ML20077F459
| ML20077F459 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/30/1994 |
| From: | Devan M, Harbison L, Moore K BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20077F456 | List: |
| References | |
| 77-2226, 77-2226-00, BAW-2226, NUDOCS 9412140117 | |
| Download: ML20077F459 (131) | |
Text
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OVIAHA 3UB_IC 30WER JISTRICT
- ort Cahoun Station Jnit l\ um]er 1 Evaluation f Irradiated Capsule W-275 l 1
REACTOR VESSEL MATERIALS IRRADIATION SURVElLLANCE PROGRAM NOVEMBER 1994
' $f[ It$bb 0>$bk) E5 ,
4 ANALYSIS OF CAPSULE W-275 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION UNIT NO. 1 Reactor Vessel Material Surveillance Program --
BAW-2226 BWNT Document No. 77-2226-00 Prepared for:
Omaha Public Power District i
by B&W NUCLEAR TECHNOLOGIES Lynchburg, Virginia 740Sf M. J. DeVan 7ho)1.1 Date Materials and Structural Analysis This report was reviewed and was found to be an accurate description of the work reported.
$N WYS 7-20 W L. S. Harbison Date Materials and Structural Analysis-Verification of independent review.
h Ds
- t. E."M56re, Manager 78 Dafe Materials and Structural Analysis This report has been approved for release.
W. R. Gray, L ----
m Manager kV Dath
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SUMMARY
.This report describes the results of the testing of the specimens from the third capsule (Capsule W-275) of the Omaha Public Power District's Fort Calhoun Station Unit No. I reactor vessel surveillance program which was removed in October 1993 at the end of cycle.14 after 13.57 effective full power years of operation. The objective of the program is to monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor vessel materials by:
testing and evaluation of tension and Charpy impact specimens. The reactor vessel surveillance program was designed in accordance with the requirements of 10CFR50, Appendix H, and ASTM Specification E185-66.
The results of the tension tests and Charpy impact tests indicated that the
. materials exhibited normal behavior relative to the estimated neutron fluence exposure.
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CONTENTS Pace
- 1. INTRODUCTION .......................... 1-1
- 2. BACKGROUND ........................... 2-1
- 3. SURVEILLANCE PROGRAM DESCRIPTION ................ 3-1
- 4. PRE-IRRADIATED TESTS ...................... 4-1
- 5. POST-IRRADIATED TESTS . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Visual Examination and Inventory ........... 5-1 5.2. Thermal Monitors ................... 5-1 5.3. Tension Test Results ................. 5-1 5.4. Charpy V-Notch Impact Results . . . . . . . . . . . . . 5-2
- 6. DOSIMETER MEASUREMENTS ..................... 6-1 6.1. Introduction ..................... 6-1 6.2. Dosimeter Preparation . . . . . . . . . . . . . . . . . 6-1 6.3. Quantitative Gamma Spectrometry . . . . . . . . . . . . 6-2 6.4. Dosimeter Specific Activities . . . . . . . . . . . . . 6-2
- 7. FLUENCE ANALYSIS ........................ 7-1 7.1. Introduction ..................... 7-1 7.2. Discrete Ordinates Analysis . . . . . . . . . . . . . . 7-2 7.3. Neutron Dosimetry . . . . . . . . . . . . . . . . . . . 7-6 7.4. Projections of Pressure Vessel Exposure . . . . . . . 7-12
- 8. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . . . . . . 8-1 8.1. Pre-Irradiation Property Data . . . . . . . . . . . . . 8-1 8.2. Irradiated Property Data ............... 8-1 8.2.1. Tensile Properties ............ 8-1 8.2.2. Impact Properties . . . . . . . . . . . . . 8-2
- 9. REFERENCES ........................... 9-1 APPENDICES A. Reactor Vessel Surveillance Program Background Data and Information . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
- 8. Pre-Irradiated Tensile Data . . . . . . . . . . . . . . . . . . . B-1 C. Pre-Irradiated Charpy Impact Data . . . . . . . . . . . . . . . . C-1 D. Tension Test Stress-Strain Curves . . . . . . . . . . . . . . . . D-1 iii
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List of Tables l l
l Table Paae 1 3-1 Specimens in Surveillance Capsule W-275 . . . . . . . . . . . . 3-2 3-2 Chemical Composition and Heat Treatment of Surveillance Materials 3-3 3-3 Neutron Flux Dosimeters . . . . . . . . . . . . . . . . . . . . 3-4 3-4 Composition and Melting Points of Thermal Monitors ...... 3-4 5-1 Conditions of Thermal Monitors in Capsule W-275 . . . . . . . . 5-3 5-2 Irradiated Tensile Properties of Base Metal and Weld Metal from Capsule W-275 . . . ... . . . . . . . . . . . . . . . . . . . . 5-4 5-3 Charpy Impact Data From Irradiated Base Metal Plate 04802-2, Heat No. A1768-1, Longitudinal Orientation .......... 5-5 5-4 Charpy Impact Data From Irradiated Base Metal Plate D4802-2, Heat No. A1768-1, Transverse Orientation ........... 5-5 5-5 Charpy Impact Data From Irradiated Base Metal Plate D4802-2, Heat-Affected-Zone, Heat No. A1768-1 ............. 5-6 5-6 Charpy Impact Data From Irradiated Weld Metal, 305414/39512 ... 5-6 5-7 Charpy Impact Data From Irradiated Standard Reference Material .
Heat No. A1008-1, Longitudinal Orientation .......... 5-7 l 6-1 Quantifying Gamma Rays .................... 6-4 6-2 Specific Activities of Capsule W-275 Dosimetry, Fort Calhoun Station Unit No. 1 . . . . . . . . . . . . . . . . 6-5 6-3 Isotopic Fractions and Weight Fractions of Target Nuclides .. 6-6 6-4 Titanium Dosimetry Measurements for Capsule W-275, Fort Calhoun Station Unit No. 1 . . . . . . . . . . . . . . . . 6-7 6-5 Copper Dosimetry Measurements for Capsule W-275, Fort Calhoun Station Unit No. 1 . . . . . . . . . . . . . . . . 6-8 6-6 Nickel Dosimetry Measurements for Capsule W-275, Fort Calhoun Station Unit No. 1 . . . . . . . . . . . . . . . . 6-9 6-7 Iron Dosimetry Measurements for Capsule W-275, Fort Calhoun Station Unit No. 1 . . . . . . . . . . . . . . . . 6-10 6-8 Uranium-238 Dosimetry Measurements for Capsule W-275, Fort Calhoun Station Unit No. 1 . . . . . . . . . . . . . . . . 6-11 7-1 Calculated Fast Neutron Exposure Rates at the Surveillance Capsule Center . . . . . . . . . . . . . . . . . . 7-14 7-2 Calculate Fast Neutron Exposure Rates at the Pressure Vessel Clad / Base Metal Interface . . . . . . . . . . . 7-17 7-3 Relative Radial Distribution of 4(E > 1.0 MeV)
Within the Pressure Vessel Wall . . . . . . . . . . . . . . . . 7-20 7-4 Relative Radial Distribution of $(E > 0.1 MeV)
Within the Pressure Vessel Wall . . . . . . . . . . . . . . . . 7-21 7-5 Relative Radial Distribution of dpa/sec Within the Pressure Vessel Wall . . . . . . . . . . . . . . . . 7-22 7-6 Nuclear Parameters Used in the Evaluation of Neutron Sensors . 7-23 7-7 Monthly Thermal Generation During the First 14 Fuel Cycles of the Fort Calhoun Reactor ................... 7-24 7-8 Measured Sensor Activities and Reactor Rates Surveillance Capsule W275 ......................... 7-26 7-9 Measured Sensor Activities and Reaction Rates Surveillance Capsule W265 ......................... 7-27 iv a
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List of Tables (continued)
Tables Paae 7-10 Measured Sensor Activities and Reactor Rates Surveillance Capsule W225 ......................... 7-28 7-11 Summary of Neutron Dosimetry Results Surveillance Capsule W275 7-29 1 7-12 Summary of Neutron Dosimetry Results Surveillance Capsule W265 7-30 7-13 Summary of Neutron Dosimetry Results Surveillance Capsule W225 7-31 7-14 Comparison of Measured and Ferret Calculated Reaction Rates At The Surveillance Capsule Center Surveillance Capsule W275 . 7-32 7-15 Comparison of Measured and Ferret Calculated Reaction Rates At The Surveillance Capsule Center Surveillance Capsule W275 . 7-32 7-16 Comparison of Measured and Ferret Calculated Reaction Rates At The Surveillance Capsule Center Surveillance Capsule W225 . 7-33 7-17 Adjusted Neutron Energy Spectrum At The Center of Surveillance Capsule W275 . . . . . . . . . . . . . . . . . . . 7-34 7-18 Adjusted Neutron Energy Spectrum At The Center of Surveillance Capsule W265 . . . . . . . . . . . . . . . . . . . 7-35 7-19 Adjusted Neutron Energy Spectrum At The Center of Surveillance Capsule W225 . . . . . . . . . . . . . . . . . . . 7-36 7-20 Comparison of Calculated and Measured Neutron Exposure Levels For fort Calhoun Surveillance Capsules .. . ..... .... 7-37 7-21 Neutron Exposure Projections at Key Locations on the Pressure Vessel Clad / Base Metal Interface .... . ...... .... 7-40 7-22 Neutron Exposure Values for Use in the Generation of Heatu Cooldown Curves . . . . . . . . . . . . . . . . . . . . . p/ ... 7-41 8-1 Tensile Properties of the Fort Calhoun Station Unit No. 1 Reactor Vessel Surveillance Materials . . . . . . . . . . . . . 8-3 8-2 Observed Vs. Predicted Changes for Irradiated Surveillance Material 30 ft-lb Transition Temperature - 1.38 x 10 a n/cm' 1
. 8-4 8-3 Observed Vs. Predicted Changes for Irradiated Surveillance Material Upper-Shelf Energy - 1.38 x 10 n/cm' .. . . .... 8-5 8-4 Comparison of Capsules W-225, W-265, and W-275 Charpy Test Results . . . . . . . . . . . . . . . . . . . 8-6 A-1 Capsule Assembly Identification . . . . . . . . . . . . . . . . A-2 A-2 Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials - Fort Calhoun Station Unit No. 1 ............ ..... . .... A-3 A-3 Type and Quantity of Specimens Contained in Each Irradiated Capsule Assembly ........... ... .... . .... A-4 B-1 Tensile Properties of Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1, Longitudinal ...... .. ... . .... B-2 B-2 Tensile Properties of Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1, Transverse .. .... . .. ... . .... B-2 B-3 Tensile Properties of Unirradiated Base Metal Plate D4802-2, Heat-Affected-Zone, Heat No. A1768-1 .. . . . . ... .... B-3 B-4 Tensile Properties of Unirradiated Weld Metal, 305414/3951 .. B-3 C-1 Charpy Impact Data From Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1, Longitudinal Orientation . . ... . .... C-2 v
List of Tables (continuedl Tables Pace C-2 Charpy Impact Data From Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1, Transverse Orientation ........... C-2 C-3 Charpy Impact Data From Unirradiated Base Metal Plate D4802-2, Heat-Affected-Zone, Heat No. A1768-1 ............. C-3 C-4 Charpy Impact Data From Unirradiated Weld Metal, 305414/3951 . C-3 List of Fiaure, Fiaure Pace 3-1 Reactor Vessel Cross Section Showing Reactor Vessel Cross Section Showing Location of RVSP Capsules in Fort Calhoun Station Unit No. 1 ........................... 3-5 3-2 Typical Surveillance Capsule Assembly Showing Location of l Specimens and Monitors ..................... 3-6 3-3 Typical Surveillance Capsule Tensile - Monitor Compartment Assembly (Three Per Capsule) .................. 3-7 ;
3-4 Typical Surveillance Capsule Charpy Impact Compartment Assembly j Four Per Capsule) . . . . . . . . . . . . . . . . . . . . . . . . 3-8 5-1 Photographs of Thermal Monitor Melt Wire Capsules a's Removed from Surveillance Capsule W-275 ................... 5-8 5-2 Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces - Base Metal, Longitudinal Orientation .... 5-9 5-3 Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces - Base Metal Heat-Affected-Zone . . . . . . . 5-10 5-4 Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces - Weld Metal, 305414/3951 . . . . . . . . . . 5-11 5-5 Charpy Impact Data for Irradiated Base Metal Plate D4802-2, Heat No. A1768-1, Longitudinal Orientation . . . . . . . . . . 5-12 5-6 Charpy Impact Data for Irradiated Base Metal Plate D4802-2, Heat No. A1768-1, Transverse Orientation ........... 5-13 5-7 Charpy Impact Data for Irradiated Base Metal Plate D4802-2, Heat-Affected-Zone, Heat No. A1768-1 ............. 5-14 5-8 Charpy Impact Data for Irradiated Weld Metal 305414/3951 ... 5-15 5-9 Charpy Impact Data for Irradiated Standard Reference Material, Heat No. A1008-1, Longitudinal Orientation .......... 5-16 5-10 Photographs of Charpy Impact Specimens Fracture Surfaces Base Metal Plate D4802-2, Heat No. A1768-1, (LT) ....... 5-17 5-11 Photographs of Charpy Impact Specimens Fracture Surfaces Base Metal Plate D4802-2, Heat No. A1768-1, (TL) ....... 5-18 5-12 Photographs of Charpy Impact Specimens Fracture Surfaces Base Metal Plate D4802-2, Heat No. A1768-1, (HAZ) . . . . . . . 5-19 5-13 Photographs of Charpy Impact Specimens Fracture Surfaces Weld Metal 305414/3951 .................... 5-20 vi
d List of Fioures (Continued)
Fioure Paae 5-14 Photographs of Charpy Impact Specimens Fracture Surfaces Standard Reference Material, Heat No. A1008-1, (LT) . . . . . . 5-21 i' 7-1 Plan View of a Reactor Vessel Surveillance Capsule . . . . . . . 7-42 .
8-1 Comparison of Inirradiated and Irradiated Charpy Impact Data.
Curves for Base Metal Plate D4802-2, Heat No.'A1768-1, Longitudinal Orientation .................... .
8-7 2 Comparison of Initradiated and Irradiated Charpy Impact Data Curves for Base Metal Plate D4802-2, Heat No. A1768-1, Transverse Orientation ..................... 8-8 8-3 Comparison of Inirradiated and Irradiated Charpy Impact Data Curves for Base Metal Plate D4802-2, Heat-Affected-Zone Heat No. A1768-1 ........................ 8-9 8-4 Comparison of Inirradiated and Irradiated Charpy Impact Data Curves for Weld Metal, 305414/3951 .............. 8-10 A-1 Location and Identification of Materials Used in the Fabrication of Fort Calhoun Station Unit No. 1 Reactor Pressure Vessel ... A-5 C-1 Charpy Impact Data From Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1, Longitudinal Orientation ........... C-4 C-2 Charpy Impact Data From Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1, Transverse Orientation ............ C-5 C-3 Charpy Impact Data From Unitradiated Base Metal Plate D4802-2, Heat Affected-Zone, Heat No. A1768-1 .............. .
C-6 C-4 Charpy Impact Data From Unirradiated Weld Metal 305414/3951 . . . C-7 .
D-1 Tension Test Stress-Strain Curve for Base Metal Plate D4802-2, Specimen No. 103, Tested at 70F . . . . . . . . . . . . . . . . . D-2 D-2 Tension Test Stress-Strain Curve for Base Metal Plate D4802-2, Specimen No. 102, Tested at 250F ................ D-2 0-3 Tension Test Stress-Strain Curve for Base Metal Plate 04802-2, Specimen No. 1EA, Tested at 550F ................ D-3 D-4 Tension Test Stress-Strain Curve for Base Metal Plate D4802-2, HAZ, Specimen No. 4EA, Tested at 70F .............. D-3 ,
0-5 Tension Test Stress-Strain Curve for Base Metal Plate D4802-2, HAZ, Specimen No. 4EJ, Tested at 250F . . . . . . . . . . . . . . D-4 D-6 Tension Test Stress-Strain Curve for Base Metal Plate D4802-2, HAZ, Specimen No. 4EE, Tested at 550F , . . . . . . . . . . . . . D-4 0-7 Tension Test Stress-Strain Curve for Weld Metal 305414/3951, Specimen No. 3EK, Tested at 70F . . . . . . . . . . . . . . . . . D-5 D-8 Tension Test Stress-Strain Curve for Weld Metal 305414/3951, Speciren No. 3J1, Tested at 250F ................ D-5 D-9 Tension Test Stress-Strain Curve for Weld Metal 305414/3951, Specimen No. 3DK, Tested at 550F ................ D-6 i vii
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- 1. INTRODUCTION This report describes the specimen test results from the third reactor vessel surveillance capsule (Capsule W-275) discharged from the Omaha Public Power District's Fort Calhoun Station Unit No. 1 (FC1). The capsule was removed and evaluated after being irradiated in the reactor vessel as part of the reactor vessel surveillance program (RVSP). The objective of the program is to monitor the effects of neutron irradiation on the mechanical properties of reactor vessel materials for the 40-year design life under actual plant operating conditions.
The FCI RVSP was designed in accordance with ASTM Standard E185-66' and fabricated by Combustion Engineering.
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- 2. BACKGROUND l The ability of the reactor pressure vessel to resist fracture is the primary
' factor in ensuring the safety of the primary system in light water-cooled reactors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to the highest levels of neutron irradiation.
The general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels such as SA-533, Grade B, used in the fabrication of the FCI 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 the ductile-to-brittle transition temperature accompanied by a reduction .in the Charpy upper-shelf energy value.
Appendix G to 10CFR50, " Fracture Toughness Requirements,"2 specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of light water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitations for operation of the RCPB. The fracture toughness and operational requirements are specified to provide adequate safety I margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10CFR50 became effective on August 16, 1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effective date.
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Appendix H to 10CFR50, " Reactor Vessel Material Surveillance Program Requirements,"' defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor l vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against non-ductile fracture throughout its service life.
l A method for guarding against brittle fracture in reactor pressure vessels is l described in Appendix G to the ASME Boiler and Pressure Vessel (B&PV) Code, l Section III, " Nuclear Power Plant Components."* This method utilizes fracture mechanics concepts and the reference nil-ductility temperature, RTuor, which is defined as the greater of the drop weight nil-ductility transition temperature 5
(per ASTM E208 ) or the temperature that is 60F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion. The RTuor of a given material is used to index that material to a reference stress intensity factor curve (Km curve), which appears in Appendix G of ASME B&PV Code Section III. The K m curve is a lower bound of dynamic and crack arrest fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Km curve, allowable stress intensity factors can be obtained for the material as a function of temperature. The operating limits can then be determined using these allowable stress intensity factors.
i The RTuor and, in turn, the operating limits of a nuclear power plant, can be adjusted to account for the effects of radiation on the properties of the reactor vessel materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which surveillance capsules containing prepared Charpy and tensile specimens of the reactor vessel materials are 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 RTuor to adjust it for radiation embrittlement. The adjusted RTuor is used to index the material to the Km curve which, in turn, is used to set operating limits for the 2-2 1
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nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.
