , P . EF ' Date Project Technical Manager This report has been reviewed for technical content and accuracy.
7/20 A4 M. J. B'evan (Material Analysis) 6 //c/c Date M&SA Unit Verification of independent review.
h6 YbWJ b N$ (kW UK. E. Moore, hanagerM Q lb Q3 Date M&SA Unit This report is approved for release.
f$ksnw T. L. Baldwin, P.E.
GkC/93 Date Program Manager j BW!!nE?ti%r
l I
APPENDIX A Reactor Vessel Surveillance Program Background Data and Information l
A-;
B W!!EEVa % ur
- 1. Material Selection Data The data used to select the materials for the specimens in the surveillance program, an accordance with E185-66, are shown in Table A-1. The locations of these materials within the reactor vessel are shown in Figure A-1.
- 2. Definition of Beltline Reaion The beltline region of Calvert Cliffs Unit I was defined in accordance with the definition given in ASTM E185-73.
- 3. Caosule Identification The capsules used in the Calvert Cliffs Unit I surveillance program are identified below by identification, location, and original target fluence.'
Capsule Capsule Approximate Target Removal Locat i on'*) Refueling Fluence, n/cm 2 1 97 5 3.6 x 10'8 2 104 14 8.8 x 10'8 3 284* 23 1.5 x 10'8 4 263* 30 2.2 x 10'8 5 277" 35 2.5 x 10'8 6 83" 40 3.1 x 10'8
- 4. Specimens Per Surveillance Capsule The type and quantity of each material contained in each surveillance capsule is shown in Table A-2.
- ~'
BWitnWitM;w
Table A-1. Unirradiated Impact Properties and Residual Element Content Data 2
of Beltline Reoion Materials - Calvert Cliffs Unit NO, 1e Charpy impact Data, Lonattudinal Material 30 50 35 Fabricator ft-lb, MLE, USE, Chemistry. wt%
Material ident., Beltline Drop wt ft-lb, RTuoy Region Location ft-lb F Cu N1. P 5 Code Heat No. Material Type T uoy, F F F F 07206-1 C4351-2 5A533, Gr. B Intermed. Shell -10 --- --- ---
90 +20 0.11 0.55 0.011 0.014 07206-2 C4441-2 SA533, Gr. B Intermed. Shell -30 --- --- ---
81 -30 0.12 0.64 0.0!! 0.014 D1206-3 C4441-1 SA533, Gr. B Intermed. Shell +10 --- --- --
112 +10 0.12 0.64 0.011 0.016 07207-1 C4420-1 SA533, Gr. B Lower Shell 0 --- --- ---
77 +10 0.13 0.54 0.010 0.016 D1207-2 B8489-2 SA533, Gr. B Lower Shell -10 --- --- ---
90 -10 0.11 0.56 0.009 0.014 07207-3 B8489-1 5A533 Gr. B Lower Shell -20 --- --- ---
81 -20 0.11 0.53 0.008 0.014 9-203 33A277/ ASA Weld / Middle Circum. -80 --- --- ---
158 -80 0.23 0.23 0.014 0.011
?
w 3922* Linde 0091 2-203-A,-8,-C 20291/12008 ASA Weld / Intermed. Lengit. -60 --- --- ---
110- -50 0.21 0.88 0.010 0.008 3833* Linde 1092 3-203-A,-B,-C 21935/ ASA Weld / Lower Longit. --- --- --- --- ---
-56 0.21 0.69 0.015 0.011 3869* Linde 1092
- Weld wire heat and flux lot identifiers. .
Si a
h an Sc' op-L*D
Table A-2. Tvoe and Quantity of SDecimens Contained in Each Irradiation Capsule Assembly Base Metal Weld Metal Correl.
Target (Heat No C-4441-11 (33A277/3922) HAZ (Heat Materi al'"
Capsule Fl uence' Impact No. C-4441-1 Total Specimens location (n/cm') L T Tensile Impact Tensile Impact Tensile Impact Impact Tensile 97 3.6 x 10 12 12 3 12 3 12 3 --
48 9 104" 8.8 x 10'8 12 --
3 12 3 12 3 12 48 9 284* 1.5 x 10 12 12 3 12 3 12 3 --
48 9 263" 2.2 x 10 12 --
3 12 3 12 3 12 48 9 277 2.5 x 10 12 12 3 12 3 12 3 --
48 9
{ 83* 3.1 x 10 R R 3 12 3 12 3 --
48 _9 TOTALS 72 48 18 72 18 72 18 24 288 54
Adjusted to nearest value attainable during scheduled refueling.
