ML20085J733
| ML20085J733 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/31/1991 |
| From: | Aadland J, Lowe A, Nana A BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20085J719 | List: |
| References | |
| BAW-2140, NUDOCS 9110290326 | |
| Download: ML20085J733 (118) | |
Text
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BAW-2140 s August 1991 a
ANALYS13 0F CAPSULE S 4 WISCONSIN ELECTRIC POWER COMPANY I POINT BEACH NUCLEAR PLANT UNIT N0, 2
-- Peactor Vessel Material Surveillance Program --
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BAW-2140 August 1991 s
i ANALYSIS OF CAPSULE S WISCO: SIN EL ECTRIC POWER COMPANY POINT BEACH h0 CLEAR PLANT UNIT NO. 2
-- Reactor Vessel Material Surveillance Program --
by A. L. Lowe, Jr., PE J. D. Aadland A. D. Nana B&W Nuclear Service Company and S. L. Anderson -
Westinghouse Electric Corporation (Neutron Dosimetry Evaluations -
Section 6)
B&W Document No. 77-2140-00 (See Section 10 for document signatures)
! B&W NUCLEAR SERVICE COMPANY Engineering and Plant Services Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 B W H E E1 8 0; m
SUMMARY
This report describes the results of the examination of the fourth capsule (Capsule S) of the Wisconsin Electric Power Company, Point Beach Nuclear Plant Unit No. 2 reactor vessel surveillance program. The objective of the program is to monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor vessel materials by the testing and evaluation of tension and Charpy inapact specimens. The program was based on the requirements of ASTM Specification E185-66.
The capsule received an average fast fluence of 3.47 x 10 I9 n/cm' (E > 1.0 MeV) and the predicted peak fast fluence for the reactor vessei T/4 location at the end of the ixteenth cycle (14.8 EFPY) is 1.06 x 10 I9 n/cm' (E > 1 HeV). Based on the calculated fast flux at the vessel wall, an 80% load factor, and the planned fuel management, the projected fast fluence that the Point Beach Nuclear Plant Unit 2 reactor pressure vessel inside surface will receive in 40 calendar years of operation is 2.92 x 10 I9 n/cm' (E > 1 HeV) and the corresponding T/4 fluence is calculated to be 1.93 x 10 I9 n/cm' (E > 1 MeV). The peak calculated RT at T/4 vessel wall location is 262F at 32 EFPY per Regulatory Guide 1.99, NDT Rev. 2. Likewise, the T/4 vessel wall upper-shelf energy is calculated to decrease to 41 ft-lbs per Regulatory Guide 1.99, Rev. 2.
The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure. The Charpy impact data results <
for the base metal forging materials exhibited the characteristic shift to higher temperature for the 30 ft-lb transition temperature and a decrease in upper-shelf energy. The charpy impact data results for the weld metal and correlation material also 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 HDT and the decrease in upper-shelf properties due to irradiation are conservative.
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1 CONTENTS Page
- 1. INTRODUCTION .........,........... . . . . . . 1-1
- 2. BACKGROUND ......... . . . . . . . . . . . . . . . . . . . 2-1
- 3. SURVEILLANCE PROGRAM DESCRIPTION . . . . . . . . . . . . . . . . . 3-1
- 4. PRE-IRRADIATION TESTS . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Tension Tests . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2. Impact Tests . . . . . . . . . . . . . . . . . . . . . . . . 4-1
- 5. POST-lRRADIATION TESTING . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Visual Examination and Inventory . . . . . . . . . . . . . . 5-1 5.2. Thermal Monitors . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3. Tension Test Results . . . . . . . . . . . . . . . . . . . . 5-1 5.4. Charpy V-Notch Impact Test Results . . . . . . . . . . . . 5-2 5.5. Wedge Opening Loading Specimens . . . . . . . . . . . . . . . 5-3
- 6. NEUTRON DOSIMETRY EVALUATIONS . . . . . . . . . . . . . . . . . . . 61 6.1. Introduction . . . . . , . . . . . . . . . . . . . . . . . . 6-1 6.2. Heutron Transport Analysis Methods . . . . . . . . . . . . . 6-2 6.2.1. Reference Forward Calculation . . . . . . . . . . . . 6-3 6.2.2. Cycle Specific Adjoint Calculations . . . . . . . . . 6-4 6.2.3. Results of Neutron Transport Calculations . . . . . . 6-6 6.3. Neutron Dosimetry Evaluation Methodology .......... 67 6.3.1 Determination of Sensor Reaction Rates . . . . . . . 6-7 6.3.2. Least Squares Adjustment Procedure . . . . . . . . . 6-9 6.3.3. Results of Dosimetry Evaluations . . . . . . . . . . 6-13 6.4. Vessel Exposure Projections . . . . . . . . . . . . . . . . 6-13 6.4.1. Baseline Exposure at the End of Cycle 14 . . . . . . 6-14 6.4.2. Exposure Accrued During Cycles 15 and 16 . . . . . . . 6-15 6.4.3. Projection of Future Vessel Exposure . . . . . . . . 6-15 6.4.4. Exposure of Specific Beltline Materials . . . . . . . 6-16 6.5. Uncertainties . . . . . . . . . . . . . . . . . . . . . . . . 6-17
- 7. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . . . . . . . 7-1 7.1. Pre-Irradiation Property Data . . . . . . . . . . . . . . . . 7-1 7.2. Irradiated Property Data . . . . . . . . . . . . . . . . . . 7-1
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Contents (Cont'd)
Page 7.2.1. Tensile Properties . . . . . . . . . . . . . . . . . 7-1 7.2.2. Impact Properties . . . . . . . . . . . . . . . . . . 7-2 7.3. Reactor Vessel fracture Toughness . . . . . . . . . . . , , . , 7-4 7.4. Neutron Fluence Analysis ....... . . . . . . . . . . 7-6
- 8. SUMMArt 0F RESULTS .........................B1
- 9. SURVElltANCE CAPSULE REMOVAL SCHEDULE . . . . . . . . . . . . . . . 9-1
- 10. CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 APPENDIXES A. Reactor Vessel Surveillance Program Background Data and Inforntion . A-1 B. Pre-lrradiation Tensile Data ................... B-1 C. Pre-Trradiation 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 S . . . . . . . . . . . . . . . 3-2 3 2. Chemical Compositions and Heat Treatment of Surveillance Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 5-1, sensile Properties of Point Beach Unit 2 Capsule S, Base Metal and Weld Metal Irradiated to 3.47 x 10,n/cm 2(E > 1 MeV) .. . . 5-4 5-2. Charpy Impact Data from Point Beach Unit 2, Capsule 5, Base Metal Forging 122W195 val, Irradiated to 3.47 x 10 n/cm2 (E > 1 MeV) . 5-5 5-3. Charpy Impact Data from Point Beach Unit 2, Capsule S, Base Metal Forging 123V500VA), Irradiated to 3.47 x 10 n/cm2 (E > 1 MeV) . 5-5 l 5-4. Charpy Impact Data from Point Beach Unit 2, Capsule S, Base Metal Forging Heat- Af fected Zone Material, Irradiated 2
to 3.47 x 10 n/cm (E > 1 MeV) . . . . . . . . . . . . . . . . 5-6 5 5. Charpy Impact Data from Point Beach Unit 2, Capsule S, Weld Metal WF-193, Irradiated to 3.47 x 10 n/cm 2 (E > 1 MeV) . . . . . . . 5-6 5-6. Charpy Impact Data from Point Beach Unit 2, Capsule S, Correlation Monitor Material, Heat No. A-1195-1, Irradiated to 3.47 x 10 n/cm' (E > 1 MeV) . . . . . . . . . . . . . . . . 5-7 6-1. Calculated Fast Neutron Exposure Rates at the Center of the 33 Degree Surveillanco Capsules Core Midplane Evaluation ... . . 6-20 6-2. Calculated Integrated Fast Neutron Exposure at the Center of 33 Degree Surveillance Capsules Core Midplane Evaluation . .. . . 6-21 ,
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Tables (Coni'Al Table Page 6-3. Derived Exposure Rates from the Capsule S Dosimetry Evaluation . 6-22 r
5 4. Derived Neutron Exposure Rates and Integrated Exposure I
Experienced by Sur'veillance Capsule S . . . . . . . . . . . . . . 6 23 g
6-5. Calculated Spectrum Averaged Reaction Cross-Sections and Exposure Parameter Ratios at the 33.0 Degree Surveillarce Capsule locations . . . . . . . . . . . . . . . . . . . . . . . . 6-24 6-6. Comparison of FERRET Results With Exposure Parameters Based on the Spectrum Averaged Cross Section Approach .. . ... . . . . 6-25 6-7. Maximum Fast Neutron Exposure of Point Beach Unit 2 Beltline Circumferential Weld (SA-1484) .. . .. ... . . . , . . . . . 6-26 6-8. Maximum fast Neutron Exposure of Point Beach Unit 2 Intermediate Shell forging (123V500) . . . . . . . . . . . . . . . . . . . . . 6-27 6-9. Maximum Fast Neutron Exposure of Point Beach Unit 2 Lower Shell Forging (122W195) . . . . . . . . . . . . . . . . , , . . . . . 6 28 7-1. Comparison of Point Beach Unit 2, Capsule S, Tension Test Results . 7-7 7-2. Summary of Point Beach Unit 2 Reactor Vessel Surveillance Capsules Tensile Test Results . . . . . . . . . . . . . . . . . . 7-8 7-3. Observed Vs. Predicted Changes for Point Beach Unit 2, Capsule S g Irradiated Charpy Impact Properties -
3.47 x 10
- n/cm (E > 1 MeV) . . . . . . . . . . . . . . . . . .
2 7-10 7-4. Summary of Point Beach Unit 2 Reactor Vessel Surveillance Capsules Charpy Impact Test Results . . , . . . . . . . . . . . . 7-11 7-5. Evaluation of Reactor Vessel End-of-Life (32 EFPY) Fracture Toughness - Point Beach Unit 2 .. ... .... . ..... . . 7-12 7-6. Evaluation of Reactor Vessel Pressurized lhermal Shock Criterion for 32 EFPY - Point Beach Unit 2 . . . ... . . . . . 7 13 7-7. Evaluation of Reactor Vessel End-of-Life (32 EFPY) Upper-Shelf i Energy - Point Beach Unit 2 . . . . . . . . . . . . . . . . . . . 7-14 B-1. Tensile Properties of Unirradiated Base Metal Forging Material, Heat 123V500 val . . . . . . . . . . . . . . . . . . . . . . . . . . B-2 B - 2 .- Tensile Properties of Unirradiated Base Metal Forging Material, Heat 122W195 val . . . . . . , , . . . . . . . . . . . . . . . . . . B-2 B-3. Tensile Properties of Unirradiated Weld Metal, WF-193 . . . . . . . B-2 C-1. Charpy impact Data from Unirradiated Base Metal Forging Material, Heat 122W195 val . . . . . . . . . . . . . . . . . . . . . . . . . . C-2 C-2. Charpy Impact Data from Unirradiated Base Metal forging Material, Heat 123V500 val . . . . . . . . . . . . . . . . . . . . . . . . . C-3 C-3. Charpy Impact Data from Unirradiated Base Metal Forging Heat-Affected Zone Material - Heat 122W195 val . . . . . . . . . . . C-4 C-4. Charpy Impact Data from Unirradiated Correlation Monitor Material, SA533, Grade B, Class 1, Heat A-1195-1 . . .. . .... . . . . . C-5 C-5. Charpy Impact Data from Unirradiated Weld Metal, WF-193 . . . . . . C-6 r
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list of Fiqures Figure Page 3-1. Reactor Vessel Cross Section Showing Location of Capsule S in Point Beach Unit 2 . . . . . . . . . . . . . . . . . . . . . . . 3-4 3-2. Loading Diagram for Test Specimens in Capsule S . . . . . . . . 3-5 5-1. Charpy impact Data for Irradiated Base Metal Forging, Heat 122W195 val . . . . . . . . . . . . . . . . . . . . . . . . 5-8 5-2. Charpy impact Data for Irradiated Base Metal Forging, Heat 123V500 val . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5-3. Charpy impact Data for Irradiated Base Metal Fo71ng heat-Affected Zone Material, Heat 122W195 val . . . . . . . , . . 5-10 5-4. Charpy Impact Data for Irradiated Weld Metal WF-193 . . . . . . 5-11 5-5. Charpy Impact Data for Irradiated Correlation Monitor Material, HSST PL-02, Heat No. A-1195-1 . . . . . . . . . . . . 5-12 5-6. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal Forging Heat No.122W195 val . . . 5-13 5-7. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal Forging Heat No. 123V500 val . . . 5-14 5-8. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surf aces - Weld Metal WF-193 . . . . . . . . . . . . . 5-15 5-9. Photographs of Charpy impact Specimen Fracture Surfaces -
Base Metal Forging, Heat 122W195VA1 ...... ... . . . . . 5-16 5-10. Pnotographs of Charpy Impact Specimen fracture Surfaces -
Base Metal Forging, Heat 123V500 val . . . . . . . . . . . . . . . 5-17 5-11. Photographs of Charpy impact Specimen Fracture Surfaces -
Base Metal Forging, Heat 122W195 val . . . . . . . . . . . . . . . 5-18 5-12. Photograpns of Chtrpy impart Specimen Fracture Surfaces -
Weld Metal, WF-103 ............... . . . . . . . 5-19 5-13. Photographs of Cnarpy Impact Specimen Fracture Surfaces -
Correlation Moritor Material, HSST Plate-02 . . . . . . . . . . . 5-20 6-1. Reactor Geometry Showing a 45' R,e Sector . . . . . . . . . . . . 6-29 6-2. Internal Surveillance Capsule Geometry ........ .. . . 6-30 6-3. Fast Neutron Fluence (E > 1.0 MeV) as a function of Azimuthal Angle at the Inner Radius of the Beltline Circumferential Weld . 6-31 6-4. Fast Neutron Fluence (E > 0.1 MeV) as a Function of Azimuthal Angle at the Inner Radius of the Beltline Circumferential Weld . 6-32 6-5. Iron Atom Displacements [dpa] as a function of Azimuthal Angle at the Inner Radius of the Beltline Circumferential Weld .. . . 6-33 A-1. Location and Identification of Materials Used in the fabrication of Point Beach Unit 2 Reactor Pressure Vessel . . . . . . . . . . . A-4 A-2. Location of Surveillance Capsule Irradiation Sites in Point Beach Unit 2 Reactor Vessel (Lead Factors for the Capsules Shcwn in Parentheses are for the Original fuel Management) . . . . A-5 C-1. Charpy impact Data from Unirradiated Base Metal Forging liaterial, Heat ll2W195 val . . . . . . . . . . . . . . . . . . . . . . . . . . C-7 C-2. Charpy Impact Data from Unirradiated Base Metal Forging Material, 4 Heat 123V500 val . . . . . . . . . . . . . . . . . . . . . . . . . . C-8 i
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Figure Page C-3. Charpy impact Data from Unirradiated Base Metal forging Material, Heat- Affected Zone, Heat 122W195 val . . . . . . . . . . . . . . . . C-9 C-4. Charpy Impact Data from Unirradiated Weld Metal, WF-193 . . . . . . C-10 C-5. Charpy impact Data from Unirradiated Correlation Monitor Material, SA533, Grade B, Class 1, Heat A-1195-1 . . . . . . . . . . . . . . C-Il D-1. Tension Test Stress-Strain Curve for Base Metal Forging Heat 122W195VA1, Specimen No. E8, Tested at 70F . . . . . . . . . D-2 D-2. Tension Test Stress-Strain Curve for Base Metal forging Heat 122W195 val, Specimen No. E9, Tested at 300F . . . . . . . . . D-2 0-3. Tension Test Stress-Strain Curve for Base Metal Forging Heat 122W195VA1, Specimen No. E7, Tested at 550F . . . . . . . . . D-3 D-4. Tension Test Stress-Strain Curve for Base Metal forging Heat 123V500 val, Specimen No. V9, Tested at 70F . . . . . . . . . . D 3 D-5. Tension Test Stress-Strain Curve for Base Metal Forgino Heat 123V500 val, Specimen No. V8, Tested at 300F . . . . . . . , . D-4 D-6. Tension Test Stress-Strain Curve for Base Metal forging Heat 123V500 val, Specimen No. V7, Tested at 550F . . . . . . . . . D-4 D-7. Tension Test Stress-Strain Curve for Weld Metal WF-193, Specimen No. W9, Tested at 70F . . . . . . . . . . . . . . . . . . D-5 0-8. Tension Test Stress-Strain Curve for Weld Metal WF-193, Specimen No. W7, Tested at 300F , . . . . . . . . . . . . . . . . . D-5 D-9. Tension Test Stress-Strain Curve for Weld Metal WF-193, 5pecimen No. WB, Tested at 550F . . . . . . . . . . . . . . . . . . D-6 4
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- 1. INTRODUCfl0N This report describes the results of the examination of the fourth capsule (Capsule S) of the Wisconsin Electric Power Company, Point Beach Nuclear Plant Unit No. 2 (Point Beach Unit 2) teactor vessel material surveillance program (RVSP). The caosule was removed and evaluated after being irradiated in the Point Beach Nuclear Pl ard VI.it-2 as part of the reactor vessel materials surveillance program (WCAP-7712).' lhe capsule experienced a fluence of 3.47 x 10 I9 n/cm 7 (E > 1 lieV), which is the equivalent of approximately 39 effective full power years' (EFPY) operation of the Point Beach Nuclear Plant Unit 2 reactor vessei inside surface. The first capsule (Capsule V) from this program was removed and examined after the first period of operation or 1 52 EFPY; the results are reported in a Battelle report.' The second capsult. (Capsule T) was removed and examined after irradiation for 4 cycles, or 3.45 IfPY; tne results are reported in WCAP-9331.8 The third capsule (Capsule R) of the program was removed and evaluated after 6 cycles, or 5.1 EFPY and the results reported in WCAP-W35.*
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 Point Beach Nuclear Plant Unit 2 was designed and furnished by Westinghouse Electric Corporation (W) as described in WCAP-7712' and is 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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1 l 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 neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of such low-alloy ferritic steels as SA508, Class 2. forging material used in the fabrication of the Point Beach Unit 2 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 steeis 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,"' specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of water-cooled power reactors, and provides specific guidelines for determining the pressure-temperature limitations on operation of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during any condttion 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 13, 1973, the requirements are applicable to ail 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 Requirements,"5 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 reacurs resulting frcm exposure to neutron irradiation and the thermal envin nment. fracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel, These data will permit deterinination of the conditions under which the vessel can be operated with adequate safety margins agaicst fracture throughout its service life.
