ML19290F145
| ML19290F145 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/31/1979 |
| From: | Byrne S, Koziol J, Schoenbrunn A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML19290F143 | List: |
| References | |
| TR-O-MCM-001, TR-O-MCM-1, NUDOCS 8003180183 | |
| Download: ML19290F145 (122) | |
Text
{{#Wiki_filter:_ OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit no.1 evaluation OF = = irradiated capsule w-22s REACTOR VESSEL MATERIALS IRRADIATION SURVEILLANCE PROGRAM _tmPOWER SYSTEMS COMBUSTION ENGINEERING, INC. 8003180lh
TR-0-MCM-001 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION UNIT NO. 1 POST-IRRADIATION EVALUATION OF REACTOR VESSEL SURVEILLANCE CAPSULE W-225 May 1979 Prepared by: h' - Date: fb d /927 S. T. Byr e ', Co afzan Engineer y Approved by: J. f/N / Dz.te: f v7/ /IJf Koziol,p rdgram tana'ger [ Approved by: e f[ m Date: gewp M[/[ff A. G. StrrosTibrunn, OPPD Project Engineering Services Manager Combustion Engineering, Inc. Nuclear Power Systems Windsor, Connecticut
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QA Status: 'ierded The safety re'ated desig, in ctmation codained in this document has been reviewed 3nd satisfies (where ap;;hcat:e) the .teil.3 C0n!3.'e1 01 Chack-
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I TABLE OF CONTENTS Section Title Page No. I Summary 1 II Introduction 3 III Surveillance Program Description 4 IV Capsule Withdrawal and Disassembly 16 V Test Results 18 VI Data Analysis 67 VII Reactor Coolant System Pressure - Temperature 73 Limitation Curves VIII References 84 Appendix A Tensile Tests - Description and Equipment A-1 Appendix 8 Charpy Impact Tests - Description and Equipment B-1 Appendix C Instrumented Charpy V-Notch Data Analysis C-1 I I I I I I I ii I
ist of Tables Table No. Title _ Page No. III-l Reactor Vessel Beltline Plates 5 III-2 Reactor Vessel Beltline Welds 6 III-3 Reactor Vessel Beltline Plates Chemical Analysis 7 III-4 Surveillance Plate and Weld Metal Chemical 8 Analysis III.5 Fort Calhoun Reactor Vessel Surveillance 14 Capsule Removal Schedule III-6 Type and Quantity of Specimens in 225 Capsule 15 IV-1 Mechanical Test Specimens Removed from 225 17 Capsule V-1 Composition and Melting Points of Temperature 19 Monitor Materials V-2 Neutron Flux Monitors 21 V-2 Ft. Calhoun Iron Flux Attenuation Monitors, 26 Compartment P',14 V-4 Ft. Calhoun Iron Flux Attenuation Monitors, 26 Compartment 2441 _ V-5 Ft. Calhoun Iron Flux Attenuation Monitors, 27 Compartment 2473 V-6 Ft. Calhoun Flux Spectrum Monitors, Compartment 27 2414 V-7 Ft. Calhoun Flux Spectrum Monitors, Compartment 28 2441 V-8 Ft. Calhoun Flux Spectrum Monitors, Compartment 28 2473 i V-9 Flux Monitor Activities 31 V-10 Fort Calhoun Fast Neutron Flux and Fluence Values 34 V-ll Iron Flux Monitors 37 V-12 Charpy V-Notch Impact Resul ts for Fort Calhoun 39 Standard Reference Material iii 1 - - - -
I List of Tabies (Cont'd.) Table No. Title Page No. V-13 . Post-Irradiation Tension Test Properties 44 V-14 Pre-Irradiation Tension Test Properties 45 V-15 Charpy Impact Results, Base Metal 46 V-16 Charpy Impact Results, Weld Metal 47 V-17 Charpy Impact Results, HAZ 48 VI-1 Summary of Toughness Property Changes 71 C-1 Instrumented Charpy Test, Base Metal C-3 C-2 Instrumented Charpy Test, Weld Metal C-4 C-3 Instrumented Charpy Test, HAZ C-5 C-4 Instrumented Charpy Test, SRM C-6 C-5 Toughness Property Change: Based on C-7 Instrumented Charpy Impact Test I I I I I I 1 iv I
List of Fiaures Fiqure No. Title Page No. III-l Surveillance Capsule Assembly 10 III-2 Charpy Impact Compartment Assembly 11 III-3 Tensile-Monitor Compartment Assembly 12 III-4 Location of Surveillance Capsule Assemblies 13 i V-1 Efficiency Calibration 23 V-2 Efficiency Calibration 24 V-3 ANISN Geometry 33 V-4 Iron Flux Wire Housing 36 V-5 Post-Irradiation Charpy Impact Properties, 41 Standard Reference Material V-6 Trend Curve Analysis, Standard Reference 42 Material Data, Comparison with HSST Data V-7 Stress-Strain Record, Base Metal, 72F 49 V-8 Stress-Strain Record, Base Metal, 250F 49 V-9 Stress-Strain Record, Base Metal 550F 50 V-10 Stress-Strain Record, Weld Metal, 72F 50 V-ll Stress-Strain Record, Weld Metal, 250F 51 V-12 Stress-Strain Record, Weld Metal, 550F 51 V-13 ctress-Strain Record, HAZ, Metal, 72F 52 V-14 S^ress-Strain Record, HAZ, Metal, 250F 52 V-15 Sti ess-Strain Rec.ord, HAZ, Netal, 550F 53 V-16 Fraccure Surface of Irradiated Tension 54 Specincens V-17 Charpy Impact Energy, Base Metal 55 V-18 Charpy Lateral Expansion, Base Metal 56 v
List of Figures (Cont'd.) Figure No. Title Page No. V-19 Charpy Shear Fracture, Base Metal 57 V-20 Charpy Impact Energy, lleld Metal 58 V-21 Charpy lateral Expansion, Weld Metal 59 V-22 Charpy Shear Fracture, Weld Metal 60 V-23 Charpy Impact Energy, HAZ 61 V-24 Charpy Lateral Expansion, HAZ 62 V-25 Charpy Shear Fracture, HAZ 63 V-26 Fracture Surfaces, Impact Specimens, 64 Base Metal V-27 Fracture Surfaces, Impact Specimens, 65 Weld Metal V-28 Fracture Surfaces, Impact Specimens, 66 HAZ VI-l Predicted NOTT Shift for the Fort Calhoun 72 Reactor Vessel Beltline VII-l 5 Year RCS Pressure - Temperature Limitations, 76 Heatup VII-2 5 Year RCS Pressure - Temperature Limitations, 77 Cooldown VII-3 5 Year RCS Pressure - Temperature Limitations, 78 Heatup, Critical VII-4 5 Year RCS Pressure - Temperature Limitations, 79 Cooldown, Critical VII-5 10 Year RCS Pressure - Temperature Limitations, 80 Heatup VII-6 10 Year RCS Pressure - Temperature Limitations, 81 Cooldown VII-7 10 Year RCS Pressure - Temperature Limitations, 82 s Heatup, Critical VII-8 10 Year RCS Pressure - Temperature Limitations, 83 Cooldown, Critical vi
i i List of Figures (Cont'd.) Figure No. Title 'aqe No. A-1 Tensile Test System A-2 A-2 Typical Tensile Specimen A-3 A-3 Location of Tensile Specimens in Base A-4 Metal A-4 Location of Tensile Specimens in Weld A-4
> Metal A-5 Location of Tensile Specimens in HAZ A-6 B-1 Charpy Impact Test System B-4 B-2 Typical Charpy V-Notch Impact Specimen B-5 B-3 Location of Charpy Specimens in Base Metal B-6 B-4 Location of Charpy Specimens in Weld Metal B-7 B-5 Location of Charpy Specimens in HAZ B-8 C-1 ICV Load vs. Temperature Curves, Base Metal C-8 C-2 ICV Load vs. Temperature Curves, Weld Metal C-9 C-3 ICV Load vs. Temperature Curves, HAZ C-10 C-4 ICV Load vs. Temperature Curves, SRM C-ll vii
I.
SUMMARY
The first surveillance wall capsule (W-225) was removed from the Fort Calhoun reactor vessel in October 1977 after 2.6 effective full power years of reactor operation. The surveillance test specimens and monitors were evaluated at C-E's Windsor, Connecticut laboratory facility. Post-irradiation evaluation of the temperature monitors indicated that the irradiation temperature was between 536 F and 558 F. Analysis of the neutron threshold detectors provided a capsule fluence of 5.1 x 10 18 n/cm2 (E>l M2V), which corresponded to a maximum fluence at the inside surface of the reactor vessel of 4.5 x 10 18 n/cm2 , Radiation induced changes in the tensile and impact properties were determined for the base metal (longitudinal orientation), weld metal and heat-affected zone surveillance materials. Transition temperature shifts ranged from 60 F for the base metal to 238 F for the weld metal. The upper shelf impact energy after irradiation was in excess of 50 ft-lb for each of the surveillance materials, ranging from 119 ft-lb for the base metal to 64 ft-lb for the weld metal. The measured shift and percent decrease in shelf energy were in agreement with the predictions based on Regulatory Guide 1.99, Revision 1. The weld metal exhibited the greatest toughness property change consistent with having the highest residual copper content (0.35 w/o). The post-irradiation tensile properties exhibited the same general trends as the toughness properties; the yield strength of the weld metal increased 35% versus 14 to 18% for the base metal and HAZ, respectively. Ductility changes with irradiation were generally of a smaller magnitude such that the ductility was sustained near the pre-irradiation levels. For example, total elonaation for the weld metal ranged from 20 to 23% after irradiation as compared to 22 to 28% before irradiation. 1
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I g Future projections of transition temperature shift for the Fort Calhoun reactor vessel beh;line materials will be based on the g Guidelines of Regulatory Guide 1.99, Revision 1. The RCS operating a limit curve; for operation through January 1980 (4.3 EFPY) were adjusted based on revised estimates of transition temperature shift and neutrco fluence. Limit curves were also developed for 4.3 to 8.3 EFPY operation. i I I l I I i I i I 1 1 4 I
N- - - - - II. Introduction i The purpose of the Fort Calhoun surveillance program is to monitor the radiation induced changes in the mechanical properties of ferritic materials in the reactor vessel beltline during the operating lifetime of the reactor vessel. The surveillance program includes the determination of the preirradiation (baseline) strength and toughness properties and periodic determinations of the property changes following neutron irradiation. These property f changes are used to verify and Update the operat, limits (heat-up and cool down pressure / temperature limit curves) for the primary system. The Fort Calhoun Surveillance program (I) is based upon ASTM E185-66, " Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors". The pre-irradiation (baseline) evaluation results from the Fort Calhoun reactor vessel surveillance materials are described in C-E report TR-0-MCD-001.(2) The following report describes tha results obtained from evaluation of irradiated materials from capsule W-225 which was removed from the reactor in October 1977. I i l 1 I I i r
I III Surveillance Program Description The Fort Calhoun reactor pressure vessel was designed and fabricated by Combustion Engineering, Inc. The reactor vessel beltline, as defined by 10CFR50, Appendix H, consists of the six plates used to fann the lower and intermediate shell courses in the vessel, the included longitudinal seam welds and the lower to intermediate shell girth seam weld. The plates were manufactured from SA533 Grade B Class 1 quenched and tempered plate. The heat treatment consisted of austenization at 1575125F for four hours, water quenching and tempering at 1225 1 25F for four hours. The ASME Code qualification test plates were stress relieved at 1150 1 25F for forty hours, and furnace cooled to 600F. The longitudinal s and girth seam welds were fabricated using E8018-C3 manual arc electrodes and Mil B-4 submerged arc weld wire with Linde 124 and s Linde 1092 flux. The post weld heat treatment consisted of a twelve hour 1150 1 25F stress relief heat treatment followed by furnace cooling to 600F. The beltline materials (3) are identified , in Tables III-l and III-2. The chemical analysis ( ) of the six beltline plates is given in Table III-3. The materials included in the surveillance program were selected to represent the beltline materials from the reactor vessel. The base metal surveillance material, plate D-4802-2, was selected from the six beltline plates on the basis of the highest initial drop weight NDTT. The , heat treatment of the surveillance plate duplicated that of the reactor vessel ASME Code qualification test plates. The surveillance weld material was fabricated by welding plate D-4802-1 to plate D-4802-3 using the same weld procedure used for the intermediate to lower shell girth seam weld. The same type of filler wire and flux was used. The post-weld heat treatment consisted of a forty hour stress relief at I PS 1 25F followed by furnace cooling to g 600F. The surveillance heat-affected zone material was fabricated W, by welding plate D-4802-2 to plate D-4802-3 in the same manner as the surveillance weld material with the same postweld heat treatment. The chemical analysis of the surveillance plate and weld ( ) is given in Table III-4. 4 I
TABLE III-l REACTOR VESSEL BELTLINE PLATES Location Piece Number Code Number Heat Number Supplier Intennediate 436-02B D-4802-1 C-2585-3 Lukens Shell Intermediate 436-02A D-4802-2 A-1768-1 Lukens Shell Intermediate 436-02C D-4802-3 A-1768-2 Lukens Shell Lower Shell 436-03B D-4812-1 C-3213-2 Lukens Lower Shell 436-03A D-4812-2 C-3143-2 Lukens Lower Shell 436-03C 0-4812-3 C-3143-3 Lukens 5
I TABLE III-2 REACTOR VESSEL BELTLINE WELDS i Location Weld Seam NA Wire Heat No. Flux Type Flux Batch Intermediate 2-410A 51989 Linde 124 3687 Shell Longi-tudinal Seam Intermediate 2-410B M/A JBFG* - - g Shell Longi- E tudinal Seam Intermediate 2-410C 51989 Linde 124 3687 Shell Longi-tudinal Seam Lower Shell 3-410A 12008 Linde 1092 3774 Longitudinal 13252 Linde 1092 3774 Seam 27204 Linde 1092 3774 Lower Shell 3-410B M/A E0AG* - - Longitudinal Seam Lower Shell 3-410C 12008 Linde 1092 3774 Longitudinal 13252 Linde 1092 3774 Seam 27204 Linde 1092 3774 Intennediate 9-410 20291 Linde 1092 3833 to Lower Girth Seam
- Manual shielded metal arc electrode (all others automatic submerged arc wire).