'" Reference material correlation monitors.
'*' Weld wire / weld flux lot combination.
Q L = Longitudinal R
T = Transverse o!
E 80 Rb l,'
Figure A-1. Location and Identification of Materials Used.in 8the-Fabrication.
of Calvert Cliffs Unit 1 Reactor Pressure Vessel REACTOR VESSEL BELTLINE MATERIALS NOT SHOWN INTERMEDIATE SHELL ' r n ennmnne,n
. WELD SEAM No. 2-203C p LOWER SHELL -
WELD SEAM No. 3-203B WELD SEAM No. 3-203C PLATE No. D-7207-pg, 'gcNg y p
/ %
42"10 ). f 30"1D OUTLET 1 INLET NOZZLE L NOZZLE UPPER TO INTERMEDIA SHELL GIRTH SEAM :- INTERMEDIATE SHELL .
WELD No. 8-203 "f/LONGITUDINALWELD
- SEAM No. 2-203B INTERMEDIATE SHELL -- a lNTERMEDIATE SHELL . i p
PLATE No. D-7206-1 p PLATE No. D-7206-3 INTERMEDIATE SHELL -h j INTERMEDIATE TO. LOWER LONGITUDINAL WELD , ; SHELL GIRTH SEAM .
SEAM No. 2-203A - -
WELD No. 9-203 INTERMEDIATE SHELL $
PLATE No. D-7206-2 LOWER SHELL 4
- * ~ ~
LOWER SHELL PLATE No. D-7207-1 LOWER SHELL ~_ ~-
LONGITUDINAL WELD SEAM No. 3--203A
]g
-, , _ . 1 f:
$$i!!!!!!!!!!!!!!!!!!!!!!b REACTOR VESSEL l
Oh5 R COM ANY
b APPENDIX B Pre-Irradiation Tensile Data B-1 BWs* FEE?afaar
Table B-1. Tensile Properties of Unirradiated Shell Plate Material . Heat No. C-4441-1. Lonaitudinal Test Reduction Elongation, Specimen Temp., Strenath. ksi of Area, Total /Unif.
No. F Yield
- Ultimate % %
IJC RT 71.0/69.2 92.7 71.4 28.0/9.62 IJ3 RT 72.3/71.0 94.5 71.4 26.0/8.72 IJ5 RT 74.1/71.6 94.4 71.4 26.0/9.25 IK3 250 64.9/62.5 84.6 69.4 23.0/8.53 IJJ 200 63.7/61.2 88.9 71.4 26.0/8.47 IKl 250 64.9/64.3 85.4 71.4 25.0/9.03 IJU 250 67.4/65.6 87.4 71.4 24.0/9.7 IKP 550 66.1 90.6 67.3 23.0/9.07 IK6 550 65.5 90.8 67.3 23.0/9.64 1KE 550 65.5 90.3 65.3 24.0/9.09
- Lower and upper yield strengths.
Table B-2. Tensile Properties of Unirradiated Shell Plate Material. Heat No. C-4441-1, Transverse Test Reduction Elongation, Specimen Temp., Strenath. ksi of Area, Total /Unif.
No. F Yield
- Ultimate % %
2J1 RT 72.3/69.2 92.1 67.3 27.0/10.0 2J3 RT 70.4/68.6 92.1 69.4 26.0/9.98 2J5 RT 71.0/68.6 92.4 69.4 27.0/9.65 2JB 250 66.5/64.3 86.0 65.3 23.0/9.7 2JP 250 65.5 87.0 59.1 21.0/8.52 2J7 250 66.1/63.7 85.1 69.4 21.0/7.77 2JU 550 66.1 91.6 65.3 27.0/8.44 2K1 550 66.7 91.8 61.2 25.0/10.94 2K3 550 66.1 90.5 65.3 23.0/9.07
- Lower and upper yield strengths.
$ TEC L$ gly
i Table B-3. . Tensile Propert'ies of Unirradiated Shell' ,j Plate HA2 Material. Heat No. C-4441-1 q Test. .
Reduction Elongation, l
Specimen Temp., Strenath, ksi of Area,. Total /Unif. !