A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Cod 2,Section III,
" Nuclear Power Plant Components."' This method utilizes fracture mechanies concepts and the reference nil-ductility temperature, RTNDT, which is defined as the greater of the drop weight nil-ductility transition temperature (per ASTM E-208) or the temperature that is 60F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion. The RT NDT f a given material is used to index that material to a reference stress intensity factor curve (K IR curve),
which appears in Appendix G of ASME Section III. The K !R curve is a lower bound of dynamic, static, and crack arrest fracture toughness results obtained from several heats of pressure vessel steel. When a g;ven material is indexed to the K
IR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
The RT NDT and, in turn, the operating limits of a nuclear power plant, can be adjusted to account for the effects of radiation on the properties of the reactor vessel materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which a surveillance capsule 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. lhis adjusted RT NDT is used to index the material to the K IR curve which, in turn, is used to set operating limits for the nuclear l
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power plant. These new limits take into account the effects of irradiation on the reactor vessel naterials.
Appendix G,10CFR50, also requires a minimum Charpy V-notch upper-shelf energy of 75 f t-lbs for all beltline region materials unless it is demonstrated that i lower values of upper-shelf fracture energy will provide an adequate margin for deterioration as the result of neutron radiation. No action is required for a material that does not meet the 75 ft-lb requirement provided the irradiation deterioration does not cause the upper-shelf energy to drop below 50 ft-lbs. The regulations specify that if the upper-shelf energy drops below 50 ft-lbs, it must be demonstrated in a manner approved by the Office of Nuclear Regulation that the lower values will provide adequate margins of safety.
When a reactor vessel f ails to meet the 50 ft-lb requirement, a program must be submitted for review and approval at least three years prior to the time tne predicted fracture toughness will no longer satisfy the regulatory requirements.
The program must address the following:
A. A volumetric examination of 100 -arcent of the beltline materials that do not meet the requirement.
B. Supplemental fracture toughness data as evidence of the fracture toughness of the irradiated beltline materials.
C. Fracture toughness analysis to demonstrate the existence of equivalent margins of safety for continued operation.
If these procedures do not indicate the existence of an adequate margin of safety, the reactor vessel beltline may be given a thermal annealing treatment to recover the fracture toughness properties of the materials.
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- 3. SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for Point Beach Unit 2 comprises six surveillance capsules designed to monitor the effects of neutron and thermal environment on the materials of the reactor pressure core region. The capsules, which were inserted into the reactor vessel before initial plant startup, were positioned inside the reactor vessel between the thermal shield and the vessel wall at the locations shown in Figure 3-1. The six capsules, designed to be placed in holders attached to the thermal shield are positioned near the peak axial and azimuthal neutron flux. WCAP-7712 includes a full description of the capsule locations and design. During the sixteen cycles of operation, Capsule 5 was irradiated in the 33" position as shown in figure 3-1.
Capsule S was removed during the sixteenth refueling shutdown of Point Beach Un.
- 2. The capsule contained Charpy V-notch impact test specimens fabrica* om the two base metals (SA508, Class 2), one heat-affected-zone, a weld icetal and a correlation monitor. Tension test >pecimens were fabricated from the two base metals and the weld metal, in addition, specimens were included for determining the fracture toughness of the base metals and the weld metal. lhe 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 Figure 3-2. The chemical composition and heat treatment of the surveillance material in capsule S are described in Table 3-2.
All test specimens were machined from the 1/4-thickness (1/4T) location of the plate material. Charpy V-notch and tension test specimens were cut from the surveillance material such that they were oriented with their longitudinal axes
. parallel to the principal working direction (hoop direction of the shell ring forging) and the Charpy V-notch oriented perpendicular to the principal working direction.
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Capsule S contained dosimeter wires, described as follows:
Qqsimeter Shieldina U03 Cd0 Np0 2 Cd0 Ni None Co-Al Cd Co-Al None fe None Cu Hone Thermal monitors of low-melting alloys were included in the capsule. The alloys and their melting points are as follows:
Allov Meltina Point. F 97.5% Pb, 2.5% Ag 579 97.5% Pb, 1.75% Ag, 0.75% Sn 590 Table 3-1. Specimens in Survpillance Caosule S Number of TesLJpecimens Per Caosula_
Material Description Tension CVN(*) Impact WOL(b)
Base Metal:
forging No. 122W195 val 3 12 3 Forging No. 123V500 val 3 12 3 HAZ --
8 --
Weld Metal: WF-193 3 8 3 Correlatien Monitor:
HSST Plate 02 _- _.B ..
Total per Capsule 9 48 9 (a)CVN denotes Charpy V-notch.
(b)WOL denotes wedge opening loading.
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Table 3-2. Chemical Composition and Heat Treatment of Surveillance Materials Chemica]_C.pp_gosition. w/o Forging i;rging HSST Weld Metal Element 123V505 val 122W195 val Plate 02" (WF-193)*
C 0.20 0.22 0.22 0.08 Mn 0.65 0.59 1.48 1.40 P 0.009 0.010 0.012 0.014 S 0.009 0.008 0.018 0.013 Si 0.24 0.23 0.25 0.55 Ni 0.71 0.70 0.68 0.59 Cr 0.35 0.33 ---
0.07 Mo 0.59 0.60 0.52 0.39 Cu 0.088 0.051 0.14 0.25 Heat Treatment (d)
Heat No. Tema. F line d Coolina Forging - 1550 9.5 Water Quenched 123V500 val 1200 12 Air Cooled 1125 12 Furnace Cooled forging - 1550 8 Water Quenched 122W195 val 1200 12 Air Cooled 1125 12 Furntce Cooled Weld Metal 1125 11.5 Furnace Cooled Correlation 1650 - 1700 4 Air Cooled Monitor, 1575 - 1625 4 Water-Quenched Plate 02 1200 - 1250 4 Furnace Cooled 1125 - 1175 40 Furnace Cooled to 600 F I") Chemical analy* sis by Westinghouse of surveillance proorsm test plate B7835-1.
(b) Chemical analysis from ORNL-4313."
(c) Chemical analysis by Westinghouse of surveillance program test weld metal.'
(d) Post-weld heat treatment data per WCAP-7712, 3-3 13W!!sefef45.v
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- 4. PRE-IRRADIATION TESTS Unirradiated material whs evaluated for two purposes: (1) to establish a baseline of data to which irradiated properties data could be referenced, and (2) to determine those materia's properties to the extent practical from available material, as required f.x :ompliance with 10CFRdd, Appendixes G and H.
The pre-irradiated sp uimens were tested by Westinghouse as part of the development of the f oint Beach Unit 2 surveillance program. The details of the testing procedu"e" are described in WCAP-7712 and are summarized here to provide continuity.
L l _ Tension Tests Tension te :t spetimens were fabricated from the reactor vessel shell forging and weld w.t. 1. The specimens were 4.23 inches long with a reduced section 1.255 inches long by 0.250 inch in diameter, lhey were tested on a universal test nachine. An extension device with a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accordance with the applicable requirements of ASTM A370 68.9 For each material type and/or condition, six specimens in groups of two were tested at room temperature, 300 and 600F. All test data for the pre-irradiation tensile specimens are given in Appendix B.
4.2. Lmpact Tes.tji Charpy V-notch impact tests were conducted in accordance with the requirements 10 on an impact tester certified to meet Watertown standards.II of ASTM E23-72 Test specimens were of the Charpy V-notch type, which were nominally 0.394 inch square and 2.165 incW long.
Impact test data for the unirradiated baseline reference materials are presented in Appendix C. Tables C-1 through C-5 contain the basis data that are plotted 4-1 BWUK"E?isLar
in Figures C-1 through C-5. These data were replotted and re-evaluated to be consistent with the irradiated Charpy curves and evaluations.
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- 5. POST-IRRADIATION TESTING 5.1. Visual Examination and Inventory The capsule was inspected and visual examination confirmed the markings as those of Capsule S. 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. The weld metal wedge opening loading (WOL) specimens were stored for future disposi-tion.
5.2. Thermal Monitors Surveillance Capsule S contained five temperature monitors, one in each holder block which also contained dosimeters. The upper, middle and lower holder blocks each contained a thermal monitor designed to melt at 579F and the top middle and bottom middle holder blocks each contained a thermal monitor designed to melt at 590F. The holder blocks were radiographed for evalaation. The monitors in the top middle, middle, and bottom holder blocks had not melted. The monitor in the top holder block had melted and the monitor in the bottom middle holder block had probably softened but not melted entirely. From these data, it was concluded that the irradiated specimens had been exposed to a maximum temperature grmter than 579F during the reactor vessel operating period but less than 590F. This is not significantly greater than the nominal inlet temperature of SS8F, and is considered acceptable for inclusion of the data in the general. pool of irradiated surveillance data. There appeared to be a small temperature gradient along the capsule length.
5.3. Tension Test Results The results of the post-irradiation tension tests are presented in Table 5-1.
Tests were performed on specimens at room temperature, 300F and 550F. They were 5-1 BW#47 Fahr
tested on a 55,000-lb load capacity universal test machine at a cross head speed of 0.005 inch ner minute to yield point and thereafter 0.040 inch per m,nute.
A 4 pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accordance with the applicable requirt 'ents of A51h A370 '/7.12 f or each material type and/or condition. =
specimens were tested at room temperature, 300 and 550f. The tension compression load cell used had a certified accuracy of better than 40.5% of full scale (25,000 lb). Photographs of the tension test specimen fractured surfaces are presented in Figures 5-6, 5 7, and 5 8.
In general, the ultimate strength and yield strength of the material increased with a corresponding slight decraase 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 in within the range of changes to be expected for the radiation environment to which the specimens were exposed.
The results of the pre-irratiiation tension tests are presented in Appendix B.
5.4. Charov V Notch impact Tost Resulti The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 5-2 through 5 6 and Figures 5-1 through 5-5. Phot 1 graphs of the Charpy specimen fracture surfaces are presented in Figures 5 9 thiough 5 13. The Charpy V-notch impact tests were conducted in accordance with the requirements of ASTM E23-86 I3 on an impact tester certified to meet Watertown standards.II 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.
The results of the pre irradiation Charpy V notch impact tests are given in Appendix C.
5-2 13W###fa %v
5.5. Wedge Openino loadina_Spagjn.n_
The weld retal wedje opening l u4ing spacimens were not tested. Two specimens
( were stried for testing at a future date. Two specimens were transferred to the B&W (wners Group Master Integrated Reartor Vessel Surveillance program for t further irradiation."
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Table 5-1. g Capsule S. Base Metal Tensile and WeldProperties of PointtoBeach Metal Irradiated 3.47 xUnit 10 g, n/cm' (E > 1 MeV)
Strenoth, psi Fracture Elonnation. % Rc tion Specimen Test Temp, load, Stress, Strength, in Area, No. F Yield Ultimate lbs osi psi Uniform Total %
Base Metal Foraino. Heat 122W195 val 70 79.700 100,600 3130 195,100 63,800 9.3 22.8 67.3 i E8 300 73,000 93,100 2920 181,800 59,500 9.1 20.6- 67.3 E9 81,000 103,600 3380 198,800 68,800 8.2 20.2 65.4 E7 550 Base Fetal Forcina. Heat 123V500VA1 74,300 94,700 2790 189,100 56,800 9.7 24.7 70.0 V9 70 80,500 96,500 2860 215,800 58,300 8.2 '2.4 73.0 V8 300 70,100 95,100 3070 176,200 62,600 8.9 22.3 64.5 V7 550 f
l Weld Metal. WF-193 96,900 110,600 4000 180,400 81,500 10.3 21.7 54.8 W9 70 300 89,400 103,000 3970 165,1C0 80,900 8.8 18.3 51.0 l W7 83,700 102,000 4300 154,900 87,600 6.1 15.4 43.4 l g W8 550 EE EE N!
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} Forging 122W195 val, Irradiated to 3.47 x 10 n/cm' (E > 1 MeV)
Absorbed lateral Shear Specimen Test Temp., Energy, Expansion, fracture, 8
No. F ft-lb 10 in. %
E27 -50 17.0 11 0 E33 -25 36.0 28 0 E35 0 36.5 27 0 E30 25 37.5 27 10 E28 35 20.0 16 10 E25 70 70.0 51 50 E29 125 107.0 73 70 E31 175 132.0* 84 100 E3r, 200 141.5* 90 100 E34 3C0 130.0* 83 100 E26 400 135.0 87 100 E32 550 127.5 82 100
- Values used to determine upper-shelf energy value per AS1H E185.