I e I I
TABLE III-3 REACTOR VESSEL BELTLINE PLATES CHEMICAL ANALYSIS Element D-4802-1 D-4802-2 D-4802-3 D-4812-1 D-4812-2 D-4812-3 Si .23 .23 .24 .24 .26 .25 S .015 .014 .012 .012 .013 .011 P .011 .009 .009 .009 .010 .010 Mn 1.27 1.43 1.50 1. 31 1.33 1.30 C .21 .22 .29 .22 .26 .22 Cr .08 .04 .05 .18 .06 .06 Ni .56 .48 .51 .60 .56 .56 Mo .49 .50 .53 .54 .52 .51 V <.001 <.001 .002 .002 .002 .002 Cb <.01 <.01 <.01 <.01 <.01 <.01 B .0004 .0003 .0004 .0006 .0006 .0002 Co .007 .007 .008 .009 .007 .007 Cu .12 .10 .11 .12 .10 .10 Al .020 .030 .024 .029 .038 .027 W .02 .02 .02 .02 .02 .01 Ti <.01 <.01 <.01 <.01 <.01 <.01 As <.01 <.01 <.01 <.01 <.01 4.01 Sn .002 .002 .002 .002 .001 .001 Zr .002 .002 .002 .002 .002 .002 N .009 .009 .010 .007 .008 .007 2 1 7
E TABLE III-4 SURVEILLANCE PLATE AND WELD METAL CHEMICAL ANALYSIS Weight Percent Plate Weld Element D-4802-2 D-4802-1/D-4802-3 Si .23 .14 S .014 .011 P .009 .013 Mn 1.43 1.57 g C .22 .14 5 Cr .04 .03 Ni .48 .60 Mo .50 .50 V <.001 .002 Cb <.01 <.01 B .0003 .0002 Co .007 .014 Cu .10 .35 E Al .030 .009 5 W .02 .02 Ti <.01 <.01 As <.01 <.01 Sn .002 .007 l Zr .002 .002 N .009 .012 2 I I I 8 i
Drop weight, Charpy impact and tension test specimens were machined from the surveillance materials as described in reference 1. In addition to the surveillance material specimens, Charpy impact specimens were machined from a section of plate 01 from the Heavy Section Steel Technology (HSST) program to serve as standard reference material (SRM). The surveillance and SRM test specimens were enclosed in six capsules for irradiation in the Fort Calhoun reactor vessel. The surveillance capsule assembly is shown in Figure III-1. Each assembly consists of four compartments containing Charpy impact specimens (Figure III-2) and three compartments (Figure III-3) containinc tensile specimens and monitors (flux and temperature). Each capsule is positioned in a holder tube attached to the reactor vessel cladding to irradiate the specimens in an environment which duplicates as closely as possible that experienced by the reactor vessel. Capsule locations are shown in Figure III-4. The axial portion of each capsule is bisected by the midplane of the core. The circumferential locations were selected to coincide with the peak flux regions of the reactor vessel. The withdrawal s hedule for the surveillance capsules is given in Table III-5. It was based on the requirements of 10CFR50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements," and an estimated end-of-life adjusted reference temperature in excess of 200F. The type and quantity of test specimens contained in the 225 capsule are given Table III-6. 9
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I TABLE III-5 I FORT CALHOUN REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE Removal Refueling Capsule Removed Sequence Schedule EFPY* Preferred Alternate 1 2.6 225 85, 95 or 275 2 10 45 265 3 17 85, 95, 275 or 225 4 24 85, 95, 275 or 225 5 Standby Any of remaining capsules 6 Standby Any of remaining capsules
- EFPY - Effective full power years I
I I I I I I I 14 I I
TABLE III-6 TYPE AND QUANTITY OF SPECIMENS IN 225 CAPSULE Material Charpy Impact Tensile Base Metal 12 3 (longitudinal) Weld Metal 12 3 Heat-Affected Zone 12 3 Standard Reference 12 - Material Total 48 9 15
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I IV. CAPSULE WITHDRAWAL AND DISASSEMBLY I The Fort Calhoun 225 surveillance capsule was removed from the reactor vessel during the October 1977 refueling outage (September 30, 1977, shutdown). Removal was accomplished by attaching a special tool to the capsule lock assembly to disengage the latches E and withdraw the capsule. The five remaining capsules were 5 inspected using an under-water video system. The inspection revealed that the capsules were securely locked in position. The 225 capsule was transferred to the spent fuel pool where it was sectioned into lengths for insertion into a shipping cask. Sectioning was accomplished by drilling to separate the wedge assembly halves, leaving the specimen compartments intact. The surveillance capsule was shipped to Neutron Products, Inc. in g Dickerson, Maryland, for inspection, disassembly and specimen 5 removal in the hot cell facility. No unusual features or damage were revealed by visual inspection. A remote control circular saw was used to open the capsule compartments. Each compartment was identified and inspected prior to cutting, and the contents were removed and verified against the original loading records. An inventory of the mechanical test specimens removed from the 225 capsule is given in Table IV-1. I I I I I I 16 I
TABLE IV-1 MECHANICAL TEST SPECIMENS REMOVED FROM 225 CAPSULE Compartment Material and Number Specimen Type Specimen Identification 2414 HAZ Tensile 4EC, 4D2, 4EK 2424 HAZ Charpy 463, 45L, 45U 42B, 42U, 452 41B, 43P, 42Y 461, 41Y, 446 2435 SRM Charpy 55M, 55Y, 575 573, 55P, 563 56J, 55T, 56A 566, 56D, 565 2441 Base Metal Tensile 1DU, 1D1, lEL 2451 Base Metal Charpy 14T, 144, ISD 167, 13C, 13D 15C , 12A, 13B 112, 117, 14M 2463 Weld Metal Charpy 341, 32J, 32E 31 E , 317, 31 B 32U, 336, 33T 352, 312, 31C 2473 Weld Metal Tensile 3ES, 3EC, 3ET 17
I V. TEST RESULTS A. Irradiation Environment
- 1. Temperature Monitors Each Tensile-Monitor Compartment (Figures III-l and III-3) in the Capsule assembly contained a set of four temperature monitors to provide an indication of the maximum temperature in the capsule during irradiation. The composition and melting point of each eutectic alloy monitor is given in Table V-1. Each monitor consisted of a helix of the eutectic alloy and a stainless steel weight encapsulated in a quartz tube. Each set of four temperature monitors was inserted into a stainless steel housing, and the temperature monitors were irradiated in the top, middle and bottom surveillance g capsule compartments. 5 Post-irradiation examination of the temperature monitors was performed in the hot cell. Once the monitor housing was extracted from the capsule compartment, each temperature monitor a s identified by length. Each monitor was inspected to determire vhether the eutectic alloy helix had been crushed by the weight. Only the 80% Au-20% Sn alloy melted, indicating that the capsule temperature exceeded 536 F, but was less than 558 F (the next higher monitor melting point).
The same behavior was exhibited by each of the three sets of monitors, indicating a relatively uniform maximum temperature profile along the length of the surveillance capsule.
- 2. Neutron Dosimetry I
Each Tensile-Monitor compartment (Figures III-l and III-3) in the capsule assembly contained one set of neutron flux I 18
TABLE V-1 COMPOSITION AND MELTIfiG POINTS OF TEfiPERATURE MONITOR MATERIALS Composition fielting Temperature (Weight ";) F 80 Au, 20 Sn 536 90 Pb, 5 Sn, sag 558 97.5 Pb, 2.5 Ag 580 97.5 Pb, 0.75 Sn,1.75 Ag 590 19
I monitors and one set of flux attenuation monitors. The flux monitors are described in Table V-2. Each flux monitor was g encapsulated in a stainless steel sheath (except for the 5 sulfur which had a quartz sheath); in addition, cadmium covers were placed around the uranium, nickel and copper monitors which have competing thermal activities. Each set of seven ilux monitors was inserted into a stainless steel housing, one set for each of the top, middle and bottom l surveillance capsule compartments. The flux attenuation monitors are composed of five iron g wires encapsulated in a stainless steel sheath and positioned 5 at three different distances from the core within a stainless steel housing. One set of flux attenuation monitors was inserted in each of the top, middle and bottom capsule compartments. The flux monitors were removed from the capsule compartments in the hot cell. Each monitor was inspected and its position in the housing verified by the number of grooves in the stainless steel sheath. The monitors were then repackaged and shipped to C-E's Windsor, Connecticut facility for radiochemical analysis.
