ML20246E855
| ML20246E855 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/30/1989 |
| From: | Aadland J, Lowe A, Nana A BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20246E853 | List: |
| References | |
| BAW-2074, NUDOCS 8905120043 | |
| Download: ML20246E855 (105) | |
Text
{{#Wiki_filter:- - _ _ _ _ - _ _ _ _ _ _ . __ BAW-2074 April 1989 ANALYSIS OF CAPSULE RSI-F SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO UNIT-1
-- Reactor Vessel Material Surveillance Program --
l Babcock &Wilcox a Mcoermott company isF A88# $5li2812 F' FDC
BAW-2074 April 1989 I I ANALYSIS OF CAPSULE RS1-F SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO UNIT-1
-- Reactor Vessel Material Surveillance Program -- .I by A. L. Lowe, Jr., PE J. D. Aadland A. D. Nana I L. Petrusha W. R. Stagg I
I I I B&W Document No. 77-1174844-00 I (See Section 12 for document signatures) l I BABC0CK & WILC0X Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 Babcock &Wilcom a McDermott company
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SUMMARY
This report describes the results of the examination of the third capsule (RS1-F) of the Sacramento Municipal Utility District Rancho Seco Unit-1 l. reactor vessel surveillance program. The objective of the program is to g' 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 impact specimens. The program was designed ] in accordance with the requirements of 10CFR50, Appendix H, and ASTM Specifi- I cation E185-73. The capsule received an average fast fluence of 1.42 x 10 19 n/cm2 (E > 1.0 MeV) and the predicted fast fluence for the reactor vessel T/4 location at the end of the seventh cycle is 1.68 x 1018 n/cm2 (E > 1 MeV). Based on the i calculated fast flux at the vessel wall, an 80% load factor, and the planned fuel management, the projected fast fluence that the Rancho Seco Unit-1 reactor pressure vessel inside surface will receive in 40 calendar years of operation is 8.87 x 10 I0 n/cm2 (E > 1 MeV). The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure. The Charpy impact data l results exhibited the characteristic shift to higher temperature for the 30 ft-lb transition temperature and a decrease in upper-shelf energy. These j results demonstrated that the current techniques used for predicting the l change in both the increase in the RT NDT and the decrease in upper-shelf properties due to irradiation are conservative. The recommended operating period was extended to 24 effective full power years as a result of the fourth capsule evaluation. These new operating limitations are in accordance l] I with the requirements of Appendix G of 10CFR50. A low upper-shelf fracture analysis in accordance with 10CFR50, Appendix G, demonstrated that the low l upper-shelf weld metals will not restrict normal plant operations for at I least 32 EFPY. i I' 1 Babcock &Wilcox 3 a McDermott company 3 J
I LI I 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
.I 4.2. Impact Tests . . . . . . . . . . . . . . . . . . . . . . . 4-1
- 5. 5-1 I
POST-IRRADIATION TESTS . . . . . . . . . . . . . . . . . . . . . 5.1. Thermal Monitors . . . . . . . . . . . . . . . . . . . . . 5-1 5.2. Tension Test Results . . . . . . . . . . . . . . . . . . . 5-1 5.3. Charpy V-Notch Impact Test Results . . . . . . . . . . . . I 5.4. Compact Fracture Tests . . . . . . . . . . . . . . . . . . 5-2 5-2
- 6. NEUTRON FLUENCE . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1. Introduction . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2. Vessel Fluence . . . . . . . . . . . . . . . . . . . . . . 6-4 6.3. Capsul e Fl uence . . . . . . . . . . . . . . . . . . . . . . 6-5 6.4. Fluence Uncertainties . . . . . . . . . . . . . . . . . . . 6-6
- 7. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . . . . . . 7-1 7.1. Pre-Irradiation Property Data . . . . . . . . . . . . . . . 7-1 7.2. Irradiated Property Data . . . . . . . . . . . . . . . . . 7-1 7.2.1. Tensile Properties . . . . . . . . . . . . . . . . 7-1 I 7.3.
7.2.2. Impact Properties . . . . . . . . . . . . Reactor Vessel Fracture Toughness . . . . . . . . . . . . .
. . . . . 7-2 7-4 I 8. DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY PRFSSURE -
TEMPERATURE LIMITS . . . . . . . . . . . . . . . . . . . . . . . 8-1
- 9. LOW UPPER-SHELF FRACTURE TOUGHNESS ANALYSIS . . . . . . . . . . . 9-1 9.1. Background . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.2. Charpy Upper-Shelf Impact Energy . . . . . . . . . . . . . 9-2 9.3. Material Fracture Toughness Properties . . . . . . . . . . 9-3 I 9.4. Analytical Method and Acceptance Criteria . . . . . . . . . 9-3 9.5. Fracture Analysis . . . . . . . . . . . . . . . . . . . . . 9-4 ,
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[ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
e Contents (Cont'd) ! t Page 9.6. Summary . . . . . . . . . . . . . . . . . . . . . . . . . . 9-4
- 10.
SUMMARY
OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . 10-1
- 11. SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . . . . . . . . . . . . . 11-1
- 12. CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . 12-1 APPENDIXES A. Reactor Vessel Surveillance Program Background Data and Information . . . . . . . . . . . . . . . . . . . . . . . A-1 B. Pre-Irradiation Tensile Data . . . . . . . . . . . . . . . . . . . B-1 g C. Pre-Irradiation Charpy Impact Data . . . . . . . . . . . . . . . . C-1 g D. Fluence Analysis Methodology . . . . . . . . . . . . . . . . . . . D-1 E. Capsule Dosimetry Data . . . . . . . . . . . . . . . . . . . . . . E-1 F. References . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 List of Tables Table 3-1. Specimens in Surveillance Capsule RSI-F ............3-2 3-2. Chemical Composition and Heat Treatment of Surveill ance Material s . . . . . . . . . . . . . . . . . . . . . 3-3 E 5-1. Tensile Properties of Cggsule RS1-F Base Metal and Weld Metal B Irradiated to 1.42 x 10 n/cm (E > 1 MeV) . . . . . . . . . . 5-3 5-2. Charpy Impact Data From Capsule RSI-F, Base Metal, E Transvers990riengation,HeatNo.C5062-1Irradiatedto g, 1.42 x 10 n/cm (E > 1 MeV) . . . . . . . . . . . . . . . . . 5-3 5-3.
Charpy Impact Data From Capsule RSI-F HeagAffecged Zone Metal, 5-4. Heat No. C5062-1, Irradiated to 1.42 x 10 CharpyImpactDataFromgapsulgRS1-FWeldMetalWF-193 Irradiated to 1.42 x 10 1 n/cm (E > 1 MeV) . . . . . . . . . . n/cm (E > 1 MeV) . . 5-4 5-4 l' 6-1. Surveillance Capsule Dosimeters . . . . . . . . . . . . . . . . 6-7 El 6-2. Rancho Seco Unit 1 Reactor Vessel Fast Flux . . . . . . . . . . 6-7 5 6-3. Calculated Rancho Seco Unit 1 Reactor Vessel Fluence . . . . . . 6-8 6-4. Surveillance Capsule RSI-F Fluence, Flux, and DPA . . . . . . . 6-9 m 6-5. Estimated Fluence Uncertainty . . . . . . . . . . . . . . . . . 6-9 g! 7-1. Comparison of Capsule RSI-F Tension Test Results . . . . . . . . 7-6 i 7-2. Summary of Rancho Seco Reactor Vessel Surveillance Capsules
. . . . . . . . . . 7-7 7-3.
Tensile Test Results . . . . . . . . . . . fSI-F 7-8 ObservedVs.PredictedChangesforCggsule Charpy Impact Properties - 1.42 x 10 n/cm (E > 1 MeV) . . . . Irradiated 7-4. Summary of Rancho Seco Reactor Vessel Surveillance Capsules Charpy Impact Test Results . . . . . . . . . . . . . . . . . . . 79
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Tables (Cont'd) Table Page 7-5. Evaluation of Reactor Vessel End-of-Life Fracture Toughness and Pressurized Thermal Shock Criterion - Sacramento Municipal Utility District, Rancho Seco Unit-1 . . . . . . . . . . . . . . 7-10 7-6. Evaluation of Reactor Vessel End-of-Life Upper Shelf Energy - Sacramenco Municipal Utility District, Rancho Seco Unit-1 . . . . 7-11 8-1. Data for Preparation of Pressure-Temperature Limit Curves for Rancho Seco Unit-1 -- Applicable Through 24 EFPY . . . . . . . . 8-5 9-1. Input Data for Deformation Plasticity Failure Assessment Diagram . . . . . . . . . . . . . . . . . . . . . . . 9-5 9-2. Failure Assessment Data Points . . . . . . . . . . . . . . . . . 9-6 A-1. Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials -- Rancho Seco Unit 1. . . . . . . A-3 A-2. Test Specimens for Determining Material Baseline Properties . . . A-4 A-3. Specimens in Surveillance Capsules (Designation A, C, and E) . . . A-5 A-4. Specimens in Surveillance Capsules (Designation B, D, and F) . . . A-5 B-1. Tensile Properties of Unirradiated Shell Plate Material, Heat No. C5062-1 . . . . . . . . . . . . . . . . . . . . . . . . . B-2 B-2. Tensile Properties of Unirradiated Weld Metal, WF-193 . . . . . . B-2 C-1. Charpy Impact Data From Unirradiated Base Material, Transverse Direction, Heat No. C5062-1 . . . . . . . . . . . . . . C-2 C-2. Charpy Impact Data From Unirradiated Base Material, HAZ, Transverse Direction, Heat No. C5062-1 . . . . . . . . . . . C-3 C-3. Charpy Impact Data From Irradiated Weld Metal, WF-193 . . . . . . C-4 D-1. Flux Normalization Factor . . . . . . . . . . . . . . . . . . . . D-7 D-2. Rancho Seco Unit 1 Reactor Vessel Fluence by Cycle . . . . . . . . D-8 E-1. Detector Composition and Shielding . . . . . . . . . . . . . . . . E-2 E-2. Measured Specific Activities (Unadjusted) for Dosimeters in Cap:ule RSI-F . . . . . . . . . . . . . . . . . . . . . . . . . . E-2 E-3. Dosimeter Activation Cross Sections, b/ atom . . . . . . . . . . . E-3 l List of Fiaures Figure 3-1. Reactor Vessel Cross Section Showing Location of Capsule RS1-F in Rancho Seco Unit-1 . . . . . . . . . . . . . . . . . . . . . 3-4 l 3-2. Reactor Vessel Cross Section Showing Location of Rancho Seco ) Capsule RSI-F in Davis-Besse Unit 1 ..............3-5 3-3. Loading Diagram for Test Specimens in Capsule RS1-F ......3-6 t 5-1. Charpy Impact Data for Irradiated Plate Material, Transverse Orientation, Heat No. C5062-1 . . . . . . . . . . . . . . . . . 5-5 } Charpy Impact Data for Irradiated Plate Material, Heat-Affected 5-2. Zone, Heat No. C5062-1 . . . . . . . . . . . . . . . . . . . . . 5-6 5-3. Charpy Impact Data for Irradiated Weld Metal, WF-193 . . . . . . 5-7 6-1. General Fluence Determination Methodology . . . . . . . . . . . 6-2
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U 1 Fiaures (Cont'd) Figure 6-2. Fast Flux, Fluence and DPA Distribution Through Reactor Page I Vessel Wall . . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 6-3. Azimuthal Flux and Fluence Distributions at Reactor Vessel Inside Surface . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 8-1. Predicted Fast Neutron Fluence at Various Locations Through Reactor Vessel Wall for 24 EFPY - Rancho Seco Unit-1. . . . . . 8-6 3 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal 5 Operation - Heatup, Applicable for First 24 EFPY - Rancho Seco Uni t-1 . . . . . . . . . . . . . . . . . . . . . . . 8-7 8-3. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Cocidown, Applicable for First 24 EFPY - Rancho Seco Unit-1 . . . . . . . . . . . . . . . . . . . . . . . 8-8 8-4. Reactor Vessel Pressure-Temperature Limit Curves for Inservice E Leak and Hydrostatic Tests, Applicable for First 24 EFPY - 3 Rancho Seco Unit-1 . . . . . . . . . . . . . . . . . . . . . . . 8-9 { 9-1. Failure Assessment Diagram for Rancho Seco Unit-1 Reactor Vessel a Based on Weld Metal WF-70 . . . . . . . . . . . . . . . . . . . 9-7 g 9-2. Safety Factors Vs. Crack Extension for Rancho Seco Unit-1 Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . . . 9-7 A-1. Location and Identification of Materials Used in the E Fabrication of Rancho Seco Unit-1 Reactor Pressure Vessel . . . . A-6 5 A-2. Location of Longitudinal Welds in the Upper and Lower Shell Courses . . . . . . . . . . . . . . . . . . . . . . . . . . A-7 E C-1. Charpy Impact Data From Unirradiated Base Metal, g Transverse Orientation . . . . . . . . . . . . . . . . . . . . . . C-5 C-2. Charpy Impact Data From Unirradiated Base Metal, HAZ, Transverse Orientation . . . . . . . . . . . . . . . . . . . . . . C-6 C-3. Charpy Impact Data From Unirradiated Weld Metal . . . . . . . . . C-7 D-1. Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the Capsule . . . . . . . . . . . . . . . . . . . D-9 E D-2. Rationale for the Calculation of Neutron Flux in the 5 Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . D-10 D-3. Plan View Through Reactor Core Midplane (Reference R-0 Cal cul ati on Model ) . . . . . . . . . . . . . . . . . . . . . . . . D-11 I I I I I
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- 1. INTRODUCTION This report describes the results of the examination of the third capsule (RS!-F) of the Sacramento Municipal Utility District Unit-1 (Rancho Seco) reactor vessel material surveillance program (RVSP). The capsule was removed and evaluated after being irradiated in the Davis-Besse Unit I reactor as part of the Integrated Reactor Vessel Materials Surveillance Program (BAW-1543A).1 The irradiation exposure in Davis-Besse Unit 1 plus the previous irradiation in Rancho Seco is the equivalent of 12.5 EFPY or approximately 21 calendar years of exposure in the Rancho Seco reactor vessel. The capsule experienced a fluence of 1.42 x 10 19 n/cm2 (E > 1 MeV), which is the equiva-lent of approximately 51 effective full power years (EFPY) operation of the Rancho Seco reactor vessel. The first capsule (RSI-B) from this program was removed and examined after the first year of operation; the results are reported in BAW-1702.2 The second capsule (RSI-D) was removed and examined after irradiation in Davis-Besse Unit-1 as part of the Integrated Reactor Vessel Materials Surveillance Program; the results are reported in BAW-1792.3 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 Rancho Seco Unit-1 was designed and furnished by Babcock & Wilcox (B&W) as described in BAW-10006A4 and conducted in accordance with BAW-1543A.1 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 pressut e vessel.