No. F- Yield
- Ultimate % %
4JE RT- 68.6/66.7 87.6 73.01 26.0/8.37 l 4J6 RT 69.8/67.4 84.6- /6.87 ;
4JA RT .68.6/66.1 83.3 --
. /5.0 ,
4K3 250 69.6/68.4 81.9 74.0 21.0/7.25' ;
80.6
- 4JK 250 64.3/62.5 7515 /6.25 4JM 250 62.4/60.0 79.4 74.0 *-f6.54 1 4JL 250 63.7/62.5 80.5 75.5 *'/6.43 4JJ 250 63.7/62.5 80.4 73.5 ** /6.25 :;
4J8 250 61.2/60.6 75.0 74.0 /5.37 .
4KT 550 65.4 84.6 64.0 *'/6.62. .
4K5 550 66.6 84.6 72.0 . 22.0/7.1 1 66.0 66.0
- 4KM 550 84.6 ./6.28 ;
- Percent reduction in area (R.A.) and/or percent total elongation (TE) were not' .;
obtainable because the specimen broke outside of the gage length. j l
Table B-4. Tensile Properties of Unirradiated Weld' Metal 33A277/3922 -- !
Test Reduction- Elongation, !
Specimen Temp., 'Strenath, ksi -of Area, Total /Unif. 3 No. F Yield
- Ultimate % % ;
3JA RT 77.1/74.1 88.5 71.4- 27.0/10.0 3JE RT 79.6/74.1 87.8 73.4 30.0/10.05 .;
3J3 RT 82.0/75.9 89.3 .73.4 30.0/9.45 i 3JM 250 73.5/69.8 83.3 75.5 26.0/8.5 4 3K6 250 72.3/68.6 81.6 71.4 25.0/8.47- 3 3K1 250 71.0/68.0 81.7 71.4 '25.0/7.84 3L2 550 65.5 84.0 71.4 27.0/10.41. :
3KD 550 67.4 84.8 71.4 28.0/9.73 1 3KP 550 68.0 85.8 69.4 25.0/8.5 ?
3XB 550 66.7 84.0 69.4 25.0/9.69 q
- Lower and upper yield strengths. 'l 1
i 1
u Bl%B&WNUCLEAR ecumawares 1
3 APPENDIX C Pre-Irradiation Charpy Impact Data
~'~
B Ws*fa W af45 -
l i
o Table C-1. Charpy Impact Data From'Unirradiated Base Material,
- Lonoitudinal Orientation. Heat No. C-4441-1 '
Absorbed Lateral Shear-
. Specimen. Test' Temp, Energy, 'Expagsion, -Fracture, ID F. 'ft-lb 10 in. %-
IIT -80 6.8 5 0 13L -40 12.0 13 10 15M -40 15.4- 16 10 12J 0 22.4 22 20 12C 0 28.4- 28 20 11U- 40 50.5 46 35:
15Y ~40 75.6' 60 40-IIC 70- 59.5 50 50.
IIP 70 70.5 56 55 llE 70 89.0 ~68 60-12E' 80 67.3 52 55-13J 80- 96.3 70 70 144 120 -110.0 78 80 15K 120 118.5 84 80 12D 160 130.8 90. 100 162 160 142.5 97 100 152 210 137.5 90 1100 13Y :210 139.0 91 100
. Table C-2. Charpy Impact Data From Unirradiated Base Material, Transverse Orientation. Heat No.'C-4441-1 Absorbed Lateral 3 hear Specimen. Test Temp, Energy, 'Expagsion, Fracture, ID F ft-lb- 10 .i n . %
22E -80 3.3- 6 ~0 24A' -40 6.8 11' 10~
21L -40 11.0 13 10 24Y -0 19.0 25 20.
22B 0 24.0 24'15-21E 40 37.3 42 30 22J 40 39.0 36 30' 22K 70 58.3 51- 40 233 '70 60.0 52 40 22T 70 68.5 57 50 23P 80 63.0 51 50
-214 80 70.0- 59 50 21C 120 83.5- 68. 75 23D 120 88.4 72 75 22B 160 97.0 79 100 24B 160 102.5 74 90 21P 210 107.2 80 100 257 210' 118.3- '84 100 SW#4L"c*o." r
. _ . _ --_J:_.__ __ l '
Table C-3. Charpy Impact Data from Unirradiated Base Metal, -
Heat-Affected Zone. Heat No. C-4441-1 Absorbed Lateral Shear Specimen Test Temp, Energy, Expagsion, Fracture, ID F ft-lb 10 in. %
467 -150 6.5 9 0 43D -120 48.2 31 20 .