- a. Radiochemical Analysis Radiochemical analysis of the flux monitors was performed in accordance with C-E Procedure 00000-FMD-401, Rev. O, November 1, 1978 (" Standard Method for the Analysis of Radioisotopes in Reactor Irradiation Surveillance Detectors and Flux Distribution Monitors"). Each Monitor was removed from its sheath and inserted in a glass vial. Recovery of the uranium, titanium and cadmium shielded monitors was complicated by oxidation and contamination of the monitors.
I 20 I
TABLE V-2 NEUTRON FLUX MONITORS Material Reaction Threshold Eneray (Mev) Half-Life Uranium U238(n,f) Csl37 0.7 30.2 years Titanium Ti46(n,p) Sc46 8.0 84 days Iron F.54(n,p) Mn54 4.0 314 days Uranium U238(n,f) Csl37 0.7 30.2 years (Cadmium Shielded) Nickel Ni 8(n,p) CoS8 5.0 71 days (Cadmium Shielded) Copper CuS3(n,a) Co60 7.0 5.3 years (Cadmium Shielded) Sulfur S32(n p) P 2 2.9 14.3 days 21
I The uranium foil had converted to a black powder, assumed to be U380 . Therefore, instead of using a simple gravimetric measurement, the amount of uranium recovered was determined by atomic absorption spectroscopy. The titanium wires were found broken into several pieces, but otherwise they presented no handling or counting problems. The cadmium shield on the copper and nickel wires had apparently melted and fused to the wire during irradiation. The cadmium shields were mechanically removed by stripping, scraping and filing. Final monitor weights were based on elemental analysis using atomic absorption spectroscopy. The remainder of the samples for radiochemical analysis were prepared using standard methods. Counting was performed with a 4096 channel gamma spectrometer system coupled with a lithium-drifted gemanium detector. g The system was calibrated at 0.5 Kev per channel to span the y gamma energy range from 0.05 to 2 Mev. Efficiency calibration was performed using eight (8) gamma energies emitted from an flBS traceable mixed isotope standard. Detector efficiency curves were detennined by a least squares analysis of eight (8) plotted efficiency poir.;s. Detector efficiency curves used in the surveillance capsule analysis are given in Figures V-1 and V-2. Sulfur monitors could not be analyzed due to the complete decay of phosphorous-32 during the elapsed time from end of g irradiation to analysis. 5 Physical constants used in the calculation of radioisotope activity levels are as follows: Isotope Half-Life Gamma Energy (Mev) Intensity Cobalt-58 71.3 days 0.810 0.9944 Cobalt-60 5.26 years 1.173 0.9988 Cesium-137 30.0 years 0.662 0.846 Manganese-54 314 days 0.835 1.000 Scandium-46 83.8 days 0.889 1.000 22 g
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I 2d I I
Results for the iron flux attenuation monitors are listed in Tables V-3 through V-5. Flux spectrum monitor activity levels are presented in Tables V-6 through V-8. All values are decay corrected to the time of reactor shutdown, 0948, September 30, 1977. The uncertainty listed with each result is the 2-sigma counting error only. An additional error of 1,20% for uranium monitors and 1,5% for all other monitors is estimated from volumetric / gravimetric operations and from the certified uncertainties of calibration isotopes. The shutdown activities determined from gamma ray emission rates were calculated as follows: A= "P EWBC (exp-At) where: A = shutdown activity in disintegrations per minute per milligram of material (dpm/mg) Np = radioisotope net counts per minute E = full energy peak effiency (counts per gamma ray emitted) W = weight of monitor sample (milligrams) B = radioisotope gamma ray branching r>tio (gamma rays per disintegration) C = correction for coincident or random summing a = radioisotope decay constant t = elapsed time between plant shutdown and counting 25
I TABLE V-3 FT. CALHOUN IRON FLUX ATTENUATION MONITORS COMPARTMENT 2414 Counting Monitor Number of Weight Decay Interval Measured Activity at End Material Grooves (mg) (Days) (Seconds) Isotope Irradiation (0pm/mg) 5 Iron 1 26.0 479.02 2000 54 Mn 1.629 1 0.013 x 10 5 Iron 2 26.4 479.02 2000 54 Mn 1.618 1 0.013 x 10 5 Iron 3 26.4 479.15 2000 54 Mn 1.645 1 0.013 x 10 5 Iron 4 26.1 479.17 2000 54 Mn 1.788 1 0.013 x 10 5 Iron 5 26.4 479.20 2000 54 Mn 1.531 1 0.012 x 10 I TABLE V-4 FT. CALHOUN FLUX ATTENUATION MONITORS COMPARTMENT 2441 Counting a Monitor Number of Weight Decay Interval Measured Activity at End 5 Material Grooves (mg) (Days) (Seconds) Isotooe Irradiation (Dom /mg) 5 Iron 1 26.8 479.22 2000 54 Mn 1.596 1 0.013 x 10 5 Iron 2 26.2 479.93 2000 54 Mn 1.625 1 0.013 x 10 5 Iron 3 26.0 479.96 2000 54 Mn 1.605 1 0.013 x 10 5 Iron 4 26.5 479.99 2000 54 Mn 1.722 1 0.013 x 10 5 Iron 5 26.0 480.02 2000 54 Mn 1.497 1 0.012 x 10 I I I I 26
TABLE V-5 FT. CALHOUN FLUX ATTENUATION MONITORS COMPARTMENT 2473 Counting Monitor Number of Weight Decay Interval Measured Activity at End Material Grooves (ma) (Days) (Seconds) Isotope Irradiation (Dpm/mg) 5 Iron 1 26.4 476.09 2000 54 Mn 1.524 1 0.012 x 10 5 Iron 2 26.5 476.15 2000 54 Mn 1.554 1 0.012 x 10 5 Iron 3 26.1 476.17 2000 54 Mn 1.519 1 0.012 x 10 5 Iron 4 26.4 476.19 2000 54 Mn 1.636 1 0.012 x 10 5 Iron 5 26.3 476.22 2000 54 Mn 1.384 1 0.011 x 10 TABLE V-6 FT. CALHOUM FLUX SPECTRUM MONITORS COMPARTMENT 2414 Counting Monitor Number of Weight Decay Interval Measured Activity at End Material Grooves (mg) (Days) (Seconds) Isotope Irradiation (Dpm/mg) Uranium 1 24.00 475.96 3000 137 Cs 5.90 1 0.05 x 10 4 Titanium 2 8.1 483.02 4000 46 Sc 5.51 10.70 x 10 5 Iron 3 26.6 472.98 2000 54 Mn 1.598 1 0.019 x 10 Uranium 4 (Shielded) 5 16.85 475.21 3000 137 Cs 1.76 1 0.04 x 10 Nickel 6 (Shielded) 6 23.6 474.94 2000 58 Co 3.303 1 0.050 x 10 Copper 3 (Shielded) 7 29.9 474.17 3000 60 Co 6.55 1 0.22 x 10 27
I TABLE V-7 FT. CALHOUN FLUX SPECTRUM MONITORS COMPARTMENT 2441 Counting l Monitor Number of Weight Decay Interval Measured Activity at End 5 Material Grooves (mg) (Days) (Seconds) Isotooe Irradiation (Dpm/mg) 4 Uranium 1 25.33 475.92 3000 137 Cs 4.94 1 0.05 x 10 4 Titanium 2 14.1 475.10 4000 46 Sc 5.38 1 0.50 x 10 5 Iron 3 26.6 474.23 2000 54 Mn 1.648 1 0.019 x 10 Uranium 4 E (Shielded) 5 7.33 475.26 3000 137 Cs 1.72 1 0.05 x 10 g Nickel 6 (Shielded) 6 21.9 474.97 2000 58 Co 3.364 1 0.053 x 10 Copper 3 (Shielded) 7 2 /. 3 474.14 3000 60 Co 6.65 1 0.23 x 10 TABLE V-8 FT. CALHOUN FLUX SPECTRUM MONITORS COMPARTMENT 2473 Counting I Monitor Number of Weight Decay Interval Measured Activity at End 5 Material Grooves (mg) (Days) (Seconds) Isotope Irradiation (Dom /mg) 4 Uranium 1 30.08 476.00 3000 137 Cs 3.73 1 0.04 x 10 4 Titanium 2 14.5 475.17 4000 46 Sc 5.47 1 0.53 x 10 5 Iron 3 26.1 474.26 2000 54 Mn 1.542 1 0.018 x 10 Uranium 4 (Shielded) 5 14.50 475.88 3000 137 Cs 1.45 1 0.04 x 10 l Nickel 6 (Shielded) 6 23.0 475.00 2000 58 Co 3.052 1 0.049 x 10 Copper 3 (Shielded) 7 25.0 474.06 3000 60 Co 6.27 1 0.23 x 10 I 28 g
- b. Threshold Detector Analysis The SAND-II(4) and ANISN( ) computer codes were used to calculate the fast flux and fluence at the surveillance capsule assembly location and at the reactor vessel.