l ( The surveillance program for Rancho Seco was designed in accordance with E185-665 and thus is not in compliance with 10CFR50, Appendixes G6 and H7 I since the requirements did not exist at the time the program was design. ) Because of the difference, additional tests and evaluations were required to ensure meeting the requirements of 10CFR50, Appendixes G and H. The recom-1 ! 1-1 } Babcock & Wifcox ) a McDermott company
I. mendations for the future operation of Rancho Seco included in this report do comply with these requirements. I I E I I I I I I I I I I I I 1-2 I Babcock &Wilcox 3 a McDermott company g
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- 2. BACKGROUND The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water-cooled reactors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of I such low-alloy ferritic steels as SA533, Grade B1, used in the fabrication of the Rancho Seco reactor vessel, are well characterized and documented in the literature. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. The most significant mechanical property change in reactor pressure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the Charpy upper shelf energy value.
g Appendix G to 10CFR50, " Fracture Toughness Requirements,"6 specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of I water-cooled power reactors, and provides specific guidelines for determining the pressure-temperature limitations on operation of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including antici-pated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10CFR50 became effective on August 13, 1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on I the effective date. Appendix H to 10CFR50, " Reactor Vessel Materials Surveillance Program I Requirements,"7 defines the material surveillance program required to monitor ! I 2-1 1 I Babcock &WHcom a McDermo:t company i
I changes in the fracture toughness properties of ferritic materials in the l reactor vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Fracture l toughness test data are obtained from material specimens withdrawn periodi-cally from the reactor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life. A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code, Section III, " Nuclear Power Plant Components."8 This method utilizes fracture mechanics concepts and the reference nil-ductility temperature, RTNDT, which is defined as the greater of the drop weight nil-ductility transition l 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 of a given material is used to index that material to a reference NDT stress intensity factor curve (K IR curve), which appears in Appendix G of , ASME Section III. The K IR curve is a lower bound of dynamic, static, and crack arrest fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K IR cune, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined l' 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 l reactor vessel materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which a surveillance capsule contain-ing prepared specimens of the reactor vessel materials is periodically , removed from the operating nuclear reactor and the specimens are tested. The increase in the Charpy V-notch 30 ft-lb temperature is added to the original ! RT to adjust it for radiation embrittlement. This adjusted RT is used NDT NDT to index the material to the K curve which, in turn, is used to set IR operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials. g 2-2 I Babccck & Wilcox i a McDermott company i
10CFR50, Appendix G, also requires a minimum Charpy V-notch upper-shelf energy of 75 ft-lbs for all beltline region materials unless it is demon-strated that lower values of upper-shelf fracture energy will provide an adequate margin for deterioration as the result of neutron radiation. No action is required for a material that does not meet the 75 ft-lb requirement provided the irradiation deterioration does not cause the upper-shelf energy to drop below 50 ft-lbs. The regulations specify that if the upper-shelf energy drops below 50 ft-lbs it must be demonstrated in a manner approved by the Office of Nuclear Regulation that the lower values will provide adequate margins of safety. When a reactor vessel fails to meet the 50 ft-lb requirement, a program must be submitted for review and approval at least three years prior to the time the predicted fracture toughness will no longer satisfy the regul atory requirements. The program must address the following: A. A volumetric examination of 100 percent of the beltline materials that do not meet the requirement. B. Supplemental fracture toughness data as evidence of the fracture toughness of the irradiated beltline materials. C. Fracture toughness analysis to demonstrate the existence of equiva-lent margins of safety for continued operation. If these procedures do not indicate the existence of an adequate margin of safety, the reactor vessel beltline may be given a thermal annealing treat-ment to recover the fracture toughness properties of the materials. )- I ! ) l l 2-3 Babcock & WHcom a McDermott company
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- 3. SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for Rancho Seco comprises six surveillance capsules designed to monitor the effects of neutron and thermal environment on the materials of the reactor pressure vessel core region. The capsules, which were inserted into the reactor vessel before initial plant startup, were I positioned inside the reactor vessel between the thermal shield and the vessel wall at the locations shown in Figure 3-1. The six capsules, originally designed to be placed two in each holder tube, are positioned near the peak axial and azimuthal neutron flux. BAW-10006A4 includes a full :
description of the capsule locations and design. After the capsules were removed from Rancho Seco in 1976 and included in the integrated RVSP, they were scheduled and irradiated in the Davis-Besse Unit-1 reactor as described in BAW-1543AI . During this period of irradiation, capsule RSI-F was irra-diated in the bottom position in holder tube WX for four cycles and the top position in holder tube WZ for one cycle as shown in Figure 3-2. Capsule RSI-F was removed during the fifth refueling shutdown of Davis-Besse. This capsule contained Charpy V-notch impact test specimens fabri-I cated from one base metal (SA533, Grade B1), or.e heat-affected-zone, and a weld metal . Tension test specimens were fabricated from the base metal and the weld trictal only. The fracture toughness specimens were fabricated from the weld metal. The specimens 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 RSI-F are described in Table 3-2. All test specimens were machined from the 1/4-thickness (1/4T) location of the plate material. Charpy V-notch and tension test specimens were cut from the surveillance material such that they were oriented with their longitudi-nal axes either parallel or perpendicular to the principal working direction. 1I I 3-1 Babcock & Wifcox I. a McDermott company ____-__m
I Capsule RSI-F contained dosimeter wires, described as follows: Dosimeter Wire Shieldina U-Al alloy Cd-Ag alloy Np-Al alloy Cd-Ag alloy Nickel Cd-Ag alloy j 0.66 wt % Co-Al alloy Cd-Ag alloy 0.66 wt % Co Al alloy None Fe None Thermal monitors of low-melting metals and alloys were included in the l capsule. The metals and alloys and their melting points are as follows: ' Allov Meltina Point. F 90% Pb, 5% Ag, 5% Sn 558 I ! 97.5% Pb, 2.5% Ag 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Cadmium 610 Lead 621 Table 3-1. Specimens in Surveillance Caosule RS1-F Number of Test Specimens 5 1/2TCompt 5 Material Description Tension CVN Impact Fracture Weld Metal, WF-193 2 12 8 HAZ(b) Heat No. C5062-1, Transverse - 12 - Base Metal Heat No. C5062-1, Transverse 2 J2 . l Total Per Capsule 4 36 8 i a) Compact fracture toughness specimens not precracked. I (b)HAZ denotes heat-affected zone. 3-2 I! i Babcock & Wilcox 3l a McDermott company gi l 1 .- - - - - - - - - - - - _ _ - _ - - _ _ - - _ .
- _ _ _ _ - __-_ - __ -_ - _ _ - _ __ _ _-___ _ ___ _ _ _ _ _ _ - _ _ _ _ - _ - _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ - _ - _ Table 3-2. Chemical Composition and Heat Treatment' of Surveillance Materials Chemical Composition, w/o Heat N WeldMygpl Element C5062-1yg) WF-193 , C 0.20 0.09 Mn 1.26 1.49 P 0.013 0.016 S 0.017 0.016 Si 0.15 0.52 L Ni- 0.60 0.59 Cr 0.14 0.07 i
~
Mo 0.55 0.40 I Cu 0.12 0.31 4 Heat Treatment Heat No. Temo. F Time, h Coolina
- C5062-1(c) 1550-1600 4.5 Brine quench 1225 5.0 Brine quench 1100-1150 28.0 Furnace cooled WF-193(c) 1100-1150 28.0 Furnace cooled ;
(a)Per Certified Materials Test Report 1 L (b)Per Licensing Document BAW-1500P 9 (c)Per Licensing Document BAW-182010 I i h. I 3-3 Babcock &Wilcom a McDermott company
e I Figure 3-1. Reactor Vessel Cross Section Showing Location of Caosule RS1-F in Rancho Seco Unit-1
- Surveillance Capsule Holder Tube -- Capsules RS1-C, j
/ RS1-D J h ~ ~~~ l r %
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l I I I 3-4 I l Babcock &Wilcox g a McDermott company 3
Figure 3-2. Reactor Vessel Cross Section Showing Location of Rancho Seco Capsule RSl-F in Davis-Besse Unit 1 I n Surveillance Capsule Holder - ""' Tubes t I
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- 4. PRE-IRRADIATION TESTS Unirradiated material was evaluated for two purposes: (1) to establish a baseline of data to which irradiated properties data could be referenced, and (2) to determine those materials properties to the extent practical from available material, as required for compliance with Appendixes' G and H of 10CFR50.
4.1. Tension Tes_t1 Tension test specimens were fabricated frem the reactor vessel shell course forging and weld metal. The specimens were 4.25 inches long with a reduced section 1.750 inches long by 0.357 inch in diameter. They were tested on a 55,000-lb load capacity universal test machine at a crosshead speed of 0.050 inch per minute. A 4-pole extension device with .a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accor-dance with the applicable requirements of ASTM A370-77.II For each material type and/or condition, six specimens in groups of three were tested at both room temperature and 580F. The tension-compression load cell used had a certified accuracy of better than +0.5% of full scale (25,000 lb). All test data for the preirradiation tensile specimens are given in Appendix B. 4.2. Imoact Tests Charpy V-notch impact tests were conducted in accordance with the require-ments of ASTM Standard Methods A370-77 II and E23-8212 on an impact tester certified to meet Watertown standards. Test specimens were of the Charpy I V-notch type, which were nominally 0.394 inch square and 2.165 inches long. V Prior to testing, specimens were temperature-controlled in liquid immersion baths, capable of covering the temperature range from -85 to +550F. Speci-mens were removed from the baths and positioned in the test frame anvil with tongs specifically designed for the purpose. The pendulum was released 4-1 l Babcock & WHcom a McDermott Company
I , manually, allowing the specimens to be broken within 5 seconds from their removal from the temperature baths. Impact test data for the unirradiated baseline reference materials are presented in Appendix C. Tables C-1 through C-3 contain the basic data that are plotted in Figures C-1 through C-3. l II I Il l I I. I I I1 I I I I I I 4-2 Babcock &Wilcox a McDermott company ( _ ___--_____--_--- - -_-__--__-- _ _ _ _
- 5. POST-IRRADIATION TESTS i
5.1. Thermal Monitors 1 Capsule RS1-F contained three temperature monitor holder tubes, each containing five fusible alloy wires with ' melting points ranging from 558 to 621F. All the thermal monitors at 558, 580, 588 and 610F locations had melted while the monitor at' the 621F location showed no signs of melting in l all three holder tubes. From these observations, it was concluded that the capsule had been exposed to a peak temperature in the range of 610 to 621F during the reactor operating period. These peak temperatures are attributed ; to operating transients that are of short durations and are judged to have l insignificant effect on irradiation damage. Short duration operating i transients cause the use of thermal monitor wires to be of limited value in determining the maximum steady state operating temperature of the surveil- l lance capsules; however, it is judged that the maximum steady state operating temperature of specimens in -the capsule was held within 25F of the 1/4T vessel thickness location temperature of 559 to 577F. It is concluded that the capsule design temperature may have been exceeded during operating transients but not for times and temperatures that would make the capsule q 1 data unusable. ! 5.2. Tension Test Results The results of the postirradiation tension tests are presented in Table 5-1. Tests were performed on specimens at room temperature and at 580F using the j l same test procedures and techniques used to test the u'nirradiated specimens l F (Section 4.1). In general, the ultimate strength and yield strength of the material increased with a corresponding slight decrease in ductility as , I i 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 1 l i 5-1 Babcock &Wilcox A McDermott Company
I material properties changed is within the range of changes to be expected for the radiation environment to which the specimens were' exposed. The results of the pre-irradiation tension tests are presented in Appendix B. 5.3. Charov V-Notch Impact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 5-2 through 5-4 and Figures 5-1 through 5-3. The test procedures and techniques were the same as those used to test the unirradiated- specimens (Section 4.2). The data show that the materials exhibited a sensitivity to irradiation within the values to be expected from their chemical compos'ition and the fluence to which they were l exposed. 5.4. Comoact Fracture Tests The compact fracture specimens fabricated from the weld metal, which were a part of the capsule specimen inventory, were tested by an approved single specimen J-integral testing procedure. The results of the testing of these specimens are reported in a separate report. I I' I I l I I' I I 5-2 I Babcock & Wilcox a McDermott company _____a
. Table 5-1. Tensile Properties of Capsule RSI-{gBasegetaland Weld Metal Irradiated to 1.42 x 10 n/cm (E > 1 MeV)
Strenath, osi Elonaation. % Red'n. Specimen Test Temp, . in Area, No. F Yield Ultimate Uniform Total % Base Metal. C5062-1. Transverse LL 603 70 63,300 88,200 11.9 25.3 58.5 LL 610 580 62,000 83,700 10.6 22.6 56.7 Weld Metal. WF-193 MM 003 70 83,900 98,600 11.6 25.7 58.5 MM 010 580 72,600 90,574 9.5 18.0 37.6 Table 5-2. Charpy Impact Data From Capsule RSI-F, Base Metal, Transverse 0Qentat{on,HeatNo.C5062-1 Irradiated to 1.42 x 10 n/cm (E > 1 MeV) Test Impact Lateral Shear l Specimen Temperature Energy Expansion Fracture ID F ft-lbi Inch % LL 629 40 13.0 0.013 20 LL 643 70 29.0 0.028 20 i LL 659 100 48.5 0.045 50 LL 636 120 53.5 0.050 80
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l LL 660 150 82.0 0.069 90 LL 630 180 75.5 0.064 100 LL 684 180 84.0 0.075 100 LL 617 240 '74.0 0.065 100 LL 682 300 77.5 0.068 100
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l Table 5-3. CharpyImpactDataFromCapsuleRSI-FHeagAffecgedZoneMetal, i Heat No. C5062-1, Irradiated to 1.42 x 10 n/cm (E > 1 MeV) Test Impact Lateral Shear Specimen Temperature Energy Expansion Fracture ID F ft-lbs Inch % i LL 344 -60 21.5 0.020 40 LL 339 -25 22.0 0.029 50 ] LL 301 0 37.5 0.030 50 l LL 321- 40 63.0 0.049 80 LL 388 LL 373 70 120 73.0 78.0 0.058 0.067 85 100 l{ LL 364 180 67.5 0.061 100 LL 369 240 92.0 0.076 100 LL 307 300 71.5 0.063 100 LL 361 360 83.0 0.057 100 LL 304 450 72.0 0.060 100 LL 329 550 65.0 0.057 100 Table 5-4. Charpy Impact Data From Capsul g RS1-F Weld Metal WF-193 Irradiated to 1.42 x 10 n/cm2 (E > 1 MeV) Test Impact Lateral Shear Specimen Temperature Energy Expansion Fracture i ID F ft-lbs Inch % ' l MM 023 70 15.0 0.015 5 ' MM 026 120 25.0 0.025 60 MM 013 150 28.0 0.026 65 i MM 006 180 36.0 0.041 70 MM 054 MM 090 240 300 48.0 47.5 0.048 0.043 90 100 l MM 031 360 48.0 0.045 100 MM 029 405 46.5 0.050 100 MM 076 450 41.5 0.049 10 E MM 032 550 48.0 0.047 100 5 f-i 5-4 I l l Babcock & WHcox 3 l a McDermott company 5, l
I Figure 5-1. Charpy Impact Data for Irradiated Plate Material, l Transverse Orientation. Heat No. C5062-1 100 , ,
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- 6. NEUTRON FLUENCE 6.1. Introduction The neutron fluence (time integral of flux) is a quantative way of expressing the cumulative exposure of a material to a pervading neutron flux over a specific period of time. Fast neutron fluence, defined as the fluence of neutrons having energies greater than 1 MeV, is the parameter that is pre-sently used to correlate radiation induced changes in material properties.