Table 5 3. Charpy impact Data from Point Beach Unit 2," Capsule 5, Base Metal forging 123V500 val, Irradiated to 3.47 x 10 n/cm' (E > 1 MeV) _
Absorbed lateral Shear Specimen Test Temp., Energy, Expansion, Fracture, No. F ft-lb 10' in. %
V36 -50 3.5 3 0 V32 -10 16.0 12 0 V34 0 3.5 2 0 V30 20 61.5 49 30 V33 40 81.5 63 40 ,
V35 70 104.0 73 70 V26 100 150.0* 84 100 V31 125 178.0* 81 100 V29 200 163.0* 90 100 V25 300 148.0 87 100 V28 400 170.0 80 100 V27 550 153.5 75 100
- Values used to determine upper-shelf energy values per ASTM E185.
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i Table 5-4. Charpy impact Data from Point Beach Unit 2. Capsule S, Base Metal forging Heit Affected Zone Haterial, Irradiated to 3.47 x 10" n/cm' (E > 1 HeV)
Absorbed lateral Shear Specimen Test Temp., Energy, Expansion, fracture, No. I ft-Ib 10' in. %
H18 70 21.5 18 20 H21 70 156.5 87 100 H22 140 107.0 71 100 H23 200 45.5 30 60 H19 270 140.5 84 100 H17 300 80.0 55 100 H2O 400 133.0 89 100 H24 550 66.0 54 100 Opper sheff energy value could not be determined Table 5 5. Charpy impact Data from Point Beach Unit 2, Capsule 5, Weld Metal WF 193, Irradiated to 3.47 x 10" n/cm' (E > 1 MeV)
Absorbed lateral Shear Specimen Test Temp., Energy, Expansion. Fracture, 8
No. I ft-lb 10 in. %
W19 150 22.5 19 40 W23 200 21.5 17 30 W21 225 28.0 25 50 W22 250 38.0 39 70 W17 270 44.0* 46 85 W18 300 36.0 33 100 l W20 400 43.5* 41 100 W24 550 39.5 37 100
- Values used to determine upper shelf energy value.
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Table 5-6. Charpy impact Data from Point Beach Unit 2, Capsulo S, Correlation Monitor Material.' Heat No. A 1195 1, Irradiated to 3.47 x 10" n/cm (E > 1 MeV)
Absorbed lateral Shear Specimen Test Temp., Energy, Expansion, fracture, No. F ft lb 10' in. %
R18 125 12.0 10 10 R22 175 13.5 11 45 R20 200 35.5 30 40 R17 250 61.0 47 C0 R21 300 56.0 49 90 R24 350 82.0* 69 100 R23 400 86.0* 71 100 R19 550 84.0* 70 100
- Valued used to determine upper shelf energy value per ASTM E185."
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- 6. NEUIRON DOSIMETRY EVALVA110NS 6.1. IntI9dEllRD Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed ii) the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The formir requiremert is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.
The latter information is derived solely from anelysis. The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, is has been suggested that an exposure model that accounts for differences in neutr s energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall. Because of this potential shif t away from a threshold fluence toward an energy dependent damage function for data correlation. ASTM Standard Practice E853, 6-1 B WitEEVEL % 1 1
"Anelysis and Interpretation of Light Water Reactor Surveillance Results,""
recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in AS1H Standard Practice EE93, " Characterizing Neutron Exposures in ferrith Steels in Terms of Displacements per Atom."" The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials."" This section provides the results of the neutron dosimetry evaluations performed in conjunctit,n with the analysis of test specimens contained in surveillance Capsule S. Fast neutron exposure parameters in terms of f ast neutron fluence (E
> 1.0 MeV), fast neutron fluence (E > 0.1 HeV), and iron atom displacements (dpa) are established for the capsule irradiation history. These results are then used in conjunction with measurements from prior surve111ar..:e capsule withdrawals and reactor cavity dosimetry irradiations to project the integrated exposure of the vessel wall. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vestel are provided.
L2. Neuttpn Transngrt Ana]Ysis Metho.d1 Fast neutron exposure calculations for the reactor geometry were cartied out using both forward and adjoint discrete ordinates transport techniques. A single forward calculation provided the relative energy distribution of neutrons and gamma rays for use as input to neutron dosimetry evaluations as well as for use in relating measurement results to the actual exposure at key locations in the pressurc vessel wall. A series of ad.ia%t calculations, on the other nand, established the means to compute abr31ute wx osure i) .e values using fuel cycle specific core power distributions; he., pro iding a direct comparison with all dosimetry results obtained over the a re ing history of the reactor. In combination, the absolute cycle spe;ific data from the adjoint evaluations together with relative neutron energy spectre distributions from the forward calculation provided the meanc 'o:, 6-2 BW!!nn?i%
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- 1. Evaluate neutron dosimetry from reactor cavity and surveillance capsule locations.
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- 2. Enable a direct comparison of analytical prediction with measurement.
- 3. Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel.
- 4. Establish a mechanism for projection of pressure vessel expostre as the design of each new fuel cycle evolves.
5J.J_,__ Reference Forward Calculation A plan view of the reactor geometry at the core midplane elevatior is shown in Figure 6 1. Since the reactor exhibits 1/8 ccre symmetry only a 0-45 degree sector is depicted. In addition to the core, reactor internals, pressure vessel, and the primary biological shield, the model also included explicit representa-tions of the surveillance capsules, the pressure vessel cladding, and the mirror insulation located external to the vessel. A description of a single surveillance capsule attached to the thermal shield is shown in Figure 6-2. From a neutronic standpoint, the inclusion of the surveillance capsules and associated support structures in the analytical model is significant. Since the presence of the capsules and structure has a marked impact on the magnitude of the neutron flux as well as en the relative neutron and gamma ray encrgy spectra at dosimetry locations within the capsules, a meaningful comparison of measurement and calculation can be made only if these perturbation effects are properly accounted for in the analysis. The forward transport calculation for the reactor model depicted in Figures 6-1 and 6-2 was carried out in R,6 geometry using the DOT two-dimensional discrete ordinates code" and the SAILOR cross-section library." The SAILOR library is a 67 group coupled neutron-gamma ray ENDFB-IV based data set produced specifical-ly for light water reactor applications. In these analyses, anisotropic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an S8 order of angular quadrature. The reference forward calculation was normalized to a core midplane power density characteris-tic of operation at a thermal power level of 1518 MWt. 6-3 BW!!nn? ale %mr
The spatial core power distribution utilized in the reference forward calculation was derived from statistical studies of long-term operation of Westinghouse 2-loop plants, inherent in the development of this reference core power distribution was the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery, furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Due tt, the use of this bounding spatial power distribution, the results from the reference forward calculation established conservative exposure projections for reactors of this design operating at 1518 MWt. Since it is unlikely that actual reactor operation would result in the implementation of a power distribution at the nominal + 2a level for a large number of fuel cycles and, further, because of the widespread implementation of low leakage fuel management strategies, the fuel cycle specific calculations for this reactor result in exposure rates well below these conservative predictions. 5,2.2. Cycle Specific Ad.ioint Calculalinai All adjoint analyses were also carried out using an 58 order of angular quadrature and the P3 cross-section approximation from the SAILOR library. Adjoint source locations were chosen at each of the azimuthal locations containing cavity dosimetry (0, 15, 30, and 45 degrees) as well as at the corresponding azimuths on the pressure vessel inner radius, in addition, adjoint calculations were carried out for sources positioned at the geometric center of surveillance capsules located at 13, 23, and 33 degrees relative to the core cardinal axes. Again, these calculations were run in R,a geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, o (E > 1.0 MeV). The importance functions generated from these eleven individual adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each of the fuel cycles to date; and, l 13 Wit?MsM"an
established the means to parform similar predictions and dosimetry evaluations for all subsequent fuel cycles. Having the importance functions and appropriate core source distributions, the response of interest can be calculated as: x(R,s) = laf,ie 1(R,#,E) S(Re,E) R dR de dE where d(R,#) - Flux (E > 1.0 MeV) at radius R and azimuthal angle d. l(R,#,E) - Adjoint importance function at radius R azimuthal angle e, and neutron source energy E. S(R,#,E) - Neutron source strength at core location R,s and energy E. It is important to note that the cycle specific neutron source distributions. S(R,e,E), utilized with the adjoint importance functions,1(R,t,E), permitted the use not only of fuel cycle specific spatial variations of fission rates within the reactor core; but, also allowed for the inclusion of the effects of the differing neutron yield per fission and the variation in fission spectrum introducrd by the build-in of plutonium isotopes as the burnup of individuci fue'. assemblies increased. Although the adjoint importance functions used in these analyses were based on a response function defined by the thresholo neutron flux (E > 1.0 MeV), prior calculations have shown thst, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the exposure par. meter ratios such as [dpa/sec]/[4 (E > 1.0 MeV)] are insensitive to changing code source distribu-tions. In the application of these adjoint importance functions to the current evaluations, therefore, calculation of the iron displacement rates (dps/sec) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using the appropriate [dpa/sec]/[4 (E > 1.0 MeV)] and [4 (E > 0.1 MeV)]/[d (E > 1.0 MeV)] ratios from the reference forward analysis in conjunction with the cycle specific d (E > 1.0 HeV) salutions from the individual adjoint evaluations. 6-5 13W#enWAv
In particular, after defining the following exposure rate ratios, (dpa/ sec) q,,_ [$ (f > 1.0 #eV)) g , (Q (f < 0.1 #cV)] ($ (f > 1.0 NeV)] the corresponding fuel cycle specific exposure rates at the adjoint source locations were computed from the following relations: dpa/sec - [d (E > 1,0 MeV)]4, R, 4 (E > 0.1 MeV) = [d (E > 1.0 MeV)lg, R3 6.2.3. Results of Neutr.pn Transport Calculations The results of the plant specific calculations applicable to the irrhilation period associated with Capsule S are summarized in Tables 6-1 and 6 2. !n table 6-1, exposure rates calculateri for the first 16 fuel cycles of operation are given for the surveillance capsule position. Values for neutron flux (E > 0.1 MeV) and dpa/sec are provided along with the calculated values of neutron flux (E > 1.0 MeV). These additional exposure parameters were computed using the following ratios extracted from the reference forward transport calculation.
"U' (I ' *
- I "'Y5 dpalsec LOCATION -
FLUX (f > 1.0 MeV) FLUX (f > 1.0 MeV) 33* CAPSULE ?.50 1. 76f-21 Integrated exposures calculated for the end of each fuel cycle are given in Table 6-2. The fuel cycle lengths associated with the first sixteen fuel cycles are also included on Table 6-2. Plant specific calculations applicable to other surveillarice capsule locations, to the pressure vessel wall, and to reactor cavity sensor positions are provided in Reference 21. These additional computations along with the measured data also 6-6 ! 13W#eais h
s provided in Reference 1 were employed in conjunction with the Capsule S results to establish exposure projections for the pressure vessel wall of Point Beach Unit 2. 6.3. Neutron Dosimetry Evaluatimi Methodology [ The use of passive neutron sensors such as those included in the internal surveillance capsult and reactor cavity dosimetry sets does not yield i direct { measure of the enargy dependent neutrun flux level at the measurement location. Rather, the activation or fission process is a measure of the integrated effect that the tine- and energy dependent neutron flux has on the target material over the course of the irradiation p,aiod. An accurate assessmint of the average flux level and, hence, time integrated esposure (fluence) experienced by the sens u s may be developed from the measurements only if the sensor characteristics and the parameters of the irradiation are well known. In particular, the following variables are of interest:
- 1. The measured specific activity of each sensor.
- 2. The physical characteristics of each sensor.
- 3. The operating history of the reactor.
- 4. The energy response of each sensor.
- 5. The neutron energy spectrum at the sensor location.