The SAND-II(4) computer code is used to calculate a neutron flux spectrum from the measured activities of the flux monitors. SAND-II requires an initial flux spectrum estimate; this is calculated using ANISN.(5) The measured activities must be adjusted before they can be put into SAflD. The various steps of the procedure are described below. The measured activities were decay corrected to reactor shutdown. The foils irradiated and the shutdown activities are shown in Table V-9. Before being used by SAND, the foil activities must be converted to saturated activity with units of disinte-grations per second per target atom (dps/a). The following equation was used for the conversion:
=
M A 16.67 A sat 7; y 3 where A sat
= Saturated activity (dps/a) fi = fleasured activity at shutdown (dpm/mg)
A = Atomic weight N = Avogadro's number I = Isotopic abundance of target isotope S = Saturation factor, explained below 238 For U fission product activities, the required SAND input 238 has dimensions of fissions per second per U atom (fps /a). 29
I This is obtained by dividing A by the fractional fission sa yield of the fission product whose activity was measured. The saturation factor, S, converts the measured activity to I a saturated activity. The actual reactor operating history was used to calculate the saturation factor. The reactor was assumed to operate for several periods of constant power. Then, for each isotope, S was calculated. S=ghexp(-AT)[1-exp(-At)] j j 1 where Pi = Power of ith interval g Po = Full Power 5 A = isotope decay constant T$ = Time between end of ith operating period to reactor shutdown t$ = length of ith operating period The saturated activities are given in Table V-9. The uranium foil is shielded with cadmium to prevent thermal fissioning in any U-235 impurities. However, the cadmium cover does not prevent fast fissioning in U-235. Therefore, an unshielded uranium foil is included in the flux monitor set. The activity of the ur. shielded foil can be used to determine the amount of fissioning in the shielded uranium foil caused by U-235. As a result of this calculation, the U-238 fission rate was detemined to be 75% of the shielded uranium foil activity in Table V-9. SAND requires an initfa: estimate of the neutron flux spectrum. This initial estimate was calculated using ANISN, a one-E dimensional discrete ordinate code. The DLC-23 CASK, 22 5 I 20 g
TABLE V-9 FLUX M0ilITOR ACTIVITIES Monitor Material Measured Isotope 1 Uranium Cs 137 2 Titanium Sc 46 3 Iron Mn 54 4 Uranium (shielded) Cs 137 5 flickel (shielded) Co 58 6 Copper (shielded) Co 60 Shutdown Saturated Compartment Monitor Activity (dom /mo) Activity (dos /a)b 2414 1 5.90(+4)a 1.10(-13) 2 5. 51( +4 ) 1.08(-15) 3 1.598(+5) 6.02(-15) 4 1.76(+4) 3.29(-14) 5 3.303(+6) 9.08(-15) 6 6.55(+3) 6.36(-17) 2441 1 4.94(+4) 9.24(-14) 2 5.38(+4) 1.05(-15) 3 1.698(+5) 6.21(-15) 4 1.72(+4) 3.22(-14) 5 3.364(+6) 9.25(-15) 6 6.65(+3) 6.45(-17) 2473 1 3.73(+4) 6.97(-14) 2 5.47(+4) 1.07(-15) 3 1.542(+5) 5.81(-15) 4 1.45(+4) 2.71(-14) 5 3.052(+6) 8.39(-15) 6 6.27(+3) 6.09(-17)
- a. Denotes power of 10
- b. Uranium Foils are (fps /a) 31
I group neutron cross section library was used. The reactor geometry is shown in Figure V-3 (as-built dimensions). SAfl0 uses an iterative technique to calculate the neutron I flux spectrum. The activities of the set of flux monitors and an initial flux spectrum are the input required by SATID. Activities are calculated for each fail for the flux spectrum using the following equation e=I i a(Ej )D(E j)AE j I where a(Ej ) is reaction cross section at energy E j , barns. 0(Ej ) is the flux at E , n/cm j -s,mev AE j is width of energy band at E j, mev. The flux spectrum is adjusted by an iterative technique I until the calcul6ted and measured activities agree. The result of this is a 620 group neutron flux. The flux and fluence results are shown in Table V-10. From the ANISN case, the flux at the clad-vessel interface and at 1/4 of the vesse' thickness was determined to be 0.37 and 0.47 times the flux at the surveillance capsule. These factors were used to calculate the flux ad fluence at the vessel clad interface and 1/4 the thickness into the vessel. Taking factors from the one-dimensional ANISN is appropriate since the 225 surveillance assembly is at the azimuthal location of the maximum vessel flux. The fluences have been I extrapolated to end of cycle five and end of life. 5 Reference 3 states that the SAND code will give fluxes that are accurate to within + 10% to + 30% if the errors in the measured activities are within similar limits. From Tables I 32 3
FIGURE V-3 ANISil Geometry 1 2 3 4 5 6 7 8 Region Inside Region Name Material Radius (cm) 1 Core Homogenized core 100.0 2 Shroud Stainless steel 141.0 3 Coolant Water 142.6 4 Core Support Barrel Stainless steel 153.2 5 Coolant Water 157.0 6 Thermal Shield Stainless steel 161.3 7 Coolant Water 168.9 8 Vessel Carbon steel 179.7 Inlet Water temperature 528 F I 33
I TABLE V-10 FORT CALHOUN FAST NEUTRON FLUX AND FLUENCE VALUES Fast Flux (E>l.0 Mev) 2 location Maximum Flux (n/cm -s) Surveillance Capsule 6.3(+10)^ Vessel-clad interface 5.5(+10) 1/4 thickness of vessel 2.9(+10) Fast Fluences (E>1.0 MeV) d Location End of Cycle 3 D End of Cycle S c End of Life Surveillance Capsule 5.l(+18) 8.5(+18) 6.3(+19) Vessel-clad interface 4.5(+18) 7.4(+18) 5.5(+19) 1/4 thickness of vessel 2.4(+18) 3.9(+18) 2.9(+19)
- a. Denotes power of 10
- b. 2.6 effective full power years (EFPY)
- c. 4.3 EFPY (estimated)
- d. 32.0 EFPY I
I I I I I I 34
V-3 throu- 1-8, the 2-sigma uncertainties in tha measured activitias were less than +12%. Therefore, it is estimated that the uncertainty in the measured flux at the surveillance capsule location is about +20% to +30%. The extrapolated flux in the vessel will be slightly higher, so a reasonable value to use for the vessel uncertainty is +30%. The maximum flux in the Fort Calhoun reactor vessel is 5.5 x 2 10+10 n/cm -s, resulting in an end of life fluence of 5.5 x 2 10 +I9 n/cm . This assumes a 40 year life with an 80% capacity factor. The maximum fluence at the end of cycle five operation is 7.4 x 10 18 2 n/cm . The maximum fast flux at the 225 surveillance capsule assembly location is 6.3 x 10 2 10 n/cm -s. These values are a factor of 2.75 higher than the original design estimates (eg, 5.5 x 10 I9 n/cm2 end-of-life fluence versus the 9 sign estinate of 2.0 x 10 I9 n/cm2 ),
- c. Flux Attenuation Monitor Analysis Each of the three tensile monitor compartments contained a set of five iron flux attenuation monitors as shown in Figure V-4. Three iron wires (1 through 3) were located equidistant from the core; the remaining two wires (4 and 5) were located 0.162 inches either side of the radial midplane of the monitor block. The measured activity and calculated flux for each wire is given in Table V-ll. The neutron flux for each wire was calculated using the spectrum averaged cross section determined for the iron threshold detectors (previous section) using the following relationship:
o=f where o is the spectrum averaged cross section (b) A is the saturated activity (dps/ atom) 0 is the fast flux (n/cm -s E>lMev). 35
I I I
~ 0.638
- I
- 0.319 ~
0.162 ; I
; 2 0.162 e FE WIRE NUMBER i .
I
-G '"
3 0.162 0.394 1
- ~iii , ,
0.34 1 1.387 Y 0.162 RD
/ . >
I 0.025 x 45 CHAM TYP No. 50 (0.0700) DRILL 1150 MIN INCLUDED ANGLE 5 HOLES e re eLux HouSiNo I ALL DIMENSIONS IN INCHES FIGURE V 4 IRON FLUX WIRE HOUSING I I 36
TABLE V-Il IRON FLUX MONITORS Saturated Number of ) Activity Op (" 2 ) Compartment 3 (b) Grooves (dos /a) cm -s
-15 6.46+10(1) 2414 .095 1 6.138
-15 6.42+10 2 6.097
-15 6.52+10 3 6.198
-I 4 6.737 7.09+10
-15 6.07+10 5 5.769
-15 5.95+10 2441 .101 1 6.014
-15 2 6.123 6.06+10
-15 3 6.048 5.99+10
-15 4 6.488 6.42+10
-15 5 5.641 5.59+10
-15 +10 2473 .116 1 5.742 4.95
-15 2 5.855 5.05+10
-15 3 5.724 4.93+10
-15 4 6.164 5.31+10
-15 5 5.215 4.50+10 (1) Denotes power of ten (2) O p is flix greater than 1 Mev.
37
I The fast flux in the iron flux monitors is then just the saturated activity divided by the average cross section. The average cross section in each compartment is given in Table V-ll. In any compartment, the spread in activities of wires 1, 2, and 3 is about 2%. This fact supports an assumption made in analyzing the threshold monitors, that they all were exposed to about the same flux spectrum. The calculated fast flux for wires 1 through 3 is consistent I with the values obtained for the iron threshold detectors in each compartment. Fast flux values for the wires closer to the core (Number 4) were 6.6 to 9.6% greater than the flux at the midsection, and wires away from the core (number 5) were 6.2 to 9.6% less than the flux at the midsection. This indicated that the flux gradient through the test specimens was approximately 15%.
- d. Standard Reference Material (SRM) Analysis I
Charpy impact specimens from a standard reference material (Heavy Section Steel Technology Program, HSST Plate 01) were irradiated along with the reactor vessel surveillance materials in the 225 capsule. The SRM specimens were included t I augment the dosimetry analysis through correlation with B results from experimental data and other surveillance program data on the same material. The Charpy impact test results from the irradiated SRM specimens are given in Table V-12; the impact e1ergy data are plotted as a function of test temperature in Figure V-5. Also shown in Figure V-5 is an average curve for unirradiated material *, The radiation induced shift in the SRM transition temperature measured at the 30ft-lb energy level is 124 F.
- SRM specimens were not tested as part of the Fort Calhoun baseline evaluation.
The unirradiated transition curve in Figure V-5 was based on test results on the same material (HSST Plate 01) from other sources. (6-8) I 3e g
TABLE V-12 CHARPY V-NOTCH IMPACT RESULTS FOR FORT CALHOUN STANDARD REFERENCE MATERIAL Specimen Test Impact lateral Fracture Identification Temperature Energy Expansion Appearance Number ( F) (ft.lb) (mils) (% Shear) 55T 80 9 7 0 573 120 17 19 0 566 120 22 21 0 55M 160 38 24 20 55Y 160 38 33 20 56J 200 50 38 30 563 200 63 47 30 565 250 84 70 80 56A 250 75 50 70 560 300 92 75 90 575 350 102 80 100 55P 350 107 83 100 39
I Available irradiation data (6-16) for HSST plates 01 and 02 are shown in Figure V-6. The HSST data include specimens from the quarter thickness of the plates oriented in both the longitudinal (RW) and transverse (WR) direction. The data represent both experimental and power reactor irradiation exposures. Also shown in Figure V-6 is the neutron fluence (including the + 30% uncertainty band) obtained from analysis of the flux monitors nearest to the capsule compartment in which the SRM specimens were irradiated. The upper bound of the HSST plate irradiation data is seen to intersect the upper 18 2 bound fluence estimate (7.0 x 10 n/cm ) at the 124 F shift measured from the Fort Calhoun SRM data. The range of fluence inferred from the HSST plate trend band is 0.7 to 1.6 x 10 I9 n/cm for the 124 F SRM shift. The SRM irradiation data confirm that the Fort Calhoun 225 capsule exposure was well in exct.ss of the original target fluence of 1.8 x 10 18 n/cm2 . The information presented in 18 2 Figure V-6 confirms that the actual exposure was 5.1 x 10 n/cm or higher, thereby supporting the results from the neutron flux monitor analysis. A more explicit detemination of fluence from the SRM data is not practical because of possible differences in materials and radiation environment (dose rate and flux spectrum) between the Fort Calhoun reactor vessel and the data used to evaluate the SRM data in Figure V-6. I I I
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B. Strength and Toughness Properties
- 1. Tension Tests Tension tests were conducted in accordance with applicable ASTM standards and C-E laboratory procedures. The test method and equipment are described in Appendix A.
The three irradiated specimens from each material (base metal, weld metal and heat-affected zone) were tested at room temperature, 250 F and 550F. The tensile p,operties are listed in Table V-13, and the stress-strain curves are shown in Figure V-7 through V-15. The pre-irradiation tensile properties (2) are summarized in Table V-14 (each value average of three tests). Photographs of the fracture surface of the broken irradiated specimens are shown in Figure V-16.
- 2. Charpy V-flotch Imoact Tests Charpy V-notch impact tests were conducted in accordance with applicable ASTM standards and CE laboratory procedures.