Accordingly, the fast fluence must be determined at two locations: (1) in the test specimens located in the surveillance capsule, and (2) in the wall of the reactor vessel. The former is used in developing the correlation between fast fluence and changes in the material properties of specimens, and the latter is used to ascertain the point of maximum fluence in the reactor vessel, the relative radial and azimuthal distribution of the fluence, the fluence gradient through the reactor vessel wall, and the corresponding material properties. The accurate determination of neutron flux is best accomplished through the simultaneous consideration of neutron dosimeter measurements and analytically derived flux spectra. Dosimeter measurements alone cannot be used to predict the fast fluence in the vessel wall or in the test specimens because (1) they cannot measure the fluence at the points of interest, and (2) they provide only rudimentary information about the neutron energy spectrum. Conversely, reliance on calculations alone to predict fast fluence is not prudent because of the length and complexity of the analytical procedures involved. In short, measurements and calculations are necessary complements of each other and together they provide assurance of accurate results. Therefore, the determination of the fluence is accomplished using a combined 7 I analytical-empirical methodology which is outlined in Figure 6-1 and describ-ed in the following paragraphs. The details of the procedures and methods are presented in general terms in Appendix D and in BAW-1485P.13 f 1 6-1 Babcock &Wilcox a McDermott company __-_a
Fiaure 6-1. General Fluence Determination Methodoloav l l MEASUREMENTS OF NEUTRON DOSIMETER ACTIVITIES
#MLYTICAL DETERMINATION OF DOSIMETER ACTIVITIES NO NEUTRON F1.UX l
I I: ADJUSTED ENERGY DEPE10ENT NEUTRON E FLUX 5 I REAUDR OPERATING NEUTRON HISTORY #0 PRE-FLUENCE DICTED FUTURE OPERATION I Analytical Determination of Dosimeter Activities and Neutron Flux The rnalytical calculation of the space and energy dependent neutron flux in the test specimens and in the reactor vessel is performed with the two dimensional discrete ordinates transport code, D0TIV.I4 The calculations emplay an angular quadrature of 48 sectors (SB), a third order LeGendre polynomial scattering approximation (P3), the CASK 23E cross section set 15 with 22 neutron energy groups and a fixed distributed source corresponding to the time weighted average power distribution for the applicable irradiation period. In addition to the flux in the test specimens, the DOTIV calculation deter-mines the saturated specific activity of the various neutron dosimeters located in the surveillance capsule using the ENDF/B5 dosimeter reaction cross sections.16 The saturated activity of each dosimeter is then adjusted by a factor which corrects for the fraction of saturation attained during the I 6-2 Babcock & WHcox a McDermott company
I dosimeter's actual (finite) irradiation history. Additional corrections are made to account for the following effects: o Photon induced fissions in V and Np dosimeters (without this correc-tion the results underestimate the measured activity). e Fissile impurities in U dosimeters (without this correction the results underestimate the measured activity). e Short half-life of isotopes produced in iron and nickel dosimeters (303-day Mn-54 and 71-day C0-58, respectively). (Without this correction, the results could be . biased high or low depending on the long term versus short term power histories.) Measurement of Neutron Dosimeter Activities The accuracy of neutron fluence predictions is improved if the calculated neutron flux is compared with neutron dosimeter measurements adjusted for the effects noted above. The neutron dosimeters located in the surveillance capsules are listed in Table 6-1. Both activation type and fission type dosimeters were used. The ratio of measured dosimeter activity to calculated dosimeter activity (M/C) is determined for each dosimeter, as discussed in Appendix D. These M/C ratios are evaluated on a case-by-case basis to assess the dependability or veracity of each individual dosimeter response. After carefully evaluat- j 1 ing all factors known to affect the calculations or the measurements, an average M/C ratio is calculated and defined as the " normalization factor." The normalization factor is applied as an adjustment factor to the DOT- l. calculated flux at all points of interest. Neutron Fluence The determination of the neutron fluence from the time averaged flux requires l only a simple multiplication by the time in EFPS (effective full-power seconds) over which the flux was averaged, i.e. f q3(AT)=[g$3g AT 9 L where ) 2 fq3 (AT) = Fluence at (i,j) accumulated over time AT (n/cm ), g - Energy group index, 1 l 6-3 BatDCOCIEEn M ICOE I a McDermott company
I 2 4$3g - Time-average flux at (i,j) in energy group g, (n/cm -sec), AT - Irradiation time, EFPS. Neutron fluence was calculated in this analysis for the following components over the indicated operating time: l Test Specimens: Capsule irradiation time in EFPS Reactor Vessel: Vessel irradiation time in EFPS Reactor Vessel: Maximum point on inside surface extrapolated to 32 g effective full power years E The neutron exposure to the reactor vessel and the material surveillance specimens was also determined in terms of the iron atom displacements per atom of iron (DPA). The iron DPA is an exposure index giving the fraction of iron atoms in an iron specimen which would be displaced during an irradia-tion. It is considered to be an appropriate damage exposure index since l displacements of atoms from their normal lattice sites is a primary source of neutron radiation damage. DPA was calculated based on the ASTM Standard E693-79 (reapproved 1985).17 A DPA cross section for iron is given in the ASTM Standard in 641 energy groups. DPA per second is determined by multi-plying the cross section at a given energy by the neutron flux at that energy and integrating over energy. DPA is then the integral of DPA per second over the time of the irradiation. In the DPA calculations reported herein, the l ASTM DPA cross sections were first collapsed to the 22 neutron group struc-ture of CASK-23E; the DPA was then determined by summing the group flux times l I the DPA cross section over the 22 energy groups and multiplying by the time of the irradiation. 6.2. Vessel Fluence The maximum fluence (E > 1 MeV) exposure of the Rancho Seco Unit I reactor vessel during Cycles 5-6 was determined to be 6.90 x 10 17 n/cm2 based on a maximum neutron flux of 1.23 x 1010 n/cm2-s (Tables 6-2 and 6-3). The maximum fluence occurs at the cladding / vessel interface at an azimuthal l location of approximately 11 degrees from a major horizontal axis of the Core. I 6-4 l Babcock & Wilcox a McDermott company
Fluence data were extrapolated to 32 EFPY of operation based on two assump-tions: (1) the future fuel cycle operations'do not differ significantly from their current designs, and (2) the latest calculated (or extrapolated) flux remains constant from that time through 32 EFPY. The extrapolation was carried out in two stages, (1) from E006 to E0C7, and (2) from E00/ to 32 EFPY. In the first stage, cycle averaged fluxes are calculated using D0TIV, the current design data for cycle 7, and D0T adjoint factors for assembly-averaged power distributions. In the second stage, the 32 EFPY fluence was calculated by assuming a constant flux over the period which was equal to the average flux for cycle 7. Relative fluence and DPA (displacement per atom) as a function of radial location in the reactor vessel wall is shown in Figure 6-2. Reactor vessel neutron fluence lead factors, which are the ratio of the neutron flux at the clad interface to that in the vessel wall at the T/4, i f 2 and 3T/4 locations, i are 1.79, 3.53, and 7.30, respectively. DPA lead factors at the same locations are 1.58, 2.64, and 4.57, respectively. The relative fluence as a function of azimuthal angle is shown in Figure 6-3. A peak occurs in the fast flux (E > 1 MeV) at about 11 degrees with a corresponding value of 1.23 x 1010 n/cm 2 .3, 6.3. Caosule Fluence The capsule was irradiated for 1609.5 EFPD in the bottom holder tube position during Cycles 1-4 and in the top holder tube position during Cycle 5 of l Davis-Besse located 11 degrees off the major horizontal axis at about 202 cm from the vertical axis of the core. The capsule was also irradiated for l 170.5 EFPD in Rancho Seco-1, during cycle 1, located at the 11 degree position about 211 cm from the vertical axis of the core. The cumulative I fast fluence at the center of the surveillance capsule was calculated to be 1.42 x 10 19 n/cm2 of which 3.7% was accumulated during the Rancho Seco cycle l 1 irradiation, and 96.3% was accumulated during the Davis-Besse cycles 1-5 irradiation (Table 6-4). This fluence value represent an average value for l the various locations in the capsule. / 6-5 Babcock & WHcom a McDermott company
6.4. Fluence Uncertainties Uncertainties were estimated for the fluence values reported herein. The results are shown in Table 6-5 and are based on comparisons to benchmark experiments, when available; estimated and measured variations in input data; and on engineering judgement. The values in Table 6-5 represent best estimate values based on past experience with reactor vessel fluence ana-lyses. I 1 I I I I I I I I 6-6 Babcock & Wilcox a McDermott company
l Table 6-1. Surveillance Caosule Dosimeters Lower Energy Limit for Isotope
' Dosimeter Reactions (a) Reaction, MeV Half-Life 54Fe(n,p)54Mn 2.5 312.5 days 58Ni(n,p)58Co 2.3 70.85 days 238U (n,f)137Cs 1.1 30.03 years 237Np(n,f)137Cs 0.5 30.03 years l
(a) Reaction activities measured for capsule flux evaluation. Table 6-2. Rancho Seco Unit' 1 Reactor Vessel Fast Flux Fast Flux (E > 1 MeV), n/cm2 -s Flux n/cm2 -s (E > 0.1 MeV) Inside Surface Inside Surface Cvele (Max location) T/4 3T/4 (Max location) Cycles 1-3 1.94E;10 1.1E+10 2.5E+9 4.2E+10 (1029.5 EFPD) Cycle 4 1.30E+10 7.2E+9 1.7E+9 2.8E+10 (229 EFPD) Cycles 5-6 1.23E+10 6.90E+9 1.69E+9 2.58E+10 (648.2 EFPD) Cycle 7** 1.09E+10 0.61E+10* 1.49E+9* (340 EFPD) 8 EFPY 0.72E+10 0.40E+10* 0.99E+9* 15 EFPY i.72E+10 0.40E+10* 0.99E+9* 21 EFPY 0.72E+10 0.40E+10* 0.99E+9* { 24 EFPY 0,72E+10 0.40E+10* 0.99E+9* l 32 EFPY 0.72E+10 0.40E+10* 0.99E+9* 1
- Divide flux at inside surface by the appropriate lead factors on p. 6-8 to obtain these T/4 and 3T/4 fast flux values.
** Assumed cycle length of 340 EFPD for flux extrapolation for Cycle 7.
6-7 Babcock &Wilcox ' }: ) a McDermott company
Table 6-3. Calculated Rancho Seco Unit 1 Reactor Vessel Fluence Fast Fluence, n/cm2 (E > 1 MeV) Cumulative Inside Surface Irradiation Time (Max Location) T/4 T/2 3T/4 End of Cycle 3 (1029.5 EFPD) 1.73E+18 9.6E+17 4.9E+17* 2.2E+17 l End of Cycle 4 1.99E+18 1.1E+18 5.6E+17* 2.6E+17 (1258.5 EFPD) End of Cycle 6 2.68E+18 1.50E+18 7.59E+17 3.67E+17 m (1906.7 EFPD) g End of Cycle 7 3.00E+18 1.68E+18* 8.50E+17* 4.11E+17* (2246.7 EFPD) 8 EFPY 3.42E+18 1.91E+18* 9.69E+17* 4.68E+17* 15 EFPY 5.01E+18 2.80E+18* 1.42E+18* 6.86E+17* 21 EFPY 6.37E+18 3.56E+18* 1.80E+18* 8.73E+17* 24 EFPY 7.05E+18 3.94E+18* 2.00E+18* 9.66E+17* 32 EFPY 8.87E+18 4.96E+18* 2.51E+18* 1.22E+18*
- Calculated using these 1.0 1.79 3.53 7.30 lead factors Conversion Factors Fluence (E > 1 MeV) 1.45E-21** 1.63E-21** 1.94E-21** 2.30E-21**
to DPA
** Multiply fast fluence values (E > 1 MeV) in units of n/cm 2 by these factors to obtain the corresponding DPA values.
I I 6-8 Babcock &Wilcox g a McDermott company g
Table 6-4. Surveillance Caosule RS1-F Fluence. Flux and DPA Flux (E >21 MeV), Fluencg, c Irradiation Time n/cm s n/cm DPA RS1, Cycle 1 3.59E+10 5.29E+17 0.69E-3 (170.5 EFPD) DB1, Cycle 1-5, 9.84E+10 1.37E+19 2.01E-2 (1609.5 EFPD) Total --- 1.42E+19 2.08E-2 Table 6-5. Estimated Fluence Uncertainty Estimated Calculated Fluence Uncertainty Basis of Estimate In the capsule 15% Activity measurements, cross section fission yields, satu-I ration factor, deviation from average fluence value In the reactor vessel 21% Activity measurements, cross at maximum location for sections, fission yields, fac-cycles 1 through 6 of tors, axial factor, capsule Rancho Seco Unit I location, radial / azimuthal ex-trapolation, normalization factor I In the reactor vessel at the maximum location for end-of-life extra-23% Factors in vessel fluence above plus uncertainties for extra-polation to 32 EFPY polation l I l LI I 6-9 Babcock &Wilcox . a McDermott company l . _ _ _ _ _ _ _ _ _ _ _ . __. .____-_________w
Figure 6-2. Fast Flux, Fluence and DPA Distribution Throuah Reactor Vessel Wall 1.0 _ 0.9 y L.F. = = 1.58 0.8 - D g m \ 0.7 - D t 2 3 0.6 - c 1 8 [ L.F. = = 2.64 g 0.5
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- 7. DISCUSSION OF CAPSULE RESULTS l 7.1. Pre-Irradiation Procerty Data A review of the unirradiated properties of the reactor vessel core beltline l region materials indicated no significant deviation from expected properties except in the case of the upper shelf properties of the weld metal. Based on the predicted end-of-service peak neutron fluence value at the 1/4T vessel 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 ft-lb.