Appendix G of 10CFR50 also requires a minimum initial Charpy V-notch upper-shelf I energy of 75 ft-lbs for all beltline region materials unless it is demonstrated that lower values of upper-shelf fracture energy will provide an adequate margin for deterioration as a result of neutron irradiation. No action is required for a material that does not meet the initial 75 ft-lbs requirement provided the irradiation deterioration does not cause the upper-shelf energy to drop below 50 I ft-lbs. The regulations specify that if the upper-shelf energy drops below 50 l ft-lbs, it must be demonstrated, in a manner approved by the Office of Nuclear l Reactor Regulation, that the lower values will provide adequate margins of safety, i
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- 3. SURVEILLANCE PROGRAM DESCRIPTION The reactor vessel surveillance program for FC1 includes six capsules designed to monitor the effects of neutron and thermal environments on the materials of the reactor pressure vessel core region. The capsules, which were inserted into the reactor vessel before initial plant startup, are positioned near the inside wall of the reactor vessel at the locations shown in Figure 3-1. The capsule positions were designed to be near the peak axial and azimuthal neutron flux.
Two previous surveillance capsules from this program have been removed and evaluated. The first surveillance capsule, Capsule W-225, was removed after the third operating cycle, and the results are reported in Combustion Engineering Report TR-0-MCM-001, Revision 1.8 The second surveillance capsule, Capsule W-265, was removed and evaluated after the seventh operating cycle. The results from the second surveillance capsule are reported in Combustion Engineering Report TR-0-MCM-002.'
Capsule W-275 was removed and evaluated after the fourteenth operating cycle of FC1. The capsule contained Charpy V-notch (CVN) impact test specimens fabricated from basemetal(SA-533,GradeB, Class 1), heat-affected-zone (HAZ) material, and a submerged-arc weld metal. The base metal CVN specimens were fabricated from material in both the transverse and longitudinal orientations. Tensile test specimens were fabricated from the base metal, HAZ, and weld metal. The number of specimens of each material contained in Capsule W-275 are described in Table i 3-1, and the location of the individual specimens within the capsule are described in Figures 3-2 through 3-4. The chemical composition and heat treatment of the surveillance materials in Capsule W-275 are described in Table 3-2.
All plate and HAZ specimens were machined from the4 '/ -thickness ('/,T) location of the plate material. Weld specimens were machined throughout the thickness of 3-1 )
.the weldment. The CVN and tensile test specimens were cut from the surveillance materials such that they were oriented with their longitudinal axes either parallel or perpendicular to the principal working direction.
There are three sets of nine neutron flux dosimeters per surveillance capsule; one set each is located in the top, middle, and bottom of the capsule. The neutron dosimeters contained in each set in Capsule W-275 are listed in Table 3- 1 3.
There are three sets of four thermal monitors of low-melting alloys also located at the top, middle, and bottom of the capsule. The eutectic alloys and their melting points are listed in Table 3-4.
Table 3-1. Specimens in Surveillance Capsule W-275 Number of Test Specimens Material Description Tension CVN Impact Base Metal Plate 0-4802-2 (Heat No. A1768-1) longitudinal 3 12 Transverse -
6 Heat-Affected-Zone 3 12 Weld Metal (305414/3951)* 3 12 Standard Reference Material -
6 (HSSTPL-01)
Total 9 48
- Weld wire heat number and flux lot identifiers.
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Table 3-2. Chemical Composition and Heat Treatment of Surveillance Materials Chemical Composition, wt%
Standard Reference Plate D-4802-2'd Weld Metal'd Material *'
Element (Heat No. A1768-1) (305414/3951) (HSST PL-01)
C 0.22 0.14 0.22 Mn 1.43 1.57 1,48 P 0.009 0.013 0.012 S 0.014 0.011 0.018 Si 0.23 0.14 0.25 Ni 0.48 0.60 0.68 Cr 0.04 0.03 ---
Mo 0.50 0.50 0.52 Cu 0.10 0.35 ---
Heat Treatment Heat No. Temp., F Time, h Cooling Base Metal, A1768-1 1575125 4 Water Quenched 1225125 4 Air Cooled 1150125 40 Furnace Cooled to 600F Weld Metal (305-414/3951) 1125125 40 Furnace Cooled to 600F Standard Reference Material 1685115 4 Air Cooled 1 (HSST PL-01) 1600150 4 Water Quenched 1250125 4 Air Cooled j Furnace Cooled to 600F 1150125 40 Chemical analysis by Combustion Engineering of surveillance program !
(a) test plate.
(b) Chemical analysis from ORNL-4314.8 1
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Table 3-3. Neutron Flux Dosimeters Threshold Material Shielding Reaction Energy (MeV) Half-Life 238 U (n,f) "'Ce Uranium None/Cd 1.1 284 days Titanium None Ti (n,p) **Sc 3.0 83.8 days Iron None Fe (n,p) . Mn 2.5 312.5 days Nickel Cd asNi (n,p) seco 2.3 70.9 days Copper Cd 83 Cu M)
- Co 6.1 5.27 years Sul fur None 32 S (n,p) 2p 2.3 14.3 days Table 3-4. Composition and Meltina Points of Thermal Monitors Alloy Composition, wt% Melting Point, F 80 Au, 20 Sn 536 90 Pb, 5 Sn, 5 Ag 558 97.5 Pb, 2.5 Ag 580 1 97.5 Pb, 0.75 Sn, 1.75 Ag 590 l
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- c. . - - . - . .. - -.
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"- Figure 3-1. Reactor Vessel Cross Section Showing Location of RVSP Caosules in Fort Calhoun Station Unit No. 1 i
A PLANT I NORTH ~ ~ ~ ~ 's Outlet Nozzle
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/ l l \ Nozzle
,,J+-core Shroud l_ /j
(\ -
- -Core Support Barrel
. Vessel Thermal Shield rVessel 0
' 8 95 'I s, / eactorvessel R / 26
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I /yessel Vessel- /
0 8 85 _ , 275 l, '
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Vessel \ e 49 \ I n -- .--> .
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Figure 3-2. Typical Surveillance Capsule Assembly Showing location of Soecimens and Monitors 1.ock Assembly Wedge Coupling Assembly i
1
- Extension Assembly -
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Tensile-MonRor-- - 'l Compartmr.nt
$p u -
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'" !> Charpy impact Compartments i
$I Tensile-Monitor Compartment %s 9 ;
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s: > Charpy impact Compartments k
s Tensile-Monitor j
Compartment (
h 3-6
Figure 3-3. Typical Surveillance Capsule Tensile - Monitor Comoartment Assembly (Three Per Caosule)
- 1 Stainless Steel Tubing l Wedge Coupling - End Cap y ,
I
., Threshold Detector l Flux Attenuation Monitor .
~
Flux Attego Monitor ,
3 Muartz Tubing Temperature Monito -
g l '. ,,
Stainless SteelTubin !I,' , Weight Threshold Detector - LowMelting Alloy
%N* --Temperature Monitoi Flux Spectrum Monitor- Housing Flu r ,N C b b fd q Stainless Steel Tubing Tensile Specimen
- Th esho d Dete or Split Spacer-
%y Tensile Specimen Housing -
Mectangular Tubing
. ~ Wedge Coupling - End Cap a
3-7
Figure 3-4. Typical Surveillance Capsule Charpy Impact Compartment Assembiv (Four Per Caosule)
/ -Wedge Coupling - End Cap l
Charpy impact Specimens
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% i Spacer. %* ,s m ,
s I
N .
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" Rectangular Tubing k -Wedge Coupling - End Cap N
3-8 l
- 4. PRE-IRRADIATION TESTS Unirradiated material was evaluated for two purposes: (1) to establish baseline data to which irradiated properties data could be compared; and (2) to determine those material properties as required for compliance with 10CFR50, Appendices G and H.
The unirradiated specimens were tested by combustion Engineering as part of the development of the FCI surveillance program. The details of the testing procedures are described in Combustion Engineering Report TR-0-MCD-001' and the results are summarized in Appendices B and C of this report.
4-1
i I
- 5. POST-IRRADIATION TESTING 5.1. Visual Examination and Inventory The contents of Capsule W-275 were inventoried and found to be consistent with the surveillance program report inventory. All specimens were visually examined.
During the capsule disassembly, it appeared that the contents in compartments 4 and 6 had experienced abnormal conditions during irradiation. In compartment 4 a loss of integrity of the quartz capsules containing the temperature monitors was observed, and the tensile specimens and dosimeters were discolored. In compartment 6, the CVN specimens were covered with black scale and bonded together. In addition, rust (bright orange color) was also noted in the V-notch area of the specimens. To determine whether any degradation had occurred in the notch areas of the specimens in compartment 6, the V-notch of these CVN specimens were examined under a stereomicroscope at a magnification of 73.5X. This inspection revealed that all the CVN specimens in compartment 6 were acceptable for subsequent impact testing per ASTM E23.'
5.2. Thermal Monitors Surveillance Capsule W-275 contained three temperature monitor holder blocks each containing four fusible alloys with different melting points. The capsule temperature monitors were inspected and the results are tabulated in Table 5-1.
As discussed in the preceding paragraph, the temperature monitors in compartment 4 are not reported due to the loss of integrity of the quartz capsules.
From these data, it was concluded that the irradiation specimens in Capsule W-275 had been exposed to temperatures between 536F and 557F during the reactor vessel operating period. There appeared to be no signs of a temperature gradient along the capsule length.
5-1 i...... .. .
5.3. Tension Test Results ,
The results of the post-irradiation tension tests are presented in Table 5-2.
Tests were performed on specimens at 70, 250, and 550F. The tests were performed on a 55,000-lb load capacity MTS servohydraulic computer-controlled universal test machine. All tension tests were run using stroke control witi, an initial actuator travel rate of 0.0075 inch per minute through the yield point.
Following specimen yielding, an actuator speed of 0.060 inch per minute was used.
The test conditions were in accordance with the applicable requirements of ASTM E8" and ASTM E21.'" The specimen fracture surfaces are shown Figures 5-2 through 5-4.
In general, the ultimate and yield strength of each material increased with a corresponding decrease in ductility as compared to the unirradiated test results.
Both these effects were the result of neutron irradiation damage.
5.4. Charny V-Notch Impact Results The results of the CVN impact testing are shown in Tables 5-3 through 5-7 and Figures 5-5 through 5-9. Photographs of the CVN specimen fracture surfaces are presented in Figures 5-10 through 5-14. The impact testing was performed in accordance with the applicable requirements of ASTM E23 on a Satec S1-1K Impact tester certified to meet NIST ("Watertown") standards.'
The CVN impact data for each material indicated an increase in transition temperature and drop in upper-shelf energy as a result of neutron irradiation exposure.
5-2
Table 5-1. Conditions of Thermal Monitors in Caosule W-275 Capsule Melt Post-Irradiation Segment Temperature Condition Compartment 1 536F Melted (Top) 558F Unmelted 580F Unmelted 590F Unmelted Compartment 4 536F N/A*
(Middle) 558F N/A*
580F N/A*
590F N/A*
Compartment 7 536F Melted l
(Bottom) 558F Unmelted l 580F Unmelted 590F Unmelted
- - Temperature monitors could be located due to the loss of integrity of the quartz capsules that contained the temperature monitors.
l I
5-3
Table 5-2. Irradiated Tensile Properties of Base Metal and Weld Metal from Caosule W-275 Strenath Fracture Properties Elonaation Reduction Specimen Test Temp, Yield, Ultimate, Load, Stress, Strength, Uniform, Total, in Area, No. F osi osi lbs osi osi % % %
Base Metal. A1768-1. Lonaitudinal 103 70 79,500 101,700 3231 188,000 65,800 12.1 26.8 65.0 102 250 74,300 94,600 3233 176,000 65,900 9.63 21.5 62.5 IEA 550 67,000 91,700 3208 162,000 65,300 9.62 20.0 59.6 Base Metal HAZ. A1768-1 4EA 70 73,600 95,600 3316 161,000 67,600 5.90 16.5 58.0 72,100 92,200 3926 149,000 79,300 5.38 11.2 46.8
{ 4EJ 4EE 250 550 69,700 94,100 3908 147,000 79,600 5.67 12.4 45.8 Weld Metal. 305414/3951 3EK 70 99,300 113,100 3950 180,000 80,500 10.0 22.6 55.4 3J1 250 92,800 107,400 3875 188,000 79,000 9.80 20.4 58.0 3DK 550 88,500 103,900 4152 161,000 84,600 7.99 16.9 47.6
l Table 5-3. Charpy Impact Data From Irradiated Base Metal Plate D4802-2, Heat No. A1768-1. Lonaitudinal Orientation Specimen Test Temp.,
Impact Energy, Lateral Expansion, Shear Fracture f No. F ft-lb mils %
14B 40 15.5 15 10 146 70 26 20 20 16B 70 50.5 16 35 160 100 25 24 40 14P 120 32.5 31 50 15J 140 49.5 43 50 ISB 160 70.5 57 65 16A 180 73 63 85 --
16C 220 88.5 72 95 14E 260 108* 83 100 137 300 107* 88 100 165 350 106* 85 100
- Values used to determine upper-shelf energy value per ASTM E185."
Table 5-4. Charpy Impact Data From Irradiated Base Metal Plate D4802-2, Heat No. A1768-1. Transverse Orientation Test Impact lateral Shear Specimen Temp., Energy, Expansion, Fracture No. F ft-lb mils %
21K 70 20.5 17 10 21M 100 28.5 28 25 228 220 63.5 53 70 22J 300 83* 70 100 235 350 89.5* 80 100 .
248 400 91.5* 81 100
- - Values used to determine upper-shelf energy value per ASTM E185.12 5-5
l i
Table 5-5. Charpy Impact Data From Irradiated Base Plate D4802-2, Heat-Affected-7one. Heat No. A1768-1 Test Impact lateral Shear Specimen Temp., Energy, Expansion, Fracture No. F ft-lb mils %
42E -20 8.5 7 0' 44C 0 25 18 30 43A 40 19.5 15 15 43T 70 20.5 17 10 ;
42A 100 62 51 65 42T 120 44 38 60 411 150 59 46 85 445 180 88* 71 100 412 220 74* 62 100 45T 280 58* 61 100 46D 280 62.5* 64 100 432 350 86.5* 76 100
- - Values used to determine upper-shelf energy value per ASTM E185.
Table 5-6. Charov Impact Data From Irradiated Weld Metal. 305414/3951 Test Impact lateral Shear Specimen Temp., Energy, Expansion, Fracture No. F ft-lb mils %
33K 70 16.5 12 0 32M 100 14.5 11 10 31A 120 20 16 30 316 150 22 20 25
- 31J 180 27.5 24 35 l 314 220 38.5 37 75 33J 260 39.5 36 75 32K 300 53.5 52 95 l 33M 350 49.5 55 90 313 400 57.5* 58 100 311 420 65* 63 100 32L 450 58.5* 54 100
- - Values used to determine upper-shelf energy value per ASTM E185.12 5-6
l Table 5-7. Charpy Impact Data From Irradiated Standard Reference Material. Heat No. A1008-1. Lonaitudinal Orientation Test Impact Lateral Shear Specimen Temp., Energy, Expansion, Fracture No. F ft-lb mils %
56T 70 6.5 6 0 56E 100 12.5 13 5 564 220 47 46 50 572 300 53.5 52 95 56K 350 90.5* 78 100 56B 400 89.5* 83 100
- - Values used to determine upper-shelf energy value per ASTM E185.
5-7
l J
l Figure 5-1. Photographs of Thermal Monitor Melt Wire Capsules
! as Removed From Surveillance Caosu'a W-275 t
536*F
\ f u .c a.: a C P u s a w e T r:. ~ ,;. Y = ~ : e - o. . + ,
w.mxmom. rgn i
g o.m . '
JJ.i , ;-
l t i:. . . .'.;& n
, ;* a, i
a;m:du f9 .-
n-r."'.558'F i A1 s2 n c .,x:p r m w e n~.r..a -
580*F WMd ML,.,,er, g w ;g n - ,f;q
- m. . ..
90*F a
. ' o %M'. '
ss ,k a V t'%'g ay e.
c_ . . .. . ,
"~ ~
Y, 536*F
.; w .
. a :..
? - V f g n p fe. xvo. w , a..<4 . , +p,34:y. .(
. ., f
. . . . ~.8 ' ' N 5 A.?.rfh. b ,Qn4
,e e ~ ^ "'~~ 'fr r a ,% 558'F A7 ' ~ " ' " ~~ - -- - "was- ~~*' -" ^**>m -u (Bottom) WQp kv
.D, 580*F vu;:.2 . 3 we . ,x m ~c u ... , + v ~ ,. . c. .. . ..
590*F .
> l 1
l' 4 % , L. shed:&hh_ : ? f.,p ,=s_ %
' [v $Q l
, y% ,.. ,' , [' ?q , _ q. *4 _
5-8
Figure 5-2. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal, Loncitudinal Orientation
,. .mw
. ;y;, gy;;
y.;
- ,. 1 !
,*t. <=..y. s
,b
. 5 ,, 3;$ p .a
.A'
=wmirm T .
l l
Specimen 1D3 (70*F)
--n w r-- :n 3 m - .. . . . y < r, --w p . .. , ;. . - 1 yp q , >
l
- ,. .__; :.* % s. -
~
8 Specimen 1D2 (250*F)
Specimen 1EA (550*F)
R, n.ui. a g, . . . .-.
Specimen 103 (70*F) Specimen 1D2 (250*F) Specimen 1EA (550*F) 5-9
1 I
Figure 5-3. Photographs of Tested Tension Test Specimens and Correspondir,;J Fractured Surfaces - Base Metal Heat-Affected Zone _ , . _ _ _ , ,
~ ~mg.g;_y:
.. A
- as_aa.w u ~,,a. w .. >
, Specimen 4EA (70*F) l l Specimen 4EJ (250*F) i
- c. . . a -,,.aa~. -
Specimen 4EE (550'F) c.. d3 , , . . . - . . . . -
y y . ..;-, ..
Specimen 4EA (70'F) Specimen 4EJ (250*F) Specimen 4EE (550*F) 5-10
l:
Figure 5-4. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Weld Metal 305414/3951 Specimen 3EK (70'F)
.. ,. ,. y ,,.., n _ - , .. , . . . .
,, y . j,'*f Y, .
'_ .i Specimen 3J1 (250*F) v-
,1
.- .~
. . A '.
Specimen 3DK (550*F)
Specimen 3EK (70*F) Specimen 3J1 (250*F) Specimen 3DK (550'F)
\
5-11 !
Figure 5-5. Charpy Impact Data for Irradiated Base Metal Plate D4802-2, Heat No. A1768-1. Lonaitudinal Orientation 100 Y
- rs -
e .,
l
$M u.
e l
s e a
e 25 -
t/J 0
- 100 0 100 200 h awAl 500-0.1 j 0.08 -
a C 0.08 -
0.04 E -
.e .0.02 3
0.100 0 100 200 300 400 500 l
- 1. 220 T """
- '~
~ T,(35 MLE) +126F i
130 T,(50 FT.LS) +13sF -
l
' T"(30 FT.LB) +esF 4 100 -
I at CvUSE (AVG.) 107 FT.LS8 140 RT -
NDT -
l 1 120 -
( k * .