424 - 80 23.0 23 20 44P - 80 50.0 40 50 465 - 40 36.5 46 35 450 - 40 84.0 56 40 441 0 81.0 64 75 434 0 92.5 60 50 422 40 120.0 82 95 45B 40 130.0 84 90 l 443 60 64.0 49 55 442 60 66.7 54 60 44B 60 135.0 90 98 l 44E 80 91.0 69 -
65 1
417 80 145.5 85 100 44T 120 131.5 86 100 41M 120 133.5 86 100 42M 160 107.5 79 100 45P 160 125.3 86 100 Table C-4. Charoy Impact Data from Unirradiated Weld Metal. 33A277/3922 Absorbed lateral Shear Specimen Test Temp, Energy, Expagsion, Fracture, ID F ft-lb 10 in. %
334 -120 4.2 8 0 346 - 80 10.0 12 10 36C - 80 15.3 15 10 35A - 40 38.6 35 30 34Y - 40 44.5 38 35 33P - 20 74.0 60 40 345 - 20 91.8 67 50 344 - 20 100.7 71 65 356 0 62.0 52 60 332 0 85.0 63 50 361 40 132.8 90 85 31Y 40 133.0 100 90 36A 80 140.0 95 100 32L 80 142.0 97 100 34U 120 153.9 98 100 31J 120 156.0 100 100 34T 160 158.4 101 100 l 322 160 160.6 97 100
'~
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Figure C-1. Charpy Impact Data From Unirradiated Base Metal (Plate). Lonaitudinal Orientation. Heat No. C-4441-1 ifC
- e
- 75 -
W p 30 _______ '______________________
E .
6 25 -
, a l f f t t t f t f f f } i 0 -
,O.10 ,
Z
[0.00 O.06 - .
w ..
0.04 -
s .
0.02 -
2 l f t f f i t l t f t f t
_J O
200 w, CALVERT CLIPT'S. LNIT I CAPSLLE OMIT IED . T POT TCV (35 >LE) 'M T
CV (50 FT-LB) +42F T #
] CV (30 FT-LB) s
[ C y .ugg g gyg) 4 3an- ne ,
~
O
~
{ 120 i
e 4 400 -
e Z 00 -
W e H e k 80 - .
t 40 -
MATERIAL SA533- GF- Of 20 -
m NOT my m, C4441-f .
1 I I f f I t i i f i I 1
-100 -50 0 50 100 150 200 250 300 350 400 450 300 550 600 TEST TD'PERATURE, F C-4 PS B&WNUCLEAR ,
13 WSERVICE COMPANY
Figure C-2. Charpy Impact Data From Unirradiated Base Metal (Plate). Transverse Orientation. Heat No. C-4441-1 im e e M
. 73 -
t, 30 _______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
E
@25 e
i I I I I I I I I I I I I I O i
,0.10 2
- e g'O.OO .
0.06 -
w .
0.04 O.02 -
- 2 f I I I I J I I I I I I I I O
l l 200 PLANTS CALVE'MT CLIFTS, UNIT l CAPSLLE, BASEL!T T M2T
ggo TCV (35 MI) *37 T
CV (50 FT-LB) *W T 'M j ] CV (30 FT-LD) i c y_ygg gayo) m t-n.
h _
RT M3T g
{ 120 .
R .
e ,. -
8 Wm W
W e foO 40 -
MATERIAL SA-533, Sr. BI ORIENTATION TR 6 20 -
FLUENCE NOT w.AT no. c444'-'
' I I I I I I I I I I I I 0 -
-100 -50 0 50 100 150 200 250 300 350 400 450 F3.0 550 000 TEST TEMPERATURE. F i
C-5 m
eswnuctran Q WSERVICE COMPANY 2
Figure C-3. Charpy Impact Data From Unirradiated Heat-Affected-Zone Base Metal. Heat No. C-4441-1 100 :
.M 75 - e s
gso - _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
L E
Q 25 .
a O
O.10 0.00 -
8.06 gt 0 -
g .