I This weld was selected for inclusion in the surveillance program in accordance with the criteria in effect at the time the program was designed for Rancho Seco. The applicable selection criterion was based on the unirradiated properties only. L 2_. Irradiated Property Dati l l 7.2.1. Tensile Properties Table 7-1 compares irradiated and unirradiated tensile properties. At both l room temperature a; 'levated temperature, the ultimate and yield strength changes in the base as a result of irradiation and the corresponding changes in ductility _.e within the limits observed for similar materials. l There is some strengthening, as indicated by increases in ultimate and yield strengths and decreases in ductility properties. All changes observed in the base metal are such as to be considered within acceptable limits. The changes at both room temperature and 580F in the properties of the base metal are not as large as those observed for the weld metal, indicating a lesser sensitivity of the base metal to irradiation damage. In either case, the
- g changes in tensile properties are insignificant relative to the analysis of iE the reactor vessel materials at this time period in the reactor vessel service life.
- I 7-1 Babcock & Wilcox I a McDermott company
A comparison of the tensile data from previously evaluated capsules (Capsules RS-1B and RS-ID) with the corresponding data from the capsule reported in ! i this report is shown in Table 7-2. The currently reported capsule experienc-ed a fluence that is approximately 3.5 times greater than the first capsule. ! The general behavior of the tensile properties as a function of neutron irradiation is an increase in both ultimate and yield strength and a decrease in ductility as measured by both total elongation and reduction of area. The most significant observation from these data is that the weld metal exhibited greater sensitivity to neutron radiation than the base metal. 7.2.2. 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 metal is in relatively good agreement with the value predicted using Regulatory Guide 1.99, Rev. 18 2 and the predicted value is conservative. It would be expected that these values would exhibit good agreement when it is considered that the data used to develop Regulatory Guide 1.99, Rev. 2, was taken at the 30 ft-lb tempera-ture. 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 i 2 and the predicted value is also conservative. The shift being 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 conser-vative value. The data for the decrease in Charpy USE with irradiation showed good agree- l ment with predicted values for for both the base metal transverse directions and the weld metal. However, the comparison of the measured data with the predicted value is to be expected in view of the lack of data for medium , or l 7-2 l Babcock & WHcox A MrDermott comparty i
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 Rancho Seco Unit-1 with the corresponding data from the capsule reported in this report is shown in Table 7-4. The currently reported data experienc-ed a fluence that is three and a half times greater than the first capsule. The base metal exhibited transition temperature shifts at the 30 ft-lb levels for the latest capsule that were similar in magnitude to those of the previous capsule. The corresponding data for the weld metal also showed further increase at the 30 ft-lb level as compared to the previously reported increase at the 30 ft-lb level. This may be related, in part, to a further decrease in the upper-shelf energy. Both the base metal and the weld metal exhibited decreases in the upper-shelf values similar to the previous capsules. The weld metal in this capsule exhibited a decrease similar to the weld metal in the previous capsule. These data confirm that the upper-shelf drop for this weld metal have not reached saturation as observed in the results of capsules evaluated by others. This behavior of Charpy USE drop for this weld metal should not be con::idered indicative of a similar behavior of upper-stelf region fracture toughness properties. This behavior indicates that otter reactions may be taking place within the material besides simple neutron damage. Verification of this relationship must await the testing and evaluation of the data from compact fracture toughness test specimens. Results from other surveillance capsules also indicate that RTNDT estimating i curves have greater inaccuracies than originally thought. These inaccuracies are a function of a number of parameters related to the basic data available at the time the estimating curves are established. These parameters may include inaccurate fluence values, poor chemical composition values, and variations in data interpretation. The change in the regulations requiring the shift measurement to be based on the 30 ft-lb value has minimized the ) errors that resulted from using the 30 ft-lb data base to predict the shift l behavior at 50 ft-lbs. i 7-3 Babcock & Wilcox a McDermott cornpany
I: The design curves for predicting the shift will continue to be modified as more data become available; until that time, the design curves for predicting ' the RT NDT shift as given in Regulatory Guide 1.99, Revision 2, are considered adequate for predicting the RT NDT shift of those materials for which data are not available. These curves will be used to establish the pressure-tempera- g) l ture operational limitations for the irradiated portions of the reactor j vessel until the time that new prediction curves are developed and approved. The lack of good agreement of the change in Charpy upper-shelf energy is Il ! j further support of the inaccuracy of the prediction curves. Although the prediction curves are conservative in that they generally predict a larger decrease in upper-shelf energy than is observed for a given fluence and copper content, the conservatism can unduly restrict the operational limita-tions. 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 Touchness An evaluation of the reactor vessel end-of-life fracture toughness and the l pressurized thermal shock criterion was made and the results are presented in [ Table 7-5. The fracture toughness evaluation shows that the controlling weld metal may have a T/4 wall location end-of-life RT NDT f 231F based on Regulatory Guide 1.99, Revision 2, with a margin of 56F. This predicted shift is excessive since data from an Integrateo Reactor Vessel Surveillance Program surveil- i lance capsules exhibit measured RT NDT significantly less for comparable ' fluence values. It is estimated that the end-of-life RT NDT shift will be g significantly less than the value predicted using Regulatory Guide 1.99, 3 Revision 2. This reduccd shift will permit the calculation of less restric-tive pressure temperature operating limitations than if Regulatory Guide 1.99, Revision 2, was used. l The pressurized thermal shock evaluation demonstrates that the Rancho Seco Unit I reactor pressure vessel remains below the screening criterion limits. l The increased margin below the screening criteria is the result of improved g 7-4 I l Babcock & Wilcox a McDermott company
hel management which significantly lowers the reactor vessel inside surface fl uence. Therefore, the decrease in RT values indicate good fuel manage-PTS ment must be continued to assure that no additional corrective actions, as required by the PTS regulations, will be necessary prior to license expira-tion. An evaluation of the reactor vessel end-of-life upper-shelf energy for each of the materials used in the fabrication was made and the results are presented in Table 7-6. This evaluation was made because the weld metals used to fabricate the reactor vessel are characterized by relatively low-upper-shcif-energy and high copper contents; and, consequently, are expected to be sensitive to neutron radiation damage. Two methods were used to evaluate the radiation induced decrease in upper-shelf energy; the method of Regul atory Guide 1.99, Revision 2, which is the same procedure used in Revision 1, and the method presented in BAW-180319 which was developed specifically to address the need for an estimating method for this class of weld metals. The method of Regulatory Guide 1.99, Revision 2, show that all of the weld metals used in the fabrication of the beltline region of the reactor vessel will have an upper-shelf energy below 50 ft-lbs prior to the 32 EFPY design life based on the T/4 wall location. Regulatory Guide 1.99 method predicts 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 none of the reactor vessel material upper-shelf energies will decrease to below 50 ft-lbs during the vessel design life. The uncertainties of the procedures used to evaluate the matericls upper- ! shelf energies necessitates a conservative approach to the problem to insure that the requirements of 10CFR50, Appendix G, are satisfied. Therefore, a l fracture analysis based on the most limiting weld metal was performed to I insure that the operating limitations would not compromise the specified , margins of safety. The details of this fracture analysis are described in Section 9. l l 7-5 Babcock &Wilcox a McDermott company
I 1 Table 7-1. Comparison of Caosule RS1-F Tension Test Results Room Temo Test Elevated Temo Test l) Ilnirr Irrad Unirr Irrad* Base Metal -- C5062-1, Transverse Fluence, 10 19 n/cm2 (E > 1 MeV) 0 1.42 0 1.42 ; Ultimate tensile strength, ksi 83.8 88.2 82.9 83.7 0.2% yield strength, ksi 63.9 69.3 57.5 62.0 l l Uniform elongation, % 16 12 18 11 Total elongation, % 27 25 24 23 Reduction of area, % 66 59 57 57 Weld Metal -- WF-193 Fluence, 10 I9 n/cm2 (E > 1 MeV) 0 1,42 0 1.42 Ultimate tensile strength, ksi 83.5 98.6 80.6 90.6 0.2% yield strength, ksi 67.5 83.9 61.6 72.6 Uniform elongation, % 16 12 14 10 l Total elongation, % 29 26 21 18 Reduction of area, % 63 59 52 38
*The test temperature is 580F.
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Table 7-3. OMerved Vs. Predicted Changes for Cgsule SI-F Irradiated Charpy Impact Properties - 1.42 x 10 n/cm (E > 1 MeV) Predicted - Ii! Material Observed Predicted RG 1.99/2 (a) RG 1.99/2+M(b) Increase in 30 ft-lb Trans. Temo.. F I{' Base Material (C5062-1) Transverse 68 91 125 Heat-Affected Zone (C5062-1) 35 91 125 Weld Metal (WF-193) 166 216 272 Decrease in Charov USE. ft-lb Base Material (C5062-1) Transverse 11 20 N.A. Heat-Affected Zone (C5062-1) 18 21 N.A. Weld Metal (WF-193) 20 31 N.A. (a)Mean value per Regulatory Guide 1.99, Revision 2, May 1988. (b)Mean value per Regulatory Guide 1.99, Revision 2, May 1988, plus E margin (20). 5 N. A. - Not applicable. I I I I Il l 7-8 I i Babcock & Wilcox 5j a McDermott company g<
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- 8. DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY PRESSURE - TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) of Rancho Seco are established in accordance with the requirements of 10CFR50, Appendix G. The methods and criteria employed to establish operating pressure and temperature limits are described in topical report BAW-10046A, Rev. 2.25 The objective of these limits is to prevent nonductile failure during any normal operating condition, including anticipated operation occurrences and system hydrostatic tests. The loading conditions of interest include the following:
- 1. Normal operations, including heatup and cooldown.
- 2. Inservice leak and hydrostatic tests.
- 3. Reactor core operation.
The major components of the RCPB have been analyzed in accordance with 10CFR50, Appendix G. The closure head region, the reactor vessel outlet nozzle, and the beltline region have been identified as the only regions of the reactor vessel (and consequently of the RCPB) that regulate the pressure-temperature limits. Since the closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt [ preload), this region largely controls the pressure-temperature limits of the first several service periods. The reactor vessel outlet nozzle also affects l the pressure-temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle, l which can be two to three times the membrane stresses of the shell. After the first several years of neutron radiation exposure, the RT NDT f the beltline region materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-temperature limits of the RCPB. For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained l 8-1 Babcock &Wilcox A MCDermott Comparty
through a point-by-point comparison of the. limits imposed by the closure head region, the outlet nozzle, and the beltline region. The maximum allowable pressure is taken to be the lowest of the three calculated pressures.