- W Im -
D +
a e0 -
E e
-.E so -
........................... 4 ........ . . . . . . . . . . . . . . .
40 -
= . . . . . . . . . . . . . . . . . . 9. ..... . . .. . . . ... .. . . .... . .. . . . . , MATE RIAL SA433 0R. 51 ~
20 -
HEAT NO. A17881 (LD 0 ' ' ' ' '
100 0 100 200 300 400 500 Temperature, F 5-12
i -
l r
F Figure 5-6.- Charpy Impact Data for Irradiated Base Metal Plate D4802-2, Heat No. A1768-1. Transverse Orientation 100 Y
Ts - -
50 -
w e 25 - * .-
O.100 0 100 200 300 400 500 0.1 0.00 - -
g 0.0. - -
F 0.04 LL e
1 0.02 0.100 0 100 200 300 400 800 220 T NOT """
- ~ T,(35 MLE) +185F
~
100 7,(SO FT.L8) + 170F -
T,(30 FT43) +10SF
~
~ CvuSE(AYtL) es FT.LS$
$ 140 Ri ggy - -
120 - -
LLI 100 - -
CL n - -
.E s0 - -
40 -
........................... ..... ..............................4 MATERIAL SA.533 OR. 21 "
20 -
, HEAT NO. A17881 frL) -
0 ' ' ' ' '
100 0 100 200 300 400 500 ,
Temperature, F 5-13 9
, ~- -% ,--3--,m-9.v+, y-- -,e ,r+r----. - - - - -
1 Figure 5-7. Charpy impact Data for Irradiated Base Metal Plate D4802-2, Heat-Affected-Zone. Heat No. A1768-1 100 rs -
50 -
k.
u i
e *
&g 0.200 100 0 100 200 300 400 800 g 0.1 f 6H -
a -
\
-
- 8
$ 0.0s -
g u ' O.04 .......................................
- 10. -
n 0.200 100 0 100 200 300 400 800 220 '
T """
T"(30 PT.LS) +40F E 100 CvUSE (AVO.) 74 PT.LBS 4 140 nt noT - -
120 -
m 100 -
U.-.0 CL Sw -
40 -
.............................................................. MATERIAL SA.533 GR. 51 ~
n -
HEAT NO. A1788-1 (HAZ) .
0 ' ' ' ' ' '
200 100 0 100 200 300 400 500 Temperature, F ,
S-14
Figure 5-8. Charpy Impact Data for Irradiated Weld Metal, 305414/3951 100 .
g,.
eu u.
- a. *
- a. u - -
a m .
!200 100 0 100 200 300 400 600 0.1
, e gm e o.a.
ai 1
u 0.04 ............................................................ ...- .- _ ................-
0.02 -
3 9200 100 0 100 200 300 400 000 220 T NDT """
- + 234F
~
~ T,(35 MLE) 100 T,,(50 FT.LB) + 306F___. .
T,(30 FT.LB) +191F A 160 -
=r- CvUSE (AVG.) 60 FT.LBS 140 RT - -
NDT 120 -
8 MJ 100 - -
De - .
, c. e0 t- E .
- 60 - -
I 40 -
MATERIAL ASA/ UNDE 1092 ~
~
- HEAT NO. 305414/3951 '~
l 0
200 100 0 100 200 300 400 $00 Temperature, F 5-15
Figure 5-9. Charpy Impact Data for Irradiated Standard Reference 1 Material. Heat No. A1008-1. Loncitudinal Orientation l l
l l
i 100 Y
[ 75 - -
g2 - -
6
$U
.c M
l O.100 0 100 200 300 400 500 30.1
@ 0.00
'Ei
@ 0.06 .
W 0.04 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...........................-
E - -
.o 0.02 i
b 0 l 100 0 100 200 300 400 tb 220 T NDT ""
l * +183F
~
~Tu (35 MLE) 180 7,(50 FT.LB) + 22sF -
T"(30 FT.LB) + 184F
.O 160 CvtJSE (AVG.) 90 FT.LSS 140 RT -
NDT -
,E120 -
l c
W 1M - -
1 U ca 30 -
1 O.
.E. so -
l ........................ _
...........?
" ~
~ ~
HEAT NO. A10061 (LT) 0 ' ' ' ' '
100 0 100 200 300 400 500 Temperature, F l
l 5-16
Figure 5-10. Photographs of Charpy Impact Specimens Fracture Surfaces -
Base Metal Plate D4802-2. Heat ,,,,mr.
No. A1768-1. (LT) '
?hS?*$$ , N ..l? Wh { .
I e. f .
J .;
- i p$
g :;
} ' ,{ p... ,> [.[' .
.p+- - ~
4
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- N Specimen No.148. Test Temperature +40'F Specimen No.15B, Test Temperature +160*F Specimen No.146. Test Temperature +70'F Specimen No. ISA, Test Temperature +180*F I
f
. ,. ; > .w g . g. ya :..- . t AR . p , l . ,, s . .
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"Q .19 ; .g; n.
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' % % *3 ,
g., y' c,.o c . s ,* ny .
Specimen No.168, Test Temperature +70*F Specimen No.16C, Test Temperature +220*F
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I. q-
- 1,, / ,.
.. . ., . v ..
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2 .Y :(S #' Uk .
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Specimen No.16D, Test Temperature +100*F Specimen No.14E, Test Temperature +260*F
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- s. .
g i Speermen No.14P, Test Temperature +120*F Specirnen No.137, Test Tamperature +300'F
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3 .. #. w.-
Speermen No.15), Test Temperature +140'F Specimen No.165. Test Temperature +350*F 5-17
Figure 5-11-. Photographs of Charpy Impact Specimens fracture Surfaces -
Base Metal Plate D4802-2, Heat No, A1768-1, (TL) 4 hMfi.JMT
,M$QS 'r,Mp }W l%* I,
? .';: w u . 3.:tJi%.QV Ir{g@.$.
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- 5i*6 .
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. y.4
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-t .}/5-s:. 3f ynsb >y.e.<*g.?)4 Spectmen No. 21K, Test Temperature +70*F p- - ,, ,,
b } p ';.r' _
~Q' .,,=. sA
.;: . .x x ?.
c g 54 w >, . , .3
, L ~ '. . y '
"[y K 7 d4 >*
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i,. q t' . *.
}
Specimen No. 21M, Test Temperature +100'F Specimen No. 22A, Test Temperature +220*F Elli Specimen No. 22), Test Temperature +300*F Specimen No. 235, Test Temperature +350*F kg. .
.[
} : 3 h
- Jqi . 'g s ** l [, g .',' '..t..
. ?
- 4. ,k .
Specimen No. 248. Test Temperature +400*F 5-18
Figure 5-12. Photographs of Charpy Impact Specimens Fracture Surfaces -
Base Metal Plate 04802-2. Heat No. A1768-1. (HAZ)
.m . . . -
Ms.e: .
Specimen No. 42E, Test Temperature -20*F l
Specimen No. 411. Test Temperature +150*F
~hl [
Specimen No 44C, Test Temperature 0*F Specimen No. 445, Test Temperature +180*F l
l(*
M lliEEE Specimen No. 43A, Test Temperature +40*F
)
Specimen No. 412, Test Temperature +220*F l ' - :
4 ;
V;[5j p
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i.
.$ ~gpc t n
. [,TAE i x1 :-
Specimen No 43T, Test Temperature +70*F Specimen No. 45T, Test Temperature +280*F g
W '
l
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f t . .
,7
, ,f.- 1.
- )
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Specimen No. 42A, Test Temperature +100*F Specimen No. 46D, Test Temperature +280*F l
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- =' " ~ . , ! k' h l* ']
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- Spec [ men No. 42T, Test Temperature +120*F Specimen No. 432. Test Temperature +350*F 5-19 s
l l
- ~ - - - - - - - - - - - - - -l
Figure 5-13. Photcgrapns of Charpy Impact Specimens Fracture Surfaces -
Wel d __Me t al 305414/3351 -- -
l e.w.. =; f, A .- ,. y w. .
. 1
. .. . .a ry L, ; -- ....3,[ ;
_T?'h_ (. f
, W.*f ; '
j .
s 1
Specimen No. 33K, Test Temperature 70*F Specimen No. 33), Test Temperature +260*F
.] w ll ., ..
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f ,
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- ,W .,y b'>._ , a. ,;,_~ ,,J. ' [5 , -
l Specimen No. 32M, Test Temperature +100*F Specimen No. 32K, Test Temperature +300*F I.- .
e 4 T' %.;
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8 h'[d,.' Q . tg [ . '.)hk<0 >%Q
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Specimen No. 31 A, Test Temperature +120*F Specimen No. 33M, Test Temperature +350*F i - -
.T .
l_ e.,
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I' * * : . -
., 'l , . . _s
- 4. .
. l& .__
Specimen No. 316, Test Temperature +150*F Specimen No. 313, Test Temperature +400*F
. ** .' . . ,4 ., "h.l l . . ' ~ 3 7., 7 - . . ..
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Specimen No. 31J, Test Temperature +180*F Specimen No. 311, Test Temperature +420*F
'L y .
<- : .- a, i' b _ ..
Ag . .g m,op.&y. "g $;[.
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- '9 ,'.
- 9 _,
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'[. . . . ' . . . ,
Specimen No. 314. Test Temperature +220*F Specimen No 32L. Test Temperature +450*F 5-20
Figure 5-14. Photographs of Charpy Impact Specimens Fracture Surfaces -
Standard Refer _ gage Material. Heat No. A1008-1. (LT)
Specimen No. 56T, Test Temperature +70*F h
Specimen No. 56E. Test Temperature +100*F Specimen No. 564, Test Temperature +22t :'F
- R.T
- . .
- l.i ',
% pig ( A3 ' ' . : c. ,t 1 T . :? idk' Spectmen No. 572, Test Temperature +300*F Spec: men No. 56K, Test Temperature +350*F 7 .
3\fw ,
e .z , ' :c :;,y.*n'
- 'h yf .C.l$..-
, .g l,, f
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. ,.. y' : ' . , . ! ,y p a?.. ya , p k . .;,,..J -
Spee! men No. 569, Test TemperatJre +400*F
~5-21
- 6. 00SIMETER MEASUREMENTS 6.1. Introduction l The dosimeter set consists of twenty-seven dosimeters made up of shielded Cu, l
8 shielded Ni, unshielded Ti, unshielded Fe, shielded and unshielded 0 attenuation monitors as described in Table 3-3. The three sets of five iron {
wires used as flux attenuation monitors were not analyzed at this time and were placed in storage. Sulfur dosimeters encapsulated in quartz tubing were also included in the set, but were not analyzed due to the short half-life of the analyte P. Only two sulfur monitors, top and bottom, remained in the quartz tubes compartments. Four sulfur monitors were lost when the quartz tubes lost their intensity. Each dosimeter was contained in one of three stainless steel holder blocks that were installed in various positions in the capsule assembly. ,
The dosimeters were delivered in vials identified by labels consisting of the position of the holder block in the capsule assembly and the number of i
identification grooves on the stainless steel encapsulation.
6.2. Dosimeter Preparation Vials were prepared for the dosimeters by labeling them with identifications that indicated their positions in the holder blocks and the number of identification grooves on the stainless steel encapsulation. For example, the one capsule in
~
the top block was labeled FC-275 ,1 7G Sh. When the analyte nuclides were determined by gamma scanning, the identifications were appended accordingly. For example, FC-275 , 1 7G Sh Cu. This identification code stands for FC1, Capsule W-275, first (or top) holder block, seven identification grooves, cadmium shielded, copper dosimeter.
The stainless steel fission powder capsules were clamped in a metalworking vise which was mounted on two lead bricks in a hood. A flat mill bastard file was used to file the capsules open. The cadmium-covered wires had been crimped at 6-1
the ends so that the wires had to be removed by cutting through the shield with diagonal cutters and removing the wires.
The dosimeters were cleaned by washing in reagent acetone, aad blotting dry with a laboratory towel. Each dosimeter was measured with a certified micrometer caliper and weighed on a certified analytical balance. Each was then mounted in the center of a PetriSlide* with double-sided tape.
The exact oxide composition of the uranium dosimeters was uncertain. It was not possible to correct for self-absorption of the powder, therefore it was necessary to dissolve them and put them into geometries for which the gamma spectrometer was calibrated. TLis was the 20cc liquid scintillation vial geometry. The uranium dosimeters were dissolved in 8R HNO 3 acid and diluted to 20 ml in the same acid in a pre-weighed 20cc scintillation vial. The total uranium content was measured by inductively coupled plasma atomic emission spectroscopy (ICP).
6.3. Quantitative Gamma Spectrometry Each of the dosimeters, in the PetriSlide" (point source), or 20cc vial geometry, was given a 300 second preliminary count on the 31% PGT gamma spectrometer. This provided information to best judge the distance at which to count the dosimeter to obtain a minimum of 10,000 counts in the photopeak of interest while keeping the counter dead time below 15%. It also provided qualitative identification of the dosimeters. This identification was made from the presence of the gamma rays in Table 6-1. The spectra confirmed the identities of the dosimeters.
The spectra was then measured quantitatively at the appropriate counting positions and for the appropriate count times determined from the preliminary counts.
6.4. Dosimeter Specific Activities The dosimeter specific activities are shown in Table 6-2, and the associated elemental weight fractions of the dosimeters and the isotopic fractions of the target nuclides are listed in Table 6-3. The detailed calculations are shown in Tables 6-4 through 6-8.
6-2
1 1
Ordinarily, the analyte nuclide associated with the ""U dosimeters is CS which has a 30 year half life. The specific activity is determined by counting the 661.59 kev primary peak. Incons#-tencies found during intercomparison of dosimeter specific activities, however, led to the discovery that there was significant silver contamination in the uranium dosimeters. Activation of the naturally occurring silver isotope, ' 8AG, results in the production of the radionulide " AG, which decays with a 657.69 kev gamma ray, and competes with the 661.59 kev cesium gamma ray. Silver count rate is higher than cesium count rate for the upper and middle dosimeter, and about 30% in the bottom dosimeter.
The total count rate (cesium + silver) is of no practical value since the fraction of the total count rate that is attributable to cesium alone is unknown.
To resolve the contamination problem, another analyte nuclide was selected for reanalysis. The criteria imposed include that the nuclide should be a fission product, no interference from other nuclides, enough counts (from the specimen),
and appropriate abundance and half-life. After careful comparison, '"Ce was chosen. The specific activity for *80 using Ce as the analyte is presented in Table 6-3. The results are consistent with the observat' ions from the other capsule dosimeters.
l l
I l
l 6-3
i l
l l
1 Table 6-1. Quantifyina Gamma Rays
]
l l
Dosimeter Analyte Iron Mn 0 834 kev from Fe Titanium d'SC 01121 kev from Ti Nickel Co 0 811 kev from 58 Ni Copper e Co 01332 kev from Cu, very low activity compared to Co wires, wire has coppery color Uranium-238 1"Ce 0 657 kev from fission 1
6-4
Table' 6-2. Specific Activities of Capsule W-275 Dosimetry, Fort Calhoun Station Unit No. 1 Specific Dosimeter' Shielded Target Analyte Activity Identification (Yes/No) Nuclide Nuclide (yci/gm Target)
FC-275,1 2G Ti No Ti-46 Sc-46 116.5 FC-275,4 2G Ti No Ti-46 Sc-46 147.9 FC-275,7 2G Ti No Ti-46 Sc-46 135.6 FC-275,1 7G Sh Cu Yes Cu-63 Co-60 10.18
- FC-275,4 7G Sh Cu Yes Cu-63 Co-60 9.058 FC-275,7 7G Sh Cu Yes Cu-63 .Co-60 7.410-FC-275,1 6G Sh Ni Yes Ni-58 Co-58 770.9 FC-275,4 6G Sh Ni Yes Ni-58 Co-58 723.8 FC-275,7 6G Sh Ni Yes Ni-58 Co-58 643.2 FC-275,1 3G Fe No Fe-54 Mn-54 618.2 "
FC-275,4 3G Fe No Fe-54 Mn-54 570.9 FC-275,7 3G Fe No Fe-54 Mn-54 509.6 FC-275,1 5G Sh U-238 Yes U-238 Ce-144 31.87 FC-275,4 SG Sh U-238 Yes U-238 Ce-144 29.50 FC-275,7 SG Sh U-238 Yes U-238 Ce-144 26.19 6-5
Table 6-3. Isotonic Fractions and Weiaht Fractions of Taraet Nuclides Isotopic Weight Target Fraction of Fraction of Dosimeter Nuclide Target Target Element Iron **Fe 0.0570 0.99975 Titanium Ti 0.0768 0.99793
' Nickel "Ni 0.6739 0.99951 Copper es cu 0.6850 0.99999 Uranium-238 238g 1.0000 I C P'*)
Inductively Coupled Plasma Atomic Emission Spectroscopy.
l 6-6 y r --m--,,, -- - y .-- . , - - , , _ _--.-.----.- ,.---.c r- - - r-- -
-m -- - - -- , , ,
Table 6-4. Titanium Dosimetry Measurements from Capsule W-275, Fort Calhoun Station Unit No.1 Post-Irred. Average Wire Attenuation Target Analyte Weight Shielded Weight Wire Diameter Length Coefficient Dosimeter Type Nuclide Nuclide Fraction (fes/No) (gm) (cm) (in) Element 5 FC 273,1 2G Wire TI-46 Sc-46 ').0766 No 0.0191 0.0546 -5/8 Ti 2.520E-01 FC 275,4 2G Wire TI-46 Sc-46 0.0766 No 0.0144 0.0533 -5/8 Ti 2.520E-01 FC 275,7 2G Wire TI-46 Sc-46 0.0766 No 0.0143 0.0525 -5/8 Ti 2.520E-01 Detector Sc-46 Activity Geometry Self Abs. Corrected Systematic Distance Activity Error Offset Wires Abs. Activity Error Dosimeter (cm) (Act) (%) Factor V2 Factor (sci /gm) (1)
FC 275,1 2G 2.286 1.655E-01 0.73 0.9766 0.98942908 1.0057 8.923E+00 4.50 m FC 275,4 2G 2.286 1.585E-01 0.75 0.9772 0.98966947 1.0056 1.133E+01 4.50 b FC 275,7 2G 2.286 1.444E-01 0.77 0.9775 0.98982937 1.0055 1.039E+01 4.50 SCl/gm Target Error Error Desimeter Ti-46 (scl/sm) (%)
FC 275,1 2G 1.165E+02 4.135E+00 5.73 FC 275,4 2G 1.479E+02 5.679E+00 5.92 FC 275,7 2G 1.356E+02 5.255E+00 5.94
Table 6-5. Copper Dosimetry Measurements from Capsule W-275, Fort Calhoun Station Unit No. 1 Post-Irrad. Average Wire Attenuation Target Analyte Weight Shielded Weight Wire Diameter Length Dosimeter Coefficient Type Nuclide Nuclide Fraction (Yes/No) (gs) (ca) (in) Element a FC 275,1 7G Wire Cu-63 Co-60 0.6850 Yes 0.0032 0.0161 -5/8 Cu 4.544E-01 FC 275,4 7G Wire Cu-63 Co-60 0.6850 Yes 0.0054 0.0229 -5/8 Cu 4.544E-01 FC 275,7 7G Wire Cu-63 Co-60 0.6850 Yes 0.0058 0.0212 -5/8 Cu 4.544E-01 Detector Co-60 Activity Geometry Self Abs. Corrected Systematic Distance Activity Error offset Wires Abs. Activity Error Dosimeter (cm) (ACl) (%) Factor v2 Factor (Aci/gm) (%)
FC 275,1 7G 2.286 2.208E-02 0.70 0.9930 0.99683570 1.0031 6.970E+00 4.50 m FC 275,4 7G 2.286 3.303E-02 0.64 0.9901 0.99551608 1.0044 6.205E+00 4.50
[n FC 275,7 7G 2.286 2.905E-02 0.68 0.9908 0.99584528 1.0041 5.076E+00 4.50 ACl/g:n Target Error Error Dosimeter Cu-63 (Aci/gm) (%)
FC 275,17G 1.018E+01 1.256E+00 13.14 FC 275,4 7G 9.058E+00 7.315E-01 9.25 FC 275,7 7G 7.410E+00 6.023E-01 9.29
l l Table 6-6. Nickel Dosimetry Measurements from Capsule W-275,.