In this section the procedures used to determine sensor specific activities, to develop reaction rates for individual sensors from the measured specific activities and the operating history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates are described. 6.3.1. Delermination of Sensor Reaction Ratei The specific activity of each of the radiometric sensors was determined using established ASTM procedures [16, 22-31). Following sample preparation and weighing, the specific activity of each sensor was determined by means of a lithium drifted germanium, Ge(t.i), gamma spectrometer. These analyses were performed either by direct counting of each of the individual wires; or, as in the case of U-238 and Np-237 fission monitors, by direct counting preceded by dissolution and chemical separation of cesium from the sensor. 6-7 13WfafME%v
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The irradiation history of the reactor over its operating lifetime was obtained from NUREG 0020 " Licensed Operating Reactors Status Summary Report."" In particular, operating data were extracted from that report on a monthly basis from reactor startup to the end of the current evaluation period. For the sensor sets utilized in surveillance capsu!, irradiations, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. Having the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to full power operation at 1518 MWt were determined from the following equation: A R= NfY o (P /P,,,) C j j (1 - e '/} [e ") where: R - Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P,,, (rps/ nucleus). . I No - Number of target element atoms per gram of sensor. i f - Waight fraction of the target isotope in the sensor material. Y Number of product ato.ms produr.ed per reaction. P - Average core power level during irradiation period j (MW). 3 P,,, - Maximum or reference core power level of the reactor (MW). C, - Calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted average 4 (E > 1.0 MeV) over the entire irradiation period. A - Decay constant of the product isotope (sec"). t,- Length of irradiation period j (sec). to - Decay time following irradiation period J (sec). SW##efef45=r i i
and the summation is carried out over the total number of monthly intervals comprising the total irradiation period, in the above equation, the ratio P/P,,, accounts for month by month variation of power level withia a given fuel tycle. The ratio C, is calculated for each fuel cycle using the adjoint transport methodology and accounts for the chang 9 in sensor reaction rates caused by variations in flux level aue to changes in core power spatial distributions from fuel cycle to fuel cycle. For a single cycle irradiation C 3 1.0. However, for multiple cycle irradiations, particularly those employing low leakage fuel management the additional C, correction must be utilized. Prior to using the measured reaction rates in the least squares adjustment procedure discussed in Section 6.2.2, corrections were made to the U-238 foil measurements to :ccount for the presenr.e of U 235 impurities in the sensors as well as to adjust for the build-in of plutonium isotoras over the course of the irradiation. Likewise, corrections were made to both U-238 and Np-237 sensors to account for gamma ray induced fission reactions occurring over the course of the irradiation. These corrections were location and fluence dependent and were derived from a combination of data from the reference forward transport calculation and the cycle specific adjoint analyses. In performing the dosimetry evaluations for the internal surveillance capsule, the sensor reaction rates measured at the lo'ations shown in figure 6-2 were indexed to the geometric center of the cap:,ules prior to the use in the spectrum adjustment procedure. This procedure required correcting the measured reaction rates by the application of analytically determined spatial gradients. For the Point Beach Unit 2 surveillance capsules, the gradient correction factors for each sensor reaction were obtained from the reference forward transport calculation and were used in the multiplicative f ashion to relate individual measured reaction rates to the corresponding value at the geometric center of the surveillance capsule. 6.3.2. Least Sagares Ad_iniment Procedure Values of key fast neutron exposure parameters were d-ived fro.n the measured reaction rates using the FERRET least squares adjustment code." The FERRET 6-9 BWUMYa%r
1 1 approach used the measured reaction rate data and the calculated neutron energy spectrum at the sensor set locations as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with associated uncertainties were then obtained from the adjusted spectra. In the FERRET eva'uations, a log-normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertain-ties and correlations. In general, the measured values f are linearly related to tFe flux 6 by some response matrix A: 7j..., . g 4.,4., where i indexes the measured values belonging to a single data set s, 9 designates the energy group and a delineates spectra that may be simultaneously adjustr'. For exampic,
% 0, relates a set of measured reaction r n es R, to a single spectrum ,. (In this case, FERRET also adjusts the cross-sections.) The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the FERRET analysis of the dasimetry data, the continuous quantities (i.e., flures and cross-sections) were approximated in 53 groups. The calculated fluxes from tne reference forward calculat ^ n were expanded into the FERRET group structure using the SAND-Il code.' nis procedure was carried out by first expanding the a priori spectrui. .ne SAND-Il 620 group structure using a SPLINE interpolaticn procedure im interpolation in regions where group boundaries do not coincide. The 620 point spectrum was then easily collapsed to the group scheme used in FERRET. 6-10 BWs%"v'Ma%v ) l 1
The cross sections were also collapsed into the 53 energy-group structure using SAND 11 with calculated spectra (as expanded to 620 groups) as weighting ( functions, The cross sections were taken from the ENDf/B-V dosimetry file Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross settions were neglected due to cata and code limitations, but this omission does not significantly impact the results of the adjustment. For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight. In some cases, as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form is employed: M,,, = R * + R, R,. P,,, where R, specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional uncertainties R, specify additional random uncertainties for group g that are correlated with a correlation matrix: Poo ,.(1-6)6 er,+ee ~I9'97 gp . The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 3 (e specifies the strength of the latter term). For the a priori calculated fluxes, a short-range correlation of y - 6 groups was used. This choice implies that neighboring groups are strongly correlated whan e is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R. E. Maerker. Maerker's results are closely duplicated when y - 6. For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties. BWiina?a%r
l In performing the least squares adjustment with the FERRET code, the input spectra from the reference forward transport calculation were normalized to the measured fe-54 (n p) Mn-54 reaction rates to remove any constant calculation to measurement bias and, thus, to permit the adjustment to take place on a relative basis. The specific normalization factors for individual evaluations depended on the location of the sensor set es well as on the neutron flux level at that location. The specific assignment of uncertainties in the measured reaction rates and the input (a priori) spectra used in the FERRET evaluations was as follows: Reaction rate uncertainty 5% Flux normalization uncertainty 30% Flux group uncertainties (E > 0.0055 MeV) 30% (0.68 ev < E < 0.0055 MeV) 58% (E < 0.68 ev) 104% Short range correlation (E > 0.0055 MeV) 0.9 (0.68 ev < E < 0.0055 MeV) 0.5 (E < 0.68 ev) 0.5 flux group correlation range (E > 0.0055 MeV) 6 (0.68 ev < E < 0.0055 MeV) 3 (E < 0.68 ev) 2 ' It should be noted that the uncertainties listed for the upper energy ranges extend down to the lower range. Thus, the 58% group uncertainty in the second range is made up of a 30% uncertainty with a 0.9 short range correlation and a range of 6, and a second part of. magnitude 50% with a 0.5 correlation and a range of S. These input uncertainty assignments were based on prior experience in using the FERRET least squares adjustment approach in the analysis of neutron dosimetry from surveillance capsule, reactor cavity, and benchmark irradiations. The 6-12 BW!!na?%fe%r l,
values are liberal enough to permit adjustment of the input spectrum to fit the measured data for all practical applications. 6.3.3. Results of Dosimetry Evaluations Results of the FERRET evaluation for surveillance Capsule S are provided in Tables 6-3 and 6-4. In Table 6-3, the derived exposure rates experienced by the test specimens along with data illustrating the fit of the adjusted spectra to the measurements are provided. A summary of the exposure rates and integrated exposures of the capsule is given in Table 6-5. As a consistency check on the exposure results obtained with the adjustment procedure, a comparison was made with results calculating using tne :;pectrum averaged reaction cross-section approach. From the reference forward calcula-tion, the appropriate spectrr averaged cross-sections calculated for the 33.0 degree surveillance capsule position were extracted and are listed in Table 6-5. Also included in Table 6-5 are the exposure parameter ratios required to calculate Flux (E > 1.0 MeV), Flux (E > 0.1 MeV), dpa/3ec, and Flux (E < 0.414 ev). The results of the consistency check are shown in Table 6-6. An examination of Table 6-6 indicates excellent agreement between the least squares adjustment approach and the spectrum averaged cross-section evaluation for all neutron reactions. 6.4. Vessel Excosure Pro.iections In this section the measurement results provided in Section 6.3 are combined with the results of the neutron transport calculations and prior surveillance capsule and reactor cavity dosimetry results documented in Reference 21 to provide the best estimate neutron exposure of the beltline region of the Point Beach Unit 2 reactor pressure vessel through the completion of Cycle 16. Based on the continued use of the Cycle 16 fuel loading pattern incorporating part length aafnium absorbers, projections of future vessel exposure to 32 and 48 effective full power years of operation are also provided. Data pertinent to the maximum exposure experienced by the upper and lower shell forgings and the beltline circumferential weld are highlighted. In essence, an approach using analytically determined gradient information to extrapolate measurement results to locations of interest within the pressure 6-13 BW##Nahr
vessel is based on the assertion that the measured values of exposure rates in the reactor cavity or from surveillance capsules represent the best available neutron fluy data for the irradiation period in question and, further, on the assumption that the analytically determined radial distribution functions provide accurate representations of the spatial gradients that exist among the measurement locations and points of interest within the pressure vessel wall. This method is analegous to the common practice of normalizing a cycle specific forward neutror. transport calculation to available measurements from either surveillance capsule or reactor cavity dosimetry programs, This approach provides accurate assessments of vessel exposure with associated uncertainties for periods of operation during which continuous monitoring has occurred. In the case of Point Beach Unit 2, the cavity dosimetry program' providing a complete spatial mapping of a sector of the beltline regon of the pressure vessel was installed at the start of Cycle 15. Additional conitoring was limited to the four scheduled surveillance capsule withdrawals. The dosimetry data from these capsules provide measurement information at , single point within the reactor geometry for the four extended irradiation periods, but cannot be used to establish a verification of the exposurc of the vessel at azimuthal locations far removed from the measurement point. Therefore, in order to establish a baseline exposure of the pressure vessel applicable to the onset of the reactor cavity measurement program documented ir Reference 21, all available core midplane measured data were combined with fuel cycle specific transport calculations to provide best estimate exposures for the first 14 cycles of operation. The reactor cavity measurements were then used directly to provide the continuous monitoring capability for Cycles 15 and beyond. 6.4.1. Baseline Exposure at the End of Cycle 14 In Reference 21, comparisons of calculated and measured exposure rates for the four surveillance capsule dosimetry sets and for the eight cavity dosimetry sets that were located on the core midplane are given. From Reference 21, it was noted that, considering all of the midplane data, the calculated exposure values underpredicted measurement by factors of 0.854, 0.899, and 0.900 for 4 (E > 1.0 MeV), & (E > 0.1 MeV), and dpa/sec, respectively. The corresponding la standard 6-14 B W!!nEV1M L r
deviations in these averages of the twelve sample data sets were 19.0%, i 9.7%, t and i 8.9%. L in developing the best estimate baseline exposure for the Point Beach Unit 2 reactor pressure vessel these ratios were employed as bias factors to scale the cycle specific neutron transport calculations documented in Reference 21. In particular, the following bias factors were employed to establish the baseline exposures of the vessel wall: ti'C Bias e (E > 1.0 MeV) 1.171 e (E > 0.1 tieV) 1.112 dpa 1.111 6.4.2. Exposure Accrued Durina Cycles 15 and 16 To assess the incremental exposure resulting from the Cycles 15 and 16 s irradiation, the measured results from the reactor cavity multiple foil sensor sets were directly extrapolated to the vessel clad / base metal interface using the analytically derived gradient data from Reference 21. Again, as noted in reference 21, exposure distributions through the vessel wall, can be developed using these surface exposures and analytically determined radial distribution functions. This exposure information, applicable through the end of Cycle 16, was derived from an extensive set of measurements and assures that embrittlement gradients can be established with a minimum uncet tainty. Further, as the monitoring program continues and additional data become available, the overall plant specific data base for Point Beach Unit 2 will expand resulting ia reduced uncertainties and an improved accuracy in the assessment of vessel condition. 6.4.3. Proiection of Future Vessel Exoosure At the end of Cycle 16, the Point Beach Unit 2 reactor had accrued 14.8 effective full power years (EFPY) of operatien. In order to establish a framework for the assessment of future vessel condition, exposure projections to 32 and 48 EFPY were developed. These temporal extrapolations into the future were based on the 6-15 B W sY s N a fai m
assuraption that the measured data from the Cycle 16 irradiation were representa-tive of all future fuel cycles. That it, that future fuel designs would incorporate the low leakage fuel management concept including part length hafnium absorbers designed tn providt ,sx reduction measures at the maximum exposure locations along the beltline circumferential weld. Examination of these projected exposure levels establishes the long term effectiveness of flux reductica measured incorporated to date and can be used as a guide in assessing strategies for future vessel exposure management. The validity of these projections for future operation will be confirmed via the continued cavity l monitoring program. 6.4.4. Exposure of Specific Beltline Materials The beltline region of the Point Beach Unit 2 reactor pressure vessel is comprised of an intermediate shell forging (Heat 123V500), a lower shell forging (Heat 122W195), and a circumferential weld (SA-1484) joining the two ring forgings. The circamferential weld is centered 15.06 inches below the axial midplane of the active core; while the intermediate shell forging extends upward to an elevation 8.44 inches atove the active fuel and the lower shell forging extends downward to an elevation 39.87 inches below the bottom of the active fuel. The raximum neutron exposure experienced by each of these beltline materials can be extracted from the data provided in Reference 21. 6.4.4.1. Circunferential Weld (SA-14841 The current (End of Cycle 16) and projected maximum exposures of the beltline circumferential weld are listed in Table 6-7 and illustrated graphically in Figures 6-3 through 6-5. In this table and the accompanying figures, the weld - exposure is expressed in terms of e (E > 1.0 MeV), o (E > 0.1 MeV), and dpa. - In developing the exposure profiles for the circumferential weld, it is noted that, although the flux reduction afforded by the Cycle 16 fuel loading pattern l with part length hafnium absorbers has lessened the exposure rates within the 0-15 degree azimuthal sector, the maximum exposure point on the weld remains at the 0 degree azimuth throughout the service life of the unit. However, the magnitude l of the projected exposures are significantly lower than would be the case had the flux reduction measures not been implemented. 6-16 BW!!Mn%
) i
! 1.4.4.2. Intermediate Shell Foraina (Heat 123V5001 The current and projected maximum exposures of the intermediate shell forging are ! given in Table 6-8. Again, all three exposure parameters are provided. In the case of the intermediate forging, it can be noted from Reference 21, that, due to the introduction of the part length absorbers and the corresponding reduction in exposure rates in the vicinity of the circumferential weld, the axial locatinn of the maximum exposure at the 0 and 15 degree azimuthal angles shifts from an elevation near core midplane to an elevation approximately 2.5 ft. bove core midplane as the lifetime of the unit increases. Corresponcir.g variations at the 30 and 45 degree azimuths are less evident. Since the maximum exposure point for the intermediate shell forging is variable due to the flux reduction measures, these values are not illustrated graphically, but are presented only in tabular form.
6.4.4.3. Lower Shell Foraina (Heat 122W1951 The current cnd projected exposures for the lower shell forging are listed in Table 6-9. As in the case of the intermediate forging, all three exposure parameters are tabulated. in the case of the lower shell forging, the part length absorbers cause the maximum exposure location at 0 and 15 degrees to shift from the top of tne forging to a position 3.5 feet below the active core midplane. However, the absorbers have a negligible iirpact at the 30 and 45 degree azimuths, resulting in the maximum exposure location remining at the top of the forging adjacent to the circumferential weld. Again, due to this shift in the maximum exposure elevation, the data applicable to the lower shell forging are not illustrated graphically, but, rather, are presented only in tabular form. 6.5. Untertainties in Exposure Proiections The overall uncertainties associated with the exposure rates ir.d integrated exposures determined for Point Beach Unit 2 stem from two basic sources; the accuracy of the neutron flux measurements at the sensor set locationh and the accuracy of the radial gradient projections derived from the use of the transport code. Based on the least squares adjustment approach used in the FERRET analyses the 1 sigma uncertainties in the measured data were as follows: 6-17 BWUh;En%r
]
I la Ui. certainty-Otpaglg Cavitv-Flux (E > 1.0 MeV) 5% 6%
-Flux (E > 0.1 MeV) 13% 15%
dpa/sec 9% 12% ( These values represent uncertainties derived from the reaction rate measurements and from the least squares fit of the output spectrum to the measured data. As additional data is obtained from the ongoing measurement program, the knowledge l
-of- the neutron spectra at the measurement locations will increase and the uncertainties in the measured exposure parameters will be reduced somewhat.