- The limit curves for Rancho Seco are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of the twenty-fourth EFPY. The twenty-fourth EFPY was selected because it represents a logical sequence from the previous analysis. The surveil-lance capsule was withdrawn at the end of the refueling cycle when the estimated capsule fluence corresponds to approximately the inside surface end-of-life value. The time difference between the withdrawal of this surveillance capsule and future operating requirements provides adequate time for re- establishing the operating pressure and temperature limits for j subsequent periods of operation beyond the current surveillance capsule withdrawal. i
( The unirradiated impact properties were determined for the surveillance beltline region materials in accordance with 10CFR50, Appendixes G and H. ; For the other beltline region and RCPB materials for which the measured properties are not available, the unirradiated impact properties and residual j elements, as originally established for the beltline region materials, are ' listed in Table A-1. The adjusted reference temperatures are calculated by 1 adding the predicted radiation-induced RT and the unirradiated RT NDT. The NDT predicted RT is calculated using the respective neutron fluence and copper NDT and nickel contents. Figure 8-1 illustrates the calculated peak neutron fluence at several locations through the reactor vessel beltline region wall. The supporting information for Figure 8-1 is described in Section 6. The ) neutron fluence values of Figure 8-1 are the predicted fluences that have been demonstrated (Section 6) to be conservative. The design curves of 1 Regulatory Guide 1.99, Rev. 2, were used to predict the radiation-induced RT NDT values as a function of the material's copper and nickel content and 1 neutron fluence. l The neutron fluences and adjusted RT values of the beltline region NDT materials at the end of the twenty-fourth full-power year are listed in Table ( 8-1. The neutron fluences and adjusted RT NDT values are given for the 1/4T and 3/4T vessel wall locations (T - wall thickness). The assumed RTNDT Of i l ( 8-2 Bat > cock & Wilcox 1 a McDermott company
the closure head region and the outlet nozzle steel forgings is 60F, in accordance with BAW-10046, Rev. 2. The chemistry factors for two of the weld metals in the beltline region were recalculated in accordance with the procedures described in Regulatory Guide 1.99, Revision 2, Section 2. The data used to calculate a new chemistry factor for weld metal WF-154 was obtained from the B&WOG Integrated Reactor Vessel Surveillance Program.1 The data for weld metals WF-112 and WF-193, both of which have the same wire heat (Heat No. 406L44) as WF-154. A summary of the available data is as follows. Capsule Weld Metal Fluence, n/cm2 RTNDT, F Reference OCI-E WF-112 1.50E+18 78 26 0CI-A WF-ll2 8.95E+18 191 27 OCI-C WF-ll2 9.86E+18 185 28 AN1-E WF-193 7.27E+17 105 29 AN1-A WF-193 1.03E+19 151 30 RSI-B WF-193 3.99E+18 99 2 RSI-D WF-193 6.60E+18 152 3 DB1-LG1 WF-112 8.21E+18 204 31 The analysis of these data produced a new chemistry factor for WF-154 of 180. Similarly, the data used to calculate a new chemistry factor for weld metal WF-70 was obtained from the B&WOG-IRVSP and included irradiated WF-70 data and data for WF-209-1 which has the same weld wiro heat (Heat No. 72105) as t WF-70. A summary of the available data is as follows. I Capsule Weld Metal Fluence, n/cm2 RTNDT, F Reference DB1-LG1 WF-70 6.63E+18 135 31 0CII-C WF-209-1 1.02E+18 45 32 0CII-A WF-209-1 3.37E+18 114 33 OCII-E WF-209-1 1.21E+19 179 34 0CIII-A WF-209-1 8.10E+17 48 35 0CIII-B WF-209-1 3.12E+18 64 36 lhe analysis of these data produced a new chemistry factor for WF-70 of 148. l 8-3 Babcock &WHcox a McDermott company
I This reactor vessel contains welds made with weld wire Heat No. 72105 and, therefore, is required to include in the analysis the atypical weld metal as directed by the NRC in their approval of BAW-10144A. A chemistry factor was calculated for the atypical weld based on the approach described in BAW-10144 and the procedures defined in Regulatory Guide 1.99, Revision 2. A summary of the available atypical weld data is as follows. Capsule Fluence, n/cm2 RTNDT, F Refereace CR3-B 1.17E+18 28 37 CR3-C 6.56E+18 122 38 CR3-D 7.50E+18 119 39 CR3-F 1.08E+19 120 40 The analysis of these data produced a chemistry factor for the atypical weld metal of 123. These three chemistry factors were used as shown in Taule 8-1 to determine the correct Reference Temperature values for the calculation of the pressure-temperature operating limits. l Figure 8-2 shows the reactor vessel's pressure-temperature limit curve for l normal heatup. This figure also shows the the core criticality limits as required by 10CFR50, Appendix G. Figures 8-3 and 8-4 show the vessel's pressure-temperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively. All pressure-temperature limit curves arc applicable up to the twenty fifth EFPY. Protection against , 1 nonductile failure is ensured by maintaining the coolant pressure below the ' upper limits of the pressure-temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to go critical until the pressure-temperature combinations are to the right of the ll criticality limit curve. To establish the pressure-temperature limits for i protection against nonductile failure of the RCPB, the limits presented in Figures 8-2 through 8-4 must be adjusted by the pressure differential between g the point of system pressure measurement and the pressure on the reactor s: vessel controlling the limit curves. This is necessary because the reactor vessel is the most limiting component of the RCPB. I' Babcock & Wilcox g
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- 9. LOW UPPER-SHELF FRACTURE TOUGHNESS ANALYSIS L1. Backaround Paragraph VI. A.1,10CFR50, Appendix G, specifies that the reactor beltline materials must have Charpy upper-shelf energy of no less than 50 ft-lb throughout the life of the vessel. Otherwise, it must be demonstrated in a manner approved by the Director of Nuclear Regulation that lower values of upper-shelf energy will provide margins of safety against fracture equivalent to those required by Appendix G of the ASME Code.
The pressure-temperature curves presented in Section 8 are based on the linear-elastic fracture mechanics methods described in Section III, Appendix G of the ASME Code and the analytical procedures documented in topical report BAW-10046A, Rev. 2. Accordingly, the material reference toughness data (i.e. K curve) of Appendix G, with an upper-shelf value of 200 ksi M, was used IR to calculate the allowable pressures. The radiation induced reduction in fracture toughness of the weld metals progresses as the fluence in the reactor vessel increases; concurrently, the Charpy upper-shelf impact energy decreases. Certain of the weld metals in the reactor vessel are expected to reach values below 50 ft-lb at some time before E0L. At which time, an l alternate low upper-shelf fracture analysis becomes necessary. The methodo- j i logy for the elastic-plastic fracture analysis is documented in topical report BAW-10046A, Rev. 2. NRC approval of B&WOG procedures for lower upper- ) i shelf toughness fracture analysis was received in 1986. p The purpose of this section is to provide a low upper-shelf elastic-plastic i I toughness analysis based on the controlling weld in the Rancho Seco reactor vessel. The analysis is based on the flaw described in'Section III, Appendix G of the ASME Code, and using the B&WOG elastic-plastic fracture mechanics analytical methods as approved by the NRC. The material fracture toughness data is obtained from the B&WOG IRVSP (BAW-1920P).31 This analysis is ( ) 9-1 Babcock &WHcom I a McDermott company
I included to demonstrate the adequacy of the " low-upper-shelf-material" such an evaluation and provide validity to the pressure-temperature operating l limits described in Section 8. Also, it is intended to provide an assessment of the reactor vessel integrity to both the licensee and the NRC that the l reactor vessel will have adequate upper-shelf fracture toughness for the 32 EFPY design life. 9.2. Charny Upoer-Shelf Imoact Enerav All welds are designated as shown in a cross section of the beltline region of the Rancho Seco reactor vessel presented in Figure A-1. The requirement for an alternate elastic-plastic fracture analysis starts three years prior to when it is estimated that the controlling Charpy upper-shelf energy will g drop below 50 ft-lbs. Frca Table 7-6 the most limiting weld is identified as 5' WF-70. The Charpy upper-shelf energy at 32 EFPY is estimated as 43 ft-lbs using the procedure in RG 1.99, Revision 2, and to be greater than 50 ft-lbs at 32 EFPY using the procedure described in BAW-1803. It is conservative to state, based on the limited data, that the Charpy upper-shelf energy is expected to decrease to a value below the 50 ft-lb level prior to 32 EFPY. Using this approach it is not necessary to accurately predict when the weld metal drops below the 50 ft-lb regulatory value. Weld metal WF-70 is identified in Table 7-6 as the controlling weld metal to exhibit a Charpy upper-shelf energy value less than b ft-lbs, therefore, a low upper-shelf toughness fracture analysis is appropriate at this time. g Since the regulatory requirement is based on Charpy energy l evel , these 5 values are provided to determine the " trigger point" for the required low Charpy upper-shelf analysis, however, for demonstrating an adequate margin on the material toughness requires fracture toughness data, i.e. J-resistance curves, and not Charpy upper-shelf energy data. These toughness data are available from the B8WOG IRVSP. Data are available for this weld metal at a fluence of 6.6 x 10 18 n/cm2 which provides a conservative basis for demon-strating adequate upper-shelf toughness for 32 EFPY since the Rancho Seco 10 reactor vessel estimated T/4 fluence at 32 EFPY is 4.96 x 10 n/cm2 , I I 9-2 I Babcock &WHcox a McDermott company
9.3. Material Fracture Touchness properties ' To perform an elastic-plastic fracture analysis, material J-resistance curves and corresponding tensile properties of the limiting welds in the reactor vessel are needed. These material data must be representative of the particular reactor vessel fluence value and corresponding operating tempera-ture range. The operating temperature range for the reactor vessel of Rancho St.co Unit 1 is 550-570F. The estimated end-of-life fluence level at the T/4 18 thickness of the vessel wall is 4.96 x 10 n/cm2 . There are currently available several sets of irradiated fracture toughness data for WF-70 welds in the B&W data base at a fluence value of 6.6 x 10 18 n/cm2 . Therefore, instead of analyzing the reactor vessel for 24 EFPY, which would correspond to the pressure-temperature operating limit curves in Section 8, this evaluation will be performed for 32 EFPY (the end of the design life). There are two sets of fracture toughness data available for 550F. Another two sets of data obtained at 480F are also available. All four J-R curves from these data were examined and the two larger specimens were selected as the basis for this analysis. The two sets of data have nearly identical J-R curves. A conservative power law fit was made through the selected two J-R curves and the results used to perform this evaluation. There are two irradiated tensile data sets: one at 580F and the other at 450F. An evaluation of these data showed closely grouped yield und ultimate strength values. The corresponding Ramberg-Osgood parameters are calculated from these tensile data in accordance with the analytical model recommended in the f :1 Document of the ASME Code, Section XI. These calculated values are pres w d in Table 9-1. 9.4. Analytical Method and Accentance Criteria The elastic-plastic fracture mechanics analytical method, in the form of Failure Assessment Diagram (FAD), as used in this analysis is described in BAW-10046A, Rev. 2. The predicted instability point, and the J at the crack propagation initiation point is obtained from this analysis. , In the temperature range of the upper-shelf toughness region, there is a negligible contribution of KI from thermal gradient loading, therefore, only / the pressure load was considered. The present acceptance criteria using I f 9-3 Babcock &WHcox a McDermott company
i I DPFAD is summarized below and a more detailed description is presented in BAW-10046A, Rev. 2.
- 1. P is iUit > 3000 psi - where the pressure at crack initiation, Pdefinedasapressurea tion initiates.
J(a,P - P init) - J (6a R - 0.04 in.) where: a - T/4
- 2. J(a, P - 2500 x 2) < J (Aa R - 6a )c where: a - T/4 oac - crack depth at the instability point.
9.5. Fracture Analysis A fracture mechanics analysis was performed using the properties of the WF-70 weld as the most limiting weld metal in the Rancho Seco Unit I reactor , vessel. The DPFAD analysis methodology was applied using the input data described in Table 9-1 for 2500 psi pressure. The results are presented in , Figures 9-1 and 9-2 and the calculated safety margins are presented in Table 9-2. - In accordance with the acceptance criteria discussed in the preceding subsection, P (crack initiation pressure), the pressure at J -J IC' init 3 should be greater than 3000 psi. As shown in Table 9-2, the safety factor at 5 Aa - 0.04 in. (1 mm), where JR is equivalent to JIc, is 1.65 when the applied pressure is 2500 psi. The resulting P init is 4122 psi. Thus, the first acceptance criterion is met with an adequate margin. The safety factor at l the instability point (Table 9-2) is 2.25, which is greater than the required g safety factor of 2.0. Therefore, the second criterion is met. 9.6. Summary The results of this fracture analysis demonstrate that the most limiting low ! upper-shelf weld (WF-70) has irradiated fracture toughness characteristics which will assure adequate margins of safety in accordance with the require-l ments of 10CFR50, Appendix G, for fluence values equivalent to 32 EFPY operation of the reactor vessel. Consequently, the pressure-temperature limit curves calculated in Section 8 are valid for 24 EFPY without any E restrictions because of upper-shelf energy requirements. 5 I 9-4 I Babcock &WHcox g a McDermott company g
Table 9-1. Input Data for Deformation Plasticity Failure Assessment Diaaram n (Ramberg-Osgood Constant).......... 10.7860 Alpha (Ramberg-Osgood Const.)........ 1.5612 Poisson Ration....................... 0.3 E (Youngs Modulus, MSI).............. 27.45 YS (Yield Strength, KSI)............. 84.1 US (Ultimate Strength, KSI).......... 100.9 A (Crack Depth, Inches).............. 2.125 L (Crack Length, Inches)............. 12.75 T (Thickness, Inches)................ 8.5 RI (Inside Radius, Inches)........... 85.5 H1 (Calibration Function) . . . . . . . . . . . . 6.3361 P (Pressure, psi).................... 2500.0 F l a w Ty p e . . . . . . . . . . . . . . . . . . . . . . . . . . . . No. 2*
- Semi-Elliptical Surface Crack in cylinder i
) l l l l I 1 l [ 9-5 Babcoc.!c & Wilcox a McDermott company
u I Table 9-2. Failure Assessment Data Points DELTA-A Sr' Kr' S.F. 0.0400 0.2901 0.5922 1.6489 0.0800 0.2915 0.5271 1.8428 0.1200 0.2930 0.4946 1.9549 0.1600 0.2945 0.4742 2.0308 0.2000 0.2961 0.4600 2.0865 0.2400 0.2977 0.4496 2.1275 l 0.2800 0.2993 0.4416 2.1607 0.3200 0.3010 0.4354 2.1847 0.3600 0.3027 0.4305 2.2040 0.4000 0.3044 0.4266 2.2198 0.4400 0.3062 0.4234 2.2294 O.4800 0.3080 0.4209 2.2367 0.5200 0.3098 0.4188 2.2449 ! 0.5600 0.3117 0.4172 2.2483 0.6000 0.3136 0.4160 2.2503 0.6400 0.3156 0.4151 2.2510* 0.6800 0.3176 0.4144 2.2480 0.7200 0.3196 0.4140 2.2466
- Instability point.
I-I I 9-6 Babcock &Wilcom a McDermotr company
Figure 9-1. Failure Assessment Diagram for Rancho Seco Unit-1 Reactor Vessel Based on Weld Metal WF-70 1.28 Initiation Point 8.96 - Instability Point . 8.72 - Mr 8.48 - 8.24 - 8.88 8.88 8.31 8.62 8.92 1.23 1.54 Sr Figure 9-2. Safety Factors Vs. Crack Extension for Rancho Seco Unit-1 Reactor Vessel 2.58 a 2.88 - Minimum t Safety Factor of 2 Instability e Point t Initiation Point y 1.58 - Minimum Safety Factor for F aa - 0.04 in. (initiation a of crack growth) c 1.88 - l t l o P i 8.58 l-I 8.88 8.88 8.15 8.38 8.45 8.68 8.75 Delta _a, in. 9-7 ) Babcock &Wilcom j a McDermott conpany
i 10.
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in the third surveil-lance capsule, RSI-F, removed for evaluation as part of the Rancho Seco Unit-1 Reactor Vessel Surveillance Program, led to the following conclusions:
- 1. The capsule received an average fast fluence of 1.42 x 10 19 n/cm2 (E > 1.0 MeV). The predicted fast fluence for the reactor vessQ T/4 location at the end of the seventh fuel cycle is 1.68 x 10 n/cm2 (E > 1 MeV).
- 2. The fast fluence of 1.42 x 10 19 n/cm2 (E > 1 MeV) increased the RT of the capsule reactor vessel core region shell materials a mabumof166F. 1
- 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 Rancho Seco Unit-1 reactor pressure vessel inside ygurfacg will receive in 40 calendar year's operation is 8.87 x 10 n/cm (E > 1 MeV).
- 4. The increase in the RT for the shell plate material was in good agreement with that pEicted by the currently used design curves of RT versus fluence (i.e., Regulatory Guide 1.99, Revision 2),
andt$Tprediction techniques are conservative.
- 5. The increase in the RT for the weld metal was in good agreement withthatpredictedanNhepredictiontechniquesareconservative.
i 6. The low upper-shelf energy fracture analysis demonstrated that the most limiting weld metal has adequate irradiated toughness proper- l ties to assure safe operation to 32 EFPY.