l Fort Calhoun Station Unit No. I l
Post-Irred. Average Wire Attenuation Target Analyte Weight Shielded Weight Wire Diameter Length Coefficient Dosimeter Type Nuclide Nuclide Fraction (Yes/No) (ge). (ca) (in) Element A
! FC 275,1 6G Wire NI-58 Co-58 0.6736 Yes 0.0177 0.0449 ~5/8 Ni 6.092E-01 1 '
FC 275,4 6G Wire Ni-58 Co-58 0.6736 Yes 0.0238 0.0516 -5/8 Ni 6.092E-C' FC 275,7 6G Wire Ni-58 Co-58 0.6736 Yes 0.0270 0.0500 -5/8 Ni 6.092E-01 Detector Co-58 Activity Geometry Self Abs. Corrected Systematic Distance Activity Error offset Wires Abs. Activity Error >
- Dosimeter (cm) (gCI) (1) Factor V2 Factor (ACl/gn) (X)
FC 275,1 6G 7.353 9.030E+00 G.25 0.9939 0.99725337 1.0116 5.192E+02 4.50 cn FC 275,4 6G 7.353 1.137E+01 0.23 0.9930 0.99684157 1.0134 4.875E+02 4.50 to FC 275,7 6G 7.353 1.147E+01 0.20 0.9932 0.99694445 1.0129 4.332E+02 4.50 ACf/gm Target Error Error Dosimeter Ni-58 (ACf/ge) (%)
FC 275,1 6G 7.709E+02 2.956E+01 5.91 FC 275,4 6G 7.238E+02 2.458E+01 5.64 FC 275,7 6G 6.432E+02 2.187E+01 5.64 L
a P
Table 6-7. Iron Dosimetry Measurements from Capsule W-275 2 Fort Calhoun Station Unit No. 1 Post-Irred. Average Wire Attenuation Target Analyte Weight shletded Weight Wire Diameter Length Dosimeter Type Nuclide Nuclide Fraction (Yes/No)
Coefficient (gm) (cm) (f r.) Element A FC 275,1 3G Wire Fe-54 Mn-54 0.0570 0.0263 No 0.0495 -5/8 Fe 5.145E-01 FC 275,4 3G Wire Fe-54 Mn-54 0.0570 No 0.0249 0.0516 -5/8 Fe 5.145E-01 FC 275,7 3G Wire Fe-54 Mn-54 0.0570 No 0.0263 0.0495 -5/8 Fe 5.145E-01 Detector Mn-54 Activity Geometry self Abs. Corrected systematic Distance Activity Error offset Wires Abs. Activity Error Dosimeter (cm) (gCl) (%) Factor v2 Factor (ACl/gm) (%)
FC 275,13G 7.353 9.107E-01 0.49 0.9933 0.99697018 1.0108 3.524E+01 4.50 g FC 275,4 3G 7.353 7.957E-01 0.53 0.9930 0.99684157 1.0113 3.254E+01 4.50 FC 275,7 3G 7.353 7.507E-01 0.49 0.9933 0.99697018 1.0108 2.905E+01 4.50 o e ACl/gm Target Error Error Dosimeter Fe-54 (Aci/gm) (%)
FC 275,1 3G 6.182E+02 2.136E+01 5.67 FC 275,4 3G 5.709E+02 1.946E+01 5.65 FC 275,7 3G 5.096E+02 1.761E+01 5.67
, -,-,- --- - - - - - - - - , - - - - - - - - , - - . , - - - - , - - - - - - - - - - - - - , . - - - - - - - - - - - . ---,------.7-------- - - - - - - - - - - , , - - , , , ., - - - - - - - - - - . - - - - - - - - - - - , - , - - . . , .- , - - . - , . - - -- - - - - - - . - - - - - - - - -----
h
-I l
I 1
1
-1 1
i n
I i
r t
l' -, - . - . _ - - . - - . - - - . - . . _ _ . _ . . , . . _ , . . - . . _ . . . m.._--..-. - -- - - .- . . _ --_. _. ., ---- - . . - - , - - - ,
4 Table 6-8. Uranium-238 Dosimetry Measurements from Capsule W-275, Fort Calhoun Station Unit No.1 +
Weight of Aliquot Diluted Conc. of Target Analyte Weight shielded Sample Drawn to Target in Dil Dosimeter Type Nuclide Nuclide Fraction (Yes/No) (gn) (ge) (gn) Sol. (ppm) Element FC 275,1 SG Powder U-238 Ce-144 N/A Yes 29.6292 1.2869 20.0178 28.4390 U-238 j FC 275,4 SC Powder U-238 Ce-144 N/A Yes 30.2205 1.2932 20.0087 39.7740 U-238 FC 275,7 SG Powder U-233 Ce-144 N/A Yes 30.2455 1.2921 20.0041 40.139C U-238 a Attenuation Detector Cs-144 Activity Grams Self Abs. Corrected Systematic Coefficient Distance Activity Error of Target Wires Abs. Activity Error Dosimeter n (cm) (gCI) (%) U-238 V2 Factor (sCf/ge) (%)
FC 275,1 SG N/A N/A 4.177E-01 1.04 1.311E-02 N/A N/A N/A 6.73 i FC 275,4 SC N/A N/A 5.487E-01 0.87 1.860E-02 N/A N/A N/A 6.73 U FC 275,7 SG N/A N/A 4.923E-01 0.80 1.880E-02 N/A N/A- N/A 6.73 ACI/gm 1 Target Error Error Dosimeter U-238 (ACl/ge) (%)
a FC 275,1 SG 3.187E+01 4.050E-01 6.85 FC 275,4 5G 2.950E+01 3.351E-01 6.83 FC 275,7 SG 2.619E+01 2.837E-01 6.82 I
- F
l
- 7. RADIATION ANALYSIS AND NEUTRON 00SIMETRY l 7.1. Introduction j Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the pressur. vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is generally derived solely from analysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy depenuut damage function for data correlation, ASTM Standard Practice E853,
" Analysis and Interpretation of Light Water Reactor Surveillance Results,"
recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom."'" The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the 7-1
i l
l pressure vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Damage to Reacor Vessel Materials.""
l l
This section provides the results of the neutron dosimetry evaluations l performed in conjunction with the analysis of test specimens contained in I surveillance Capsule W-275, withdrawn at the end of the fourteenth operating cycle. Also included are updated evaluations of the dosimetry contained in Capsules W-265 and W-225, withdrawn at the conclusion of operating Cycles 7 and 3, respectively. These updates are based on current state-of-the-art j methodology and nuclear data; and, together with the Capsule W-275 results, )
provide a consistent up to date data base for use in evaluating material ,
properties of the FC1 reactor vessel.
l In each of the dosimetry evaluations, fast neutron exposure parameters in l terms of neutron fluence (E > 1.0 MeV), neutron fluence (E > 0.1 MeV), and l iron atom displacements (dpa) are established for the capsule irradiation l history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel wall. Also, uncertainties associated with the derived exposure parameters at the surveillance capsules and with the projected exposure of the pressure vessel are provided.
7.2. Discrete Ordinates Analysis A plo,i view of the FC1 reactor geometry at the core midplane is shown in Figure 3-1. Six surveillance capsules attached to the pressure vessel wall which are removed on an individual basis at frequencies defined in the Fort Calhoun Station Updated Safety Analysis Section 4.5 are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 45*, 85*, 105", 225*, 265*, and 275" relative to the core cardinal axis as shown in Figure 3-1. A plan view of a surveillance capsule holder attached to the pressure vessel wall is shown in Figere 7-1.
Fram a neutronic standpoint, the surveillance capsule structures are significant. The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the l water annulus between the thermal shield and the reactor vessel. In order to i determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.
l 7-2
In performing the fast neutron exposure evaluations for the FC1 surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distr %utions throughout the reactor geometry as well as to establish relative radial distributionsofexposureparameters{&(E>1.0MeV),
4(E>0.1MeV),anddpa/sec}throughthevesselwall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsules as well as for the determination of exposure parameter ratios; i.e., [dpa/sec]/[4(E > 1.0 MeV)], within the pressure vessel geometry. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux, 4(E > 1.0 MeV), at surveillance capsule positions and at several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The source importance functions generated from these adjoint analyses provided the basis for all absolute exposure calculations and comparison with measurement. These importance functions, when combined with operating cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of irradiation; and, established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the ycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also, accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased.
The absolute cycle specific data from the adjoint evaluations together with the relative neutron energy spectra and radial distribution information from the reference forward calculation provided the means to:
1- Evaluate neutron dosimetry obtained from surveillance capsules.
2- Relate dosimetry results to the neutron exposure at key locations at the inner radius and through the thickness of the pressure vessel wall.
3- Enable a direct comparison of analytical prediction with measurement.
7-3
i 4- Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
The forward transport calculation for the reactor model summarized in Figure 3-1 was carried out in R,0 geometry using the DORT two-dimensional discrete ordinates code ia and the SAILOR cross-section library'". The SAILOR library is a 47 energy' group ENDF/B-IV based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering was treated with a P, expansion of the scattering cross-sections and the angular discretization was modeled with an S, order of angular quadrature. The core power distribution utilized in the reference forward transport calculation was representative of the burnup weighted average over the first 14 cycles of operation.
All adjoint calculations were also carried out using an S order of angular quadrature and the P, cross-section approximation from the SAILOR library.
Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as at the geometric center of each surveillance capsule. Again, these calculations were run in R,0 geometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case $(E > 1.0 MeV).
Having the adjoint importance functions and appropriate core source distributions, the response of interest could be calculated as:
R(r,6) = ? ( ( I(r,0,E) S(r,0,E) r dr de dE
, J J J l r 0 E where: R(r,0) = $(E > 1.0 MeV) at radius r and azimuthal angle 6.
I(r,0,E)= Adjoint source importance function at radius r, azimuthal angle 0, and neutron source energy E.
S(r,0,E)- Neutron source strength at core location r,0 and energy E.
Although the adjoint importance functions used in this analysis were based on a response function defined by the threshold neutron flux $(E > 1.0 MeV),
prior calculations
- have shown that, while the implementation of low leakage 7-4
=
I loading patterns significantly impacts both the magnitude and spatial l distribution of the neutron field, changes in the relative neutron energy l spectrum are of second order. Thus, for a given location the ratio of J
[dpa/sec]/[$(E > 1.0 MeV)] is insensitive to changing core source distributions. In the application of these adjoint importance functions to the Fort Calhoun reactor, therefore, the iron atom displacement rates (dpa/sec) and the neutron flux $(E > 0.1 MeV) were computed on a cycle specific basis by using [dpa/sec]/[$(E > 1.0 MeV)] and
[$(E > 0.1 MeV)]/[$(E > 1.0 MeV)] ratios from the forward analysis in conjunction with the cycle specific $(E > 1.0 MeV) solutions from the individual adjoint evaluations.
Selected results from the neutron transport analyses are provided in Tables l 7-1 through 7-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the pressure vessel wall.
In Table 7-1, the calculated exposure parameters [$(E > 1.0 MeV),
'$(E > 0.1 MeV), and dpa/sec] are given at the geometric center of the two surveillance capsule positions for both the reference and the operating cycle specific core power distributions. The individual cycle data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. The reference data derived from the forward calculation are provided as a baseline exposure evaluation against which cycle specific fluence calculations can be compared. Similar data are given in Table 7-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for the reference and the Cycles 1 through 14 plant specific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum predicted exposure levels of the vessel wall itself.
l Radial gradient information applicable to $(E > 1.0 MeV), $(E > 0.1 Mev), and dpa/sec is given in Tables 7-3, 7-4, and 7-5, respectively. The data, obtained from the reference forward neutron transport calculation, are l presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure distributions through the vessel wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Tables 7-3 through 7-5.
7-5
For example, the neutron flux $(E > 1.0 MeV) at the 1/4T depth in the pressure vessel wall along the 45* azimuth is given by: -
$ 1f4 f4.5 ) = $(179.85, 45 ) F(184.37, 45 )
where: $3fc(45*) = Projected neutron flux at the 1/4T position on the 45* azimuth. ,
$(179.85,45*) = Projected or calculated neutron flux at the l vessel inner radius on the 45" azimuth. I F(184.37,45") = Ratio of the neutron flux at the 1/4T position to the flux at the vessel inner radius for the 45' azimuth. This data is obtained from Table 7-3.
Similar expressions apply for exposure parameters expressed in terms of
&(E > 0.1 MeV) and dpa/sec where the attenuation function F.is obtained from Tables 7-4 and 7-5, respectively.
7.3. Neutron Dosimetry The passive neutron sensors included in the Fort Calhoun surveillance program are listed in Table 7-6. Also given in Table 7-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of 4
the neutron energy spectrum within the surveillance capsules and in the subsequent determination of the various exposure parameters of interest
[$(E>1.0MeV),$(E>0.1MeV),dpa/sec]. The relative locations of the neutron sensors within the capsules are shown in Figures 3-2 and 3-3.
l The use of passive monitors such as those listed in Table 7-6 does not yield a l direct measure of the energy dependent neutron flux at the point of interest.
Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of 1 the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:
i -
The measured specific activity of each monitor.
7-6
The physical characteristics of each monitor.
The operating history of the reactor.
The energy response of each monitor.
The neutron energy spectrum at the monitor location.
The specific activity of each of the neutron sensors contained in Capsule W-275 is included in Section 6.0 of this report. Specific activities of sensors contained.in Capsules W-265 and W-225 were obtained from references 7 and 6, respectively. The irradiation history of.the Fort Calhoun reactor during Cycles 1 through 14 was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report"** for the applicable operating periods. The data given in NUREG-0020 is based on core follow information provided on a monthly basis by OPPD. The fine detail, i.e. monthly intervals, is necessary in performing radioactive decay corrections for each of the neutron sensors.
The irradiation history applicable to the exposure of Capsules W-275, W-265 and W-225 is given in Table 7-7.
Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full power operation were determined from the following equation:
R= ^
P No F Y{ Cj [1-e .x6] [e.x,3] ,
ref where, R =
Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P, (rps/ nucleus).
A =
Measured specific activity (dps/gm).
No =
Number of target element atoms per gram of sensor.
F =
Weight fraction of the target isotope in the sensor material. l Y =
Number of product atoms produced per reaction.
7-7 l
P,
= Average core power level during irradiation period j (MW).
Py = Maximum or reference power level of the reactor (MW).
C, = Calculated ratio of 4(E > 1.0 MeV) during irradiation period j to the time weighted average $(E > 1.0 MeV) over the entire irradiation period.
1 = Decay constant of the product isotope (1/sec).
t i
= Lengthofirradiationperiodj(sec).
t, = Decay time following irradiation period j (sec).
and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
In the equation describing the reaction rate calculation, the ratio [P,]/[P,]
accounts for month by month variation of the reactor core power level within any given operating cycle as well as over multiple operating cycles. The ratio C , which can be calculated for each fuel cycle using .the adjoint 3
transport technology discussed in Section 7.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from operating cycle to operating cycle. For a single cycle irradiation C, is normally taken to be 1.0. However, for multiple cycle irradiations, particularly those employing low leakage fuel management, the additional C, term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management.
For the irradiation history of Capsules W-275, W-265 and W-225, the flux level term in the reaction rate calculations was developed from the plant specific analysis provided in Section 7-2. Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Tables 7-8 through 7-10 for Capsules W-275, W-265 and W-225, respectively.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code **. The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as input and proceeded to adjust the group 7-8 I l
I fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured reaction rate data. The " measured" exposure parameters along with the associated uncertainties were then obtained from the adjusted !
spectrum.
In the FERRET evaluations, a log-normal least squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values f are linearly related to the flux $ by some response matrix A:
- '") #
= { A ) $ ")
8 where i indexes the measured values belonging to a single data set s, g designates the energy group, and a delineates spectra that may be simultaneously adjusted. For example, R, = [ a gg 4, 8
relates a set of measured reaction rates R, to a single spectrum 4, by the multigroup reaction cross-section o,. The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were aDproximated in a multi-group format consisting of 53 energy groups. The trial input spectrum was converted to the FERRET 53 group structure using the SAND-II code' . This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide. The 620 point spectrum was then re-collapsed into the group structure used in FERRET.
7-9
The sensor set' reaction cross-sections, obtained from the ENDF/8-V dosimetry
' file, were also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial spectrum, as expanded to 620 groups, was employed as a weighting function in the cross-section collapsing procedure.
Reaction cross-section uncertainties in the form of a'53 x 53 covariance matrix for each sensor reaction were also constructed from the information contained on the ENDF/8-V data files. These matrices included energy group to energy group uncertainty correlations for each of the individual reactions.
However, correlations between cross-sections for different sensor reactions were not included. The omission of this additional uncertainty information does not significantly impact the results of the adjustment.
Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation was taken from the center of the surveillance capsule modeled in the reference forward transport calculation. While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the ENDF/B-V data files, 1 the covariance matrix for the input trial spectrum was constructed from the following relation:
+
Mi=R2+RRiPi M n 8 8 M where R, specifies an overall fractional normalization uncertainty (i.e.,
complete correlation) for the set of values. The fractional uncertainties R, specify additional random uncertainties for group g that are correlated with a correlation matrix given by:
P88i = [1-0] 6 88 i + O e -"
where:
The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a 7-10
l i
1 H _ (g g')2 2y 2 1
group range y (6 specifies the strength of the latter term). The value of 6 is I when g = g' and 0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1.
Strong long range correlations (or anti-correlations) were justified based on information presented by R. E. Maerker". Maerker's results are closely duplicated when y = 6.