Since the ultimate goal of the cavity measurement program is the evaluation of . the exposure of the vessel itself, an additional uncertainty associated with the j ability to translate results from the measurement locattuns to the points of interest within the -vessel must be included along with the measurement I uncertainties listed above. Information pertinent tothis extrapolation 1 uncertainty has been obtained from benchmarking studies using the Westinghouse-neutron transport methodology and from several comparisons of- power reactor internal surveillance capsule dosimetry and reactor cavity dosimetry for which the irradiation history cf all sensors was the same. Based on these benchmarking evaluations the uncertainty or bias associated with ( the calculated slope through the steel vessel was estimited to-be approximately
- 5% for .all exposure parameters. Thus, the total uncertainty associated with projections at the clad / base metal interface is estimated to be as follows for each exposure parameter of interest, la ;
, Uncertainty L l Vessel IR Flux (E > 1.0 MeV) 11% b flux (E > 0.1 MeV) 20% dpa/sec 17% 6-18 SW#4,?c ci! =
Use of these values represents the bounding lo uncertainties for vessel exposure, [ since with penetration into the vessel wall the extrapolation uncertainty lessens until at the outer surface the overall uncertainty reverts simply to the measurement uncertainty. Again, as more data are accumulated from both reactor
~
savity and surveillance capsule dosimetry sets, the extrapolation uncertainty will also be reduced resulting in higher levels of accuracy in the vessel exposure projections. 6-19 BW!! seam 6v
Table 6-1. Calculated Fast Neutron Exposure Rates at the Center of the 33 Dearee Surveillance Caosules_# ore Midolane Elevation Neutron Flux (E > 1.0 MeV) (E > 0.1 HeV) 2 [n/cm sec) [n/cm'sec) dpa/sec Cycle 1 5.86E+10 2.05E+11 1.03E-10 Cycle 2 6.41E+10 2.24E+11 1.13E-10 Cycle 3 6.34E+10 2.22E+11 1.12E-10 Cycle 4 6.07E+10 2.12E+11 1.07E-10 Cycle 5 6.06E+10 2.12E+11 1.07E-10 Cycle 6 6.24E+10 2.18E+11 1.10E-10 Cycle 7 5.27E+10 1.84E+11 9.28E-ll Cycle 8 5.04E+10 1.76E+11 8.87E-11 Cycle 9 5.10E+10 1.79E+11 8.98E-11 Cycle 10 5.22E+10 1.83E+11 9.19E-11 Cycle 11 5.15E+10 1.80E+11 9.06E-Il Cycle 12 4.92E+10 1.72E+11 8.66E-11 Cycle 13 4.65E+10 1.63E+11 8.19E-ll. Cycle 14 5.07E+10 1.78E+11 8.93E-11 ! Cycle 15 4.64E+10 1.63E+11 8.17E-Il Cycle 16 4.05E+10 1.42E+11 7.13E-Il 6-20 BW!!nME! aim l
Table 6 2. Calculated Integrated fast Neutron Exposure at the Center of 33 Dearee Surveillance Caosules Core..Midolane Evaluation __ Cycle Neutron Fluence Length (E > 1.0 MeV) (E > 0.1 HeV) [EFPS] [n/cm') [n/cn') dpa EOC 1 4.81E+07 2.82E+18 9.87E+18 4.96E-03 E0C 2 3.32E+07 4.95E+18 1.73E+19 8.71E-03 E0C 3 2.75E+07 6.69E+18 2.34E+19 1.18E-02 EOC 4 2.74E407 8.35E+18 2.92E+19 1.47E-02 E0C 5 2.79E+07 1.00E+19 3.52E+19 1.77E-02 E0C 6 2.73E+07 1.17E+19 4.llE+19 2.07E-02 E0C 7 2.82E+07 1.32E+19 4.63E+19 2.33E-02 EOC 8 2.70E+07 1.46E+19 5.llE+19 2.57E-02 EOC 9 2.50E+07 1.59E+19 5.56E+19 2.79E-02 EOC 10 3.77E407 1.78E+19 6.24E+19 3.14E-02 E0C 11 2.68E+07 1.92E+19 6.73E_19 3.38E-02 E00 12 2.52E+07 2.05E+19 7.16E+19 3.60E-02 EOC 13 2.55E+07 2.16E+19 7.59E+19 3.81E-02 EOC 14 2.72E+07 2.30E+19 8.06E+19 4.05E-02 E0C 15 2.54E+07 2.42E+19 8.47E+19 4.26E-02 E0C 16 2.70E+07 2.53E+19 8.85E+19 4.45E-02 6-21 BWunEYafaar
lable 6-3. Deriyed Exnainte Rates frna.1hn JJLDuth_129]ijJn.2trLEyalua1.100
& Priori Adjusted Paraceter Value Value Mncertaint_y 4 (E > 1.0 MeV) 7.48E+10 7.44E+10 6%
4 (E > 0.1 MeV) 2.66E411 2.41E+11 13% 4 (E ( 0.414 ev) 3.58E+10 8.64E+10 19% 6 (Total) 6.58E+11 6.01E+11 13% dpa/sec 1.30E-10 1.22E-10 8% Comparison of Measured and Calculated Sensor Rention Batu Capsule S Eva)_utj_qp Rcaction Ra1.e (ros/nutl.gus) A Pricri Adjusted C/M Reaction Meantred ,_[alc. Calc. A Priori Ad. justed Cu-63 (n,a) 4.29E-17 6.39E-17 4.45E-17 1.49 1.04 s Ni-58 (n.p) 7.14E-15 6.96E-15 6.86E-15 0.98 0.96 U-238 (n,f) (Cd) 2.52E-14 2.52E-14 2.50E-14 1.00 0.99 tip-237 (n,f) (Cd) 1.88E-13 2.12E-13 1.92E-13 1.13 1.02 Co-59 (n,y) 3.78E-12 3.69E-12 3.78E 12 0.98 1.00 , i Co-59 (n,y) (Cd) 1.68E-12 1.63E-12 1.68E-12 0.97 1.00 6 22 BW!!sefSiid =v 1
Table 6-4. Derived Neutron Exposure Rates and Inte9r:.ted l _. . E x 22Lu r e E x p e r i e n c ed._b_y_S u rv e i l l a n c e C a e ntig_ji l Exposure R4te Ca9sule S (E > 1.0 MeV) [n/cm' sec) 7.44E+10 p (E > 0.1 MeV) [n/cm2 sec) 2.41E+11 dpa/sec 1.22E-10 4 (E < 0.414 ev) [n/cm' sec) 8.64E+10 4 (Total) [n/cm' sec) 6.01E+11 Integrated Exposure Capsule 5 o (E > 1 0 MeV) [n/cm'] 3.47E+19 e (E > 0.1 MeV) [n/cm'] 1.12E+20 dpa 5.69E-02 e (E < 0.414 ev) [n/cm'] 4.03E+19 e (Total) [n/cm') 2.80E+20 Irradiation Time (s) - 4.66E+08; EFPS - 14.8 EFPY 6-23 BW##efiELv
1 Table 6-5. Calculated Spectrum Averaged Reaction Cross Sections and Exposure Parameter Ratios at ti.e 33.0 Degree Surveillance Caosule locations Cross-Section (Barns)
. Raas_ tion HJ_Qeq ,.. C a psul e Cu 63 (n,a) 0.000629 Ni-58 (n.p) 0.0958 '
U-238 (n,f) (Cd) 0.345 Np-237 (n,f) (Cd) 2.90 i Reaction Rate Ratio [dpa/sec]/[ (E > 1.0 MeV)] 1.76E-21 [( (E > 0.1 MeV)]/[4 (E > 1.0 MeV)] 3.50 [d (E < 0.414 ev)]/[ (E > 1.0 MeV)] 1.15 i 6-24 BWs*fatfaMEm
Table 6 6. Comparison of FERRET Results With Exposure Parameters { Based on ib1Atq1te_AynntLCt9h_Sgtion Annogh 1 Flux (E>1.0 Hev) (n/cm' _sec) . _ _ Reaclign Capsule S. Cu-63 (n.a) 6.82E+10 Ni-58 (n.p) 7.45D10 U-238 (n f) (Cd) 7.30E410 Np-237 (n,f) (Cd) 6.4BE+10 AVERAGE 7.01E+10 Otapule S Comparisons Sigma Avg Si9ma Avg Methodolooy ,, FERRET FERRET 4 (E > 1.0 MeV) 7.01E+10 7.44E+10 0.94 4 (E > 0.1 MeV) 2.45E+11 2.41E+11 1.02 dpa/sec 1.23E-10 1.22E-10 1.01 4 (E < 0.414 ev) 8.06E+10 8.64E+10 0.92 4 P' b 6-25 BWitne?%fe%r
Table 6-7_. Maximum Fast Neutron Exposure of Point Beach Unit 2 Beltline Circumferential Weld (SA-14941- -
+-(E > 1.0 MeV) [n/cm']
. Azimuthal E00 16 Anale 14.8 (FPY 32.0 EFPY- 48.0 EFPY 0 Degrees 1.59E+19 2.ESE+19 3.37E+19 15 Degrees 9.83E+18 1.68E+19 2.34E+19
- 30. Degrees 7.41E+10 1.30E+19 1.80E+19 45 Degrees 5.61E+18 1.24E+19 1.79E+19 4 (E > 0.1 MeV) [n/cm']
~
Azimuthal EOC 16 Anale 14.8 EFPY 32.0 EFPY 48.0 EFPY 0 Degrees 4.10E+19 6.50E+19 8.74E+19 15 Degrees 2.72E+19 4.64E+19 6.43E+19 h
- 30. Degrees: 1.90E+19 3.33E+19 4.64E+19 4
45 Degrees 1.63E+19 3.00E+19 4.44E+19 l i Iron Atom Displacemants Idoal
-Azimuthal E0C 16 Anale 14.8 EFPY' 32.0 EFPY- '48.0 EFPY O Degrees. 2.50E-02 4.12E-02. 5.62E ,
15 Degrees- 1.58E-02 2.68E-02: 3.68E-02
-30 Degrees .l.16E-02 2.01E-02 2.79E-02 45-Degrees 1.02E-02 1.91E-02 2.72E-02 6-26 r SW#n%""chv L
Table 6-8 Maximum Fast Neutron Exposure of Point Beach Unit 2 Intermediate Shell Foroino (123V500) e (E > 1.0 MeV) [n/cm') Azimuthal E0C 16 Anole 14.8 EFPY 32.0 EFPY 48.0 EFPY 0 Degrees 1.60E+19 2.92E+19 4.17E+19 15 Degrees 9.87E+18 1.87E+19 2.70E+19 30 Degrees 7.45E+18 1.32E+19 1.87E+19 45 Degrees 6.63E+18 1.26E+19 1.82E+19 e (E > 0.1 MeV) [n/cm'] Azimuthal EOC 16
= Anale 14.8 EFP_1 22 0 EFPY 48.0 Eff1 0 Degrees 4.12E+19 7.57E+19 1.08E420 15 Degrees 2.73E+19 5.13E+19 7.41E+19 30 Degrees 1.91E+19 3.40E+19 4.79E+19 45 Degrees 1.64E+19 3.i?F419 4.51E+19 Iron Atom Displacements Idoal Azimuthal EOC 16 Anale 14.8 EFPY 32,0 EFPY 48.0 EFPY 0 Degrees 2.52E-02 4.85E-02 6.73E-02 15 Degrees 1.59E-02 2.95E-02 4.24E-02 30 Degrees 1.17E-02 2.05E-02 2.88E-02 45 Degrees 1.03E-02 1.93E-02 2.76E-02 6-27 i B W !!s ef af4 = v
I Table 6-9. Maximum Fast Neutron Exposure of Point Beach Unit 2 Lower Shell Foraina (122W195) e (E > 1.0 MeV) [n/cm') Azimuthal E0C 16 Anale 14.8 EFPY 12 0 EFPY 48.0 EFPY 0 Degrees 1.59E+19 2.66E+19 3.69E+19 15 Degrees 9.82E+18 1.76E+19 2.50E+19 30 Degrees 7.40E418 1.29E+19 1.81E+19 45 Degrees 6.60E+18 1.24E+19 1.78E+19 e (E > 0.1 HeV) [n/cm'] Azimuthal EOC 16 Analp 14.8 EFPY 32.0 EFPY 48.0 EFP1 0 Degrees 4.10E+19 6.90E+19 9.59E+19 15 Degrees 2.72E+19 4.85E+19 6.89E+19 30 Degrees 1.90E+19 3.31+19 4.64E+19 45 Degrees 1.63E+19 3.08E+19 4.42E+19 Iron Atom Disolacements Idoal Azimuthal E0C 16 Anale 14.8 EFPY 32.0 EFP1 48.0 EFPY I O Degrees 2.50E-02 4.39E-02 6.20E-02 t 15 Degrees 1.58E-02 2.78E-02 3.94E-02 6 30 Degrees 1.16E-02 2.00E-02 2.79E-02 45 Degrees 1.02E-02 1.90E-02 2.71E-02 6-28 BW!!sefassi=v l
Figure 6-1. Reactor Geometry Showing a 45' R,# Sector 0
- i 1
I I v..,,,, u 9 o
* % ~ %*
~l so I
_, e ti,,,,,,,, % % ~ 1 su ~, Mlu + >>
'n u ,
% D Q~,%];~,,N9;p
/
, , ,,,9 -
Ug
's'e r
~t,,
's4s /
9o111111y,4 /% E 94
- Kg bizzs % e,*s
+ 1 mawtum= ~-n-O mity semam
_ _ - . - - - - - . m I 6-29 BWUEMW?a%r
f_ioure 6-2. Internal Surveillance Caosule Geometry (13, 23, or 33) JJillifffilllllJ y \
~ E I I
- nw wirw, - - 154.59 f Charpy e charpy \
Capsula Omntar - 1 158.35 nmwtan- fcanrq e ,charryf - 157.59
- in m u n nr12 -." ," y k w mient .
6-30 SW#senifv"!=v
Figure 6-3. Fast Neutron Fluence (E > 1.0 MeV) as a function of Azimuthal Anale at the Inner Radius of the Beltline Circumferenti3] Weld Neutron Fluence (n/cm2) 1.000E + 20 _ (% e 0 0 - W- a 1.000E + 19 _
.r N
_ W , 1.000E + 18 O 10 20 30 40 50 Azimuthal Angle (Degrees)
+ 14.8 EFPY C 32.0 EFPY O 48.0 EFPY 6-31 B W!!nn?afa m - - _ - - - - - . _ - - - - - - _ - _ . _ . - . - . - . _ . - - - - - - - - - - - - - - - - . - - - - - - - . _ - . - - - . - - . - - - . - - - - _ _ . - - - _ - - - . - - - - - - , - - - - _ - - - - - - _ _ _ _ _ _ _ _ _ _ - _ _ - _- _ -- _ _n--
Figtre 6 4. Fast Neutron Fluence (E > 0.1-MeV) as a Function of Azimuthal- _
- Anale at the Inner Radius of the Beltline Circumferential Weld Neutron Fluence (n/cm2)
-[ .
3 a-
+
1.000E + 19 _. + . . t 9 1.000E + 18 O 10 20 30 40 50. Azimuthal Angle (Degrees) , 4- 14.8 EFPY O 32.0 EFPY 0 48.0 EFPY 4 6-32 B W #2,1 Tea %wv
- = . - -.
Figure 6-5. Iron Atom Displacements [dpa] as a function of Azimuthal Angle at the Inner Radius of the Beltlige Circumf.tr.eBlial Weld Displacements (dpa) { 0.1 _
<R_.
C 0 + N c g - N N I 0.01 _. 0.001 O 10 20 30 40 50 Azimuthal Angle (Degrees)
+ 14.8 EFPY O 32.0 EFPY 0 48.0 EFPY 6-33 l B W!!s a af4Lwv
)
- 7. DISCUSSION Of CAPSULE RESULTS
- 11. Pre-Irradiation Procerty Data The weld metal and a metals were selected for inclusion in the surveillance program in accordat . uith the criteria in effect at the time the program was designed for Point Beach Unit 2. The applicable selection criterion was based on the unirradiated properties only. A review of the original unirradiated properties of the reactor vessel core beltline region materials indicated no significant dwiation from expected properties except in the case of the upptr-shelf properties of the weld metal which was below the current required 75 ft-l'os . Based on thi design end-of-service peak neutron fluence value 'it the 1/4T vessel wall location and the copper content of the weld metal, it was predicted that the end-of-service Charpy upper-shelf energy (USE) will be below 50 f t-lb.