The test method and equipment are described in Appendix B. Twelve irradiated specimens from each material (base metal, weld metal and heat-affected zone) were tested at a series of terrperatures to establish the transition temperature behavior. The impact data (impact energy, lateral expansion and fracture appearance as a function of test temperature) are shown in Tables V-15 through V-17 and Figures V-17 through V-25. (Also shown in each of the figures is the unirradiated transition temperature curve from the baseline evaluation.(2)) Fracture surface photographs of the broken irradiated specimens are shown in Figures V-26 through V-28. Each impact test was instrumented. Additional data related to instrumented impact testing are presented in Appendix C. 43
TABLE V-13 POST-IRRADIATION TENSION TEST PROPERTIES FORT Call 100N SURVEILLANCE MATERIALS Yield Ultimate Elongation Test Strength Tensile Fracture Fracture Fractu Reduction (1-inch 9 age) Specimen Temp. Upper / Lower Strength Load Strength (a) Stress {g) of Area TE/UE Material Code ( F) (ksi) (ksi) (lb) _(ksi) (ksi) (%) (%) . Base Metal 1D1 72 79.0/78.2 100.4 3030 61.7 194 68.2 27/10.3 lEL 250 72.7/72.1 91.2 3080 62.7 182 65.5 24/(c) 100 550 65.2/62.7 89.3 3030 61.7 169 63.5 (c)/6.7 Weld Metal 3ET 72 105.7/101.0 114.6 3630 73.5 190 61.7 22/8.8 g 3E5 250 93.5/92.3 104.4 3010 60.9 160 61.9 23/8.5 3EC 550 91.7/86.2 103.2 3870 78.3 171 54.2 20/(c) HAZ 4D2 72 78.3/74.4 96.7 3060 61.9 194 68.0 21/5.9 4EK 250 72.9/70.6 88.8 3670 74.3 149 50.2 13/5.1 4EC 550 61.9/59.5 86.5 3000 60.7 164 62.9 18/5.5 a - Fracture strength is the fracture load divided by initial cross sectional area b - Fracture stress is the fracture load divided by final cross sectional area c - Not determined M M W W M M M M M W M M M M M
TABLE V-14 PRE-IRRADIATI0ft TEt4SI0tl TEST PROPERTIES FORT CALH0uti SURVEILLAt1CE MATERIALS Yield Ultimate Elongation Test Strength Tensile Fracture Fracture Fractu Reduction (1-inch gage) Temp. Upper / Lower Strength Load Strength (a) Stress {g) of Area TE/UE Material ( F) (ksi) (ksi) (lb) (ksi) (ksi) (%) (%) Base Metal 71 70.8/66.8 89.2 2740 55.9 187 70.1 28/10.9 250 64.5/62.7 83.0 2580 52.7 190 72.1 26/9.9 550 56.9/55.1 85.6 2780 56.7 178 68.0 24/10.3 g Weld Metal 71 78.0/74.1 90.2 2760 56.3 188 70.1 28/10.2 250 72.1/69.0 83.1 2600 53.1 177 70.1 24/8.4 550 65.9/64.1 85.2 2980 60.8 163 62.6 22/9.2 IIAZ 71 66.7/63.0 85.1 2840 58.0 168 65.5 24/10.0 250 60.0/58.5 79.8 2780 56.7 155 63.3 20/8.1 550 53.6/52.3 81.8 2980 60.4 160 62.4 21/8.7 a - Fracture strength is the Fracture Load divided by initial cross sectional area. b - Fracture stress is the Fracture Load divided by final cross sectional area.
I TABLE V-15 CHARPY V-NOTCH IMPACT RESULTS g FOR FORT CALH0'JN IRRADIATED BASE METAL (LONGITUDINAL) 5 (PLATE D-4802-2) Specimen Test Impact lateral Fracture Identification Temperature Energy Expansion Appearance E Number ( F) (Ft-lbs.) (mils) ("4 Shear) E 14M 40 14.0 16.0 10 167 40 15.5 17.0 10 117 75 25.0 24.0 20 12A 80 33.0 32.0 10 g 112 120 62.0 53.0 40 138 120 64.0 47.0 40 13C 160 80.0 67.0 60 14T 160 83.0 70.0 70 g 13D 200 92.0 70.0 80 y 144 200 119.0 84.0 100 15C 250 122.0 86.0 100 150 250 126.0 82.0 100 I I I I I I I I 46 E
2 TABLE V-16 CHARPY V-fl0TCH IMPACT RESULTS FOR FORT CALH0 Vfl IRRADIATED WELD METAL 1 Specimen Test Impact lateral Fracture Identification Temperature Energy Expansion Appearance Number ( F) (Ft-lbs.) (mils) (% Shear) 352 80 15.0 13.0 0 33T 120 20.0 21.0 0 31B 120 23.0 22.0 0 32E 160 19.0 20.0 10 31C 160 29.0 29.0 30 336 200 34.0 33.0 40 32U 250 41.0 40.0 60 341 250 46.0 44.0 70 312 300 62.0 53.0 90 32J 350 64.0 51.0 100 317 350 65.0 50.0 100 31E 400 66.0 62.0 100 47
I TABLE V-17 CHARPY V-fl0TCH Il1 PACT RESULTS FOR FORT CALH0Uti IRRADIATED HAZ METAL (BASE METAL PLATE D-4802-2) Specimen Test Impact lateral Frac ture Identification Temperature Energy Expansion Appearance Number ( F) (Ft-lbs.) (mils) (% Shear) 45L 0 17.0 16.0 0 45U 40 27.0 25.0 20 420 40 50.0 50.0 30 464 80 40.0 36.0 20 42Y 80 53.0 46.0 30 452 120 40.0 42.0 30 43P 120 90.0 73.0 40 g 463 160 55.0 53.0 30 y 41B 200 72.0 57.0 80 42B 250 93.0 79.0 100 41Y 300 88.0 62.0 100 446 300 112.0 79.0 100 I I I I I I I 48 i
R l l l l l l l l 1 I l I I 100,000 - b 80,000 _ sm 60,000 f - 40,000i_ -
%I CI 0 l l l l l l l l l l l l l 0 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STR AIN, IN/IN FIGURE V 7: STRESS STR AIN RECORD OF TENSILE TEST, BASE METAL PLATE D-4802-2, SPECIMEN No. IDI, TEST TEMPERATUR E 720F 120,000 l t l l l l l l l l l l l l 100,000 EXTENSOMETER ARMS SLIPPED -
80,000 - 's - E ss M, ~ s m N 60,000. 5 e 40,000!_ - i
~
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O l l l l l l l l I O 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STR AIN, IN/IN FIGURE V-8: STRESS STRAIN RECORD OF TENSILE TEST BASE METAL PLATE D-4802-2 SPECIMEN No. IEL, TEST TEMPERATURE i 2500F 19
120,000 j i ; j i i j j i j l j l 100,000 _ i E ~. M 80,000 I 60,000 - 40,000 _
~ ~
0 l l l l l l l l l l 0 0.04 0.08 0,12 0.16 0.20 0.24 0.28 STR AIN, IN/IN FIGURE V-9: STRESS STRAIN RECORD OF TEi4SILE TEST, BASE METAL PLATE D-4802 2, SPECIMEN No.1DU, TEST TEMPERATURE 5500F
' i 1 l l 1 l l I l I i 100,000 -
@ 80,000 . N $ 60,000 _ I 40,000 I I
= =
O l l l l l l l l l l g 0 0.04 0.08 0.12 0.16 0.20 0.24 0.28 3 STR AIN, IN/IN FIGURE V-10: STRESS STRAIN RECORD OF TENSILE TEST, WELD METAL PLATE D-4802-1,3, SPECIMEN No. 3ET, TEST TEMPERATURE 720F ec g
120.000 g g , ; ; , ;
; ; l 100,000 - -
Ui ! c. y 80,000 ._ _ w E e 60,000 - 40,000 _ - 0 I I i i I i l i l l 1 0 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STR AIN, IN/IN FIGURE V 11: STRESS STR AIN RECORD OF TENSILE TEST, WELD METAL PLATE D-48021,3, SPECIMEN No. 3ES, TEST TEMPERATURE 250 F 120,000 t i I i 1 i i i l l I 1 I i EXTENSOMETER ARMS SLIPPED 100,000 -
's s s -
%'s N E 80,000 _.
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, m' W
c: 60,000 . - 40,000 - - O l i I i i I i l I i 0 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STR AIN, IN/IN FIGURE V-12: STRESS STRAIN RECORD OF TENSILE TEST, WELD METAL PLATE D 48021,3, SPECIMEN No. 3EC, TEST TEMPERATURE 550 F L1
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VI. DATA ANALYSIS The radiation induced changes in toughness of the Fort Calhoun , surveillance materials are summarized in Table VI-1. Index temperature shifts (AT) were measured using the average curves at the 30 ft-lb level (T30), 50 ft-lb level (T50), and 35 mils lateral expansion level (T35). Upper shelf energy changes were based on the minimum impact energy corresponding to 100% shear fracture measured before and after irradiation. The unirradiated impact data were obtained from the baseline evaluation.( ) The weld metal exhibited the highest transition temperature shift (238 F) and decrease in upper shelf energy (34%) of all the surveillance materials consistent with its having the highest residual copper content (Table III-4). The measured shift is consistent with the prediction based on Regulatory Guide 1.99(I7) at the measured fluence of 5.1 X 10 18 n/cm2 . The measured shelf energy decrease, 34%, is also closely predicted by Regulatory Guide 1.99 (38% predicted). The upper shelf energy of the irradiated surveillance weldment is 64 ft-lb. The impact property changes for the base metal and HAZ were significantly less than for the weld metal. The transition temperature shift and decrease in upper shelf energy for the base metal (70 F and 13%, respectively) are consistent with the predictions based on Regulatory Guide 1.99 (46 F and 165, respectively). The measured values of the irradiated base metal indicate a high level of toughness (eg,119 ft-lb upper shelf energy) for the limiting reactor vessel beltline plate. Analysis of the weld heat-affected zone impact property changes is complicated by the excessive data scatter as evidenced in Figures V-23 through V-25. The data scatter is most likely a 67
I result of test specimen notch placement. (For the Fort Calhoun HAZ Charpy impact specimens, the notch was centered as closely as g possible to the weld fusion line.(I) ASTM E 185-73, " Surveillance 5 Tests for Nuclear Reactor Vessels," currently specifies notch root placement 1/32 inch from the weld fusion line.) Impact properties vary widely adjacent to the fusion line*. For the irradiated specimens, data scatter was further enhanced by the extent of weld metal dilution which created a gradient in copper content (high copper in the weld to low copper in the base metal). Specimens with their notch root nearest the weld fusion line would therefore exhibit lower toughness at a given test . temperature than specimens with their notch root away from the fusion line on the base metal side. The resultant transition temperature shift for the HAZ was, therefore, intemediate between that of the base metal and weld. (Note that the instrumented Charpy impact results, as shown in Figure C-3 of Appendix C, display comparatively little data scatter. The measured shift in the brittle transition temperature, TB , of 120 F is in cicse agreement with the 104F shift obtained from the standard Charpy data for the HAZ.) The upper shelf energy after irradiation, 88 g ft-lb, was maintained at the pre-irradiation level, thus ensuring 5 a high level of toughness in the weld heat-effected zone. I Post-irradiation tensile property measurements typically reflect an increase in strength and a decrease in ductility. The tensile property changes for Fort Ct 'oun are consistent with this behavior, as indicated by a comparison between Tables V-13 and V-14. For a given material, the magnitude of the strength increase and ductility decrease was generally the same over the range of test temperatures (70F to 550F). The weld metal exhibited the largest increase in yield and ultimate tensile strength (35% and ?S%, respectively), g n I
- Notch plactment 1/32 inch from the fusion line tenas to minimize data scatter in HAZ specimens.