- 7. The RT values decreased for 32 EFPY because of an decrease in
' the esEIbated E0L fluence values and are below the PTS screening !
l criteria for the current license duration. )
- 8. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in weld metal Charpy upper-shelf proper-ties due'to irradiation are conservative.
- 9. The analysis of the neutron dosimeters demonstrated that the 10-1 Babcock &Wilcox a McDermott company
I analytical techniques used to predict the neutron flux and fluence were accurate. l
- 10. The capsule design operating temperature may have been exceeded B' during operating transients but not for times and temperatures that 5 would make the capsule data unreliable.
I I I i I Il I I I I I-I Il I I 10-2 I Babcock & Wilcox a McDermott company
l l i
- 11. SURVEILLANCE CAPSULE REMOVAL SCHEDULE Based on the postirradiation test results of Capsule RS1-F the following schedule is recommended for the examination of the remaining capsules in the Rancho Seco Unit-1 RVSP:
Evaluation Schedule (a) Estimated Vesse}gFluenge, 10 n/cm Estimatedggpsule 2 EstimatedDatib,c) Capsule ID Fluence, 10 n/cm Surface T/4 Data Available RS1-E(c) 1.4 0.30 0.17 1988 RSI-C(c) 1.1 0.30 0.17 , 1988 (a)ln accordance with BAW-10100A and E-185-79 as modified by BAW-1543, Rev. 2, Addendum 1. (b) Estimated date based on 0.8 plant operation factor. (c) Capsules designated as standbys and are not to be evaluated. s ) l 1 ; l 11-1 Babcock &Wilcox a McDermott company
l l
- 12. CERTIFICATION ,
j The specimens were tested, and the data obtained from Sacramento Municipal ; Utility District, Rancho Seco Unit-1, reactor vessel surveillance Capsule RSI-F were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10CFR50, ! Appendixes G and H. SE. YNNW881 ' AT L. Lowe, Jr. , P.E.(/ # Date Project Technical Manager This report has been reviewed for technical content and accuracy. W ?
'Date L. q. Gross, P.E. (Material Analysis)
M&SA V it D. A. Nitti (Fluence Analysis) N D' ate 1 Performance Analysis Unit ex %~ K.K.Yoonp.E.(FractureAnalysis)
%MDat'e M&SA Unit i
Verification of independent review.
- 2lh
) A. D. McKim, Manager Date i M&SA Unit i This report has been approved for release. ll } ' J . F."Walters / /'Datp i f PfogramManager l [ 12-1 Babcock &WHcox l a McDermott company
._ ____...____.._._.__.___.._..__.____________O
E I APPENDIX A Reactor Vessel Surveillance Program l Background Data and Information I I I I I I I I Babcock &Wilcox a ucoermore company
I
- 1. Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with E-185-73, are shown in Table A-1. The locations of these materials within the reactor vessel are shown in Figures A-1 and A-2.
- 2. Definition of Beltline Reaion The beltline region of Rancho Seco Unit-1 was defined in accordance with the data given in BAL'-10100A.41
- 3. Caosule Identification The capsules used in the Rancho Seco Unit-1 surveillance pro. gram are iden-E tified below by identification number, type, and location. 5 Caosule Cross Reference Data Number Tvoe RS1-A III RS1-B IV :
RS1-C III RS1-D IV l l RS1-E III I RS1-F IV
- 4. Specimens for Surveillance Caosule See Tables A-2, A-3 and A-4.
I I I-1 I I A-2 Babcock &Wilcox g a McDermott company 5
, 0 0 0 0 0 0 0 0 0 0 0 y
r t s 9 3 3 0 0 7 5 5 7 4 5 i 0 1 1 1 1 1 1 1 1 1 1 m P 0 0 0 0 0 0 0 0 0 0 0 e h 0 0 0 0 0 0 0 0 0 0 0 C 5 2 2 0 0 6 2 0 6 7 2 u 1 1 1 1 1 1 2 2 1 2 2 C 0 0 0 0 0 0 0 0 0 0 0 t n e
- T t
nf oo T E 0 1 4 0 1 0 0 1 C R n1 t o nit b 0 8 2 0 - - - - - - eti ,l 1 mcn E - St 3 1 9 8 9 9 - eeU Uf l l Eeo e s S c r l e e5E, - - - - - - - - - - - arS Nv3LF - - - - - - - - - - - uo Vs C n M df o a i h . r sdc aT , een t a b 0 0 - - - - - R sa 0l - 0 - - . D 5 F - 8 8 - 0 - - - - - - s UR t - - 1 - - - - - - n d y f o ns - o r i t al - a
- a a b
sis h C l a l ) ) ic erl n - s s f i ea i t v v i tti d f C C 1 3 5 6 3 5 4 6 l u 6, 6, 4, a rar t F - 0 0 4, 2, 4, 4, 2, u eMe i 0 - 1 1 6 0 9 0 7 9 5 0 Q p t g 1 ( ( 5, 6, 3, 3, 3, 3, 3, 3, ona n e o 0 0 6 5 9 3 9 9 3 r rom L t A 5 6 4 7 4 4 1 4 4 3 4 u Pi d gm e t ea c cRr , o tF r a peo a w P 0 0 0 0 0 - - mnr I iP p T o D 1 1 1 2 2 d l r N e l DT W dte el c d t en n e a aBa n , i l a e s dfl eld n c pl t r aoi i 2 - - 9 o _ r e ndel aiWrm
- 3 2 6 - - 4 p rav tM ec - - - - - - 1 - - - 2 e i t r s ot - R
_ nau UDS Dr iet n e t o C s C e T _ 1
. s
_ t l l l l .) .) c) . . . l
- l l l l l g% c% r% g) g) c a
_ A e n e e e e e n0 r0 i 0 n% n% r i nno b h h h h oe i0 c0 o7 o3 i r i oi s s s s c1 1 l 2 l 7 c e e lit e
? l
( ( e( ( ( r t l t ga r r r r r r l r r a - b l ec l z e e e e em em d m em em e M a eR o z p p w w pa pa d a wa wa w B L o p p o o pe pe0 e oe oe o m - T N U U L L Us Us0 iMs L s Ls L o r f 2 B B B B s n r r r r o l 1 a C G G G G i irp e . , , , , t i ey 8 3 3 3 3 s tT 0 3 3 3 3 op a 5 5 5 5 5 d d d d d d M - - - - - l l l l l l m - A A A A A e e e e e e o S S S S S W W W W W W c - l a l a , o
. 1 2
1 2
- ic i . N 1 2 2 0 0 3 4 3 m rt 8 6 6 7 7 9 3 5 9 0 3 e 0 0 0 0 2 2 1 2 7 2 h ent 2 t ea 4 5 5 5 5 - - - - , - - C ad e V - - - - F F F F F F I MI H Z C 0 C C W W W W W W *
( Y" k5I l 0ia v.5_8M
II Table A-2. Test Specimens for Determining Material Baseline Properties No. of Test Soecimens Tension ,_, Compact Fracture Ej Material Description 70F 600F"' CVN Impact Toughness 5' Heat LL Base metal Transverse direction 3 3 15 -- Longitudinal direction 3 3 15 -- Heat-affected zone (HAZ) Transverse direction 3 3 15 -- l Longitudinal direction 3 3 H = Total 12 12 60 -- Heat MM Base metal Transverse direction 3 3 15 -- Longitudinal direction 3 3 15 -- Heat-affected zone (HAZ) Transverse direction 3 3 15 -- g Longitudinal direction 3 3 H _- E Total 12 12 60 -- Longitudinal direction 3 3 15 8 (0.5 TCT) i (a) Test temperature to be the same as irradiation temperature. I. I' I A-4 I. Babcock &WHcox E) a McDermott company 5) )
Table A-3. Soecimens in Surveillance Caosules (Designation A. C and El No. of Test Specimens Material Description Tension CVN Impact Weld metal 2 12 Weld, HAZ Heat LL, transverse - 12 Heat MM, transverse - 6 Base metal Heat LL, transverse 2 12 Heat MM, transverse - 6 Correlation material - 6 Total per capsule 4 54 Table A-4. Specimens in Surveillance Caosules (Designation B. D and F) No. of Test Soecimens ,_, Material Description Tension CVN Impact 0.5 TCTi"> Weld metal 2 12 8 Weld, HAZ Heat LL, transverse - 12 - Base metal ! Heat LL, transverse 2 12 . ) Total per capsule 4 36 8 I (8) Compact fracture toughness specimens precracked per ASTM E399-72. I l } l A-5 Babcock & Wilcox a McDermott company
I i Figure A-1. Location and Identification of Materials Used in Fabrication of Rancho Seco Unit-1 Reactor Pressure Vessel I r
- I 1 t I
r
- ZV42G1 N0ZZLE BELT Wr 233(100%)
i =- WF -29(1005 ) C 062-2 (BOTH SEAMS) WF 154(100%) ( _ l l WF 2g , _
-- C 5070 1 LOWER SHELL (1005) C 5070 2 me WF701.0. (73%)
WF290.0. (275) f _
- WF 233 (1005) 123X264 val DUTCHMAN I
I I A-6 I Babcock &Wilcox g a McDermott company g
Fioure A-2. Location of Longitudinal Welds in udder and Lower Shell Courses
% W
- 3.45*
4 W {! I Z -f-gX UPPER SHELL lE l I 4.05 w Y W ( 4 " 14.40*
-X Z_ j -
LOWER SHELL w 14.77* Y A-7 Babcock & WHeon a McDermott company
l i l l
)
{ l APPENDIX 8 Pre-Irradiation Tensile Data i 1 i B-1 Babcock &Wilcox a McDermott company
l l l Table B-1. Tensile Properties of Unirradiated Shell Plate Material. Heat No. C5062-1 l Test l l Specimen Temp, Strenath, osi Elonaation % Red'n of 5 No. F Yield Ultimate Uniform Total Area % Base Metal. Transverse LL-606 73 62.8 83.4 16.2 26.9 64.4 LL-607 73 64.1 83.8 16.7 27.1 67.3 LL-619 73 64.7 84.1 16.4 27.9 66.4 Mean 73 63.9 83.8 16.4 27.3 66.0 Std dev'n 73 0.97 0.35 0.25 0.53 1.48 LL-621 580 57.5 83.4 18.2 26.4 53.6 LL-622 581 58.1 82.5 15.9 19.9 60.2 g LL-623 582 56.9 82.8 19.2 26.1 57.7 g Mean 580 57.5 82.9 17.8 24.1 57.2 Std dev'n 580 0.60 0.46 1.69 3.67 3.33 Heat-Affected Zone. Transverse LL-301 73 62.5 83.8 9.9 21.5 64.4 LL-302 73 60.6 81.9 9.7 20.5 66.0 LL-303 73 65.0 84.7 10.5 19.0 63.0 Mean 73 62.7 83.5 10.0 20.3 64.5 Std dev'n 73 2.21 1.43 0.42 1.26 1.50 LL-304 580 68.8 81.6 8.1 16.6 58.5 LL-305 582 64.1 81.3 9.4 16.4 59.2 LL-306 580 61.9 82.5 10.0 17.9 57.7 Mean 580 64.9 81.8 9.2 17.0 58.5 g Std dev'n 580 3.52 0.62 0.97 0.81 0.75 E Table B-2. Tensile Properties of Unirradiated Weld Metal. WF-193 Test Specimen Temp, Strenath, osi Elonaation. % Red'n of No. F Yield Ultimate Uniform Total Area % MM-002 73 67.5 83.1 16.2 30.4 63.7 MM-014 73 67.5 83.8 16.2 27.5 62.3 Mean 73 67.5 83.5 16.2 29.0 63.0 Std dev'n 73 0 0.49 0 2.05 0.99 MM-015 581 61.3 81.9 14.9 21.0 51.7 MM-017 576 61.6 80.0 13.6 21.6 55.9 MM-018 581 61.9 80.0 12.5 18.9 49.3 Mean 580 61.6 80.6 13.7 20.5 52.3 l Std dev'n 580 0.30 1.10 1.20 1.42 3.34 m B-2 I Babcock & Wilcox g a McDermott company l
kt APPENDIX C Pre-Irradiation Charpy Impact Data 1 C-1 Babcock &Wilcox a McDermott company
Table C-1. Charpy Impact Data From Unirradiated Base Material, Transverse Direction. Heat No. C5062-1 Test Absorbed Lateral Shear l Specimen Temp, Energy, Expagsion, Fracture, a No. F ft-lb 10 in. % LL-637 - 79 5.5 4 0 LL-669 - 40 14.0 13 0 LL-625 - 2 30.5 30 5 LL-650 0 30.0 29 5 LL-653 0 25.5 26 5 LL-654 + 39 45.0 40 15 l LL-672 + 70 66.0 56 25 4 LL-689 + 70 56.0 54 20 LL-681 + 74 65.0 60 60 LL-661 +130 93.0 72 100 LL-645 +218 81.0 73 100 LL-699 +582 94.0 78 100 I B i C-2 Babcock & Wilcox a McDermott company
Table C-2. Charpy Impact Data From Unirradiated Base Material, HAZ. Transverse Direction. Heat No. C5062-1 Test Absorbed Lateral Shear Specimen Temp, Energy, Expagsion, Fracture, No. F ft-lb 10 in. % LL-318 - 80 20.0 17 5 LL-306 - 79 21.5 21 15 LL-351 - 59 43.0 25 15 LL-359 - 40 61.0 50 70 LL-332 - 2 67.5 53 100 LL-322 0 60.0 49 70 ; LL-349 0 64.5 48 45 LL-376 + 40 93.0 72 100 LL-385 + 74 97.0 71 100 LL-360 +130 89.5 70 100 LL-327 +213 83.0 75 100 LL-379 +280 101.5 77 100 LL-370 +338 122.5 83 100 LL-380 +590 132.0 81 100 t ) l C-3 Babcock &WHcom a McDermott company
I' Table C-3. Charov Imoact Data From Irradiated Weld Metal. WF-193 Test Absorbed Lateral Shear Specimen Temp, Energy, Expagsion, Fracture, No. F ft-lb 10 in. % MM-066 - 79 21.5 21 10 MM-062 - 40 25.0 28 20 MM-064 - 40 15.0 14 0 MM-073 - 2 39.0 42 30 MM-071 0 34.5 34 20 MM-074 0 35.0 39 50 MM-051 + 40 51.0 53 80 4 MM-083 + 70 56.0 59 90 MM-086 + 70 56.0 57 95 MM-087 + 74 66.0 66 100 MM-081 +129 66.0 69 100 MM-019 +218 68.5 69 100 MM-059 +338 74.5 79 100 l MM-025 +590 67.5 79 100 I 1 l l C-4 I Babcock &Wilcox a McDermott company
Figure C-1. Charpy Impact Data From Unirradiated Base Metal. Transverse Orientation 100 , , - i: i a i 0 75 - 5 - y 50 - j b 2s - g 0 - - i a i O 10 i i i
~0 - "
5 08
~
0.06 -
-. ~
0.04 - E a ~ 5 0.02 - 5 t P f I I i 110 . . i i i i
- DATA SURtARY -
100 - T,37 -10r Tey (35 mtt) .21r , _
~
90 -T ey (50 rt-La) +45r m Tey (30 n La) +4F * ~
]; 80 Cy -USE (Avo) 90 ri-tes RT,g7
-10r ,
5 70 - x . 5 - y 60 - B " w y 50 - U - g u0 -
~
30
~
20 - MAttRIAL SA-533.GR.B1(T)._ 10 - FLutNet WE HEAT No. C-5062-1 (TL) i e i t ^ i > 0 200 400 500 600
-100 0 100 200 Test Temperature, F C-5 Babcock &Wilcox a McDermott company
Figure C-2. Charpy Impact Data From Unirradiated Base Metal . HAZ. Transverse Orientation 100 ; -
, ^ ,: - , -
, i M
75 g , 3 b 50 - B m Es - g 25 - g 0 0.10 , , , i i i 5 *
- g 0.08 -
3 50.06 -
= .