0.04 -
- Q
. =
k O.02 -
5 7 t t f f t t i J O f t f f f 200 PLANT, CALVERT CLIFFS, t. NIT I m y. BASEL!PC 100 - TW W '
Tcv(55 u ) -stF
-27 lm - Tcv 150 n-Ln)
T -5&
3 cv (30 n-us) ,
f 140 -
CV-ust (Ave) ssoft-th. . j a7 -
g- a , .~ .
{ 120 i
4 400 -
F . ,
9 go - o .1 W
H k -
k- so - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
40 -
MATERIAL' SA-533. Gr. Bl CRIENTATICN HAZ 20 .
m NODE EAT NO. 04443*O t t t ! ! f t f I I ' I I O
-150 -100 -so o 50 100 850 200 250 300 350 400 450- 500. 550-TEST TDPERATLRE. F C-6 EDRUB&W NUCLEAR .
EPMLAfECHNOLOGIE8
Fiaure C-4. Charov Imoact Data From Unirradiated Weld Metal. 33A277/3922 soo M
. n -
k e B so _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _
e _____. ,I s ,, _
1 g
C O.10 : :!
.i.
g'O.00 -
W .
2 -
8o.os .
0.04 O.02 -
E ' ' ' ' ' ' ' '
J O
200
-PLANTe CALVERT CLIF75, UNIT I m , BASEL!tC T -W ICV (35 ME)~ *&
100 -
T "#
CV (50 FT-LB)
T cv (30 rT-La) ' - " !
iso ISO H-the j g c v-usE (AVG)
-acr l t aT eer im -
b
{ 120 3
e too - .
e 80 -
b k .o
. 40 -
MTERIAL WELD-L! TOE 0091 ORIENTATION 20 "
FUENCE NONE W.AT NO. 3 m /3922
' i t f i f f f i f f t 0
-150 -100 -50 0 50 100 150 200 250 300 350 .400 450_ E !EiO TEST TEMPERATURE. F i
.j C-7 -i ss w nuCs. san BWSERVICE COMPAM ,
- . . . . . . - . ~ , .. . .. . . . . . . _ _ ___________________1.__________________________.____ ____________ ________ _________ 2 ________________*_ ______ __,___ o
I i
l APPENDIX D Tension Test Stress-Strain Curves l
l l
i 1
1 1
i l
i D-1 BGW NtJCLEAR i BWSERVICE COMPANY l
1 Figure D-1. Tension Test Stress-Strain Curve for Base Metal Plate j Heat C-4441-1. Specimen No. 1KA. Tested at 70F i f
y Sr.ec i men: 1KA Test feno.: 70 f I 21 C1 Strength TIcId: 933r
\ - d a>
I l
- UTS
- ll 63.. -
I i
ll k.:: ,
i
- - d ~a l
.- o7 1
\
EN , :l i ! E i f 4
,E f NN t : P t T g
k I d d 0.00 0.04 0.08 0.12 0.16 0.20 0.24 0.28 0.32 Engineering Strain Figure D-2. Tension Test Stress-Strain Curve for Base Metal Plate Heat C-4441-1. Specimen No. IKM, Tested at 250F l
l No Data - See Section 5.3 1
1' l
l l
0*2 aswwucu:aa BWscavicecomany 1'
Figure D-3. Tension Test Stress-Strain Curve for Base Metal Plate Heat C-4441-1. Specimen No. 1K7. Tested at 550F g Specimens IK7 Test Temp.: 550 F f 287 Cl 2 S t r eng t h - d*
field: 82G44.
U15: It0567
\
E li \
\\ -
s4 t
d o .
1 .
- P d l NN T : P e
T.
d ma . r,,
~
, , , l i 1 1 .
d o 0.00 0.04 0.08 0.12 0.16 0.20 0.24 0.28 0.32 '
Engineering Strain Figure D-4. Tension Test Stress-Strain Curve for Base Metal Heat-Affected ,
Zone. Heat C-4441-1. Specimen NO. 4J5 Tested at 70F d Specimen: 4J5 Test Temp.: 70 Ff 21 Cl 2 Strength - d Tield: 91260. *
(175; 98326.