68 I
consistent with the highest copper content occurring in the weld. In contrast, base metal and HAZ strength property changes were 50 to 60?' less than for the weld metal. The resultant post-irradiation room temperature yield strengths ranged fron 78,000 psi for the base metal and HAZ to 106,000 psi for the weld. Reductions in ductility of the irradiated surveillance materials were relatively small; the total elongation af ter irradiation was typically 20% or better. The radiation exposure for the surveillance materials from the 225 capsule was determined to be 5.1 X 10 18 n/cm based on analysis of.the neutron threshold detectors. This contrasts with the original target fluence of 1.8 X 10 18 n/cm for the 2.63 EFPY (effective full power year) exposure of the capsule. The fluence measurements were corroborated by the analysis of the standard reference material data (Figure V-6), and by the agreement between Regulatory Guide 1.99(l7) predictions and measured shif ts and fluence for the base metal and weld metal. The measured fluence for the 225 capsule, 5.1 X 10 18 2 n/cm , was therefore employed in subsequent analyses. Figure VI-l was developed to provide a means of predicting trans-ition temperature shift of the controlling beltline material to adjust the Fort Calhoun reactor coolant system pressure-temperature
, operating limit curves. The figure is based on the upper bound shift curve from Regulatory Guide 1.99. Revision 1(17) using the measured value of transition temperature shift for the weld metal and the calculated neutron fluence. The shift in the 30 ft-lb index temperature for the weld metal was used since it provided the most accurate measure of the transition temperature shift.
This approach has been used previously( 0} when the proximity of the irradiated 50 ft-lb level to the upper shelf energy (64 ft-lb) causes the 50 ft-lb shift to be exaggerated. The accuracy of the 30 ft-lb shift value (238F) is also supported by its conse"vaH , relative to the shift in the 35 mils lateral expansion index 69
I temperature (229F) and the brittle transition temperature (187F) from the instrumented Charpy impact results (Appendix C). Predicted transition temperature shift in Figure VI-l is presented as a function of the maximum neutron fluence at the inside surface of the reactor vessel. Both the shift at the vessel inside surface (Vessel ID) and at the quarter thickness (1/4t) position in the vessel wall can be determined from this figure. The 1/4t curve was derived by reducing the vessel ID fluence by a factor of 47" to account for the attenuation of neutrons through the thickness of the vessel wall. The predicted shift for the weld metal at end-of-life is 350F at I9 2 the 1/4t location based on a surface fluence of 5.5 X 10 n/cm and 32 EFPY. The correspondin end-of-life shelf energy predicted using Regulatory Guide 1.99(17 is in excess of 50 ft-lb. These predicted values of transition temperature shift and shelf energy decrease will be reviewed as subsequent post-irradiation evaluations of Fort Calhoun surveillance capsules become available. I I I I I I I I 70 I
TABLE VI-l CALHOUN
SUMMARY
OF TOUGHNESS MATERIALSPROPERTY CllAtlG n/cm Q FOR FORT 2 SURVEILLANCE (550F, 5.1x10 IRRADIATION) Upper Shelf Shelf Energy T30( F) AT30( F) T50( F) AT50( F) T35( F) AT35( F) Energy (ft-lb) Change (%) Material Base Metal 22I ^) Sl I8) 34(*) 137.5(#) (longitudinal) 82 60 120 69 94 60 119 13 Weld Metal -28(*) 4(#) -15 I8) 97.5 IO) 210 238 262 258 214 229 64 34 IIAZ -76(a) -28(d) -51I ^) 82I ") 28 104 89 117 63 114 88 0 SRM 27(^) 54I ^) Sl fd) 128 f# (longitudinal) 151 124 190 136 181 130 102 21 a - unirradiated values 5
500
; j l t 1 400 FORT CALHOUN _
300 SURVEILLANCE WELD 200 VESSEL ID g h 1/4t m L e 100 90 80 70 60 . 50 I I I I l l 1018 2 4 6 8 10 19 2 4 6 NEUTRON FLUENCE, N/CM2 (E > 1 MEV) FIGURE VI-1 PREDICTED NDTT SHIFT FOR THE FORT CALHOUN REACTOR VESSEL BE.LTLINE M M M M M M M M M M M M M M M M M m M
VII REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITATION CURVES The pressure-tempr rature (P/T) limitation curves for the Fort Calhoun reactor coolant system (RCS) are currently described in Figures 2.lA & B and 2.2A & B of the Technical Specifications, kendment No. 22 (April 1, 1977, AEC Docket No. 50-285, Operating License No. DPR-40). Those P/T curves were established for two to five years operation, and were based on a boltup temperature of 82 F, a lowest service temperature of 162 F, and an NDTT shift of 95 F. The boltup and lowest service temperatures remain the same throughout the plant life because the components related to these temperatures are not subject to fast neutron flux and resultant NDTT shifts. The transition temperature (NDTT) shift will vary during plant life to reflect radiation induced property changes of the reactor vessel beltline materials. The 95 F NDTT shift was based on an estimated neutron fluence of 2.7 x 10 18 n/cm2 after five years of reactor operation and the NDTT shift design curve given in the Technical Specifications. Analysis of the neutron flux monitors from the 225 capsule indicates that the actual neutron fluence at the end of five 18 2 years
- will be 7.4 x 10 n/cm on the inside surface of the 18 reactor vessel (3.9 x 10 n/cm at the quarter thickness in the vessel) as shown in Table V-10. Furthermore, the measured NDTT shift for the surveillance weld metal was found to exceed the design curve prediction, requiring the implementation of Regulatory Guide 1.99(17) for determining predicted values of NDTT shift (Figure VI-1). Therefore, the pressure-temperature curves for 4.5 to 5 years (3.8 to 4.3 EFPY) operation were developed to reflect the post-irradiation surveillance test results. Revision of the P/T curves was performed in accordance with the guidelines of the Fort Calhoun Technical Specifications, Amendment No. 22.
Five years operation corresponds to the end of Cycle 5, January 1980, which is equivalent to 4.3 effective full power years. 73
I The NDTT shift values were detennined for the weld metal (the controlling beltline material) using the extrapolation method described in Regulatory Guide 1.99( "). The estimated maximum neutron fluence at the inside surface of the reactor vessel af ter five years operation (approximately 4.3 effective full power years) is 7.4 x 10 l0 n/cm2 . This corresponds to an NDTT shift (from Figure VI-1) at the quarter thickness location of 209 F. In accordance with the Technical Specifications, the limit lines (Figures 2-1A & B and 2-2A & B) were shifted parallel to the horizontal temperature axis in the direction of increasing temperature a distance equivalent to the increase in NDTT shift, or 114 F (209 F less 95 F). The boltup and lowest service temperatures remain the same as previously described. The resultant heatup and cooldown P/T limit curves for non-critical operation of the RCS are shown in Figures VII-l and VII-2; these curves should replace Figures 2-1A and 2-1B in g the current Techni mi Specifications. The heatup and cooldown 3 P/T curves for RCS operation with the reactor critical are shown in Figures VII-3 and VII-4; these curves should replace Figures 2-2A and 2-2B in the current Technical Specifications. Based on an assessment of the capability of the Fort Calhoun RCS to accomodate the 4.5 to 5 year P/T curves, no aspects of the system design were identified which would preclude full-power operation within the constraints of the B P/T curves. During plant heat-up and cooldown, however, the 5 following revisions to the Technical Specifications should be made to assure adequate low temperature overpressure protection: Pa ragraph Page 2.1.6(4) 2-15 Both PORVs should be operable at the low setpoint whenever the cold leg temperature is less than 348 F. 74 I I
2.3(3) 2-22 Whenever the RCS cold leg temperature is below 275 F...at least two (2) HPSI pump control switches shall be placed in pull-stop. 2.3(3) 2-22 Whenever the RCS cold leg temperature is below 230 F...all three (3) HPSI pump control switches shall be placed in pull-stop. In addition, RCS cooldown should be restricted to a maximum decrease of 20 F (Tc ) in any one hour period whenever the cold leg temperature is less than 215 F. (Note that the temperatures listed above do not include instrument error). The RCS P/T curves for 5 to 10 years operation were developed as described for the 4.5 to 5 year curves. The estimated ximum neutron fluence at the inside surface of the reactor ssel after ten years operation (approximately 8.3 effective full power years) is 1.4 x 10 l9 n/cm2 . This corresponds to an NDTT shift (from Figure VI-1) at the quarter thickness location of 269 F. The limit lines of Figures VII-l through VII-4 were shifted parallel to the horizontal temperature axis a distance equivalent to the increase in NDTT shift, or 60 F (269 F less 209 F). The boltup and lowest service temperature remain the same as previously described. The resultant heatup and cooldown P/T limit curves for non-critical operation of the RCS out to 10 years are shown in Figures VII-5 and VII-6. The heatup and cooldown P/T limit curves for RCS operation out to 10 years with the core critical are shown in Figures VII-7 and VII-8. 75
I I 3000 I NON-OPERATIONAL SYSTEM LEAK g CHECK AND HYDROSTATIC TEST g 2500 I f $ 2000-E NON CRITICAL HEATUP N = a I E s N 1500
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I VIII REFERENCES
- 1. "Recormtended Program for Irradiation Surveillance of the Fort Calhoun Reactor Vessel Materials," Combustion Engineering, Inc. , February 25, 1969, transmitted by letter CE-750-10ll, March 26,1969.
- 2. " Omaha Public Power District, Fort Calhoun Station Unit No.
1, Evaluation of Baseline Specimens, Reactor Vessel Materials Irradiation Surveillance Program," TR-0-MCD-001, March 11, 1977.
- 3. A. G. Schoenbrunn, "NRC Questionaire on Reactor Vessel Materials," Combustion Engineering Letter CE-18074-357, August 18, 1977.
E
- 4. SAND Users Manual, AFWL-TR 67-41, September 1967.
- 5. ANISN Users Manual, K-1693, March 1967.
- 6. W. J. Stelzman and R. G. Bergren, " Radiation Strengthening and Embrittlement ir, Heavy Section Steel Plates and Welds,"
0RNL-4871, June 1973.
- 7. J. R. Hawthorne, " Post-Irradiation Dynamic Tear and Charpy-V Performance of 12-in. Thick A533B-1 Steel Plates and Weld Metal, " Nuclear Engineering and Design, 17 (1971), pp. 116-130.
- 8. R. A. Wullaert and J. W. Sheckherd, " Evaluation of the First Maine Yankee Accelerated Surveillance Capsule, Dynatup g Technical Report CR 75-317, Effects, Technology, Inc. , Santa 5 Barbara, California, August 15, 1975.
I 3 e4
- 9. A. L. Lowe, et. al., " Analysis of Capsule OCI-F from Duke Power Co., Oconee Unit 1 Reactor Vessel Material Surveillance Program, "8AW-1421, Babcock & Wilcox, Lynchburg, Va., August 1975. ,
- 10. J. S. Perrin, et. al., " Point Beach Nuclear Plant Unit No. 2 Pressure Vessel Surveillance Program Evaluation of Capsule ,
V," June 10, 1975.
- 11. J. S. Perrin, et. al., "Surry Unit No. 1 Pressure Vessel Irradiation Capsule Program: Examination and Analysis of Capsule T," June 24, 1975.
- 12. E. B. Norris, " Reactor Vessel Material Surveillance Program for Turkey Point Unit No. 4, Analysis of Capsule T," SWRI Project No. 02-4221, June 14, 1976.
- 13. A. L. Lowe, et. al, " Analysis of Capsule OCI-E, Duke Power Co. Oconee Nuclear Station Unit 1 - Reactor Vessel Materials Surveillance Program," BAW-1436, September 1977.