5 0.04 - B 3 - 5" 0.02 - 0 220 i i i i a i
- DATA
SUMMARY
~
200 - T,37 -19F Tcy (35 ntt) %r
~
180 -Tg (50 FT-LB)-19I l Tcy (30 FT-La)-52r ~
-a 160 Cy -USE (Avo) 96 ri-tes 5 RT,37
-10r
, 140 -
_8 5 120 - R l g Iz - _ E
- w . ~
80 l 40 -
* ~
E MATERIAL SA-533,Gt.B(HAl) E 20 - FLutnct NONE HEAT No. C-5062-1 l t 1 I i 1 0 300 400 500 600
-100 0 100 _ 200 ;
Test Temperature, F C-6 I Babcock &Wilcox a ucoermott company ;
l Fiaure C-3. Charov Impact Data From Unirradiated Weld Metal 1
^ ^ * ^ :
100 , i i i i I t . 75 - \ ? I B - y 50 -
- i
& 25 . .
i i i t i 1 0.10 i i i i i i E
- e g 0.08 -
2
- e a e -
l 50,06 - l ll 5 0.04 - l5 r 5 0.02 lg e lE 0 l l 110 , , i i i i
- DATA
SUMMARY
100 -T NDT tev (55 ") - E ~
;g 90 -Tg (50 FT-Ls) +3bI o
~- Tcv (M FT-Ls) -W "
; B0 Cy -USE (Avr) b6 II-IDS RT,37 _
g e
- E
, B ~ 3 g 60 - 2 o - w E 50 -
- em e
l 40 30 lE 20 e PATERut Veld Metal-Linde 80 ll - 10 - FLUENCE None HEAT No. 9 3
' I 0
-100 0 100 200 300 Test Temperature, F 400 500 600 I C-7 Babcock &Wilcox ll g a McDermott company l
APPENDIX D Fluence Analysis Methodology l 1 D-1 Babcock &Wilcox a McDermott company
I
- 1. Analytical Method A semiempirical method is used to calculate the capsule and vessel flux. The method employs explicit modeling of the reactor vessel and internals and uses an average core power distribution in the discrete crdinates transport code !
D0TIV, version 4.3. 00TIV calculates tne energy and space dependent neutron flux for the specific reactor under consideration. This semiempirical method ll is conveniently outlined in Figures D-1 (capsule flux) and D-2 (vessel flux). The two-dimensional transport code D011V was used to calculate the energy-and space-dependent neutron flux at all points of interest in the reactor system. DOTIV uses the discrete ordinates method of solution of the Boltz-mann transport equation and has multi-group and asymmetric scattering g capability. The reference calculational model is an R-0 geometric represen- E tation of a plan view through the reactor core midplane which includes the core, core liner, coolant, core barrel, thermal shield, pressure vessel, and concrete. The material and geometry model, represented in Figure D-3, uses one-eight core symmetry. In order to include reasonable geometric detail within the computer memory limitations, the code parameters are specified as l P3 order of scattering, S8 quadrature, and 22 energy groups. The P3 order of scattering adequately describes the predominately forward scattering of neutrons observed in the deep penetration of steel and water media, n g demonstrated by the close agreement between measured and calculated dosimeter 5 activities. The S8 symmetric quadrature has generally produced accurate results in discrete ordinates solutions for similar problems, and is used routinely in the B&W R-0 DOT analyses. Flux generation in the core was represented by a fixed distributed source which the code derives based on a 235 U fission spectrum, the input relative power distribution, and a normalization factor to adjust flux level to the desired power density. l Geometrical Configuration l For modeling purposes, the actual geometrical configuration is divided into three parts, as shown in Figure D-3. The first part, Model "A," is used to generate the energy-dependent angular flux at the inner boundary of Model "B," which begins at the outer surface of the core barrel. Model A includes D-2 I Babcock & WHcom a McDermott company _ _ _ _ _ _ _ _ _ _ _ _ a
a detailed representation of the core baffle (or liner) in R-0 geometry that has been checked for both metal thickness and total metal volume to ensure that the 00T approximation to the actual geometry is as accurate as possible for these two very important parameters. The second, Model B, contains an explicit representation of the surveillance capsule and associated compo-27 nents. The B&W Owners Group's Flux Perturbation Experiment verified that the surveillance capsule must be explicitly included in the D0T models used for capsule and vessel flux calculations in order to obtain the desired accuracy. The magnitude of the perturbations in the fast flux due to the presence of the capsule was determined in the Perturbation Experiment to be as high as 47% at the capsule center and as high as 10% at the inner surface of the reactor vessel. Detailed explicit modeling of the capsule, capsule holder tube, and internal components is therefore incorporated into the D0T calculational models. The third, Model "C," is similar to Model B except that no capsule is included. Model C is used in determining the vessel flux in quadrants that do not contain a surveillance capsule; typically these quadrants contain the azimuthnl flux peak on the inside surface of the reactor vessel. An overlap region of approximately 32.5 cm or 17 radial intervals is specifi-ed between Model A and Models B or C. The width of this overlap region, which is fixed by the placement of the Model A vacuum boundary and the Model B boundary source, was determined by an iterative process that resulted in close agreement between the overlap region flux as predicted by Models A and B or C. The outer boundary was placed sufficiently far into the concrete shield (cavity wall) that the use of a " vacuum" boundary condition does not l cause a perturbation in the flux at the points of interest. Macroscopic Cross Sections Macroscopic cross sections, required for transport analyses, are obtained ( with the mixing code GIP. Nominal compositions are used for the structural metal s . Coolant compositions were determined using the average boron concen-tration over a fuel cycle and the bulk temperature of the region. The core I region is a homogeneous mixture of fuel, fuel cladding, structure, and coolant. ( D-3 I Babcock & WHcox } a McDermott company
I The cross-section library presently used is the (22-neutron group and 18-gamma group) CASK 23E coupled set. The dosimeter reaction cross sections are f based on the ENDF/B5 library, and are listed in Table E-3. The measured and calculated dosimeters activities are compared in Table D-1. Distributed Source The neutron population in the core during full power operation is a function of neutron energy, space, and time. The time dependence is accounted for in the analysis by calculating the time-weighted average neutron source, i.e. l the neutron source corresponding to the time-weighted average power distribu-tion. The effects of the other two independent variables, energy and space, are accounted for by using a finite but appropriately large number of g discrete intervals in energy and space. In each of these intervals the B neutron source is assumed to be invariant and independent of all other variables. The space and energy dependent source function can be considered as the product of a discretely expressed " spatial function" and a magnitude coefficient, i.e. Sv 43g - [v/K PD 3 ,x [RPD $3X g] (D-1) Y Y Bl magnitude spatial B l where:
= Energy-and space-dependent neutron source, n/cc-sec, Il Sv$3g v/K -
=
Fission neutron production rate, n/w-sec, Average power density in core, w/cc, l PD RPD - Relative power density at interval (1,j), unitless,
$3
- Fission spectrum, fraction of fission neutrons having energy Xg in group "g," ll, i - Radial coordinate index, j - Azimuthal coordinate index, g - Energy group index. g ll The spatial dependence of the flux is directly related to the RPD distribu- gl tion. Even though the entire (eighth-core symmetric) RPD distribution is i modeled in the analysis, only the peripheral fuel assemblies contribute I'
D-4 Babcock &Wilcox 5 a McDermott company F" j
significantly to the ex-core flux. The axial average pin-by-pin RPD distri-bution is calculated on a quarter-core symmetric basis for 8 to 12 times during each core cycle for the entire capsule irradiation period. The time-weighted average RPD distribution is used to generate the normalized space and energy dependency of the neutron source. Calculations for the energy and space dependent, time-averaged flux were performed for the midpoint of each D0T interval throughout the model. Since the reference model calculation produced fluxes in the R-0 plane that are averaged over the core height, an axial correction factor was required to adjust these fluxes to the capsule elevation. The factor used (1.14) was prescribed in BAW-1485P.22 1.1. Capsule Flux and Fluenct Calculation As discussed above, the D0TIV code was used to explicitly model the capsule assembly and to calculate the neutron flux as a function of energy within the capsule. The calculated fluxes were used in the following equation to nbtain calculated activities for comparison with the measured data. The calculated activity for reaction product Dj, in (pCi/gn) is: N fj bo(E)p(E)bFj(1-e t
-Ai d) e "Ai(T 7j) (D-2) n Dj - (3.7 x 104)An E j where:
N - Avogadro's number, An - Atomic weight of target material n, fj = Either eight fraction of target isotope in n-th material or the fission yield of the desired isotope, l l n(E) = Group-averaged cross sections for material n (listed in Table E-3) p(E) - Group averaged fluxes calculated by D0TIV analysis, Fj - Fraction of full power during j-th time interval, tj Aj - Decay constant of the ith isotope, T - Sum of total irradiation time, i.e., residual time in reactor, and the wait time between reactor shutdown and counting times, T j - Cumulative time from reactor startup to end of j-th time period. l D-5 Babcock & Wilcox l a McDermott company l
I tj = length of the j-th time period l Adjustments were made to the calculated dosimeter activities to correct for the effects listed below: Short half-life adju. aents to Ni and Fe dosimeter activities , 238 237 Photofission adjustments to 0 and Np dosimeter activities 238 Fissile impurity adjustments to U dosimeter activities After making these adjustments the calculated dosimeter activities were used ; with the corresponding measured activities to obtain the flux normalization factors: Dj (measured) C1- , Dj (calculated) These normalization factors were evaluated, averaged, and then used to adjust the calculated test specimen flux and fluence to be consistent with the dosimeter measurements. Additionally, the normalization factor was used to update the average normalization factor which had been derived from previous analyses. The updated normalization factor was then used to adjust the calculated vessel flux and fluence. The flux normalization factors are given in Table D-1.
- 2. Vessel Fluence Extrapolation For past core cycles, fluence values in the pressure vessel are calculated E
as described above. Extrapolation to future cycles is required to predict 5 the useful vessel life. Three time periods are considered in the extrapola-tion: 1) operation to date for which vessel fluence has been calculated,
- 2) future fuel cycles for which PDQ calculations have been performed, and 3) future cycles for which no analyses exist.
For the Rancho Seco Unit 1 analysis, time period 1 is through cycle 6, time l period 2 covers cycle 7, and time period 3 covers from the end of cycle 7 through 32 EFPY. The flux and fluence for time period 2 was estimated by l calculating the vessel flux using an adjoint-DOT calculational procedure with g the appropriate assembly-average power distributions and integrating these values over time period 2. The extrapolation of the fluence through time E E, D-6 I Babcock &WHcox a McDermott Company _____.._______._____w
I period 3 was accomplished by assuming that the average flux during period 3 was equal to the average flux for period 2 (cycle 7). lI Table D-1. Flux Normalization Factor Flux Measured Activity, (a) Calculatefb) Activity, Normalization i l PCi/a #Ci/a Factor 54Fe(n,p)54Mn 1229.13 1464.52 .958(c) 58Ni(n,p)58Co 2176.25 2624.21 1.030(d)
- I 238u (n,f)137Cs 17.27 16.03 1.077
'05.44 99.40 1.061 237Np(n,f)137Cs .