The uncertainties associated with the measured reaction rates included both statistical (counting) and systematic components. The systematic component of the overall uncertainty accounts for counter efficiency, counter calibrations, irradiation history corrections, and corrections for competing reactions in the individual sensors.
Results of the FERRET evaluations of the Capsules W-275, W-265 and W-225 dosimetry are given in Tables 7-11 through 7-13, respective-ly. The data summarized in these tables include fast neutron exposure evaluations in terms-I of 4(E > 1.0 MeV), 9(E > 0.1 MeV), and dpa. Summaries of the fit of the adjusted spectrum to the measurements for each of the capsules are provided in Tables 7-14 through 7-16. In general good results were achieved in the fits of the adjusted spectra to the individual measured reaction rates. The adjusted spectra from the least squares evaluations are given in Tables 7-17 -
through 7-19 in the FERRET 53 energy group structure.
A summary of the measured and calculated neutron exposure of Capsules W-275, W-265, and W-225 is presented in Table 7-20. Comparisons are provided for both the average capsule exposure and the maximum capsule exposure. The maximum exposure levels are noted to occur at the top dosimetry position. The data listed in Table 7-20 indicate that, for the fast neutron evaluations, the ratios of calculation to the maximum measured values were in the range of 0.92 to 1.22 for $(E > 1.0 MeV), 0.87 to 1.11 for $(E > 0.1 MeV), and 0.94 to 1.18 for dpa/sec. For the set of three surveillance capsules, the average C/M ratios were 1.08, 1.02, and 1.08 for 4(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec, respectively. The associated lo standard deviation for each of these C/M ratios is 14%, 13%, and 12%, respectively.
7-11
7.4. Pro.iections of Pressure Vessel Exposure Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 7-21. Alongwiththecurrent(13.6EFPY) exposure, projections are also provided for exposure periods of 20 EFPY and 32 EFPY. In computing these vessel exposures, the calculated values from Table 7-2 were scaled by the average measurement / calculation ratios (M/C) observed from the evaluations of dosimetry from Capsules W-275, W-265 and W-225 for each fast neutron exposure parameter. This procedure resulted in bias factors of 0.87, 0.93, and 0.87 being applied to the calculated values of 4(E > 1.0 MeV),
4(E >0.1 MeV), and dpa, respectively. Projections for future operation were based on the assumption that the average exposure rates characteristic of the Cycles 10-14 irradiation would continue to be applicable throughout plant life.
The overall uncertainty associated with the best estimate exposure projections at the pressure vessel wall depends on the individual uncertainties in the measurement process, the uncertainty in the dosimetry location, and on the uncertainty in the extrapolation of results from the measurement points to the point of interest in the vessel wall. For Fort Calhoun Station, an extrapolation uncertainty of 5% has been combined with a 14% uncertainty in the plant specific measurement / calculation bias factor derived from the three surveillance capsule sets to produce a net uncertainty of 15% in the projected exposure of the pressure vessel wall. This 15% uncertainty applies at the lo level for 4(E > 1.0 MeV).
In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the Fort Calhoun Station reactor coolant system, exposure projections to 20 EFPY and 32 EFPY were also employed. Data based on both a 4(E > 1.0 MeV) slope and a plant specific dpa slope through the vessel )
wall are provided in Table 7-22. In order to access RT, vs fluence curves, l dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions I were defined by the relations:
$(1/4T) = 4(OT) dpa(1/4T) dpa(OT) l 7-12
and
$(3/4T) = 4(o7) dra(3/4rj dpa(07)
Using this approach results in the dpa equivalent fluence values listed in Table 7-22.
l 1
l f
7-13
TABLE 7 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER 6(E >1.0 MeV) In/ce,"-secl _
j CAPSULE LOCATION 275*/265*_ 225*
Reference 3.73x10' -5.24x10' '
Cycle 1 4.84x10' 6.58x10' -
Cycle 2 4.66x10' 6.45x10' Cycle 3 4.94x10' 7.71x10' l Cycle 4 4.70x10' 6.62x10' l l
Cycle 5 4.47x10' 6.85x10' Cycle 6 5.12x10' 7.05x10' Cycle 7 5.33x10' 7.09x10' j Cycle 8 3.63x10' 4.93x10' Cycle 9 2.66x10' 4.69x10' Cycle 10 4.88x10' 3.22x10' i Cycle 11' 3.00x10' 5.55x10' Cycle 12 2.94x10' 4.67x10' Cycle 13 2.64x10' 4.88x10' Cycle 14 2.84x10' 3.39x10' i
1.
l l
L l
7-14
-___ __________ _ ________---_~ _ . _ _- ae--. . .- - -.-- .. -_-
TABLE 7-1(Continued)
CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER 6(E > 0.1 MeV) in/cm*-secl _
CAPSULE LOCATION 275?/265?_ 225!
Reference 8.74x10' 1.28x10" Cycle 1 1.13x10" 1.61x10" Cycle 2 1.09x10" 1.58x10" Cycle 3 1.16x10" 1.88x10" ,
Cycle 4 1.10x10" 1.62x10" I Cycle 5 1.05x10" 1.68x10" Cycle 6 1. 20x10" 1. 73x10" Cycle 7 1.25x10" 1. 74x10" Cycle 8 8.51x10' 1. 21x10" Cycle 9 6.23x10' 1.15x10" Cycle 10 1.14x10" 7.89x10' Cycle 11 7.03x10' 1. 36x10" Cycle 12 6.89x10' 1.14x10" Cycle 13 6.18x10' 1.19x10" l
Cycle 14 6.66x10' 8.28x10' 7-15
)
TABLE.7-1 (Continued)
CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER doa/sec CAPSULE LOCATION 275*l265?_ 225?
Reference 5.81x10'" 8.23x10 "
Cycle 1 7. 54x10 " 1.03x10"
4 Cycle 2 7.25x10'" 1.01x10 4
Cyc1e 3 7. 70x10'" 1.21x10 4
Cycle 4 7. 32x10~" 1.04x10 Cycle 5 6. 96x10'" 1.08x10
4 Cycle 6 7.98x10'" 1.11x10 4
Cycle 7 8. 31x10'" 1.11x10 Cycle 8 5. 66x10'" 7. 79x10'" -
Cycle 9 4.14x10'" 7.36x10" Cycle 10 7. 60x10'" 5.06x10 "
Cycle 11' 4. 67x10'" 8. 72x10'"
Cycle 12 4. 58x10'" 7.34x10 "
Cycle 13 4.11x10'" 7. 66x10'"
Cycle 14 4.43x10'" 5. 32x10'"
I 1
7-16 c
l l
l l
l TABLE 7-2 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 6(E > 1.0 MeV) In/cm*-secl AZIMUTHAL ANGLE 90.0* 75.0* 60.0* 45.0 0.0" Reference 2.49x10' 2.43x10' 2.33x10' 3.36x10' 2.21x10' Cycle 1 3.22x10 2.76x10' 2.76x10' 4.31x10' 3.22x10' Cycle 2 3.12x10' 2.62x10' 2.72x10' 4.23x10' 3.12x10' ,
Cycle 3 3.30x10' 2.80x10' 3.07x10' 5.04x10' 3.30x10' l Cycle 4 3.13x10' 2.68x10' 2.79x10 4.34x10' 3.13x10 Cycle 5 2.99x10' 2.54x10' 2.81x10' 4.49x10' 2.99x10' Cyc1e 6 3.42x10' 2.89x10' 2.96x10' 4.62x10' 3.42x10' l Cycle 7 3.56x10' 3.00x10' 3.05x10' 4.65x10'" 3.56x10' l Cycle 8 2.36x10' 2.36x10' 2.41x10' 3.25x10' 2.36x10' Cycle 9 1.76x10' 1.64x10' 2.07x10' 3.09x10' 1.76x10' ;
. Cycle 10 3.30x10' 2.53x10' 1.65x10' 2.13x10' 1.03x10' I Cycle 11 1.93x10' 2.11x10' 2.55x10 3.66x10' 1.77x10' Cycle 12 1.91x10' 2.01x10' 2.30x10' 3.08x10' 1.33x10' Cycle 13 1.76x10' 1.67x10' 2.24x10' 3.22x10' 1.63x10' Cycle 14 1.91x10' 1.60x10' 1.43x10' 2.23x10' 1.06x10' l
1 7-17 l
l 4
1 TABLE 7-2 (Continued).
CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE l
l 6(E > 0.1 MeV) In/cm'-secl 1
AZIMUTHAL ANGLE 90.0* 75.0? 60.0?' 45.0" 0.0?
Reference 6.20x10' 5.55x10' 5.86x10'" 9.10x10'" 5.49x10' Cycle'1 8.02x10' 6.30x10' 6.94x10' 1.17x10" 8.02x10 Cycle 2 7.76x10' 5.97x10' 6.83x10' 1.15x10" 7.76x10' Cycle 3 8.21x10' 6.39x10' 7.72x10'" 1. 37x10" 8.21x10'"
Cycle 4 7.80x10' -6.13x10' 7.01x10' 1.18x10" 7.80x10' Cycle 5 7.44x10' 5.80x10' 7.06x10' 1.22x10" 7.44x10' Cycle 6 8.51x10' 6.59x10' 7.46x10' 1.25x10" 8.51x10' Cycle 7 8.87x10' 6.84x10' 7.66x10' 1.26x10" 8.87x10' Cycle 8 5.89x10' 5.40x10' 6.07x10' 8.81x10' 5.89x10' Cycle 9 4.38x10' 3.75x10' 5.21x10' 8.37x10' 4.38x10' Cycle 10 8.22x10' 5.77x10' 4.14x10' 5.78x10' 2.56x10' Cycle 11 4.81x10' 4.81x10' 6.40x10' 9.91x10' 4.41x10' Cycle 12 4.75x10' 4.58x10' 5.78x10' 8.36x10' 3.31x10' Cycle 13 4.38x10' 3.81x10' 5.64x10' 8.72x10' 4.06x10' Cycle 14 4.76x10' 3.65x10' 3.59x10' 6.04x10' 2.65x10' 7-18
TABLE 7-2.(continued)
CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE Iron Displacement Rate idDa/sacl AZIMUTHAL ANGLE 90.0" 75.0! 60.0" 45.0? 0.0*
Reference 4. 01x10'" 3.57x104 ' 3.77x10 '4 5.48x10'" 3. 55x10'"
Cycle 1 5.18x10'" 4. 05x10'" 4.46x10d ' 7.04x10d ' 5.18x104 '
Cycle 2 5.01x104 ' 3.84x10d ' 4.39x104 ' 6.90x104 ' 5. 01x10'"
Cycle 3 5. 30x10'" 4.11x10d ' 4.96x10d ' 8. 24 x10'" 5.30x10'"
Cycle 4 5.04x10d ' 3.94x104 ' 4.51x10d ' 7. 09x10'" 5.04x10 "
Cycle 5 4.81x104 ' c 73e.10d ' 4.54x104 ' 7. 33x10'" 4.81x104 '
Cycle 6 5. 50x10'" 4. 24x10'" 4.79x104 ' 7.55x104 ' 5.50x104 '
Cycle 7 5.73x10d ' 4.40x10d ' 4.92x10d ' 7. 60x10'" 5.73x104 '
Cycle 8 3.80x104 ' 3.47x104 ' 3. 90x10'" 5. 31x10'" 3.80x10'"
Cycle 9 2.83x10d ' 2.41x104 ' 3. 34x10'" 5.04x10'" 2.83x10'"
Cycle 10 5.31x104 ' 3.71x104 ' ?.66x10 '4
- 3. 48x10'" 1. 65x10'"
l Cycle 11 3.11x10'" 3.09x104 ' 4.11x10d ' 5.97x10 ' 4 2.85x10 "
I Cycle 12 3. 07x10'" 2.95x104 ' 3.71x104 ' 5. 04x10'" 2.14x10'"
l Cycle 13 2. 83x10'" 2.45x10'" 3. 63x10'" 5. 25x10'" 2.62x104 '
4 Cycle 14 3.07x10 ' 2. 35x10'" 2.31x10d ' . 3. 64 x10'" 1. 71x10'"
l l
1 l
l i
i l
7-19
~, - _ - - _ _ _.
TABLE 7-3 RELATIVE RADIAL DISTRIBUTION OF 4(E > 1.0 MeV)
WITHIN THE PRESSURE VESSEL WALL RADIUS (cm) 90.0" Zjaf[ 60.0! 45.0 0.0?
179.85"' 1.000 1.000 1.000 1.000 1.000 180.36 0.963 0.961 0.963 0.965 0.963 182.37 0.774 0.773 0.775 0.777 0.774 184.09 0.624 0.623 0.625 0.626 0.624 186.39 0.455 0.454 0.455 0.458 0.455 188.40 0.342 0.342 0.345 0.343 0.342 190.41 0.255 0.256 0.259 0.256 0.255 l 192.42 0.189 0.190 0.193 0.190 0.189 I 193.14 0.171 0.172 0.174 0.171 0.171 194.43 0.139 0.140 0.142 0.138 0.139 196.44 0.099 0.101 0.102 0.098 0.099 197.94 O.081 0.083 0.084 0.085 0.081 NOTES: (1) Indicates Base Metal Inner Radius (2) Indicates Base Metal Outer Radius 7-20
l TABLE 7-4 )
RELATIVE RADIAL DISTRIBUTION OF $(E > 0.1 MeV)
WITHIN THE PRESSURE VESSEL WALL I
RADIUS (cm) 90.0'..
75.f 60.0! 45.0? ,qg 179.85"' 1.000 1.000 1.000 1.000 1.000 180'.36 1.000 1.000 1.000 1.000 1.000 182.37 0.956 0.955 0.961 0.940 0.956 184.09 0.880 0.880 0.887 0.858 0.880 186.39 0.771 0.772 0.781 0.745 0.771 188.40 0.675 0.675 0.686 0.646 0.675 190.41 0.580 0.583 0.593 0.552 0.580 192.42 0.490 0.494 0.503 0.460 0.490 193.14 0.458 0.463 0.472 0.428 0.458 194.43 0.402 0.409 0.416 0.372 0.402 196.44 0.315 0.325 0.329 0.285 0.315 197.94'"' O.266 0.278 0.282 0.235 0.266 i
NOTES: (1) Indicates Base Metal Inner Radius (2) Indicates Base Metal Outer Radius 7-21 i
l
f TABLE 7-5 RELATIVE RADIAL DISTRIBUTION OF dpa/sec WITHIN THE PRESSURE VESSEL WALL RADIUS (cm) 90.0*. 15&* 60 2 45_.f ._Q.f 179.85"' 1.000 1.000 1.000 1.000 1.000 180.36 0.969 0.971 0.972 0.971 0.969 182.37 0.822 0.820 0.824 0.825 0.822 184.09 0.700 0.700 0.706 0.706 0.700 186.39 0.558 0.561 0.567 0.563 0.558 188.40 0.459 0.461 0.467 0.462 0.459 190.41 0.373 0.376 0.383 0.374 0.373 192.42 0.302 0.305 0.311 0.300 0.302 193.14 0.280 0.283 0.288 0.278 0.280 194.43 0.240 0.245 0.249 0.237 0.240 196.44 0.185 0.190 0.193 0.178 0.185 l 197.94 O.157 0.163 0.166 0.147 0.157 NOTES: (1) Indicates Base Metal Inner Radius (2) Indicates Base Metal Outer Radius i
i l
l 7-22
l l
TABLE 7-6 NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORS REACTION TARGET FISSION OF WEIGHT PRODUCT YIELD I INTEREST FRACTION HALF-LIFE (%) )
Cu-63 (n,a) Co-60 0.6917 5.271 yrs l Ti-46 (n p) Sc-46 0.0810 83.83 dys !