8~
- :: :: mw r l M
'/j 8
m.r8 ~!;
- : k
- I ;
1 :
- d -
P* ;
95 T 4 P
E c
I
.l t
Y ~,
\
l l
n" N 1
' f o
.- f: , , , , , , ,
o 0.00 0.08 0.16 0.24 0.32 0.40 0.48 0.56 0.64 Engineering Strain e 10
D-3 eswsuctuan BWscavice counur
Figure D-5. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Z0ne. Heat C-4441-1. Specimen No. 4J4. Tested at 250F y Specimen 4J4 Test Temp.: 250 Ff 121 C1
- S tr ength .
Yleid 84294. ~ 8 UTS: 90962.
8 -
l l l * -
~ .y
~. O e
-8 _
i : k i *
?
J G. :
8 ~
3 l -
su i
d w ~3 l f 4 ?
/ ,
.w R
~ '
l d '
g E 'l' 1 i f f I t t 0.00 0.04 0.08 0.!? D.16 0.20 0.24 0.28 0.32 Engineering Strain e 10
Figure D-6. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone. Heat C-4441-1. Specimen NO. 4JP, Tested at 550F 6 Specimen: 4S Test 1,.p.: 550 Ff 287 C1 2 Strength .
8
~
Yleid: 82701.
U15: 104496.
N 7 \
l
- l
\ .,
db
~. . *a E8 e .
N b l t : ;
- a. :
I I N$
h -
P t : Te 6;;,9 " E
,. P
~
.w
,i R
N i* l I i t i g 1 I 0.00 0.04 0.08 0.12 0.16 0.20 0.24 0.28 0.32 Engineering Strain D-4 aswwuctran BWscavics coueaur
figure D-7. Tension Test Stress-Strain Curve for Weld Metal 33A277/3922. S0ecimen No. 3JD. Tested at 70F g Specimen: 3JD lest fe w.: 70 F f 21 Cl
- S t r eng t h _ g
,Y l e i d: 94538. o
.. UTS: 109572. .
. do in
. =7
- a= , .
.: k A :.
- d . .
7"' :
Q G; I F r T 6? .E
?
("
N i i i , i , i g g 0.00 0.04 0.08 0.:2 0.16 0.20 0.24 0.28 0.32 Engineering Strain Figure D-8. Tension Test Stress-Strain Curve for Weld Metal 33A277/3922. Specimen No. 3JB. Tested at 250F y Specimen 3JB Test Temp.: 250 ft 121 C1
- Strength .
8
~
Yield: 87188.
U15 101212.
'N, e.
~
hU -
i' g ', .
$ i k e , .
Ma ; e
.P m dirM h [
t : T
}d w2 * [
, .w b
~
N
+ 1 1 1 l l l l d
0.00 0.04 0.09 0,12 0.16 0.20 0.24 0.28 0.32 Engineering Strain D-5 nawwuctrAn SWssavice coupaur
Figure 0-9. Tension Test Stress-Strain Curve for Weld Metal 33A277/3922. Specimen No. 3JJ. Tested at 550F 6 Specimens 3JJ Tes t Temp. : 550 FI 287 Cl U S tr eng t h Yleid: 86881.
- dm 075: 106985.
^
ll NI
~
~! .
.: : k 1 l
'" d
.E e
t
.- T.
.E
?
.N s '"
R t t o
g t t t t I g
0.00 0. 04 0.00 0.12 0.16 0.20 0.24 0.28 0.32 Engirwering Strain D-6 sawsuctraa SWscavoce coueaur
-2 f
l ... .. . .
I l
APPENDIX E References i
1 l
I l .
I
.q i
i E-1 ntusswsucuan I@ WSERVICE COMPANY l- - - - - -
- 1. Program for Irradiation Surveillance of Calvert Cliffs Unit 1 Reactor Vessel Materials, B-NLM-010, Combustion Engineering, Inc., Windsor, Connecticut, July 15, 1970.
- 2. Perrin, J. S., et al., "Calvert Cliffs Unit No.1 Nuclear Plant Reactor Pressure Vessel Surveillance Program; Capsule 263," Final Report, BMI-1280, Battelle Columbus Laboratories, December 15, 1980.
- 3. " Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Calvert Cliffs Unit 1 Reactor Vessel Materials," Combustion Engineering, CENPD-34, February 1972. '
- 4. Code cf Federal Regulation, Title 10, Part 50, Domestic Licensing of l Production and Utilization Facilities, Appendix H, Reactor Vessel Material Surveillance Program Requirements.
- 5. Code of Federal Regulation, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix G, Fracture Toughness Requirements.