Averaged: 1.032(*) (a) Average of four dosimeters wires. (b) Average at four calculated activities. (c) Average of four ratios (one for each dosimeter wire) corrected by short half-life factor of 1.134. l (d) Average of four ratios (one for each dosimeter wire) corrected by short half-life factor of 1.236. ( (*) Average of all four dosimeters was selected as the normalization constant. lI !I I lI I I D-7 Babcock &Wilcox !I a McDermett company L_________________________________________________________ _______ . _ _ _ _
4 l Table D-2. Rancho Seco Unit 1 Reactor Vessel Fluence by Cycle Vessel Fluence, n/cm2 (c) Incremental Cumulative Vessel lux, Cycles Time, EFPY Time, EFPY n/cm s Incremental Cumulative i 1-3 2.82 2.82 1.94E+10 1.73E+18 1.73E+18 4 0.63 3.45 1.30E+10 2.57E+17 1.99E+18 5-6 1.77 5.22 1.23E+10 6.90E+17 2.68E+18 7 0,93 6.15 1.09E+10 3.20E+17 3.00E+18 1.85 8.00 0.72E+10(a) 4.20E+17(b) 3.42E+18(b) ll 7.00 15.00 0.72E+10(a) 1.59E,18(b) 5.01E+18(b) 6.00 21.00 0.72E+10(a) 1.36E+18(b) 6.37E+18(b) 3.00 24.00 0.72E+10(a) 6.81E+17(b) 7.05E+18(b) 8.00 32.00 0.72E+10(a) 1.82E+18(b) 8.87E+18(b) (8) Maximum neutron flux at inside surface of reactor vessel, based on fuel cycle designs for future cycle 7, used for extrapolation of fluence to a future times, g (b) Extrapolated values. (c) Peak fluence at inside surface of reactor vessel. Ii I I' I I I D-8 Il Babcock &Wilcom E a McDermott company B'
Figure D-1. Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the Caosule EEF/B4 GOSS SECTIONS GEDETRY & QJADRATlRE PGER DISTRI-EEF/B5 DOSDETER REAC- RR KDEL A DOT BLUIONS SINCE TOR GOSS :*LiwNS CAPSlLE INSER-TION (PDCD GIP ,r
- SORREL +,
CROSS SECTIONS C3)E v
" TDE AVE DISTRI- 1 GEDEIRY & DOT 4 : BlRED SOLR2 Sy 00ADRATLRE --* KDEL A (E, R, e ) '
P0 DEL B v ir v I DDT 4 ANGULAR FUJX KDEL B : AT BARREL PCbER HISTtRY DOSDETER OF CAPSLLE
& ACTIVITIES (PRHIST CE)E) v FINAL :
CALGLATED ACTIVITIES a EASLRED DOS M ACTIVITIES AXIAL CCRRECTION FACTIR v l v v CAPSLLE FUJX NM%LIZATION FYTIR l : M/C RATID ) l D-9 Babcock &Wifcox
) McDermott company
I Figure D-2. Rationale for the Calculation of Neutron Flux in the Reactor Vessel l GE&ETRY & DJADRATIRE Fm HDEL "A" DOT P NER DISTRI-MICROSCOPIC CROSS BUTIONS SINCE SECTIONS EMF /B4 STARTUP (PDQ EMF /B5 m EQJIVALENT) v v v WCROSCDPIC CROSS SORREL CODE SECTIONS BY E-GROLP
+
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DniKIBUTED SOLRCE Sv (E, R, e ) ll ' v + DOT 4 00 DEL A) GEDETRY #D CUADRATLRE FOR DOT KDEL B OR C v v I DUT 4 KDEL B ANGLLAR FLUX AT e 4 #D/OR KDEL C +- BARREL SLRFACE l I I
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I I e D-11 Bake &Mamm a McDermott company
1 l APPENDIX E Capsule Dosimetry Data l i 1 1 ) l l l E-1 Babcock F.Wilcox a McDermott company
I Table E-1 lists the characteristics of the neutron dosimeters. Table E-2 shows the measured activity per gram of target material (i.e., per gram of uranium, nickel, etc.) for the capsule dosimeters. Activation cross sections 235 for the various materials were flux-weighted with the 0 fission spectrum shown in Table E-3. Table E-1. Detector Composition and Shieldina Detector Material % Taraet Shieldina Reaction 238 238U (n,f)137Cs U-Al 10.38% 0 Cd-Ag 237 237 Np(n,f)137Cs Np-Al 1.44% Np Cd-Ag Ni 67.77% 58Ni 59 Cd-Ag 58Ni(n,p)58Co 59 Co(n,y)60Co l Co-Al 0.66% 00 Cd 0.66*/,59Co 59 Co(n,y)60Co Co-Al None 54 54 Fe(n,p)S4Mn Fe 5.82% Fe None Table E-2. Measured Specific Activities (Unadjusted) for Dosimeters in Caosule RSI-F Dosimeter Activity. (HCi/am of Taraet) E Detector Material Dosimeter Reaction FD5 FD6 FD7 FD8 5 58 Ni Ni(n,p)58Co 2037.55 1759.92 2543.93 2363.59 Fe 54Fe(n.p)S4Mn 1215.14 977.11 1384.28 1340.00 U-Al 238U (n,f)137Cs 17.27(a) ,,, ,,, ,,, 237 Np(n,f)I37Cs Np- Al 105.44(a) ,,, ,,, ,,, l (a) Corrected average of the measured activities for all four dosimeters. I I I I E-2 I Babcock &WHcox a McDermott company
Table E-3. Dosimeter Activation Cross Sections, b/ atom (a) 237 58 54 G Energy Range, MeV Np(n,f) 238U (n,f) Ni(n,p) Fe(n,p) 1 12.2 - 15 2.323 1.051E+0 4.830E-1 4.133E-1 2 10.0 - 12.2 2.341 9.851E-1 5.735E-1 4.728E-1 3 8.18 - 10.0 2.309 9.935E-1 5.981E-1 4.772E-1 4 6.36 - 8.18 2.093 9.110E-1 5.921E-1 4.714E-1 5 4.96 - 6.36 1.542 5.777E-1 5.223E-1 4.321E-1 6 4.06 - 4.96 1.532 5.454E-1 4.146E-1 3.275E-1 7 3.01 - 4.06 1.614 5.340E-1 2.701E-1 2.193E-1 8 2.46 - 3.01 1.689 5.325E-1 1.445E-1 1.080E-1 9 2.35 - 2.46 1.695 5.399E-1 9.154E-2 5.613E-2 10 1.83 - 2.35 1.676 5.323E-1 4.856E-2 2.940E-2 11 1.11 - 1.83 1.596 2.608E-1 1.180E-2 2.948E-3 12 0.55 - 1.11 1.241 9.845E-3 1.336E-3 6.999E-5 13 0.111 - 0.55 2.352E-1 2.436E-4 5.013E-4 6.419E-8 14 0.0033 - 0.111 1.200E-2 6.818E-5 1.512E-5 0 (a)ENDF/B5 valgg that have been flux weighted (over CASK energy groups) based on a U fission spectrum in the fast energy range plus a 1/E shape in the intermediate energy range. l t t l l E-3 Babcock &Wilcox a McDermott company
l 1 E APPENDIX F References i l F-1 Babcock &Wilcox a McDermott company
I
- 1. A. L. Lowe, Jr., et al., Integrated Reactor Vessel Material Surveillance Program, BAW-1543A. Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, May {
1985, and Addendum 1, July 1987.
- 2. A. L. Lowe, Jr., et al., Analysis of Capsule RSI-B from Sacramento l Municipal Utility District, Rancho Seco Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1702, Babcock & Wilcox, Lynchburg, Virginia, j February 1982.
l
- 3. A. L. Lowe, Jr., et al., Analysis of Capsule RSl-D from Sacramento Municipal Utility District, Rancho Seco Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1792, Babcock & Wilcox, Lynchburg, Virginia, October 1983.
- 4. G. J. Snyder and G. S. Carter, Reactor Vessel Material Surveillance I Program, Revision 3, BAW-10006A. Revision 3, Babcock & Wilcox, Lynchburg, Virginia, January 1975. E' 5
- 5. American Society for Testing and Materials, Recommended Practice for E
Surveillance Tests on Structural Materials in Nuclear Reactors, A185-66, 5 November 1966.
- 6. Code of Federal Regulation, Title 10, Part 50, Fracture Toughness Requirements for Light-Water Nuclear Power Reactors, Appendix G, Fracture Toughness Requirements. ,
- 7. Code of Federal Regulation, Title 10, Part 50, Fracture Toughness E
Requirements for Light-Water Nuclear Power Reactors, Appendix H, Reactor Ei Vessel Material Surveillance Program Requirements.
- 8. 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).
- 9. K. E. Moore and A. S. Heller, Chemistry of 177-FA B&W Owners' Group Reactor Vessel Beltline Welds, BAW-1500P, Babcock & Wilcox, Lynchburg, Virginia, September 1978.
- 10. J. D. Aadland, Babcock & Wilcox Owner's Group 177-Fuel Assembly Reactor Vessel and Surveillance Program Materials Information, BAW-1820, Babcock
& Wilcox, Lynchburg, Virginia, December 1984.
F-2 I Babcock &WHcox a McDermott company
- 11. American Society for Testing and Materials, Methods and Definitions for Mechanical Testing of Steel Products, A370-77, June 24,1977.
- 12. American Society for Testing and Materials, Methods for Notched Bar Impact Testing of Metallic Materials, E23-82, March 5,1982.
l l 13. S. Q. King, Pressure Vessel Fluence Analysis for 177-FA Reactors, BAW-1485P. Revision 1, Babcock & Wilcox, Lynchburg, Va., April 1988.
- 14. B&W's Version of D0TIV Version 4.3, One- and Two-Dimensional Transport Code System," Oak Ridge National Laboratory, Distributed by the Radia-tion Shielding Information Center as CC-429, November 1, 1983.
- 15. " CASK-40-Group Coupled Neutron and Gamma-Ray Cross Section Data,"
Radiation Shielding Information Center, DLC-23E.
- 16. Dosimeter File ENDF/B5 Tape 531, distributed March 1984, National Neutron Data Center, Brookhaven National Laboratory, Upton, Long Island, NY.
- 17. American Society of Testing Materials, Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (DPA), E693-79 (Reapproved 1985).
- 18. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel Material, Reaulatory Guide 1.99. Revision 2, May 1988.
- 19. A. S. Heller and A. L. Lowe, Jr., Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds, BAW-1803, Babcock & Wilcox, Lynchburg, Virginia, January 1984.
- 20. Code of Federal Regulations, Title 10, Part 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.
- 21. K. E. Moore, et al ., Evaluation of the Atypical Weldment, BAW-10144A, Babcock & Wilcox, Lynchburg, Virginia, February 1980.
- 22. A. L. Lowe, Jr., et al . , Pressurized Thermal Shock Evaluations in Accordance With 10CFR50.61 for Babcock & Wilcox Owners Group Reactor Pressure Vessel, BAW-1895, Babcock & Wilcox, Lynchburg, Virginia, January 1986.
I ( l F-3 Babcock &Wilcox a McDermott company
I
- 23. J. D. Aadland, Babcock & Wilcox Owners Group 177-Fuel Assembly Reactor Vessel and Surveillance Program Materials Information, BAW-1820, Babcock l
& Wilcox, Lynchburg, Virginia, December 1984.
- 24. K. E. Moore and A. S. Heller, B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study, BAW-1799, Babcock & Wilcox, Lynchburg, Virginia, July 1983.
- 25. H. W. Behnke, et al ., Methods of Compliance With Fracture Toughness and Operational Requirements of Appendix G to 10CFR50, BAW-10046A. Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, June 1986.
- 26. A. L. Lowe, Jr., et al ., Analyses of Capsule OCI-E Duke Power Company Oconee Nuclear Station -- Unit-1, Reactor Vessel Materials Surveillance Program, BAW-1436, Babcock & Wilcox, Lynchburg, Virginia, September 1977.
l
- 27. J. D. Aadland, et al., Analyses of Capsule OCI-A Duke Power Company I Oconee Nuclear Station -- Unit-1, Reactor Vessel Materials Surveillance l E l Program, BAW-1837, Babcock & Wilcox, Lynchburg, Virginia, August 1984. E l
- 28. A. L. Lowe, Jr., et al . , Analyses of Capsule OCI-C from Duke Power Company Oconee Unit-1, Reactor Vessel Materials Surveillance Program, BAW-2050, Babcock & Wilcox, Lynchburg, Virginia, October 1989,
- 29. A. L. Lowe, Jr. et al., Analysis of Capsule ANI-E from Arkansas Nuclear One, Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1440, Babcock & Wilcox, Lynchburg, Virginia, April 1977.
- 30. A. L. Lowe, Jr., et al., Analysis of Capsule ANI-A from Arkansas Nuclear One -- Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1836, Babcock & Wilcox, Lynchburg, Virginia, July 1984.
- 31. A. L. Lowe, Jr., et al., Analysis of Capsule DB1-LG1, Babcock & Wilcox Owners Group, Integrated Reactor Vessel Materials Surveillance Program, BAW-1920P, Babcock & Wilcox, Lynchburg, Virginia, October 1986.
- 32. A. L. Lowe, Jr. et al . , Analysis of Capsule OCII-C from Duke Power Company Oconee Nuclear Station, Unit 2, Reactor Vessel Materials Surveillance Program, BAW-1437, Babcock & Wilcox, Lynchburg, Virginia, May 1977. l F-4 I
Babcock & WHcox g a McDerrnott company g
- 33. A. L. Lowe, Jr., et al., Analysis of Capsule OCII-A from Duke Power Company Oconee Nuclear Station, Unit 2, Reactor Vessel Materials Surveillance Program, BAW-1699, Babcock & Wilcox, Lynchburg, Virginia, December 1981.
- 34. A. L. Lowe, Jr., et al ., Analyses of Capsule OCII-E Duke Power Company Oconee Nuclear Station -- Unit-1 Reactor Vessel Materials Surveillance Program, BAW-2057, Babcock & Wilcox, Lynchburg, Virginia, October 1988.
- 35. A. L. Lowe, Jr., et al., Analysis of Capsule OCIII-A from Duke Power Company Oconee Nuclear Station, Unit 3, Reactor Vessel Materials Surveillance Program, BAW-1438, Babcock & Wilcox, Lynchburg, Virginia, July 1977.
- 36. A. L. Lowe, Jr., et al , Analysis of Capsule OCIII-B from Duke Power Company Oconee Nuclear Station, Unit 3, Reactor Vessel Materials Surveillance Program, BAW-1697, Babcock & Wilcox, Lynchburg, Virginia, October 1981.
- 37. A. L. Lowe, Jr., e_t al., Analyses of Capsule CR3-C Florida Power Corporation Crystal River Unit 3, Reactor Vessel Materials Surveillance Program, BAW-1898, Babcock & Wilcox, Lynchburg, Virginia, March 1986.
- 38. A. L. Lowe, Jr., et al . , Analyses of Capsule CR3-B Florida Power Corporation Crystal River Unit 3, Reactor Vessel Materials Surveillance l Program, BAW-1679. Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, June 1982.
- 39. A. L. Lowe, Jr., et al., Analyses of Capsule CR3-D Florida Power Corporation Crystal River Unit 3, Reactor Vessel Materials Surveillance Program, BAW-2049, Babcock & Wilcox, Lynchburg, Virginia, September l 1988.
- 40. A. L. Lowe, Jr., et al., Analyses of Capsule CR3-F Florida Power Corporation Crystal River Unit 3, Reactor Vessel Materials Surveillance Program, BAW-1899, Babcock & Wilcox, Lynchburg, Virginia, March 1986.
- 41. H. S. Palme, G. S. Carter, and C. L. Whitmarsh, Reactor Vessel Material
- Surveillance Program -- Compliance With 10CFR50, Appendix H, for Oconee-l
) F-5 Babcock &WHcox a McDermott company
1 I Class Reactors, BAW-10100A, Babcock & Wilcox, Lynchburg, Virginia, February 1975. l I I I I I I I I I I I I I I I
~
I Babcock &Wilcox a ucoermott company}}