7.2. Irradiated Proparty Data L2.1. Tensile Prgperties Table 7-1 compares irradiated properties from Capsule S with the unirradiated tensile propert as. At both room tenperaturc and elevated temperature, the ultimate and yield strength changes in the base metal as a result of irradiatien and the corresponding changes in ductility are within the limits observed for similar materials. Tiere is some strengthening, as indicated by increases in ultiste and yield strengths and decreases in ductility properties. All changes observed in the base metal are such as to be considered within acceptable limits. The changes, at both room temperature (70F) and 550F, in the properties of the base metal are not as large as those observed for the weld metal, indicating a lesser sensitivity of the base metal to irradiation damage. In either case, the changes in tensile properties are insignificant relative to the analysis of the reactor vessel materials at this time period in the reactor vessel service life. 7-1 j SW!!aWEPim l
i I . A comparison of the tensile data from previously evaluated capsules (Capsules V. T and R) with the corresponding data from the capsule reported in this report is i shown in Table 7-2. The currently reported capsule experienced a fluence that is approximately five times greater than the first capsule. The general behavior of the tensile properties as a function of neutron irradiation is an increase in both ultimate and yield strength and a decrease in ductility as measured by both total elongation and reduction of area. The most significant observation from these data is that the weld metal exhibited greater sensitivity to neutron radiation than the base metal. 7 2.2m Impact Properties The behavior of the Charpy V-notch impact data is more significant to the calculation of the reactor system's operating limitations. Table 7-3 compares the observed changes in irradiated Charpy impact properties with the predicted changes. The 30 ft-lb transition temperature shift for the base metals are in relatively good agreement with the value predicted using Regulatory Guide 1.99, Rev. 2 and the predicted value is conservative when the margin is added to the predicted value. It would be expected that thase values would exhibit good agreement when it is considered that the data used to develop Regulatory Guide 1.99, Rev. 2, was taken at the 30 ft-lb temperature. The transition temperature measurements at 30 ft-lbs for the weld metal is in good agreement with the predicted shift using Regulatory Guide 1.99, Revision 2 and the predicted value is also conservative. The shift be'.hg in good agreement with the predicted value which indicates that the estimating technique based on the Regulatory Guide 1.99, Rev. 2, are conservative for predicting the 30 ft-lb transition temperature shifts since the method requires that a margin be added to the calculated value to provide a conservative value. The data for the decrease in Charpy USE with irradiation showed good agreement with predicted values for both base metals. The weld metal decrease in Charpy USE was less than predicted. However, the poor comparison of the measured weld metal data with the predicted value is to be expected in view of the lack of data 7-2 SW##NE%v , 1
)
for medium , or high-copper-content materials at medium fluence values that were used to develop the estimating curves. A comparison of the Charpy impact data from the previously evaluated capsules from Point Beach Unit 2 with the corresponding data from the capsule reported in this report is shown in Table 7-4. The currently reported data experienced a fluence that is five times greater than the first capsule. The base metal exhibited transition temperature shifts at the 30 ft-lb levcls for the latest capsule that were similar in magnitude to those of the previous capsule. The corresponding data for the weld metal also showed no further increase at the 30 ft-lb level as compared to the previously reported increase at the 30 ft-lb level. These data confirm that the transition temperature for this weld metal may have reached a stabilized condition, or " saturation" as observed in the results of capsules evaluated by others.42 This be:.avior does not appear to be related to a further decrease in the upper-shelf energy and the change in the slope of the Charpy curve in the transition region. Both the base metal and the weld metal exhibited decreases in the upper-shelf l values similar to the previous capsules. The weld metal in this capsule exhibited a decrease similar to the weld metal in the previo;s capsule. This behavior of Charpy USE drop for the weid metal should not be considered indicative of a similar behavior of the upper-shelf region fracture toughness properties. The behavior indicates that other reactions may be taking place within the material besides simple neutron damage. Verification of this relationship must await the testing and evaluation of the data from compact i fracture toughness test specimens. Results from other surveillance capsules also indicate that RT NDT estimating curves have greater inaccuracies than originally thought. These inaccuracies are i a function of a number of parameters related to the basic data available at the time the estimatirig curves are established. These parameters may include inaccurate fluence values, inaccurate chemical composition values, and variations in data inurpretation. The change in the regulations requiring the shift measurement to be based on the 30 ft-lb value has minimized the errors that resulted from using the 30 ft-lb data base to predict the shift behavior l at 50 ft-lbs. l l 7-3 B W!!# NSPauvv
The design curves for predict'ing the shift will continue to be modified as more data become available; until that time, the design curves for predicting the RT NDT shift as given in Regulatory Guide 1.99, Revision 2, are considered adequate for predicting the RT NDT shift of those materials for which data are not available. These curves will be used to estabitsh the pressure-temperature operational limitations for the irradiated portions of the reactor vessel until the time that improved prediction curves are developed and approved. The relatively good agreement of the change in Charpy upper-shelf energy is in support of the accuracy of the prediction curves for medium copper content materials. However, for high copper content materials such as weld metal the predicted values may be too conservative. Although the prediction curves are conservative in that they generally predict a larger decrease in upper-shelf energy than is observed for a given fluence and copper content, the conservatism can unduly restrict the operational limitations. These data support the contention that the upper-shelf energy drop curves will have to be revised as more reliable data become available; until that time the design curves used to predict the decrease in upper-shelf energy of the controlling materials are considered conservative. 7.3. Reactor Vessel Fracture Touchnin An evaluation of the reactor vessel end-of life fracture toughness and the pressurized thermal shock criterion was made and the results are presented in Tables 7-5 and 7-6. The frccture toughness evaluation shows that the controlling weld metal may have a T/t wall ucation end-of-life RT NDT f 262F based on Regulatory Guide 1.99, Revisioa : including a margin of 56F. This predicted shift is excessive since i data from :urveillance capsules exhibit measured RT NDT significantly less for comparal'e rluence values. It is estimated that the end-of-life RT NDT shift will be significrotly less than the value predicted using Regulatory Guide 1.99, Revision 2. This reduced shift will permit the calculation of iess restrictive pressure-temperature operating limitations than if Regulatory Guide 1.99, Revision 2. was used. 7-4 BW!!nEVaitsm I
The pressurized thermal shock evaluation shows that the controlling acid metal may have an end of life Rip 73 of 283f based on the latest procedure for calculating the pressurized thermal shock trradiation effects. This predicted shif t is below the P15 screening criterion. It is also '>elieved to be excessive since data frem the Integrated Reactor Vessel Surveillance Program surveillance capsules exhibit measured RTNDT valves for comparable weld metals significantly less for comparable fluence values, it is estimated that the end of life RT HDT shif t wili be significantly less than the value calculated using the pressurized thermal shock procedure. These calculational procedures based on actual surveillance capsu'e data should provide justification for continued operation of the reactor vessel beyond the limits identified by the pressurized-thermal shock critertoa. An evaluatio' the reactor vessel end of-life upper shelf energy for each of ti. 'r is used in the fabrication was made and the results are presented in lable / .. This evaluation was made because the weld metals used to fabricate the reactor vessel are characterized by relatively low upper-shelf energy and high coppu contents; and, consequently, are expected to be sensitive to neutron radiation damage. Two methods were used to evaluate the radiation induced decrease in gper shelf energy; the method of Regulatory Guide 1.99, Revision 2, which is 'ne same procedure used in Revision 1, and the method presented in BAW. 1803" which was developed specifically to address the need for An estimating method for this class of weld netalc (Automatic Submerged-Arc: Mn-Ho Ni Wire /Lind9 80 Flux). The method of ,',:ylatory Guide 1.99, Revision 2, shows that Mn-Mo Ni/Linde 80 weld metals used in t ' fabrication of the beltline region of the reactor vessel will have an upper-shei f energy below 50 f t-lbs prior to the 32 EFPY design life based on the T/4 wall l' cation. Regulatory Guide 1.99 method predic.ts a decrease below 50 ft-lbs for the controlling weld metal at the vessel inside wall. However, based on surveillance data and the prediction techniques presented in BAW-1803, it is calculated that the controlling vessel weld metal upper-shelf energy will not decrease below 50 ft-lbs during the vessel design life. I 7-5 13W#f4faf46 l
)
l 1 Neutron flueg1_/20Alnti lhe neutron fluence analysis shows a sharp reduction in the neutron flux as the l result of improved fuel management schemes to lower core leakage. These new analysis calctlatec an end of life (32 EfPY) fluence value of 2.92 x 10 I9 n/cm' I (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 1.93 x 10 I9 r/cm' (E > 1 MeV). l l t 7-6 B W unEV?ilsfar
Jable 7-1. Comparison of Point Beach Unit 2. Capsula S. Tension Test Results Engm_Jemo Test 02y2 Led Temo Test UnirL 1rtSL Unirri
- Irrad **
Base Met al f.grging_{}}cjtt 122W191W1 fluence, 10 I9 n/cm' (E > 1 MeV) 0 3.47 0 3.47 Ultimate tensile strength, ksi 92.0 100.6 88.7 103.6 0.2% yield strength, ksi 70.8 79.7 63.0 81.0 Uniform elongation, % 14.0 9.3 14.1 8.2 Total elongation, % 27.1 22.8 27.7 20.2 Reduction of area, % 70.5 67.3 71.1 65.4 Base Metal foraina (Heat 123V500VAll fluence, 10 I9 n/cm' ([ > 1 MeV) 0 3.47 0 3.47 Ultimate tensile strength, ksi 80.3 94.7 78.1 95.1 0.2% yield strength, ksi 55.5 74.3 47.1 70.1 Uniform elongation, % 17.7 9.7 15.5 8.9 Total elongation, % 32.0 24.3 28.2 22.3 Reduction of area, % 72.5 70.0 71.7 64.5 Weld tielal (WF-19.11 fluence, 10 I9 n/cm' (E > 1 MeV) 0 3.47 0 3.47 Ultimate tensile strength, ksi 86.8 110.6 84.9 102.0 0.2% yield strength, ksi 71.9 96.9 63.1 83.7 Uniform elongation, % 16.2 10.3 14.7 6.1 Total elongation, % 27.1 21.7 22.9 15.4 Reduction of area, % 64.1 54.8 54.6 43.4
- Test temperature 600f.
** Test temperature - 550f.
7-7 S W !!?vnVial4 mr
Table 7-2. Sumary of Point Beach Unit 2 Reactor Vessel Surveillance Capsules Tensile Test Results Ductility. % Strenath. ksi Total Cap. Test Material I.D. Fjgence. 10 n/cm' Temp, F Ultimate C I*) Yield C I*I Elon. d I*) Reduction of Area d(* } 70.8 27.1 71 -- 92.0 Base metal -
- 00. RT --
300 88.9 68.0 -- 21.9 -- 71 -- forging (Heat 88.7 63.0 -- 27.7 -- 72 -- 122W195VA1) 500 -- V 6.1" 88 97.4 5.9 75.3 6.4 21.5 -20.7 6' - 4.2 89.2 0.3 66.4 -2.4 20.4 - 6.8 63 - 2.8 300 550* 100.2 13.0 74,9 18.9 19.6 -29.2 68 - 5.6 12.3 82.8 16.9 23.5 -13.3 73 2.8 T 8.4" 70 103.3 70 - 1.4 300 87.9 - 1.1 68.5 0.7 20.1 - 8.2 550 91.8 3.5 67.8 7.6 22.0 .z0.6 70 - 2.8 21.5" 81 94.7 2.9 78.4 10.7 22.5 -17.0 67 - 5.6 7 R 7.2 20.6 - 5.9 68 - 4.2
" 300 89.6 0.8 /2.9 550 90.6 2.1 68.2 8.3 20.6 -25.6 69 - 4.2 34.7 70 100.6 9.3 79.7 12.6 22.8 -15.9 67 - 5.6 1 S
300 93.1 4.7 73.0 7.4 20.6 - 5.9 67 - 5.6 l 103.6 16.8 81.0 28.6 20.2 -27.1 65 - 9.7 550 55.5 32.0 -- 73 -- Base metal -
- 00. RT 80.3 -- --
300 81.5 -- 59.5 -- 25.5 -- 71 - forging (Heat 47.1 28.2 -- 72 - 600 78.1 -- -- 123V500VA1) D V 6.1" 88 91.2 +13.6 67.7 22.0 23.2 -27.5 71 - 2. 7
$ 300 80.7 - 1.0 57.4 -3.5 22.3 -12.5 70
.4 g 550 C6.8 +23.9 74.1 57.3 20.9 -25.9 68 - 5.6 E l g$
R 8.6 -22.5 75 2.7 T 8. 4* 70 83.8 + 4.4 60.3 24.8 8?, 77.3 - 5.2 60.1 1.0 24.1 - 5.5 75 5.6 (); 300 82.8 + 6.0 58.5 24.2 22.7 -19.5 71 - 1.4 550
Table 7-2. Summary of Point Beach Unit 2 Reactor Vessel Surveillance Capsules Tensile Test Results (Cont'd) Ductility. % Strenoth. Asi Total Reduction Cap. Test 6%I *I Fjgence,2 Temp, F Ultimate a%I *I Yield 6%I *I Elon. 6%I *I of Area Material I.D. 10 n/cm 86.6 + 7.8 63.9 15.1 25.2 -21.3 72 - 1.4 R 21.5" 81 1.4 300 79.5 - 2.5 58.1 -2.4 22.8 -10.6 72 80.3 + 2.8 56.1 19.1 22.5 -20.2 73 1.4 550 34.7 70 94.7 +17.9 74.3 33.9 24.3 -24.1 70 - 4.1 S 2.8 300 96.5 +18.4 80.5 35.3 22.4 -12.2 73 ! 550 95.1 +21.8 70,1 48.8 22.3 -20.9 65 - 9.7 d6.8 71.9 -- 27.1 -- 64 -- Weld metal - 00. RT 64 300 80.8 -- 65.7 -- 23.4 -- (WF-193) 63.1 22.9 55 -- 84.9 -- -- 600 --
+21.9 88.6 +23.2 21.0 -22.5 53 -17.2 E> V 6. l* 88 105.8
-24.8 37 -42.2
+20.2 80.2 +22.1 17.6 l 300 97.1 550 99.7 +17.4 81.2 +28.7 16.3 -28.8 44 -20.0 T 8. 4"" --- No weld metal tension test specimens in this capsule.
21.5" 81 107.0 +23.3 95.3 +32.5 20.9 -22.9 60 - 6.3 R 99.2 +22.8 86.6 +31.E 18.8 -19.7 52 -18.8 300 550 99.8 +17.6 83.1 +31.7 16.7 -27.1 54 - 1.8 70 110.6 +27.4 96.9 e34.8 21.7 -19.9 55 -14.1 5 34.7 -20.3 300 103.0 +27.5 89.4 +36.1 18.3 -21.8 51 G3 102.0 +20.1 83.7 +32.6 15.4 -32.8 43 -21.8 550 Eg E3 EE
$$ " Change relative to unirradiated.
no CE *All 550F data compared to 600F unirradiated data. g 5 " Prior capsule fluence value: are the calculated values per WCAP-LGS's." 5
l Table 7-3. Observed Vs. Predicted Changes for Point Beach g it 2,* Capsule S, Irradiated Charpy Impact Properties - 3.47 x 10 n/cm (E > 1 MeV) Predicted - RG 1.99/Z(b) Observed With g Witho'gy Unirrad. Irrad. Diff. Margin Margin Material. Increase in 30 ft-lb Trans. Temo.. F
-4 76 77 111 Base metal forging (Heat 123V500VAI) -80 47 41 75 Base metal forging (Heat 122W195VA1) -45 2 N.D. N.D. 41 75 Heat-affected zone (Heat 122W195VAI) -80 0 231 231 231 287 Weld metal (WF-193) 194 145 135 169 Correlation material (HSST PL-02) +49 f Decrease in Charoy USE. ft-lb IE3 17 139 N.A.
Base metal forging (Heat 123V500VA1) 180 134 11 119 N.A. Base metal forging (Heat 122WI95VAI) 145 N.D. N.D. --- N.A. 84 Heat-affected zone (Heat 122W195 val) 66 44 22 32 N.A. Weld metal (WF-193) 84 40 84 N.A. Correlation material (HSST PL-02) 124 k$ Mean value per Regulatory Guide 1.99, Revision 2, May 1988."
% "Mean value per Regulatory Guide 1.99, Revision 2, May 1988," plus margin.
g($ N.A. - Not applicable. k N.D. - Not determined. r
Table 7-4. Summary cf Point Beach Unit 2 Reactor Vessel Surveillance Capsules Ch&roy impact Test Results Upper-Shelf Energy, Transition Temperature Increase. F ,_, ft-lb ,_ Predicted 30 ft-lb "# Observed Predicted"I F 30 ft-lb 30 ft-lb USE AUSE Material 10]gence,r n/cm Observed Delta W/0 Margin W/ Margin USE AUSE 50 84 180 0 151 29 Base metal forging 6.l* -50 30 30 55 89 180 0 149 31
- 8. 4" -50 (Heat 123V500VA1) 70 104 180 0 142 38 21.5" -10 70 17 139 41
-4 76 77 111 163 34.7 61 135 10 128 17 Base metal forging 6.1" -35 10 27 0 126 19 8.(" -28 17 29 44 145 (Heat 122W195VA1) 37 71 140 5 122 23 21.5" -10 35 11 119 26 47 41 75 134 34.7 2 61 84 0 74 10 Heat-affected zone 6.1" N.D. N.D. 27 78 6 73 11 44
? (Heat 122W195VA1) 8.4* 30 110 29 84 0 70 14 37 71 C 21.5" 110 190 34.7 N.D. N.D. 41 75 --- -- --- --
165 150 206 42 24 43 23 Weld metal (WF-193) 6.l* 165 166 222 56 10 41 25
- 8. 4" 150 150 36 30 211 267 48 18 21.5" 235 235 44 22 32 34 34.7 231 231 231 287 122 94 30 99 25 Correlation material 6.1" 139 90 88 97 27
- 8. 4" 154 105 97 131 49 15 IU (HSST plate 02) 38 26 90 34 157
$ 21.5" 200 151 123 40 84 40 145 135 169 84 RR 34.7 194 "Per RG 1.99, Revision 2, May 1988.