- 14. A. L. Lowe, et. al, " Analysis of Capsule OCII-6, Duke Power Co, Oconee Nuclear Station Unit 2 - Reactor Vessel Materials Surveillance Program," BAW-1437, May 1977.
- 15. A. L. Lowe, et. al., " Analysis of Capsule TMI-lE, Metropoli* n -
Edison Co., Three Mile Island Nuclear Station Unit 1 Reactor Vessel Materials Surveillance Program, " BAW-1439, January 1977.
- 16. A. L. Lowe, et. as, " Analysis of Capsule ANI-E, Arkansas Power & Light Co., Arkansas Nuclear Unit 1 - Reactor Vessel Materials Surveillance Program, "BAW-1440, April 1977.
85
I
- 17. Regulai vry Guide 1.99, Revision 1, " Effects of Residual Elements on P.'edicted Radiation Damage to Reactor Vessel Materials," April 1977.
'8. S. E. Yanichko and S. L. Anderson, " Analysis of Capsule S from the Wisconsin Electric Power Company and Wisconsin Michigan Power Company Point Beach fluclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-8739, November 1976.
I I I I I p" 86
APPENDIX A TEtlSILE TESTS - DESCRIPTI0fl AND EQUIPMErlT The tensile tests were performed using a Riehle universal screw testing machine with a maximum capacity of 30,000 lb and separate scale ranges between 50 lb and 30,000 lb. The machine s shown in Figure A-1, is capable of constant cross head rate or constant strain rate operation. The tensile testing was covered by the certificate of calibration which is included at the end of the Appendix A. Elevated temperature tests were performed in a 2-1/2" ID x 18" 1ong high temperature tensile testing furnace with a temperature limit of 1800F. A Riehle high temperature, dual range extensometer was used for monitoring specimen elongation. The tensile specimen is depicted in Figure A-2. Figures A-3 through A-5 are isometric drawings showing the orientation and location of the tensile specimens in the base metal, weld metal and heat-affected-zone, respectively. Tensile testing was conducted in accordance with ASTM Method E 8-77, " Tension Tests of Metallic Materials" and/or Recommended Practice E 21-70, "Short-Time Elevated Temperature Tension Tests of Materials," except as modified by Section 6.1 of Recommended Practice E 184-62, " Effects of High-Energy Radiation on the Mechanical Properties of Metallic Materials." Implementation of the ASTM Test Methods to the testing of irra i ated tensile specimens is described in C-E Laboratory Procedure 00000-MCM-041, Revision C, " Procedure for Tension Testing of Irradiated Metallic Materials," August 16, 1978. A-1
I
. . , . ,n ..
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f !g[- ,, Lt _ g i .} .. I [ *. 3 I I Es I I FIGURE A-1 TENSILE TEST SYSTEM WITH CONTROL CONSOLE AND ELEVATED TEMPERATURE TESTING EQUIPMENT I A-2 ppq @!@l
/ , o 1.000' l i
/
h - 14 NC 2 Thread X 0.250'Dia 0.43 8 3.00' FIGURE A-2 TYPICAL TENSILE SPECIMEN A-3
I I I
= 5" g
^
f b h ht
~~
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.2 7- /
g Transverse
' t Longitudinal M- - -- g
/ __ __
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./ y v
/
l 1 / lj / Principal Rolling Direction + l_ _ _ _ _ _ _ _ _ 3
; ll t= Plate Thickness g I
-- r L)'
[ 3 ) _e I FIGURE A-3 LOCATION OF TENSILE SPECIMENS WITHIN BASE METAL TEST MATERIAL I A-4 I
_ Weld Metal 5n a ID Side-x
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, / Vdb
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=
hh? FIGURE A-4 LOCATION OF TENSILE SPECIMENS WITHIN WELD METAL TEST MATERIAL A-5
I i 5" l jWeld Metal g ID Side s -
, /,
/ l HAZ x '
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/
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+- P '8ti VM (J lI
$O00 h Wilson Instrument Division giciite - a u, u,r.ra.c ocu r x.,t r.ot. umootruo u mac ncu r u- . m, m .ii ' " " * ' ~ l A<r4 r.c n Ch on i C.. Lie C.mipany. Inc h Mun.c vr c,1 the bat,coca a micos. Lim ied Group of Companies Cer:ifica:e of Calibration Cahbration Date October 18,1978 Machine Description Richle DS-30 Customer Combustion Engineering Serial No. RA-44372 Prospect Hill Road Windsor, Ct. Attn: Dick Vieleux Wilson Instrurnent Disisicn of Acco certifies that the machine described above has been cahorated to ASTM designation E4 using cahorated weights and/or proving rings calibrated to National Bureau of Standards Specification. TENSION Machine Range 3,000 Machine Range 30,000 __ Machint r ea :in g
. Emor Macrane reaa.ng . Errer 600 .054 6000 +.123 1200 +.275 1200.0 + .154._
1800 +.129 18000 + _.062 2400 +.099 24000 . 05 2_.] 3000 +.235 30000 .050ii N.achine Range 6,000 Machine Range Machine reaa. rig . Error Machine reading % Error l 120.0 + .195._ l 2400 +tl46 , I 3600 t+.065 4800 +.074 6000 +.079 Machirie Range 15,000 Machine Range
*;. Error MJchine rea. ling % Error Michene reading 3000 +.490 6000_. +_245.
9000 +.164 12000.__.__ _f .123_ lji O O O . +.074 Cahbrahng apparatus used
- 0* y 1 C..eaui, serial nu Cat. date Lab. no. C htration trvacer
~
2,000 10,000 1809 4354 3-9-77 11-2-77 SJT.01/10lg't SJT.01/101206 -'
,,f d.'? ,[pn o 4 .
S'#""""""""*" 30,000 4987 7-6-78 Series 10 3T&C
98-04 010 1000 gj j wiison instrument civision h's C0f.f 4LCilCU T A /LiiUL, tilti gr,Li'fA4 r, CGt4f 4EC ilCU T UI42 WJ) 4 2',11 rnente-T..wn.: u . nu,. s l A'/L Hf_.Af a fJit lf 4 4 r, Asstt e;Gf/I'A ri f, if 4C I Certificate of Calibration I Calibration Date October 18,1978 Instrument Description Riehle Extensometer Riehle Recorder Cr .omer Combustion Engineering Serial No. 1977 Prospect Hill Road R-67338 Windsor, Ct. I Wilson Instrument Division of Acco verifies that the attached graph is certification of calibration of the instru- I ment described above. This instrument was calibrated to ASTM designation EB3. 5 I I I I Recorder Extensometer Calibrator Equipment used in calibration 528864 e, - j* -- ,4 ! e .. I 9, . . ,19 ,s- w Catioration Engineer 5 ?: .$W j/ g- __c> ~ stancarcs Man 9er I A-8
APPENDIX B CHARPY IMPACT TESTS - DESCRIPTION AND EQ'IPMENT J The standard impact tests and instrumented tests were performed on a calibrated instrumented impact testing system, shown in Figure B-1. C-E's instrumented impact test equipment provides for signal retention ar.d the subsequent data analysis. The output signal from the instrumented tup is recorded by an oscilloscope. A permanent visual record was made of the load signal, as it was displayed on the oscilloscope screen, with a polaroid camera. The system consists of the following elements:
- a. A Model SI-l BLH Sonntag Universal Impact Machine with a specific-ally machined pendulum tup, instrumented with four resistance strain gages in full bridge circuit. This tup " load cell" is calibrated statically and dynamically to provide a given pounds / volt sensitivity for known settings of the balance and gain on the dynamic response system. The instrumented machine meets all impact test machine requirements of ASTM and is certified by AMMRC, the U.S. Army Materials and Mechanics Research Center (Watertown Arsenal). A copy of the certification papers is included in this Appendix.
- b. A Model 500 Dynatup dynamic response system which supplies regulated and constant dc excitation to strain gages on the pendulum tup, provides balancing, variable load sensitivity and calibration functions, and amplifies load-time signal to a +10 volt, +100 milliampere level while preserving kHz frequency response and 0.05 percent accuracy while simultaneously recording the area beneath the load-time trace.
B-1
I
- c. A photoelectric triggering device and velocometer compond of a high intensity light directed through a grid mounteu on the pendulum of the impact tester, and pcssed to a photosensor through fiber optics. A special circuit ensures accurate, reliable and fail safe triggering of the oscilloscope recorder plus an accurate display of the average velocity of the pendulum during impact.
- d. A 5103ft Dual Beam Tektronix Storage Oscilloscope with a flo. SA18tl dual-trace amplifier plug-in unit and a No. SB12N dual time base plug-in unit. Alse included is a C-58 camera with mounting adapter. This device gives a display of each test trace for visual analysis of the load-time impluse recorded by the instrument.
The stan ard Charpy specimen is described in Figure B-2. Figures B-3 through B-5 are isometric drawings showing the orientation and location of the Charpy impact specimens in the base metal, weld metal and heat-affected-zone, respectively. All ' harpy impact tests were conducted in accordance with ASTf1 Method E 23-72, " Notched Bar Impact Testing of Metallic Materials." Implementation of ASTM E23 for the testing of irradiated Charpy specimens is described in C-E Laboratory Procedure 00000-MCM-040, Revision 0, " Procedure for Instrumented Charpy Impact Testing of Irradiated Metallic Materials," July 31, 1978. The constant temperature necessary for conducting the Charpy impact tests was obtained from a series of circulating liquid baths capable of maintaining stable temperature throughout the range of -150 F to
+250 F. Any selected temperature in this range was maintained to an accuracy of 2 F. For tests above 250 F, specimens were heated in a controlled circulation furnace where temperature was maintained to an accuracy of 5 F. The temperature baths were composed of the following equipment:
I B-2 I
Two fleslab Constant Temperature Circuiating Baths - M(del TEZ 10, v:ith Model CT 150 Thermoregulators and Labline 11 inch diameter thermocups, Designated Baths 1 and 4. Medium: Ethyler.a Glycol - room temperature to 250 F. One fieslab Constant Temperature Circulating Bath - Model TEZ 10 with a Model CT 59 Thermoregulator and a Labline 11 inch diameter thermo cup. Designated Bath 2. Medium: Isopropanol - room temperature to -10 F. tieslab Portable Bath Cooler, Model PCB-2 connected. One Low Temperature Stirred Bath, one 11 inch diameter thermo cup, one Honeywell Controller and Solenoid control valves to Flexi-Cool cooling system. Designated Bath 5. Medium: Isopropanol - room temperature to -150 F. Coolant: Freon One Grieve Industrial oven, controlled air circulation. Designated Bath 3. 2 Medium: Air,100 F to 800 F. i All baths - Copper Constantan Thermocouple Honeywell Six Point Temperature Chart Recorder Digitec Thermocouple Thermometer - Model 590 TF Standard Mercury Column Thermometer Bimetallic - spring Thermometer The temperature instruments were calibrat;d in accordance with the ASME Boiler and Pressure Vessel Code, Section III, Paragraph 2360. Copies of the applicable calibration certificates are provided at the end of this appendix. B-3
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sg, 4 g> M %l <u. w " - u ~ c .c a m w a--as. ,2 rd73 g.p} I I FIGURE S-1 CHARPY IMPACT TEST SYSTEM, ASSOCIATED CONSTANT TEMPERATURE BATHS AND INSTRUMENTED CHARPY IMPACT DATA PROCESSING EQUIPMENT
'il ll l B-4 I
0 45 15" r IN 0.010' us O.394" 2.165" 0.394"s FIGURE B-2 TYPICAL CHARPY V-NOTCH IMPACT SPECIMEN B-5
I I I 5" I on I J it _- -- A l a
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I FIGURE B-3 LOCATION OF CHARPY IMPACT SPECIMENS WITHIN I BASE METAL TEST MATERIAL I I B-6
" 5" = W
/ eld Metal T eld Side VI F
_J 11/4"
/ ,
u i
/ ,
/ ,-
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I t = Plate Thickness l/ '_ A/
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FIGURE B-4 LOCATION OF CHARPY IMPACT SPECIMENS WITHIN WELD METAL TEST MATERIAL B-7
I E
= 5" " E
[ WeldID SideMetal d! / l HAZ % Base Metals i
" "-" '( 1/4 tl g
/
I e em
}N ,i i
/ / l l j
/ /
l / - I M2 j/ I l , t N7 l, I , l l l i-l t l
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- 9 g FIGURE B-5 LOCATI0tl 0F CHARPY IMPACT SPECIMEtlS WITHIN I
HEAT-AFFECTED-ZONE TEST MATERIAL I I B-8
t DEPARTMENT OF THE ARMY
.,-fyb
/ 7'
- ARMY M ATERI ALS AND MECH ANICS RESC ARCH CENTE R 9 ;5d"l'h WATERTOWN. MASS ACHUSETTS o2172 Mr. Stenton/ bht / (617)923-3231
- k. as r
': F DRXMR-MQ 20 April 1978 Combustion Engineering, Incorporated ATTN: Mr. Ray Hurlburt 1000 Prospect Hill Road h'indsor, CT 06095
Dear Mr. Hurlburt:
A set of Charpy specimens broken en the 240 ft-lb capacity Satec machine has been received for evaluation along with the completed I questionnaire. Machine Serial No. 1366 The results of the tests indicate the machine to be producing acce; table energy values at all three energy levels (see inclosed table). This machine satisfies the proof-test requirements of ASTM Standard E-23.