Fe-54 (n,p) Mn-54 0.0580 312.5 dys Ni-58 (n,p) Co-58 0.6827 70.78 dys ,
U-238 (n,f) Ce-144 0.9996 285.00 dys 4.54 U-238 (n,f) Cs-137 0.9996 30.12 yrs 6.00 )
l l
1 j
l 1
1 l
7-23
TABLE 7-7 MONTHLY THERMAL GENERATION DURING THE FIRST 14 FUEL CYCLES OF THE FORT CALHOUN REAC10R THERMAL THERMAL THERMAL THERMAL l MONTH (MW-hr) MONTH (MW-hr) MONTH (MW-hr) MONTH (MW-br) 8/73 405832 3/76 108114 10/78 420284 5/81 536108 9/73- 405832 4/76 972047 11/78 0 6/81 744754 10/73 405832 5/76 873287 12/78 113663 7/81 1052832 11/73 405832 6/76 784994 1/79 1000857 8/81 1030313 12/73 405832 7/76 983327 2/79 856102 9/81 548012 1/74 503818 8/76 911195 3/79 1042401 10/81 0 2/74 503818 9/76 764950 4/79 1011318 11/81 0 3/74 503818 10/76 4379 5/79 1045546 12/81 255590 4/74 119993 11/76 0 6/79 973249 1/82 1093510 5/74 715109 12/76 316713 7/79 1045494 2/82 876325 6/74 771709 1/77 1004543 8/79 948080 3/82 1091851 7/74 830140 2/77 935165 9/79 1013878 4/82 1040418 8/74 764679 3/77 1048140 10/79 1038441 5/82 1108564 9/74 802286 4/77 822354 11/79 755560 6/82 1073253 10/74 749420 5/77 1017763 12/79 856079 7/82 1094322 11/74 534520 6/77 1001346 1/80 574731 8/82 1105355 12/74 752838 7/77 1036240 2/80 0 9/82 945045 1/75 691354 8/77 988495 3/80 0 10/82 616137 2/75 181155 9/77 857218 4/80 0 11/82 788153 3/75 0 10/77 0 5/80 0 12/82 81786 4/75 0 11/77 0 6/80 512426 1/83 0 5/75 355761 12/77 691329 7/80 1030436 2/83 0 6/75 729058 1/78 1039764 8/80 1049153 3/83 0 7/75 848203 2/78 936251 9/80 935548 4/83 567343 8/75 916591 3/78 992903 10/80 734262 5/83 975283 9/75 456705 4/78 1008893 11/80 809219 6/83 1074484 10/75 770475 5/78 736303 12/80 837437 7/83 1110049 11/75 910140 6/78 658400 1/81 781266 8/83 1096899 12/75 853315 7/78 1021022 2/81 835279 9/83 1073898 1/76 887368 8/78 1047090 3/81 727396 10/83 1106547 2/76 540440 9/78 1011328 4/81 472040 11/83 1029986 7-24
1 l
TABLE 7-7 (Continued)
MONTHLY THERMAL GENERATION DURING THE FIRST 14 FUEL CYCLES OF THE FORT CALHOUN REACTOR 1 THERMAL THERMAL THERMAL THERMAL MONTH (MW-hr) MONTH (MW-hr) MONTH (MW-hr) MONTH (MW-br) 12/83 1108677 7/86 1003268 2/89 749342 9/91 423141 1/84 1105888 8/86 757556 3/89 1111925 10/91 735387 2/84 997883 9/86 1070908 4/89 992388 11/91 1076325 3/84 49026 10/86 1110974 5/89 581337 12/91 1111552 4/84 0 11/86 1074493 6/89 1076891 1/92 958189 5/84 0 12/86 1078715 7/89 1112612 2/92 667 6/84 0 1/87 1091037 8/89 1111876 3/92 0 7/84 457374 2/87 1003667 9/89 748608 4/92 0
, 8/84 1103037 3/87 218683 10/89 1051493 5/92 791796 9/84 1029862 4/87 0 11/89 1075166 6/92 999520 10/84 1110458 5/87 0 12/89 1110918 7/92 334190 11/84 629245 6/87 570342 1/90 1107379 8/92 700842 12/84 944281 7/87 1109338 2/90 568522 9/92 855003 1/85 1108295 8/87 1110094 3/90 0 10/92 1113359 2/85 1001136 9/87 1073962 4/90 0 11/92 1076008 3/85 1006028 10/87 1113383 5/90 21625 12/92 1112266 4/85 1072284 11/87 1077041 6/90 793203 1/93 1112002 5/85 1111353 12/87 1111574 7/90 1112248 2/93 1002476 6/85 1069152 1/88 1109355 8/90 852679 3/93 1006266 7/85 1110687 2/88 827925 9/90 868892 4/93 633532 8/85 1110355 3/88 782732 10/80 875395 5/93 981842 9/85 974987 4/88 1030843 11/90 965605 6/93 883604 10/85 0 5/88 1107622 12/90 502832 7/93 1107027 11/85 0 6/88 1041667 1/91 788243 8/93 1113611 12/85 0 7/88 1009228 2/91 841518 9/93 785297 1/86 174509 8/88 951709 3/91 778551 2/86 838106 9/88 480276 4/91 716092 3/86 943421 10/88 0 5/91 838257 4/86 1062459 11/88 0 6/91 912097 5/86 1109682 12/88 0 7/91 1111317 6/86 1075360 1/89 3555 8/91 1007756 l
l l
l 7-25
l
'l TABLE 7-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE W-275 1 I
MEASURED SATURATED REACTION 1 MONITOR AND ACTIVITY ACTIVITY RATE AXIAL LOCATION (dis /sec-am) (dis /sec-om) (ros/ nucleus) ;
l Cu-63 (n,a) Co-60 TOP 2.58x10 5 4.54x10' 6.93x10
5 HIDDLE 2.30x10 4.05x10" 6.18x10
BOTTOM 1.88x10" 3.31x10" 5. 05x10
AVERAGE 3.97x10' 6.05x10
Ti-46 (n.0) Sc-46 TOP 3.30x10 5 5.06x10 8 4. 97x10.se MIDDLE 4.19x10 5 6.42x10" 6.31x10.is BOTTOM 3.84x10' 5.89x10" 5.78x10
AVERAGE 5.79x10" 5.69x10
Fe-54 (n,0) Mn-54 8 8 TOP 1.30x10 2.29x10 3.67x10
MIDDLE 1.20x10 8 2.12x30" 3.38x10
B0TTOM 1.07x10 8 1.89x10" 3.02x10'5 AVERAGE 2.10x10e 3.36x10'5 Ni-58 (n.0) Co-58 TOP 1.92x10 7 2.91x10 7 4.16x10
MIDDLE 1.80x10 7 2.73x10 7
3.90x105 7
BOTTOM 1.60x10 2.43x10 7 3.47x10
AVERAGE 2.69x10' 3.84x10
Cd Covered U-238 (n.f) Ce-144 l
TOP 1.18x10 8 2.06x10' 1.80x10
8 8 l MIDDLE 1.09x10 1.91x10 1. 66x10
i 5 8 BOTTOM 9.69x10 1.69x10 1.47x10
8 AVERAGE 1.89x10 1. 64 x10
7-26 l
1ABLE 7-9 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE W-265 MEASURED SATURATED REACTION MONITOR AND ACTIVITY ACTIVITY RATE bXIAL LOCATION (dis /sec-am) (dis /sec-om) (ros/ nucleus)
Cu-63 (n.a) Co-60 5
TOP 1.83x10 3.89x10" 5.94x1047 MIDDLE 1.97x10' 4.19x10' 6.40x1047 5
BOTTOM 1.63x10 3.47x10' 5.29x104 '
AVERAGE 3.85x10' 5.88x10d '
Ti-46 (n.o) Sc-46 TOP 6.83x10' 7.83x10' 7.69x10d
- MIDDLE 6.42x10' 7.36x10" 7.23x10 48 BOTTOM 6.27x10' 7.19x10' 7.06x10 48 AVERAGE 7.46x10' 7.33x10d
- Fe-54 (n.o) Mn-54 TOP 2.33x10" 3.00x10" 4.79x104 "
8 MIDDLE 2.05x10 2.64x10e 4.22x104 "
8 BOTTOM 2.10x10" 2.70x10 4.32x10 45 8
AVERAGE 2.78x10 4.44x10d "
Ni-58 (n.0) Co-58 TOP 3.68x10' 4.23x10' 6.05x10d
- MIDDLE 3.53x10' 4.06x10' 5.80x104 '
BOTTON 3.17x10' 3.65x10' 5.21x10 45 AVERAGE 3.98x10' 5.68x10d
TOP 1.27x10 1.03x10' 6.77x1044 MIDDLE 1.24x10" 1.00x10' 6.61x10d
- BOTTOM 8.25x10 5 6.67x10" 4.40x10d '
AVERAGE 8.99x10' 5.92x10d
3.05x10" 2.01x104 '
MIDDLE 3.53x10" 2.86x10" 1.88x104^
8 BOTTOM 3.35x10" 2.71x10 1.79x10d
- AVERAGE 2.87x10" 1.89x104 "
7-27
TABLE 7-10 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE W-225 MEASURED SATURATED REACTION MONITOR AND ACTIVITY ACTIVITY RATE AXIAL LOCATION (dis /sec-am) (dis /sec-am) fros/ nucleus)
Cu-63 (n.a) Co-60 TOP 1.09x10' 4.23x10' 6.45x10'"
MIDDLE 1.11x10' 4.30x10' 6. 57x10'"
BOTTOM 1.04x10' 4.03x10' 6.15x10'"
AVERAGE 4.19x10' 6. 39x10'"
Ti-46 (n 0) Sc-46 5
TOP 9.18x10' 9.95x10 9.77x10"
MIDDLE 8.97x10' 9.72x10" 9.55x10"
l BOTTOM 9.12x10" 9.88x10" 9.71x104 "
l AVERAGE 9.85x10' 9.67x104 "
l Fe-54 (n.0) Mn-54 TOP 2.66x10" 3.75x10" 6.00x10
MIDDLE 2.83x10" 3.99x10" 6.39x10 45 BOTTOM 2.57x10" 3.63x10" 5.80x10 45 l AVERAGE 3.79x10" 6.06x10 45 1
Ni-58 (n 0) Co-58 TOP 5.50x10' 5.86x10' 8.36x104 "
7 MIDDLE 5.61x10' 5.97x10 8.53x10
l BOTTOM 5.09x10' 5.42x10' 7.74x10d '
AVERAGE 5.75x10' 8.21x104 "
Bare '
U-238 (n.f) Cs-137 TOP 9.83x10 5 1.82x10 7
1.20x1048 l MIDDLE 8.23x10" 1.52x10' 1.00x10'"
BOTTOM 6.22x10' 1.15x10' 7.58x10
7 l AVERAGE 1.50x10 9.86x10
l l Cd Covered l U-238 (n.f) Cs-137 l TOP 2.93x10 5 5.42x10" 3.57x10 dd MIDDLE 2.87x10 5 5.31x10" 3.50x10 dd 80TTOM 2.42x10' 4.48x10" 2.95x10d
- AVERAGE 5.07x10" 3.34x10d
- 7-28 m.,
TABLE 7-11
SUMMARY
OF NEUTRON 00SIMETRY RESULTS SURVEILLANCE CAPSULE W-275 TIME AVERAGED EXPOSURE RATES 10 UNCERTAINTY 8
4(E > 1.0 MeV) [n/cm -sec] 3.00x10' i 10%
$(E > 0.1 MeV) [n/cm'-sec] 7.74x10' i 19%
dpa/sec 4.79x10'" i 12%
INTEGRATED CAPSULE EXPOSURE la UNCERTAINTY 4(E > 1.0 MeV) [n/cm'] 1.28x10'8 1 10%
4(E > 0.1 MeV) [n/cm'] 3.31x10'8 1 19%
dpa 2.05x10** 1 12%
Note: Total Irradiation Time = 13.6 EFPY.
I l
I I
l 7-29
TABLE 7-12 SUtHARY OF NEUTRON DOSIMETRY RESULTS SURVEILLANCE CAPSULE W-265 TIME AVERAGED EXPOSURE RATES lo UNCERTAINTY
$(E > 1.0 MeV) [n/cm*-sec] 4.13x10' i 9%
4(E > 0.1 MeV) [n/cm*-sec] 1.01x10" i 18%
dpa/sec 6. 36x10'" 1 11%
INTEGRATED CAPSULE EXPOSURE 10 UNCERTAINTY 58 gg 0(E > 1.0 MeV) [n/cm*] 7.71x10 0(E > 0.1 MeV) [n/cm*] 1.89x10 38 18%
dpa 1.19x10** 1 11%
Note: Total Irradiation Time = 5.9 EFPY.
7-30
TABLE 7-13
SUMMARY
OF NEUTRON 00SIMETRY RESULTS SURVEILLANCE CAPSULE W-225 TIME AVERAGED EXPOSURE RATES 10 UNCERTAINTY 4(E > 1.0 MeV) [n/cm'-sec] 7.16x10' i 9%
$(E > 0.1 MeV) [n/cm'-sec) 1.84x10" i 18%
dpa/sec 1.09x10' i 12%
INTEGRATED CAPSULE EXPOSURE lo UNCERTAINTY 6(E > 1.0 MeV) [n/cm'] 5.53x10'8 1 9%
4(E > 0.1 MeV) [n/cm'] 1.42x10'8 1 18%
dpa 8.41x10** 1 12%
Note: Total Irradiation Time = 2.4 EFPY.
7-31
TABLE 7-14 I
COMPARIS0N OF MEASURED AND FERRET CALCULATED i REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER SURVEILLANCE CAPSULE W-275 REACTION RATE (ros/ nucleus)
AVERAGE ADJUSTED REACTION MEASUREMENT CALCULATION C/M 47 47 Cu-63 (n,a) Co-60 6.05x10 5.74x10 0.95 Ti-46 (n.p Sc-46 5.69x10d " 5.97x10ae 1.05 d 45 Fe-54 (n,p) Mn-54 3.36x10 " 3.31x10 0.98 Ni-58 (n,p) Co-58 3.84x10 45 4.02x10d " 1.05 l d
U-238 (n,f) Cc-144 (Cd) 1.34x10d
- 1.23x10 ' O.92 TABLE 7-15 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER SURVEILLANCE CAPSULE W-265 REACTION RATE (ros/ nucleus)
AVERAGE ADJUSTED REACTION MEASUREMENT CALCULATION C/M d
Cu-63 (n,a) Co-60 5.88x10d ' 5.86x10 ' 1.00 d 48 Ti-46(n,p)Sc-46 7.33x10
- 7.54x10 1.03 d 45 Fe-54 (n,p) Mn-54 4.44x10
- 4.34x10 0.98 45 45 Ni-58(n,p)Co-58 5.68x10 5.65x10 0.99 U-238(n,f)Cs137(Cd) 1.61x104 ' 1.65x104 ' 1.03 i
7-32
TABLE 7-16 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER SURVEILLANCE CAPSULE W-225 REACTION RATE (ros/ nucleus)
AVERAGE REACTION MEASUREMENT CALCULATION C/M Cu-63 (n,a) Co-60 6.39x10' 6. 49x107 1.02 Ti-46(n,p)Sc-46 9.68x10'is 9.76x10" 1.01 Fe-54 (n,p) Mn-54 6.06x10 6.07x10 1.00 Ni-58 (n,p) Co-58 8.21x105 8.14x10'" 0.99 U-238(n,f)Cs137(Cd) 2. 89x 10 2.72x10 O.94 I
1 i
7-33
- 9
TABLE 7-17 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE W-275 ENERGY ADJUSTED FLUX ENERGY ADJUSTED FLUX 8 8 (MeV) (n/cm -sec) (MeV) (n/cm_-sec) 1.73x10 ' 9.41x10" 9.12x10*' 4. 50x10" 8
1.49x10 ' 2.08x10" 5.53x10*' 4.58x10 1.35x10 7.67x10" 3.36x10" 1.47x10" 1.16x10 ' 1.62x10" 2.84x10*' 1.45x10" 8
1.00x10 ' 3.33x10 2.40x10*8 1.45x10" 8 8 8.61x10" 5.10x10 2.04x10* 4.37x10 8
7.41x10" 1.08x10 1.23x10*' 4.40x10 8 8
6.07x10" 1.38x10 7.49x10" 4.37x10 8 8
4.97x10" 2.41x10 4.54x10* 4.15x10 8 8
3.68x10" 2.56x10 2.75x10" 4.40x10 8 8
2.87x10" 4.47x10 1.67x10" 4.59x10 8
8 8 2.23x10 4.72x10 1.01x10* 4.61x10 8
1.74x10" 5.35x10 6.14x10*8 4.62x10 8 8 8 1.35x10 4.52x10 3.73x10*8 4.65x10 8 8 1.11x10 7.14x10 2.26x10*8 4.64x10
' 8 8.21x10* 7.05x10 1.37x10*8 4.61x10
- 1 8
6.39x10* 6.95x10 8.32x10" 4.58x10 8 8
4.98x10* 4.81x10 5.04x10" 4.58x10 8
8 l 3.88x10* 5.83x10 3.06x10" 4.57x10 8 8 8 3.02x10* 9.02x10 1.86x10" 4.44x10 l 8 1.83x10* 7.72x10 1.13x10" 4.27x10 8 8
1.11x10* 6.30x10 6.83x10" 4.19x10 8 8
6.74x10** 5.51x10 4.14x10" 4.97x10 8 4.09x10** 3.85x10" 2.51x10" 1.59x10' 8
2.55x10** 3.11x10 1.52x10* 2.72x10' 8
1.99x10** 2.66x10 9.24x10" 5.94x10' 8
1.50x10** 4.43x10 Note: Energy levels represent the upper bound of each group.
7-34
TABLE 7-18 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE W-265 ENERGY ADJUSTED FLUX ENERGY ADJUSTED FLUX 8
(MeV) (n/cm -sec) (MeV) (n/cm8 -sec) 8 1.73x10' 8.92x10" 9.12x10*' 5.42x10 1.49x10 ' 1.99x10" 5.53x10** 5. 55x10" 1.35x10 ' 7.49x10" 3.36x10" 1.81x10" 8 8 1.16x10 ' 1.63x10 2.84x10*' 1.77x10 8 8 1.00x10 ' 3.50x10 2.40x10* 1.77x10 8
8.61x10" 5.77x10" 2.04x10*' 5.38x10 8
7.41x10" 1. 32x10" 1.23x10*' 5.47x10 8 8 6.07x10" 1.82x10 7.49x10" 5.37x10 8 8 4.97x10" 3.37x10 4.54x10* 5.19x10 8
3.68x10" 3.61x10" 2.75x10" 5.52x10 8 8 2.87x10" 6.35x10 1.67'.10" 5.73x10 8 8 2.23x10" 6.66x10 1.01x10* 5.76x10 8 8 1.74x10" 7.49x10 6.14x10*5 5.76x10 8 8 1.35x10" 6.23x10 3.73x10" 5.76x10 8 8 1.11x10" 9.52x10 2.26x10** 5.76x10 8 8 8.21x10* 9.27x10 1.37x10" 5.74x10 8
6.39x10*' 8.87x10" 8.32x10" 5.68x10 8
4.98x10* 6.12x10 5.04x10" 5.67x10" 8
3.88x10* 3.06x10" 8 "
7.17x10 5.66x10 8
3.02x10* 1.09x10' 1.86x10" 5.47x10 8 8 1.83x10* 9.23x10 1.13x10" 5.26x10 8 '
1.11x10* 7.53x10 6.83x10" 5.14x10
- 8 8 6.74x10" 6.61x10 4.14x10" 5.99x10 4.09x10" 4.66x10" 2.51x10* 1.89x10' 8
2.55x10" 3.66x10 1.52x10" 3.23x10' 1.99x10** 3.07x10" 9.24x10" 7.05x10' 8
1.50x10" 5.28x10 Note: Energy levels represent the upper bound of each group.
7-35
=
1 TABLE 7-19 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE W-225 i ENERGY ADJUSTED FLUX ENERGY ADJUSTED FLUX (MeV) 8 (n/cm -sec) (MeV) (n/cm8 -sec) 8 1.73x10 ' 8.94x10" 9.12x10*' 8.78x10 1.49x10 ' 2.00x10 7
5.53x10* 9.27x10" 7 8 1.35x10 ' 7.64x10 3.36x10*' 8.65x10 8 1.42x10' 1.16x10' 1.71x10 2.84x10*' _
8 1. 53x10' 1.00x10' 3.81x10 2.40x10*
8.61x10" 6.65x10 a 2.04x10* 1.92x10' 8 8 7.41x10" 1.60x10 1.23x10*' 8.84x10 6.07x10" 2. 39x10" 7.49x10" 8.71x10" 8
4.97x10" 4.88x10 4.54x10" 8. 55x10" 8 8 3.68x10" 5.'/5x10 2.75x10* 9.03x10 8
2.87x10" 1.09x10' 1.67x10" 9.35x10 2.23x10" 1.21x10' 1.01x10* 9. 38x10" 8
1.74x10" 1.42x10' 6.14x10" 9.42x10 1.35x10" 1.20x10' 3.73x10*8 9.48x10" 8
1.11x10" 1.85x10' 2.26x10" 9.50x10 i 8.21x10*' 1.83x10' 1.37x10*8 1.60x10' 6.39x10*' 1.74x10' 8.32x10*8 2.82x10' 1 8
4.98x10*' 1.18x10' 5.04x10*8 9.57x10 8
3.88x10*' 1.34x10' 3.06x10" 9.25x10 8
3.02x10*' 2.02x10' 1.86x10" 9.05x10 8
1.83x10*' 1.66x10' 1.13x10** 8.80x10 8
1.11x10*' 1.33x10' 6.83x10" 8.63x10 6.74x10*8 1.15x10' 4.14x10* 1.02x10' 8
4.09x10** 8.00x10 2.51x10" 3.25x10' 8
2.55x10*8 6.12x10 1.52x10" 5.54x10' 1.99x10*8 5.11x10" 9.24x10ea 1.21x10" 1.50x10*8 8.69x10" Note: Energy levels represent the upper bound of each group.