[
- Prior capsule fluence values are the calculated values per WCAP-10638."
t 5 N.D. - Not determined.
l n l o ei 2 Wt 0 8 8 2 A a 4 0 6 6 / 4 c 1 1 1 2 g t *F / o i TL n , U , h T
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- i S o( ( o( r 6 e e .
6 l rg t l t o h u ? 4 p 7 oe e d l d d t n c l C 0 p tR t l tl el ll a o a B 0 A e c l l e h e l e l i s
- l 1 ae ta e eW SW eW u t l - - t l
en d h h 9 c a l W o b B . e e i a M A T a Ri l t wSr r 1 emu nm. r u S . ru m R r s r e I t S r B M
=
t e e 2 c ec ec e r e t a i n e r B e t n o w 2 r 0i t r ni wr oi P P M I F F e A I l NC I C LC " * '
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N EREta NE R $ a a EJ k ,
l i Table 7-6 Evaluation of Reactor Vessel Pressurized Thermal Shock Criterion for 32 EFPY - Point Beach Unit 2 l Estimat ed Paterial DL Cheetcal Fluenc e** PTS Evaluation" M derial Descriction N Instae Surface Initial Chemistry (OL $creening Heat Factor Margin RT,,, Criterion Reactor Vessel Beltilne Region location Number Type Copper Nidei n/cm 2 RT. 48 125 270 f 123V500VAX 1A DS C1.2 0.09 0.70 2.92E+19 (+ 3)* SR Intermediate Shell 48 90 270 122w195 val SA508 C1.2 0.05 0.71 2.66t+19 (+ 3)" 31 tcwer Shell l 66 194 300 CE/SAW 0.27 0.90 4.67[+18 (-56)" 233 Nozzle Belt to Inters. Shell #ct ident. > Circtra. 'deld (100%) 300 ASA/Linde 80 0.24 0.60 2.56t+19 ( 0 )" 173 66 271 Inters. Shell to tower Shell SA.1484 Cirtum. Weld (100%) N/A N/A N/A N/A N/A N/A >l 00(+16 N/A tower Shell to Dutchman Not Ident. C[/$AW Circus. Weld (100%) 7 "Per 10CfR50, 4ction 50.61 (Pressurized Shock Criterios)."
'" Materials chemical cc=cositions per BAW-2150. December 1990.**
* "Per BAW.10046P, Marth 1976.
**Per Section 6 of this report.
N/A - Not applicable N EE RR EE 52 on
Table 7-7. Evaluation of Reactor Vessel End-of-Ufe (32 EFPY) Upper-Shelf Energy - Point Beach Unit 2 f Material Esticated I Cheetcal Ect Fluence
- Estimated EDL-USE Estimated ECL-t:5E Per PG 1.99/Z* Fer BAW-1803*
Composition. Estimated ETPt Material Descrittice w/o" Inside Y/4 Wall inttfal surface location Usi inside T/4 Wall Inside T/4 Wall _$Q ff-lt u t T/4 _ Feactor vessel Heat surface Location Seeface Locatio-1 R; I.99/2 BAW - IP'w) Type Copper Nickel n/ca' m/cr' ft-lbs Eeltline Region Location Muster 142 efa N/A >32 N/A 0.70 2.92E+19 1.93E+19 180 140 123V500 val SA508 Cl.2 0.C9 intermediate Shell >32 N/A 120 122 N/A N/A 0.05 0.71 2.66E+19 1.76E+19 145 tower Shell IZZW195fAl 5ASC8 C1.2 65 68 N/A N/A >32 N/A 0.27 0.90 4.67E+18 3.CAE+18 100 Norrle Belt to Ir.ters. Not ident. CE/5AW shell Cirtue. Weld (100%) 54 -5 >32 38 41 53 5A-1484 ASA/linde 80 0.24 0.60 2.56E+19 1.69E-19 (70)" Inters. Shell to tower Shell Circus. Weld (100%) N/A N/A N/A 4A N/A N/A N/A M/A >].00E+16 >l.0CE+16 i4/A Lower Shell to Dutchman Not Ident. CE/SAW Ctrcum. Weld (100%)
*Per Segulatory Guide 1.99 Revision 2. May 1988/*
5 *Fer BAW.1803, Revis*on 1. May 1991." l
" Materials chemical compositions per BAW-2150. Decceber 1990."
*Fer Section 6 of this report.
M/A - Not Applicable E ER E
$u a
15 e n 4 mm-i-
_ _m_ . _ . _ . _ . _ - . . _ _ . _ . . . . _ _ _ _ _ _ _ _ _ _ _ . _ , 1 l I i l
- 8.
SUMMARY
OF RESULTS ; i The analysis of the reactor vessel material contained in the fourth surveillance capsule (Capsule S) removed for evaluation as part of the Point Beach Nuclear - Plant Unit 2. Reactor Vessel Surveillance Program, led to the following conclu- j sions:
- 1. The capsule received-an average fast fluence of 3.47 x 10 I9 n/cm' (E > i 1.0 MeV). The predicted peak fast fluence for the reactor vessel T/4 logttonattheendofthesixteenthfuelcycle(14.8EFPY)is1.06x 10 n/cm' (E > 1 MeV).
- 2. The fast fluence of 3.47 x 10 I9 n/cm' (E > 1 MeV) increased the RT of the capsule reactor vessel core region shell materials a maximum"Of- !
231F. ; i
- 3. Based on the calculated fast flux at the vessel wall, an 80% load ~ -
factor and the planned fuel management, the projected fast fluence that > the Point Beach Unit 2 reactor pressure vessel insig surface will , receive in 40 calendar year's operation is 2.92 x 10 n/cm' (E > 1 MeV).
- 4. The increase in the RT for the shell forging material was in good agreement with that pr00Icted by the currently used design curves of RT versus fluence (i.e., Regulatory Guide 1.99, Revision i.;, and the pr00Ictiontechniquesareconservative. -
- 5. The increase in the RT NDT for the weld metal was in good agreement with -
that predicted. i 6. The weld _ metal _ upper-shelf energy at the T/4 location, based on an evaluation of surveillance capsule results, will not decrease below 50 ; i ft-lbs prior to 32 EFPY.
- 7. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in weld metal Charpy upper-shelf properties due to irradiation /are conservative. ,
- 8. The analysis of the neutron dosimeters demonstrated that the analytical I
. techniques ured to predict the neutron flux and fluence were accurate. I 8-1 SW#MMhv ' 1-~~.- _.~..,.-m-.3 h.. .,.-,.-,..,,,-.r.,.,-,_.,,.,',.._m,_.., .,.,--mm,._,-.m,.. ,,,,.,w..,,,c.,, , _ - ....w... , , . . . . - . ,,-1.-.- - . . .
1 1 9 .. SURVEILLANCE CAPSVLE REMOVAL SCHEDULE Dased on the postirradiation test results of Capsule $ the following schedule is recomended for the examination of the remaining capsules in the Point Beach Nuclear Plant Unit 2 RVSP: EvaluationSchedule(a) Capsule Location Lead Removal ExpectedCapgu Identification Capsules (gf Factor (bI Time fluence (n/cm )lg) N 33* 2.17 Standby Not defined P 23* 2.35 Cycle 20 4.20 x 10 I9 ("I Reference reactor vessel irradiation locations, figure 3-1. (b)The factor by which the capsule fluence leads the vessels maximum inner wall fluence. IC) Based on current capsule analysis and BAW-1543, Rev. 3.I4 l l 1 9-1 SW##eF5ifsbur
- 10. CERTIFICATION The specimens were tested, and the data obtained from Wiscotisin Electric Power Company Point Beach Nuclear Plant Unit 2, reactor vessel surveillance Capsule S were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10CFR50, Appendixes G and H.
()
~
<< 7' ff/bq/99)
A. L. Lowe, Jr., P.O 7 ' Date Project Technical Manager This report has been reviewed for technical content and accuracy.
] b /<n Y V . be Y - 49{
M. J/Devan f Date M&SA Unit
+
Verification of independent review.
'safr;M'E-i b Y N'O $l K. E. Moore' Manager , Date M&SA Unit This report is approved for release.
e ~ Wo*** $ V '?/
- 1. *.. Baldwin
. Date Program Manager 10-1 BW#stfaf45=v
l APPENDIX A Reactor Vessel Surveillance Program Background Data and information A1 S Wi!#eff"os =v
- l. Material 591eclin_ Data The date used to select the materials for the specimens in the surveillance program, in accordance with E185-66, are shown in Table A 1. The locations of these materials within the reactor vessel are rhown in figure A-1.
- 2. Definition of_ Beltline Re.gion 1he beltline region of Point Beach Unit 2 was defined in accordahce with the definition given in A51M E185 82.
- 3. Cgpsule Identification The capsules used in the Point Beach Unit 2 surveillance program are identified F
belo.v by identification, location, and lead factor. , Lead factors (b) Capsule Capsul a) Tr~eviou Current identification location Analysis CI Analysis V 13* 3.37 N.A. R 13" 3.37 N.A. I 21" 1.94 N.A. P 23' l.94 2.35 5 33' l.79 2.17 N 33' l.79 2.17 _
Reference irradiation capsule locations as shawn in figure A-3.
"dThe f actor by which the capsule fluence leads the vessels maximum inner wall fluence.
'" Previous analysis as reported in WCAP-9635.
A-2 B W !!?vE?s % r
{ A d cilment.. Ear _Jur.yetilance Cagiult Cjtplules R. S. and V Material Charov 11D1111 FA forging mt.terial Heat 122W195 val 12 3 3 Heat 123V500VA1 12 3 3 Weld metal: WF-193 8 3 3 HAZ: Heat 122W195 val 8 - - Coerelation Monitor 8 - - Capsules N. P. aD.d I Materia 1 [hitrAy lentill F01 Forging material Heat 122W195 val 12 5 5 Heat 123V500 val 12 4 4 Weld metal: WF-193 8 - - HAZ: Heat 122W195 val 8 - - Correlation monitor 8 - - A-3 IBWitanfa!?our
figure A-1. !ocation and Identification of Materials Used in the fabrication of Point Beach Unit 2 Reactor Pressurf.,hssel J r mS i ( Ns h M
/
a 4 4,
= CE Weld i
Icone 4
= Intermediate Shell(Forging)
' i 123V500VA1 I
C 15.08 i Weld SA-1484 I 144' A
=: Lower Shell (Forging) g- 22W195VA 1 g
130.84' I I _ L
)
A-4 13W##J' ffs!4 =v
Figure A-2. Location of Surveillance Capsule Irradiation Sites in Point Beach Unit 2 Reactor Vessel (lead factors for the Capsules Shown in Parenthes,.gs_ are for the Oriainal fuel Manaaement) (3 37) R 270' REACTOR VESSEL (1.94) P (1.79) N -
,/
/ Y/, / -
,x THERMAL SHIELD
.l CAPS'JLE (TYP) , - 10'
~ 10*
- 57' .
180' I - -' O. T - . t : m I I\ k S (1.79)
~ T (1.94)
V (3.37) A-5 BW###42 =v
APPENDIX 0 Pre-Irradiation Tensile Data' B-1 BW##a?!Lr
Table 8-1. Tensile Properties of Unirradiattd Base Metal forgipa Material. Heat 123V500 val Test Specin en Temp. Strgnath, ost _[lgngat i on . % ___ Reduction of No. f ligld_. Ullimal_t Uniform ig1Al Area m%
-- Room Bi,050 82,100 17.2 'J J 72.7
-- Room 53,800 78,400 18.1 i) 0 72.3
- 300 63,450 9t. 500 13.3 25 v 73.4
- 300 55,500 78,400 15.1 6.e 68.0
-- 600 44,500 77.300 14.4 c,' . ; 5' . 3
-- 600 49,700 78,950 16.5 28.5 71.1 Table B-2. Tensile Properties of Unitradiated Base Metal foraina Mdyrial. Heat IR4195 val Test Specimen Temp, Strenath, osi _{lpnaation,% Reduction of No. F Yie1d Llll_brde W11f.Etu! 19141 _ Area. %
Room 71,100 92,500 14.1 27.5 70.5 Room 70,500 91,450 13.9 26.6 70.5 300 71,500 92,150 10.4 21.6 72.3 300 64,500 85,550 11.0 22.2 69.6 600 64,350 90,650 13.9 27.2 72.3 600 61,600 86,800 14.2 28.1 71.9 Table D-3. Tensile Properties of Unirradiated Weld Metal. WF-193 Test Specimen Temp, St renath. osi _Elongt11pn. % Reduction of No. F U.gld_. UlliDutig Uniform 19111 Area. %
-- Room 71,300 86,150 16.4 27.2 64.1
-- Room 72,500 87,800 15.9 27.0 64.1 300 66,800 82,000 11.4 22.8 63.1
-- 300 64,600 79,550 13.9 24.0 64.1
-- 600 63,750 85,750 13.7 21.9 54.6 600 62,500 84,000 15.6 23.9 54.6 02 S W e fvef af48 m v
) APPENDIX C Pre irradiation Charpy impact Data'
~
B W UEEY?! L r 1
Table C 1. Charpy Impact Data from Unirradiated Base Metal forginglaterial. Heat 121H191 val , Absorbed lateral Specimen Test Temp., Energy, Expansion, fracture, No. f ft lb 10 in. %
-- -100 60.5 44 33
-- 100 10.0 7 5
-100 13.5 10 9
-- 60 32.5 30 20
- 60 29.5 24 18
-- - 60 28.0 20 17
-- - 40 3.0 2 3
-- 40 12.0 10 9
- 40 52.0 42 27
- 25 45.5 31 25
- 25 13.5 10 13
- 25 71.0 59 43 10 86.0 69 53
-- 10 90.5 70 53 10 87.0 73 48
-- 40 84.5 64 52 40 90.0 70 56 40 85.5 70 48 110 150.0 94 100
-- 110 125.0 85 79 110 147.0 94 100 210 145.5 93 100 210 147.0 92 100
-- 210 147.0 93 100 C-2 BW!!?a'af46v
Table C-2. tharpy impact Data from Untrradiated Base Metal Foraina Materigl. Heat 123V5.MyAl Absorbed lateral Specimen Test lemp., Ener9y, Expansion, fracture, No. F ft lb 10' in. 7.