- If this machine is moved or undergces any major repairs or adjustrents, ti.is certificatien occcmes invaliu and the machine must ce rechecked. Removal of the pendulum, replacement of anvils or adjusting the height of drop are examples of such major repairs cr adjustments. It should be noted that if
.a specimen requires over SO*. of the machine capacity to fracture, the machine should be checked to assure that the pendulun is stra. gat, the anvils or striker have not been damaged and that all bolts are still tight. This certification is valid for one year from the date ci the test.
Sincerely,
/ f '
1 Inc1 2&x / sun PAUL h'. ROLSTON Table Chief Quality Engineering Branch XMR Form 1FL-2 1 Sept 77 B-9
COMBUSTION ENGINEERING,INC. Nuclear Laboratories g 3 INSTRUMENT CALIBRATION REQUIREMENT SHEET DATE: 10/26/78 EQUIPMENT Digital Thermocouple Thermometer AREA Rm 235-5 EL-9G INSTRUMENT READABILITY CAllBRATION CHECKEL, I MIN FUNCTION TYPE RANGE READABILITY ' ACCURACY l FREQUENCY BY Thermomet 3r Digital -313 F + 1F +1F 3 mos. to
+752 F I
I I I I I I E I I ,ee,ARee eY A~/uscs APPROVED BY . bM~L APPROVED BY
/ /'
CE 0090193 (8n31 s-10 l
COMilVSTION ENGirJrLR:TJG. INC. Nuclear Labo ator..a (fiSTRUMENT CAL!ilRATIOfJ llLUUlitLMENT ::HTE7 DATE: I-II'79 EQUIPMENT IIoneywell Temperature Recorder EL-78 AREA I '~b
=
INSTRUMENT READABILITY CALIBR ATION CHECKED 6 MIN FUNCTION TYPE RANGE READABILITY ACCURACY FREQUENCY l BY Temperature 6 point -350 F 1F +1F 3 mos. Pocorder to
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/
M[d (W APPROVED DY S "I wa APPROVED BY kl h' - / h /' B-11 CC 0u50t93 las 73)
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COM0UOTIOfJ CrlCINE EnirJG, IrJC. Nucle.ir Labor.itories l INSTRUMENT CAlluRATION REQUIREMErJT SHEET E DATE: I-I7-70 EOUIPMENT Ifoneywoll Temporattire Controller EL-120 AREA _ Rm 235-5 INSTRUMENT READABILITY CALIBRATION CHECKED MirJ FUNCTION TYPE RANGE READABILITY ' ACCURACY FREQUENCY BY W Tceperatur e Dial -350 F 1F +1F 3 me-Control to
+250 F I
I I I I I I I I I I PREPARED BY 'We / /1 M e APPROVED BY b APPROVED DY - ' -
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C._ _ APPENDIX C INSTRUMENTED CHARPY V-NOTCH unfA ANALYSIS All baseline and irradiated Charpy impact tests in this program were performed on an instrumented test system. Instrumented impact testing provides more quantitative data from a Charpy specimen which enable a more detailed analysis of the surveillance material toughness behavior. Photographs of the oscilloscope traces of load and energy versus time were taken for each test of the base plate, weld, heat-affected-zone, and standard reference material. From each trace, the general yield load (PGY), maximum load (PM), and fracture load were determined, as shown in Tables C-1 through C-4. For each material, the loads were plotted against the corresponding test temperature to generate the irradiated load / temperature diagrams. To demonstrate the effects of neutron irradiation on each material, both baseline and post-irradiation load / temperature results were plotted to-gether. These plots are shown in Figures C-1 to C-4. Three index temperatures are of interest. T , the brittle transition B temperature, corresponds to the onset of ductile fracture; below T the B fracture is completely brittle. T g , the ductility transition temperature, corresponds to the mid-transition region where the fracture has become predominantly ductile. T D, the ductility temperature, corresponds to the onset of the upper shelf energy where fracture is completely ductile. i The radiation-induced toughr. css property changes of the surveillance materials are summarized in Table C-5. Standard Charpy impact data are included with the instrumented data since each method represents a unique material property. The standard Charpy tes provides a bulk measurement of the energy to initiate and propagate a crarx through to failure of the material. In contrast, analysis of the instrumented data enables characterization of the components of the dynamic load behavior prior to material failure. The shift in the brittle transition temoerature, T , gand the ductility C-1
I transition temperature, Tg , are comparable to the shif t in the 30 ft-lb Charpy index temperature, Cv 30 The radiation induced changes in the instrumented data therefore tend to corroborate the changes detemined from the standard Charpy impact data. The value of the instrumented results was especially evident in the case of I the HAZ. The HAZ impact energy data (Figure V-23) exhibited considerable scatter, making it difficult to establish index temperatures. In contrast, the instrumented data (Figure C-3) were relatively consistent, despite the previously cited situation with notch placement (see Section VI). The capability to focus on the crack initiation event using the instrumented analysis thus provided a means of quantitatively establishing the radiation induced shift in the HAZ which would othenvise be masked using the bulk measurement approach of the shift in the total impact energy. The third parameter obtainable from the instrumented data isD T , the ductility temperature, which is given in Table C-5. TD corresponds closely with the onset of the upper shelf energy (ninimum temperature for the material to exhibit 100% shear fracture). The agreement is seen to hold for both the unirradiated and irradiated data. I The instrumented Charpy analysis substantiates the results from the standard impact tests. In particular, this approach provides a more quantitative means of measuring radiation induced property changes by analysis of the entire load record rather than using the single (bulk) measurement of impact energy. As more experience is gained with this technique, it offers the potential of providing a more quantitative measurement of toughness property changes than is possible with current impact testing. ~ E I I I C-2
TABLE C-1 IflSTRUMENTED CHARPY IMPACT TEST, FORT CALHOUN IRRADIATED BASE f1ETAL (LONGITUDINAL) ' Test Fast Fracture Specimen Temperature Yield Load, Maximum Load, Load, Identification ( F) _PGY (lb) PM (lb) PF (lb) 167 40 3400 3600 -- 14M 40 3300 3500 -- 117 75 3400 3900 -- 12A 80 3100 3800 -- 112 120 3000 4100 3900 13B 120 3200 4200 3900 14T 160 2800 3900 -- 13C 160 3000 4000 3600 13D 200 2900 4000 3500 144 200 3000 4200 -- 15C 250 2900 3900 -- ISD 250 -- No Record -- C-3
I TABLE C-2 I If1STRUMEtlTED CHARPY IMPACT TEST, FORT CALHOUN IRRADIATED WELD METAL Test Fast Fracture Specimen Temperature Yield Load, Maximum Load, Load, Identification ( F) PGY (lb) PM (lb) PF (lb) 352 80 3900 4000 -- E 31B 120 3500 4000 -- E 33T 120 3400 3900 -- 32E 160 3300 3600 -- 31C 160 3300 3800 -- 336 200 3350 4000 -- 32U 250 3100 3700 -- 341 250 3200 3800 3600 312 300 3050 3900 3400 317 350 3000 3800 -- 32J 350 3000 3900 -- 31E 400 3000 3700 -- I I I I I I I I C-4
TABLE C-3 INSTRUMEtiTED CHARPY IMPACT TEST, FORT CALHOUf1 IRRADIATED HEAT-AFFECTED ZONE Test Fast Fracture Specimen Tempe rature Yield Load, itaximum Load, Load, Identification ( F) PGY (lb) PM (lb) PF (lb) 45L 0 3600 3900 -- 42U 40 3400 4200 4100 45U 40 -- flo Record -- 464 80 3300 4200 -- 42Y 80 3400 4200 3900 452 120 3100 3800 3700 ' 43P 120 3300 4300 2700 463 160 3000 4100 4000 41B 200 3100 4300 -- 42B 250 3000 4100 -- 41Y 300 2900 3900 -- 446 300 2900 4100 -- C-5
TABLE C-4 IflSTRUMEflTED CHARPY IftPACT TEST, FORT CALHOUfl IRRADIATED STAllDARD REFERENCE MATERIAL (SRM) l Test Fast Fracture Specimen Temperature Yield Load, flaximum Load, Load, Identification ( F) PGY (lb) PM (lb) PF (lb) 55T 80 3300 3300 -- 573 120 3100 3600 -- 566 120 31C0 3800 -- 55M 160 3100 4100 -- 55Y 160 3100 4100 -- 56J 200 3000 4000 -- 563 200 3000 4300 -- 565 250 3000 4000 3500 56A 250 3000 4000 4000 56D 300 2900 3900 -- SSP 350 2800 4000 -- 575 350 2800 4100 -- I I I I I C-6 I I
-- w immer - - same ammms amme ammme ammu amme same TABLE C-5 TOUG!!ilESS PROPERTY CilAtlGES BASED Off IllSTRUMEllTED CilARPY IMPACT TEST Ma terial TB ( F) ATB ( F) T;p ( F) AT,; ( F) ACv30( F)
ACv 50 TD I I) 100 l$ ear I) Base Metal (RW) unirrad -22 -- 52 -- -- -- 120 160 irrad 30 52 90 38 60 69 200 200 Weld Metal unirrad -112 --
-30 -- -- --
80 80 irrad 75 187 180 210 238 258 340 350 ? y llAZ unirrad -140 --
-40 -- -- --
160 160 irrad -20 120 40 80 104 117 270 250 SRM unirrad -50 -- 60 -- -- -- 160 200 irrad 80 130 190 130 124 136 315 350
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