- 6. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, Appendix G, Protection Against Ncnductile Failure (G-2000).
- 7. ASTM Standard E208, " Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic S teel s , " in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 8. Lundvall, A. E. Jr., to Davis, D. K., Response to NRC Letter dated May 20,
1970,
Subject:
Calvert Cliffs Nuclear Power Plant Unit No. I and 2, Docket No. 50-317 and 50-318, Reactor Vessel Material Surveillance Program (PDR File No. 50317-979; 50318-595), December 29, 1977.
- 9. Byrne, S. T., Biemiller, E. L., and Rag 1, A., " Testing and Evaluation of Calvert Cliffs Unit I and 2 Reactor Vessel Materials Irradiation Surveil-lance Program Baseline Samples," Combustion Engineering, TR-ESS-001, January 31, 1975. *
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- 10. ASTM Standard E8, " Standard Methods of Tension Testing of Metallic Materi-als," in ASTM Sthndards, American Society for Testing and Materials, Philadelphia, PA.
- 11. 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. ]
- 12. ASTM Designation E23-72, " Method for Notched Bar Impact Testing of Metallic Materials," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 13. A. L. Lowe, Jr., et al., Evaluation of Surveillance Capsule Temperatures, BAW-2040, Babcock & Wilcox, Lynchburg, Virginia, March 1989.
- 14. ASTM Designation A370-77, " Methods and Definitions for Mechanical Testing of Steel Products," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 15. ASTM Designation E23-88, " Methods for Notched Bar Impact Testing of Metallic Materials," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 16. Standardized Specimens for Certification of Charpy Impact Specimens from j National Iristitute of Standards and Technology, Office of Standard Reference Materials, Gaithersburg, Maryland.
I
- 17. ASTM Designation E185-XX (to be released), Recommended Practice for I Surveillance Tests for Nuclear Reactor Vessels, in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA (to be published).
- 18. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel Material, Reaulatorv Guide 1.99. Revision 2, May 1988. l
- 19. Rhoades, W. A., and Childs, R. L., "An Updated Version of the DOT 4 One- and 1
Two-Dimensional Neutron / Photon Transport- Code," Oak Ridge National '
Laboratory, ORNL-5851, April 1982.
- 20. Loretz, R. A. , and Wagner, S. G., "Recent Enhancements to the . ROCS */MC*
Reactor Analysis System," Proc. International Conference on the Physics of
[SW?$cNoNN)eE I- l L -.
7..
Reactors: Operation, Design and Computation, Vol. 3, P IV-153, Marseille, France, April 1990.
- 21. Jonsson, A., and Loretz, R. A., " Historical and Recent Developments, Applications and Performance of the DIT Assembly Lattice Code," Proc.
International Topical Meeting Advances in Mathematics, Computations and Reactor Physics, Pittsburgh, Pa., April 1991.
- 22. " CASK Cross Section Library," Radiation Shielding Information Center, Oak Ridge National Laboratory, DLC-23F, Oak Ridge, Tn. 37830.
~
- 23. " SAND Neutron Flux Spectra Determination by Multiple Foil Activation l
l Iterative Method," Oak Ridge National Laboratory, CCC-ll2, Oak Ridge, Tn.
37830.
- 24. "DOSDAM 81-82 Multigroup Cross Sections in SAND 11 Format for Spectral, Integral, and Damage Analysis," Oak Ridge National Laboratory, DLC-97, Oak Ridge, Tn. 37830.
- 25. Nair, P. K. and Williams, M. L., " Pressure-Temperature Limits for Calvert Cliffs Nuclear Power Plant Unit 1," SWRI Report 06-1278-001, November 1988, Southwest Research Institute, San Antonio, Texas, 78284.
- 26. American Society of Testing Materials, Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (DPA), E693-79, (Reapproved 1985).
- 27. Coor, J. L. and Pickera, C. W., " Analysis of Baltimore Gas & Electric's Calvert Cli ffs Unit 1, 97" RVSP Dosimetry," B&W NES Report No.
92:10256NL:01, October 9, 1992.
- 28. Code of Federal Regulations, Title 10, Part 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events. *
- 29. Baltimore Gas & Electric Company Response to Generic Letter 92-01, June 30, 1992.
- 30. ASTM Standard E185-66, " Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors" in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa.
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- 31. ASTM Standard ElB5-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa.
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