-150 9.0 8 5
-150 70.0 63 42
-- -150 5.5 4 5
-- 100 16.5 16 13
-- 100 17.0 15 13
-- 100 14.5 13 9
-- 75 10.0 7 5
-- 75 6.0 4 5 75 54.0 50 30 50 83.0 70 47
-- 50 79.5 66 42
-- - 50 77.0 68 46
- 25 42,0 38 27
-- - 25 94.5 83 55
-- - 25 110.0 79 66
-- 40 100.0 70 63
-- 40 141.0 97 77
-- 40 122.0 90 71 110 185.0 90 100 110 180.5 91 100 110 172.5 89 100 210 160.0 85 100
-- 210 170,0 90 100 210 158.0 94 100 C-3 BW!!nEY5%v
lable C-3. Charpy impact Data from Unirradiated Base Mett.1 forging Heat Affected Zone Haterial Heat 122W195 val Absorbed lateral Specimen lest Temp., Energy. Expansion, fracture, No. F ft-lb 10' in. %
-100 18.0 14 13
-- -100 16.0 15 13 100 28.5 18 23
-- - 50 32.5 26 29
-- 50 24.0 20 33
- 50 60.0 42 40
-- 40 91.0 66 71
-- 40 70.0 50 56
-- 40 66.0 50 b3
-- 40 38.0 38 54 110 56.0 61 99
-- 110 57.0 54 85
-- 210 72.5 66 100
-- 210 69.0 60 100 210 188.5 80 100 210 74.0 59 100 C-4 S W # #va ? # 2=v
Table C-4. Charpy impact Data from Unirradiated Correlation Monitor Material. SA531 Grade B. Class 1. lient A-1195-1 Absorbed lateral Specimen Test Temp., Energy, Expansion, Fracture. No. f ft lb 10 in . %
-- - 50 5.0 3 9 50 5.0 5 9
-- 50 3.0 4 9
- 20 6.5 6 9
- 20 9.0 10 13 20 6.0 9 13 10 12.0 15 23 10 14.5 14 23 10 13.5 14 23 40 22.0 23 33
-- 40 36.0 32 29 40 35.0 32 29 85 58.5 51 43 85 41.5 42 41
-- 85 52.0 45 42 110 82.5 60 58 110 85.5 71 67 110 63.5 54 55 160 108.5 72 84 160 81.0 69 85 160 109.0 79 87 210 117.0 84 98 210 115.0 88 98
-- 210 121.0 67 100 300 125.0 87 100 300 117.5 83 100 300 127.0 84 100 C-5 B W # 4 e N "o d ! = v
lable C-5. Chirov implet Data from Unirradiated Weld Metal. WF-193 Absorbed lateral Specimen Test Temp., Energy. Expansion, fracture, No. F ft-lb 10~* in. %
~
100 7.0 6 9
-- -100 18.5 16 21
-- -100 6.5 5 9
- 50 28.0 28 34
- - 50 18.5 18 27
- 25 24.5 22 29
-- 25 29.0 27 31 25 30.0 27 34 10 34.5 31 43
-- 10 25.0 27 40
-- 10 35.0 36 41
-- 40 49.0 46 64 40 39.5 39 56
-- 40 40.0 40 56 110 65.0 65 100 110 58.5 58 85
-- 110 55.5 60 87
?l0 66.0 66 100 210 65.0 67 100 210 64.5 68 100 13W!!?ManLuv
I figure C 1. Charpy impact Data from Unirradiated Base Metal foraina Material . Heat 112WIPJAL___ n; , ,: ,: i i A -: i n 5 e , ~ 6 5 - e x 4 i i - y ~ I'
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figure 0-2. Charpy impact Data from Unirradiated Base Metal foraina Material. Heat 123V500 val
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-200 -bo 0 %0 200 5W 'N W' Test le-cerature, F C-8 HIIIR& W NLICiEA 9 13 W5ERVICE COMPANY
Figure C-3, Charpy impact Data from Unirradiated Base Metal Forging Material. Heat-Affected Zone. Heat 122W195 val 1
- , i. .- i i i l -
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Fiatre C-4. Charov impact Data From Unirradiated Weld Metal. WF-193 1a0 , r: - i; i i i
", 7 c, .
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! -RT, 7 2 70 e e a -
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, t t f i i 0 300 400 SGO 600
-100 0- !00 200 Test Temperature, F C-10 rs aswsuctran G WSERVICE COMPANY
Figure C 5, Charpy impact Data From Unirradiated Correlatw.. Monitor Material. SA533. Grade B. Class 1. Heat A-1195-1 x'
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-100 0 100 200 Test Temoer:ture, F C-11 95 B&W NUCLEAR ID WSERVICE COMPANY e
l l l l APP'ENDIX 0 Tension Test Stress-Strain Curves D-1 IBW!!nn?afawv
Figure D-1. Tension Test Stress-Strain Curve for Base Metal Forging Heal 122W195VA1. Specimen No. E8. Tested at 70F 140. Strength 888-Yield: 79727. UTS: 100639 800. 112, _. - 0 M . 700. Q. M - g -- 2 fe 600. e - ~ b b m . 500. m O o E ._. 400. E 3g, S O o e
. 300. .E c c W W 2e. .
-- 200.
_ 100, g, i ! i f i i i l i e, 0.00 .06 .12 .10 .24 .30 Engineering Strot. Figure D-2. Tension Test Stress-Strain Curve for Ba:e Metal Forging Heat 122W195VA1. Specimen No. E9. Tested at 300F 110. .. Strength Yield: 73048. . 70a.
~ UTS: 93059, es. 600.
X a 2
/g I 500. ,
to a
$ OS. y t- L.
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0.00 .26 .12 .la . 24 . 30 Engineering Stroin D-2 ES B&W NUCLEAR 13 SERVICE COMPANY
l Figure D-3. Tension Test Stress-Strain Curve for Base Metal Forging Heat 122W195VA1. Specimen No. E7. Tested at 550F f i40. Strength - **** Yield: 81047. UTS: 103631 _ 800. I12. .. D
. 700. Q.
/ 600. d f, ed. _. =,
A see. h m m O O
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. 100.
O.
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0.00 .00 .12 .30 .24 . 30 Engineering Stroin Figure D 4. Tension Test Stress-Strain Curve for Base Metal Forging Heat 123V500VA1. Specimen No. V9. Tested at 70F 110. Strength Yield: 74339. . 700. UTS: 94655. BB. 4 _ 600. 5E Y . 3
. sea.
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i i figure D-5. Tension Test Stress-Strain Curve for Base Metal Forging Heat 123V500VA1. Specimen No. V8. Tested at 300F l l e. Strength Yield: 80512. .. 7ee.
-.UTS: 96450.
DU-
/
~:
_ coe. o 0. Y - y g sea. , y es. E __ 4em,
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C g 44 .- 300, *C e e i 200. ic w __
- 22. .
_ 100.
- c. . I . t , i ,
i , e, 0 00 .0e .22 .30 . 24 .30 Engineering Strain Figure D-6. Tension Test Stress-Strain Curve for Base Metal Forging Heat 123V500VA1. Specimen No. V7. Tested at 550F 110. Strength Yield: 70065. . 700. UTS: 95054. f-- BB- - __ eee. 5 $ x . 3
- _ 500.
5
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- 22. -
_ 100.
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___1 - I - t . c. 2.20 .De . 2 .IB .24 . 30 Engineering Stroin D-4 B&W Nt. CLEAR WSERWCE COMPANY
Figure D-7.- Tension Test Stress-Strain Curve for Weld Metal ,
--WF-193. Specimen No. W9. Tested at 70F 1 < e.
Strength - eso. Yield: 96850. -
- UTS: 110613 .
_ see. 112. . o M +
. 7 ue, o_
-x 2
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i
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. s es.
- m. . I a l . l . l . . g, ,
0.00 .me . 12 .2s . 24 .as
- Engineering Stroin l Figure D 8. Tension Test Stress-Strain curve for Weld Metal WF-193. Specimen No. W7. Tested at 300F 14e. '
Strenoth . ' Yleid: '89402. . ome.
- UTS: 103033 ,
_ eme. 312. o
- 7ee* n.
M' .. 3
=g, 2 0-j e 4.- g b: 2 m
- _ sea, g O- ,
D C ,
- 420.
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e
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- e. ee .e4 .ma .22 .2e .2e Engineering Strain D-5 BWMRiLr
Figura 0-9. Tension Test Stress-Strain Curve for Weld Metal WF-193. Specimen No. W8. Tested at 550F 140. Strength
- 900.
Yleid: 83697. UTS: 102036
~ 8
- 112 .
U Ui - 700. y ;; _%
- 600. m' y 04* -
a g
* . s00. G CD (P C
4 e0. .E iv so. G D C 300. .E
'b 05 C C Ld
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. 100.
O. i l i 1 . l . I i g,
-0.00 . 04 .00 . 32 .16 .20 Engineering Stroin l
l l l D-6 rs s **ssw Nuct. tan ( M W sEnvsctco w Amy
l
=
APPENDIX E References E-1 _B_ _W sV X A Fei M 2 = v
)
- l. S. E. Yanichko, eLit1. , " Wisconsin Michigan Power Company and the Wisconsin Electric Power Company Point Beach Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-7712, June 1971.
- 2. J. S. Perrin, g_1_itl., " Point Beach Nuclear Plant Unit No. 2 Pressure Vessel Surveillance Program: Evaluation of Capsule V," Battelle Memorial Institute Report, June 1975.
- 3. J. A. Davidson, et al., " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9331, August 19!8.
- 4. S. E. Yanichko, et al., " Analysis of Capsule R from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9635, December 1979.
- 5. Code of Federal Regul ation, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix H, Reactor Vessel Material Surveillance Program Requirements.
- 6. Code of Federal Regulation, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix G, Fracture Toughness Requirements.
- 7. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure (G-2000).
- 8. C. E. Childress, Fabrication hitory of the First Two 12-in.-Thick ASTM A-533 Grade B, Class 1 Steel Plates of the Heavy Section Steel Technology Program, Documentary Report 1, ORNL-4313, February 1969.
- 9. ASTM Designation A370-68, " Methods and Definitions for Mechanical Testing of Steel Products," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 10. ASTM Designation E23-72, " Method for Notched Bar impact Testing of Metallic Materials." in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
E-2 BWunnVi % r
1 1
'll . Standardized Specimens for Certification of Charpy Impact Specimens from the Army Materials and Mechanics-Research Center, Watertown, Mass. 02172,-
d Attn: DRXHR-MQ.
- 12. : ASTM Designation A370-77, " Methods and Definitions for Mechanical Testing of. Steel Products," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA. -
- 13. ASTM Designation E23-86, " Methods for Notched Bar Impact Testing of
, Metallic Materials," in ASTM Standards, American Society for Testing and
' Materials, Philadelphia, PA.
- 14. S-. Fyfitch, L. B. Gross and A. L. Lowe, Jr., Master Integrated Reactor Vessel Surveillance Program, BAW-1543. Rev. 3, Babcock & Wilcox, Lynchburg, Virginia,-September 1989.
- 15. ASTM Designation E185-XX (to be rueased), Recommended Practice for _
Surveillance Tests for helear Reactor Vessels, in ASTM Standards, American E Society for Testing and Materials, Philadelphia, PA.
- 16. ASTM Designation E853-87, " Standard Practice for Analysis and Interpreta-tion Light-Water Reactor Surveillance Results," in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa.,1989.
- 17. ASTM Designation E693-79, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (DPA)," in ASTM Standards, American -Society for Testing and Materials, Philadelphia, PA.
- 18. U.S. Nuclear- Regulatory Commission, Radiation Damage to Reactor Vessel Material, ReaulatorY Guide 1.99 Revision P, May 1988.
- 19. R. G. Soltesz, et al., " Nuclear Rocket Shielding Methods, Modification,
. Upu.. ting, and Input Data Preparation - Volume 5 - Two Dimensional Discrete - Ordinates Transport Technique," WANL-PR-(LL)-034, August 1970.
- 20. SAILOR RSIC DATA LIBRARY COLLECTION DLC-76, " Coupled Shelf-Shielded, 47 Neutron, 20 Gamma. Ray, P3, Cross Section Library for Light Water Reactors.
4
- 21. S. L. Anderson and A. H. Fero, " Reactor Cavity Neutron Measurement Program
' for Wisconsin Electric Power Company Point Beach Unit 2," WCAP-12795, Rev. 1, July 1991. SWsTEEmv
l
- 22. ASTM Designation E706-87, " Standard Master Matrix for Light Water Reactor Pressure Vessel Surveillance Standards," in AslM Standards, American Society for Testing and Materials, Philadelphia, Pa.,1989.
- 23. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa., 1989.
- 24. ASTM Designation E262-86, " Standard Method for Measuring Th - .. Neutron Flux by Radioactivation Techniques," in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa.,1989.
- 25. ASTM Designation E263-88, " Standard Method for Determining Fast Neutron Flux Density by Radioactivation of Iron," in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa.,1989.
- 26. ASTM Designation E264-87, " Standard Method for Determining Fast Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa.,1989.
- 27. ASTM Designation E481-86, " Standard Method for Measuring Neutrrn Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa., 1989.
- 28. ASTM Designation E523-87, " Standard Method for Determining f ast Neutron Flux Density oy Radioactivation of Copper," in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa.,1989.
- 29. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238," in AS1H Standards, American Society for Testing and Materials, Philadelphia, Pa., 1989.
- 30. ASTM Designation E705-84, " Standard Method for Determining f ast Neutron Flux Density by Radioactivation of Neptunium-237," in ASTH Standards, American Society for Testing an6 Materials, Philadelphia, Pa.,1989.
- 31. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM Standards, American Society for Testing and Materials, Philadelphia, Pa., 1989.
E-4 l 13W##aWan 1
_____.-_..y____-___ i l
- 32. R. A. Hartfield, Licensed Operating Reactors, NUREG-0020.- Volume 14. No. 3, Nuclear Regulatory Commission, Washington, D.C., March 1990. j
- 33. E. A. Schmittroth, " FERRET Data Analysis Code," HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979.
- 34. 'W. -N. McElroy, et al., "A Computer-Automated Iterative Method of Neutron _i Flux Spectra Determined by Foil Activation," AFWL-TR-67-41, Volumes I-IV, Air Force Weapons-Laboratory, Kirkland AFB, NM, July 1967. _i I 35. Dosimster File ENDF/B5 Tape 531, distributed March 1984, National Neutron Data Center, Brookhaven National Laboratory Upton, Long Island, NY.
- 36. R. E. Maerker, as reported by F. W. Stallman, " Workshop on Adjustment Codes
,_ and _ Uncertainties - Proc. of the 4th ASTM / EURATOM Symposium on Reactor Dosimetry," NUREG/CP-0029, NRC, Washington, D.C. , July 1982.
- 37. ' A. L. Lowe, Jr., and J. W. Pegram, Correlations for Predicting the Effects j of Neutron Radiation on Linde 80 Submerged-Arc Welds, BAW-1803. Revision 1,
- B&W Nuclear Service Company, Lynchburg, Virginia, March 1991.
l-
- 38. S. L. Anderson, Lt_gl., " Adjoint Flux Program for Po;,it Beach Units 1 and
-2," WCAP-10638, December 1984.
- 39. Code ' of Federal- Regulation, Title 10. Part 50, Domestic Licensing of F Production and Utilization Fac
- lities, Section 50. 61, Fracture Tnughness '
Requirements for Protection Against Pressurized Thermal Shock Events.
- 40. C. A. Quellette,-Materials Information for Westinghouse-Designed Reactor
} Vessels Fabricated by B&W, BAW-2150, B&W Nuclear Service Company, { Lynchburg, Virginia, December 1990. I- ~41. H. S. Palme, H. W. Behnke, and W. -J. Keyworth, Methods of Compliance with Fracture Toughness and Operational Requirements of 10CFR50, - Appendix G, BAW-10046P. Rev. 1, Babccck & Wilcox, Lynchburg, Virginia,. March 1976. i
- 42. Yanisko,
- 5. E. and Chirigos, J. N., " Observations of a Steady State Effect Limiting Radiation Damage in Reactor Vessel Steels," Nuclear Engineering
. and-Design 56 (1980) p. 297-307. r 1 . I h E-5 SWs'fa rc h y
. .}}