7-36
TABLE 7-20 ;
COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVELS FOR FORT CALHOUN SURVEILLANCE CAPSULES CAPSULE W-275 AVERAGE CALCULATED MEASURED C/M 4(E > 1.0 MeV) [n/cm"] 1.68x10'8 1.28x10'8 1.31 4(E > 0.1 MeV) [n/cm'] 3.93x10 3.31x10'8 1.19 l dpa 2.60x10* 2.05x10 4 1.27 MAXIMUM CALCULATED MEASURED C/M 4(E > 1.0 MeV) [n/cm'] 1.68x10'8 1.61x10'8 1.04 18.
4(E > 0.1 MeV) [n/cm'] 3.93x10'8 4.32x10 0.91 1
4 dpa 2.60x10 2.55x10* 1.02 I
i 1
l l
1 l
l 7-37 i
l
TABLE 7-20 (Continued)
COMPARIS0N OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVELS FOR FORT CALHOUN SURVEILLANCE CAPSULES CAPSULE W-265 AVERAGE CALCULATED MEASURED C/M 8 $
4(E > 1.0 MeV) [n/cm ] 9.08x10 a 7.71x10a 1.18 l 4(E > 0.1 MeV) [n/cm'] 2.13x10'8 1.89x10'8 1.13
! dpa 1.41x10* 1.19x10* 1.18 MAXIMUM CALCULATED MEASURED C/M 8
4(E > 1.0 MeV) [n/cm ] 9.08x10 $
8.31x10 a 1.09 i
4(E > 0.1 MeV) [n/cm'] 2.13x10'8 2.00x10 1.07 dpa 1.41x10* 1.27x10* 1.11 L
l l
I I
l 7-38
TABLE 7-20 (Continued)
COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVELS FOR FORT CALHOUN SURVEILLANCE CAPSULES CAPSULE W-225 AVERAGE CALCULATED MEASURED C/M 1 58 4(E > 1.0 MeV) [n/cm*] 5.28x10 a 5.53x10 0.95
@(E > 0.1 MeV) [n/cm*] 1.29x10'8 1.42x10'8 0.91 dpa 8.29x10* 8.41x10* 0.99 MAXIMUM CALCULATED MEASURED C/M 0(E > 1.0 MeV) [n/cm*] 5.28x10 a 5.73x10'8 0.92 4(E > 0.1 MeV) [n/cm'] 1.29x10'8 1.49x10'8' O.87 4 4 dpa 8.29x10 8.78x10 0.94
'/-39
. TABLE 7-21 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 13.6 EFPY 90" 75! 60! 45* 0*
58 5 4(E>1.0MeV) 9.69x10 8.58x10 s 9.04x10 1. 35x10 8.52x10
4(E>0.1 MeV) 2.56x10 2.07x10'8 2.41x10 3.88x10 2.25x10'8 4
dpa 1.56x10* 1.26x10* 1.46x10* 2.21x10 1.37x10*
l- 20.0 EFPY j
( 90 75* 60.0? 45! 0" l 4(E>1.0 MeV) 1.35x10'8 1.20x10'8 1.26x10'" 1.86x10 1.10x10
4(E>0.1MeV) 3.56x10 2.92x10'8 3.38x10 5.36x10 2.90x10
dpa 2.17x10* 1.77x10* 2.04x10* 3.04x10* 1.76x10*
i 32.0 EFPY 90* 75* 60.0? 45? O 4(E>1.0MeV) 2.21x10'" 2.00x10'" 2.09x10 3.03x10'8 . 65x10
, 6(E>0.1MeV) 5.43x10'" 4.50x10'" 5.19x10'8 8.11x10 4.11x10'8 4
dpa 3.30x10* 2.72x10* 3.13x10 4.59x10* 2.49x10*
l l
i NOTE: Fast neutron fluence values are in units of [n/cm'].
l l
l l
( 7-40 l
\
i TABLE 7-22 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/C00LD0WN CURVES FLUENCE BASED ON $(E > 1.0 MeV) SLOPE 90! 75" 60.02 45" Of 20.0 EFPY SURFACE 1.35x10'8 1.20x10'8 1.26x10'8 1.86x10'8 1.10x10'8
$ 18 1/4 T 8.40x10 a 7.50x10 7.90x10 1.17x10'" 6.84x10sa 3/4 T 2.30x10 2.07x10 1 2.20x10 a 3.18x10 1.87x10sa 32.0 EFPY l 18 SURFACE 2.21x10'8 2.00x10 2.09x10'8 3.03x10'8 1. 65x10'8 1/4 T 1.38x10'8 1.25x10'8 1.31x10'8 1. 90x10'8 1.03x10'8 3/4 T 3.78x10 3.44x10 3.64x10 5.18x10 2.83x10
EQUIVALENT FLUENCE BASED ON dpa SLOPE 90" 75" 60* 45" Of 20.0 EFPY 8
SURFACE 1.35x10'8 1.20x10'8 1.26x10'8 1.86x10'8 1.10x10 1/4 T 9.42x10 18 i 8.43x10 a 8.93x10 1.31x10'8 7.67x10
3/4 T 3.77x10'" 3.41x10 3.64x10 5.17x10 3.07x10 18 32.0 EFPY <
SURFACE 2.21x10'8 2.00x10'8 2.09x10'8 3.03x10'8 1.65x10'8 1/4 T 1.55x10'8 1.40x10'8 1.48x10'8 2.14x10'8 1.16x10'8 3/4 T 6.19x10'" 5.66x10'8 6.02x10 8.42x10 4.63x10 18 7-41
FIGURE 7-1 PLAN VIEW OF A REACTOR VESSEL SURVEILLANCE CAPSULE I
l Vessel Base Metal }
l
- i
)
il//// vess.i c1.eein. ////////A p i i i i i i i i , i i i i, 4
/ Dosimetry /
/ Block /
/ /
/ 1 i
,,,,,,,,,,,,)
O 7-42
- 8. DISCUSSION OF CAP!ULE RESULTS 8.1. Pre-Irradiation Property Data The base metal and weld metal were selected for inclusion in the FCI surveillance program in accordance with the criteria in effect at the time the program was designed. The applicable selection criterion was based on the unirradiated properties only. A review of the original unirradiated properties of the reactor vessel core beltline region materials indicated no significant deviation from expected properties. Based on the designed end-of-service peak neutron fluence value at the hT vessel wall location and the copper content of the base metals, it was predicted that the end-of-service Charpy upper-shelf energy (CvuSE) will not be below 50 ft-lbs.
8.2. Irradiated Property Data 8.2.1. Tensile Properties Table 8-1 compares the irradiated tensile properties from Capsule W-275 with the tensile properties from the unirradiated tensile specimens. At both room temperature and 550F, the ultimate and yleid strengths change in the surveillance base metal as a result of irradiation; the corresponding changes in ductility are within the limits observed for similar irradiated materials. There is some strengthening as indicated in the increases in ultimate and yield strengths and
]
decreases in the ductility properties. The changes in tensile properties for the l surveillance weld metal at both room temperature and 550F, as a result of irradiation, are also within the observed limits for similar irradiated materials. The strengthening in the surveillance weld metal, indicated by the increases in ultimate and yield strengths and decreases in ductility, is greater than that observed in the base metal, however, these changes are within l acceptable limits.
l The general behavior of the tensile properties as a function of neutron irradiation is an increase in both the ultimate and yield strengths and a decrease in ductility as measured by the total elongation and reduction of area. ,
8-1
1 8.2.2. Impact Properties The behavior of the CVN impact data is most significant for the calculation of the reactor system's operating limitations. Tables 8-2 and 8-3 compare the observed (measured) changes in irradiated CVN impact properties from Capsule W-275 with the predicted changes per Regulatory Guide 1.99, Revision 2."
Comparisons of the unirradiated and the irradiated CVN impact curves are presented in Figures 8-1 through 8-4. The radiation-induced changes in toughness of the FC1 surveillance materials are summarized in Table 8-4.
The observed 30 ft-lb transition temperature shifts for the surveillance base metal plate in both the longitudinal and transverse orientations are in good agreement with the value predicted using Regulatory Guide 1.99, Revision 2.
The shifts, however, are not conservative. If a margin value is added to the predicted values in accordance with Regulatory Guide 1.99, Revision 2, the values for both base metal orientations become conservative when compared to the observed data. The observed percent decrease in CvuSE due to irradiation also showed relatively good agreement with the predicted values for the surveillance base metal plate. The observed Cv0SE percent decrease, however, for both the longitudinal and transverse orientations are slightly greater than the predicted values (3% and 6%, respectively).
The predicted 30 ft-lb transition temperature shift for the surveillance weld metal is 12F greater than the observed value thus indicating good agreement with Regulatory Guide 1.99, Revision 2. If the margin value is included in the predicted shift, the predicted value becomes very conservative when compared to the observed value. The observed CvuSE percent decrease for the surveillance weld metal also shows good agreement with the Regulatory Guide 1.99, Revision 2, CvuSE percent decrease by comparison.
8-2
Table 8-1. Tensile Properties of the Fort Calhoun Station Unit No. 1 Reactor Vessel Surveillance Materials i
Cap. Fluence Test I.D. g n/cm" i Material Temo. F Ultimate g Yield M Elona. M of Area M j Base Metal, ----
0.00 RT 89. 2*' 66.8*' --- 28*'
A1768-1, 70.1*' ---
550 85. 6*' ---
55.1*' --- 24*' ---
- 68. 0*' ---
Longitudinal W-275 1.28 RT 101.7 +14.0 79.5 +19.0 27 - 4.6 65.0 - 7.3 550 91.7 + 7.1 67.0 +21.6 20 -16.7 59.6 -12.4 4
Base Metal, ----
0.00 RT 85.1*' ---
63.0*' --- 24 *' ---
6 5. 5*' ---
. HAZ, A1768-1 550 81.8*' ---
52.3*' --- 21 *' ---
62.4*' ---
2$ W-275 1.28 RT 95.6 -+12.3 73.6 +16.8 16 -33.3 58.0 -11.5 550 94.1 +15.0 69.7 +33.3 12 -42.9 45.8 -26.6 i
i Weld Metal, ----
0.00 RT 90. 2*' ---
74.1 *' --- 28*' ---
70.1*' ---
- 305414/3951 550 85.2*' ---
64.1*' ---
22*' ---
62.6*' ---
W-275 1.28 RT 113.1 +25.4 99.3 +34.0 23 -17.9 55.4 -21.0 550 103.9 +21. 9 88.5 +38.1 17 -22.7 47.6 -24.0 4
! "thange relative to unirradiated tensile data.
- ! Average of data presented in Appendix B.
h
Table 8-2. Observed Vs. Predicted Changes for Irradiated Surveillance Material 30 ft-lb Transition Temperature - 1.28 r 10'8 n/cm' Difference Predicted Per R. G.1.99/2*'_ -
Observed Chemistry Without With Material Unirrad. Irrad. Diff. Factor Marcin Marcin Marcin Base Metal, A1768-1, + 22 + 95 73 65 71 34 105 Longitudinal Base Metal, A1768-1, + 35 +108 72 65 71 34 105 Transverse HAZ Metal, A1768-1 - 76 + 40 116 65 71 34 105 en A
Weld Metal, 305414/3951 - 28 +191 219 212 231 56 '287 l
I Standard Reference + 27 +168 141 ---*' --- -- ---
itaterial, A1008-1, Longitudinal "talculated per Regulatory Guide 1.99, Revision 2.
Stopper content not available.
- - . - - _. _ __.__m_ . _._ .._ __ _ _ , _ _ _ .
Table P-3. Observed Vs. Predicted Changes for Irradiated Surveillance Material Upper-Shelf Enerav - 1.28 x 10'8 n/cnt'
% Decrease Observed Predicted Material U n i r ra d_,., Irrad. % Decrease Per R . G. 1. 99/2
Base Metal, A1768-1, 141 107 24 21 Longitudinal Base Metal, A1768-1, 120 88 27 21 Transverse HAZ Metal, A1768-1 87 74 15 21 Weld Metal, 305414/3951 103.5 60 42 45 Standard Reference 128 90 30 --
Material, A1008-1, Longitudinal
"' Calculated per Regulatory Guide 1.99, Revisicn 2.
"' Copper content not available.
8-5
Table 8-4. Comparison of Capsules W-225. W-265, and W-275 Charpy Test Results Fluence ACv30 ACv50 Upper-Shelf Material Capsule n/cm',_ F F Eneray, ft-lb Decrease. %
Base Metal, A1768-1, Baseline ---- --- ---
141 ---
1 Longitudinal W-225 i
5.53x10 a 60 69 122 13 W-265 7.71x10'8 74 78 109 23 1.28x10'8 W-275 73 84 107 24 Base Metal, A1768-1, Baseline ---- --- ---
120 ---
5 Transverse W-225 5.53x10 a ___ ___ ___ ___
5 W-265 7.71x10 a 70 78 93 23 W-275 1.28x10'8 72 108 88 27 HAZ Metal, A1768-1 Baseline ---- --- ---
87 ---
?'
W-225 5.53x10'8 104 117 98 -13 W-265 7.71x10'8 76 153 76 13 W-275 1.28x10'8 116 133 74 15 Weld Metal, 305414/3951 Baseline ---- --- ---
103.5 ---
W-225 5.53x10'8 238 258 65 37 W-265 7.71x10'8 221 283 59 43 W-275 1.28x10'8 219 301 60 42 4
Figure 8-1. Comparisor, of Unirradiated and Irradiated Charpy impact Data Curves for Base Metal Plate D4802-2, Heat No. A1768-1, Lonoitudinal Orientation too at Ts - -
so . -
n.
as - -
91oo o too m soo 4o0 soo
, o.1 gm -
15 :
Co - -
lL .oS u o.o. - a ssp .
l
~
7 l h c.02 -
f' -
.5
?1oo o too 200 soo 4o0 soo 22o m -
180 -
g 1M -
140 -
,,, _ a usE 34 na. _
100 -
D gm -
E s ur _
do -
[a 73p [ -
j 2o -
/ MATERIAL _SA-533 GR. 81 ,
HEAT NO. A17681 (LT) o 10o o too 2o0 soo doo soo Temperature, F 8-7
Figure 8-2. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal Plate 04802-2, Heat No. A1768-1, Transverse Orientation 100 Y -
75 -
50 -
n.
$H -
di 0 100 200 300 400 800
!100 0.1 ym -
3 -
g 0.06 -
0,04 . A111F ,
0.02 -
a 400 500
$100 0 100 200 300 no 200 -
100 -
g 1M 140 -
'h 120 -
y a USE = 32 ft-Ibo
] ,
9 g
gu -
60 -
3jogy 40 -
4 72F ,
20 -
HEAT NO. A1768-1 (TL) 0 100 0 100 200 300 400 500 Temperature, F 8-8
Figure 8-3. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal Plate D4802-2, Heat-Affected-Zone, Heat No. A1768-1 100 Y
fn 15
\
u L
e so - -
I. 2s - -
a 9200 100 0 100 203 300 400 500 0.1 g 0.0c -
a y 0.06 - -
0.04 -
A 140F _ -
E o 0.02 - -
200 100 0 100 200 300 400 600
\b 200 - -
180 - -
g 1M - -
"h140 b
g 120 100 - E -
g A USE = 13 Mbe ,
f3 80 - P -
60 -
A 133F -
40 -
/;,,,, / -
uArtmAL. SA-533 GR. B1 j, f-2a - -
asAr no. air i <saz) 200 100 0 100 200 300 400 500 I Temperature, F 8-9
Figure 8-4. ~ Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Weld Metal. 305414/3951 100 Y -
75 -
50 -
25 -
!200 100 0 100 200 300 400 800 0.1 0.00 -
A 252F _j .
0 . ,
0.02 -
3
!200 100 0 100 200 300 800 500 220 200 -
100 -
q 1M
. 140 -
~
120 -
g A USE a 43.5 fNbs gM -
E ' -
60 -
a soir
~
O ~
A 219F _
f' MATERIAL ASAtlNDE 1092 ,
g ,
HEATNO. 306414/3951 200 100 0 100 200 300 400 500 Temperature, F 8-10
- 9. REFERENCES
- 1. American Society for Testing and Materials (xaTM) Standard E185-66,
" Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors," Philadelphia, Pennsylvania.
- 2. Code of Federal Regulations, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix G, Fracture Toughness Requirements.
- 3. Code of Federal Regulations, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix H, Reactor Vessel Material Surveillance Program Requirements.
- 4. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, Appendix G, Protection Against Nondectile Failure (G-2000).
- 5. American Society for Testing and Materials (ASTM) Standard E208,
" Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," Philadelphia, Pennsylvania.
- 6. " Omaha Public Power District Fort Calhoun Station Unit No. 1 Evaluation -
of Irradiated Capsule W-225 Reactor Vessel Materials Irradiation Surveillance Program," TR-0-MCM-001. Revision 1, Combustion Engineering, Inc., Windosr, Connecticut, August 1980.
- 7. "0maha Public Power District Fort Calhoun Station Unit No. 1 Evaluation of Irradiated Capsule W-265 Reactor Vessel Materials Irradiation Surveillance Program," TR-0-MCM-002. Combustion Engineering, Inc.,
Windosr, Connecticut, March 1984.
- 8. " Fabricated History of the First Two 12-in.-Thick A-533 Grade B, Class 1 Steel Plates of the Heavy Section Steel Technology Program," 0RNL-4313. Oak Ride National Laboratory, Oak Ridge, Tennessee, February 1969.
- 9. "0maha Public Pore district Fort Calhoun Station Unit No. 1 Evaluation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program," TR-0-MCD-001. 2 Combustion Engineering, Inc., Windosr, Connecticut, March 1977.
- 10. American Society for Testing and Materials (ASTM) Standard E23-92,
" Methods for Notched Bar Impact Testing of Metallic Materials,"
Philadelphia, Pennsylvania.
, 11. American Society for Testing and Materials (ASTM) Standard E8-91,
" Standard Test Methods for Tension Testing of Metallic Materials,"
Philadelphia, Pennsylvania 9-1
- 12. American Society for Testing and Materials (ASTM) Standard E21-92,
" Standard Test Methods for Elevated Tergerature Tension Tests of '
Hetallic Materials," Philadelphia, P0nnsylvania.
- 13. Standardized Specimens for Certification of Charphy Impact Specimens from National Institute of Standards and Technology, Office of Standard Reference Materials, Gaithersburg, Maryland.
- 14. American Society for Testing and Materials (ASTM) Standard E185-82,
" Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," Philadelphia, Pennsylvania.
- 15. ASTM Designation E853-87, " Standard Practice for Analysis and Interpretation of Light Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, _
Philadelphia, PA, 1993.
- 16. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in ferritic Steels in Terms of Displacements per Atom (dpa)",
in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
- 17. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May 1988.
- 18. ORNL RSIC Computer Code Collection CCC-543, " TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport - Version 2.7.3", May 1993.
- 19. ORNL RSCI Data Library Collection DLC-76, " SAILOR Coupled Self-Shielded, 47 Neutron, 20 Ganna-Ray, P3, Cross Section Library for Light Water Reactors", .
- 20. R. E. Macrker, et al., " Accounting for Changing Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis", Nuclear Science and Engineering, Volume 94, Pages 291-308, 1986.
- 21. NUREG-0020, " Licensed Operating Reactors Status Summary Report",
Nuclear Regulatory Commission Monthly Publication, September 1973 through October 1993.
- 22. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-THE 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
l
- 23. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by foil Activation, AFWL-TR-7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 24. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al., 1981.
- 25. U.S. Nuclear Regulatory Connission, Standard Review Plan Branch Technical Position 5-2, Revision 1, NUREG-0800, July 1981.
9-2
- 26. Code of Federal Regulations, Title 10, Part 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.
l I
l 1
1 9-3 l l
l
i l
l
l l
\
APPENDIX A Reactor Vessel Surveillance Program Background Data and Information l
l l
I i
l l
l A-1
- 1. Capsule Identification The capsules used in the FCI surveillance program are identified in Table A-1 by identification, location, and original target fluence.
Table A-1. Capsule Assemb1v Identification Capsule Capsule Removal Time Target Removal Location EFPY Fluence, n/cm' 1 W-225 2.5 5.1 x 10 58 1
2 W-265 5.9 9.0 x 10 a 3 W-275 13.6 1.6 x 10
4 W-45 20 3.3 x 10
5 W-85 21 2.0 x 10
6 W-95 27 2.5 x 10'8 7 W-225 32 3.6 x 10
8 W-265 Standby ____
9 W-275 Standby ____
- 2. Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with ASTM E185-66, are shown in Table A-2. The locations of these materials within the reactor vessel are shown in Figure A-1.
- 3. Definition of Beltline Reoion The beltline region of the FC1 reactor vessel is defined in accordance with l 10CFR50, Appendix G.
- 4. Specimens Per Surveillance Capsule The type and quantity of each material contained in each surveillance capsule is shown in Table A-3.
A-2
Table 8-2. Unirradiated Impact Properties and Residual Element Content Data of Beltline Reaion Materials - Fort Calhoun Station Unit No. 1 Fabricator Material Material Beltline Drop Wt RTm USE. Chemistry, wtt Code Heat No. Material Type Region Location T. F F ft-lb Cu ni P S D4802-1 C2585-3 SA-533 Gr.81 Intermed. Shell -50 0" 75.45 0.12 0.56 0.011 0.015 D4802-2 A1768-1 SA-533 Gr.B1 Intermed. Shell -20 +18 121 0.10 0.48 0.009 0.014 D4802-3 A1768-2 SA-533 Gr.B1 Intermed. Shell -30 0" 77.35 " 0.11 0.51 0.009 0.012 04812-1 C3213-2 % 533 Gr.B1 Lower Shell -30 0" B6.45" 0.12 0.60 0.009 0.012 D4812-2 C3143-2 SA-:,33 Gr.B1 Lower Shell -20 C" 87 " 0.10 0.56 0.010 0.013 04812-3 C3143-3 SA-533 Gr.B1 Lower Shell -30 0" 89.7" 0.10 0.56 0.010 0.011 9-410 20291/3833 ASA Weld /Linde 1092 Middle Circum. -- -56" -
0.23 0.75 0.013 0.011 2-410-A.-B.-C 51989/3687 ASA Weld /Linde 124 Intenned. Longit. -- -56" -- 0.17 0.17 0.012 0.010 Y
" 3-410- A.-8. -C 27204/3774*8 ASA Weld /Linde 1092 Lower Longit. --- -56* -
0.21 1.00 0.013 0.011 12008/3774 13253/3774 l
l (a) Estimated per NRC Branch Technical Position MTEB 5-2.
(b) Estimated per 10CFR50.61. Protection Against Pressurized Thermal Shock Events.
(c) Most limiting chemistry data reported
Table A-3. Tvoe and Ouantity of Specimens Contained in Each Irradiated Capsule Assembly Base Metal Weld Metal Heat-Affected-Zone Standard (Heat A1768-1) (305414/3951) (Heat A1768-1) Reference Impact Material Total Specimens Capsule L T Tensile Impact Ter.sile Impact Tensile Impact Impact Tensile 1
l W-45 12 12 3 12 3 12 3 --
48 9 W-85 12 --
3 12 3 12 3 12 48 9 W-95 12 6 3 12 3 12 3 6 48 9 W-225 12 --
3 12 3 12 3 12 48 9
$ W-265 12 12 3 12 3 12 3 --
48 9 W-275
_ R 6 3 R 3 R 3 6 48 _R Totals 72 36 18 72 18 72 18 36 288 54 i
n , - - - -
Figure A-1. Location and Identification Of Materials Used in the Fabrication of Fort Calhoun Station Unit No. 1_ Reactor Pressure Vessel REACTOR VESSEL BELTLINE MATERIALS NOT SHOWN INTERMEDIATE SHELL Ml" m "P"N N WELD SEAM No. 2-410C Q ...
Q LOWER SHELL 'g;,TQ:g WELD SEAM No. 3-4108 WELD SEAM No. 3 410C
%'N PLATE No. D4812-2 .
Lk 1i c- , ,
3 e , Yl0 Ih ${
OUTLET f INLET NO2ZLE I NOZZLE L
UPPER TO INTERMEDIATE
,_ / INTERMEDIATE SHELL SHELL GIRTH SEAM L = ;/ LONGITUDINAL WELD WELD No. 8410 / SEAM No. 24109 INTERMEDIATE SHELL- x INTERMEDIATE SHELL PLATE No. D4802-2 y
, , - PLATE No. D-4802-3
~-
INTERMEDIATE SHELL h INTERMEDIATE TO LOWER LONGITUDINAL WELD , SHELL GIRTH SEAM SEAM No. 2410A WELD No. 9410 INTERMEDIATE SHELL
=
PLATE No. D48021 - LOWER SHELL LOWER SHELL : PLATE No. D4812-3 PLATE No. D-48121 LOWER SHELL 7 LONGITUDINAL WELD SEAM No. 3-410-A -
g" #
E!!!!!!!!!!!!!!!!!!!!!!!!!!!!$
A-5
i APPENDIX B Pre-Irradiation Tensile Data i
\
B-1 l
\
Table B-1. Tensile Properties of Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1 Lonaitudinal_j Test Elongation Reduction Specimen Temp.,
No.
Total / Uniform of Area, F Yield
- Ultimate % %
105 71 72.2/68.0 91.1 26/10.1 69.4 IE4 71 68.6/63.7 85.5 31/11.4 71.4 1EP 71 71.6/68.6 90.9 28/11.1 69.4 10B 250 64.9/63.1 83.8 26/10.0 69.4 10J 250 63.7/61.2 81.0 26/10.2 73.5 r 106 250 64.9/63.7 84.1 25/ 9.6 73.5 1EU 550 58.8/56.3 89.3 24/10.5 67.3 10E 550 56.9/55.1 85.7 22/10.0 67.3 1E3 550 55.1/53.9 81.9 25/10.3 69.4 o - Upper and lower yield strengths.
Table B-2. Tensile Properties of Unirradiated Base Metal, Plate 04802-2, Heat No. A1768-1. Transverse _7 Test Elongation Reduction Specimen Temp.,
No.
Total / Uniform of Area, F Yield
- Ultimate % %
207 71 66.1/64.3 86.5 29/11.6 65.3 200 71 69.2/66.1 89.1 28/11.7 67.3 2E1 71 70.4/67.4 90.6 28/11.3 67.3 2E2 250 63.7/62.5 83.0 25/ 9.3 67.3 2E3 250 64.9/62.5 83.8 25/ 9.6 69.3 20C 250 62.5/60.0 80.0 26/10.9 65.3 20L 550 55.7/53.9 83.9 25/10.5 73.5 20Y 550 57.6/56.3 87.1 23/10.0 63.3 200 550 55.1/53.9 83.3 24/ 9.6 65.3
- Upper and lower yield strengths.
B-2 I
I Table B-3. Tensile Properties of Unitradiated Base Metal Plate D4802-2, Heat-Affected-Zone. Heat No. A1768-1?
Test Elongation Reduction Temp., Total / Uniform of Area, Specimen %
No. F Yield
- Ultimate %
4E3 71 64.9/61.2 84.2 21/10.1 59.2 4DJ 71 66.1/63.7 86.9 22/ 9.8 67.3 4EM 71 69.0/64.2 84.1 29/10.2 70.0 4DL 250 60.0/58.8 79.8 20/ 8.1 65.3 4DE 250 61.2 80.3 21/ 8.0 63.3 _
4DK 250 58.8/58.1 79.4 19/ 8.3 61.2 4EL 550 54.0/52.8 80.7 22/ 8.3 66.0 4E5 550 52.7/51.4 81.3 18/ 8.7 57.1 407 550 54.0/52.8 83.5 24/ 9.2 64.0
- - Upper and lower yield strengths.
Table B-4. Tensile Properties of Unirradiated Weld Metal. 305414/39511 Test Elongation Reduction Specimen Temp., Total / Uniform of Area, No. F Yield
- Ultimate % %
3EP 71 75.3/73.5 90.3 27/10.0 69.4 3EJ 71 74.7/72.2 88.9 29/10.1 69.4 3JA 71 83.9/76.5 91.4 29/10.6 71.4 3EY 250 72.3/69.8 83.5 25/ 8.3 69.4 3DU 250 69.2/68.6 83.5 22/ 8.2 69.4 3DP 250 74.7/68.6 82.4 24/ 8.7 71.4 302 550 68.0/66.1 87.0 21/ 9.0 59.2 3JK 550 63.7/62.5 83.8 22/ 9.3 63.3 3E7 550 66.1/63.7 84.7 23/ 9.3 65.3
- - Upper and lower yield :trengths.
B-3 i
APPENDIX C Pre-Irradiation Charpy Impact Data C-1
Table C-1. Charpy Impact Data From Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1. Lonoitudinal Orientation
- Test Impact Lateral Shear Specimen Temp., Energy, Expansion, Fracture No. F ft-lb mils %
120 -80 4.5 1 0 ISK -40 7 8 10 13J -40 9 8 10 15P O 20.5 21 20 13E O 25.5 24 25 11C 40 41.5 38 35 162 40 45.5 39 35 llM 80 78 62 50 151 80 79 60 50 llP 120 115 85 80 12E 120 132.5 90 85 110 160 137.5 94 100 llT 160 140 93 100 13Y 190 147.5 90 100 llE 210 133.5 90 100 15L 210 144.5 90 100 Table C-2. Charpy Impact Data From Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1. Transverse Orientation' Test Impact lateral Shear Specimen Temp., Energy, Expansion, Fracture No. F ft-lb mils %
230 -80 4 3 0 23E -40 6.5 8 10 22P -40 8.5 10 10 233 0 17 18 20 247 0 26.5 27 25 21E 40 30 30 30 21U 40 32 32 30 21C 80 60 43 40 22T 80 62.5 53 55 242 120 78.5 66 60 21B 120 96.5 74 75 22C 160 94.5 76 90 238 160 102.5 74 85 21T 190 121 85 100 220 210 122.5 83 100 24D 210 125 84 100 217 230 110.5 83 100 C-2
i l
l Table C-3. Charpy Impact Data From Unirradiated Base Metal Plate D4802-2, Heat-Affect-Zone. Heat No. A1768-l' Test Impact Lateral Shear Specimen Temp., Energy, Expansion, Fracture No. F ft-lb mils %
444 -160 3 1 0 42L -140 13.5 11 10 423 -115 28 23 25 43C -80 20 20 25 46A -80 34.5 28 25 44Y -40 28.5 32 35 448 -40 70.5 50 40 43D 0 83.5 68 80 431 0 110 73 80 45E 40 30.5 30 45 415 40 101.5 67 80 417 80 113.5 81 80 450 120 65.5 60 85 416 120 129 84 90 45Y 160 82 73 100 467 160 87 68 100 466 200 93 67 100 Table C-4. Charov Imoact Data From Unirradiated Weld Metal. 305414/3951' Test Impact lateral Shear Specimen Temp., Energy, Expansion, Fracture No. F ft-lb mils %
364 -120 5.5 4 0 35U -80 5.5 9 10 32Y -80 18.5 17 15 332 -40 20.5 21 30 31L -40 30 28 35 35E O 49.5 41 40 33E O 55 47 50 35T 40 66.5 55 65 31Y 40 74 64 70 33C 80 97.5 83 100 324 80 105.5 86 100 35P 120 92.5 81 100 33D 120 105 89 100 346 160 105.5 83 100 31U 160 115 89 100 C-3
Figure C-1. Charpy Impact Data for Unirradiated Base Metal Plate D4802-2, '
Heat No. A1768-1. lonaitudinal Orientation 100 -.
Y
- p 75 -
- 3 D
ga i
EH J::
- e V) /
0.100 0 100 200 300 400 500 g 0.1
[ 0.0e _
a
@ 0.06 0.04
/
6 -
.o. 0.02 -
3 0
- 100 0 100 200 300 400 500 220 T 20F NOT
~ Yn (35 MLE) +31F -
180 Tu (50 FT.LB) + 41 F T"(30 FT LB) + 22F
.O 160 ~
r CvUSE (AVG.) 141 FT.LBS..
,120 -
C s W 100 -
U R* -
e
-m -
40 -
. .. .. . . . . . . . . . . . . . . . . . . . . . . . MATERIAL SA 533 OR. B1 ~
20 -
FLUENCE NONE ~
HEAT NO. A17681 (LT) 0 ' ' i t i
- 100 0 100 200 300 400 500 Temperature, F C-4
1 Figure C-2. Charpy Impact Data for Unirradiated Base Metal Plate D4802-2, Heat No. A1768-1. Transverse Orientation ;
100 8
rs - - -
80 -
3g . -
0.100 0 100 200 300 400 800 0.1 s .
0.00 -
0 - -
n h a04
.06 0.02 -
0.100 0 100 200 300 400 500 220 T NDT *#
- + 44F
~
~ T,y(35 MLE) 100 T,y(50 FT LO) +70F -
T8'(30 FT.LB) + 36F J3 180 -
CvuSE (AVG.) 120 FT L89
- 0. 140 RT HDT +18F -
f120 s ,00 .
u o e0 -
Q.
.E. e0 - -
40 -
. . . . . . . . . . . . . . ....... ... ...... ...... MATERIAL 8A 533 GR. Si ~
20 - -
HEATNO. A17881 (TL) 1 0
-- i i i i i l
100 0 100 200 300 400 500 Temperature, F C-5 a
Figure C-3. Charpy Impact Data for Unirradiated Base Metal Plate D4802-2, Heat-Affected-Zone. Heat No. A1768-1 100 :
Y. . .
g 18 - -
w eu -
s gn - . . .
.c V) .
200 1b 0 1bo 200 300 dbo 500 0.1 g au .
to + . . .
@ 0.06 a .
5 0.04 O .
e 0.02 sa b
200 100 0 100 200 300 400 500 220 NOT OF
- 62F
~
~ T,(35 MLE) 180 T,(50 FT.LB) 28F .
~ T,(30 FT LB) 76F h
~
CvDSE (AVO.) . 87 FT.LBS 140 RT + 24E_,_ -
NDT h"120 100 - * # -
ti
- M 80 - * -
4 .
.E so . -
0 -
.. ..,..... MATERIAL SA 533 OR. 81
- n -
e ..
HEAT NO. A17681 (HAZ) 0
-200 100 0 100 200 300 400 500 Temperature, F C-6
I Figure C-4. Charpy Impact Data for Unirradiated Weld Metal, 305414/3951 100 Y -
p 7s -
- 3 D
ew u.
- m. e m * -
e 25 -
C .
M 9200 100 0 100 200 300 400 000 g 0.1 8 .
g a0e -
@ 0.06 ,
ca.
ON ~ . . . . . . . . . ... .... . . . . . . . . . . . . . . . . . .............................................~
E - . -
.c am .
b 400 500 o.200 100 0 100 200 300 220 T NOT OF
~
200 18F T,(35 MLE) j30 T,(50 FT.LB) +4F -
T (30 FT.LB) 28F -
D 160 7 o"vuSE (AVO.) 103.5 FT.L8S
.140 - RT NOT OF 120 -
6 ,w -
u -
m 30 -
C2.
E m -
40 -
MATERLAL ASA/ Unde 1092 ~
20 -
HEAT NO. 305414/3951 0
200 100 0 100 200 300 400 500 Temperature, F C-7
l APPENDIX D Tension Test Stress-Strain Curves e
0-1
Figure D-1. Tension Test Stress-Strain Curve for Base Metal Plate 04802-2. Specimen No. 103. Tested at 70F Specimeni 103 feet Temp.: 70 FI 21 Cl b Strength -
y field: 79527.
U15: 101695.
~
[
~
b
<S .
- j h
- : a b l l pd NN
- e
- ,' .g t . .
Y 5
.w e
r" g ?
d d , . . .
0.00 0.0i1 0.08 0.12 0.16 0.20 0.24 0.28 0.32 Engine.,-ing Strain Figure D-2. Tension Test Stress-Strain Curve for Base Metal Plate 04802-2. Specimen Np. 102. Tested at 250F SP*cI*eni 102 Tent Temp.: 250 F( 121 Cl d
".d S treng th - d Tleid 74257.
UIS: 94626.
8 -
y f -
- 5. .
.t P
S i 35 c : P
.i.'. rj -
.I
.w 7
Pd i i i e i i i g g 0.00 0.04 0.00 0.12 0.16 0.20 0.24 0.28 0.32 Engineering Strain D-2
_a Figure D-3. Tension Test Stress-Strain Curve for Base Metal Plate D4802-2. Specimen No. 1EA. Tested at 550F d Specimen IER Tes t T..p. i $50 Ff 287 C1
% S treng th _
I1eIdi 66998. O' ufS 91698.
N -
db
- G-
"8 .:
J : b 2 .
PS .
sm T i ? -=
.t
. T.
Es -
.?
e l
gi-d r l
I f I I i f I 0.00 0.04 0.08 0,12 0.16 0.20 0.24 0.28 0.32 Engineering sic.in Figure D-4. Tension Test Stress-Strain Curve for Base Metal Plate 04802-2. HAZ. Specimen No. 4EA. Tested at 70F Test Temp.: 70 Fi 21 Cl g Specimen 4EA *
- Strength E Tleidi 73636.
U15: 95883.
b-sh 55 f $
- t 2 .
d .
go . i. -
g5 I P T
P .'
Id : .!
e r; "
d b
. t I I I f i 1 0.00 0.04 0.00 0.12 0.16 0.20 0.24 0.28 0.32 Engineering Stealn D-3
~
1 l
1 l
l Figure D-5. Tension Test Stress-Strain Curve for Base Metal Plate 04802-2. HAZ. Soecimen No. 4EJ. Tested at 250F g Specimen 4EJ Tes t Temo.: 250 Fi 121 Cl ,
- Strength d m
l l
TieIdi 72073.
UTS: 92209.
i
$N i C;
x 8 j
! J A. !
P8 ~! dS I I P
.E i I
[d '.j .5 I
d*
! # i i t i I .
O. O 0.00 0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16 Engineering Strain Figure D-6. Tension Test Stress-Strain Curve for Base Metal Plate D4802-2. HAZ. Specimen No. 4EE. Tested at 550F Test Temp.: $50 F t 287 C) g Specimen 4EE
- Strength _ d fleid: 69736. o U1$r 94071. ,
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