ML20070D004
| ML20070D004 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/30/1982 |
| From: | Shaun Anderson, Shogan R, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17256A416 | List: |
| References | |
| WCAP-10086, NUDOCS 8212140420 | |
| Download: ML20070D004 (83) | |
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ANALYSIS OF CAPSULE T FROM THE ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR PLANT REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM (WCAP-10086) r EPRI RESEARCH PROJECT 1021-3 TOPICAL REPORT
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l S. E. Yanichko L S. L. Anderson R.P.Shogan
. R. G. Lott APRIL 1982 Prepared by WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division J P.O. Box 355 Pittsburgh, Pennsylvania 15230 T. R. Mager, Principal Investigator l Prepared for l
ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillv,iew Avenue Palo Alto, California 94304 - l T. U. Marston, Project Manager 8212140420 821208 DR ADOCK 05000
LEGAL NOTICE .s This report was prepared by Westinghouse Electric Corporation (WESTINGHOUSE) as an account of work sponsored by the C!ectric Power Research Institute, Inc. (EPRI). Neither EPRI, members of EPRI, nor WESTINGHOUSE, nor any person >- acting on behalf of either:
- a. Makes any warranty or representation, express or implied, with respect to the ~
accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or
- b. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report. ,_
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l TABLE OF CONTENTS Title Page Section 1
SUMMARY
OF RESULTS 1- 1 INTRODUCTION 2- 1 2 3 BACKGROUND 3- 1 4 DESCRIPTION OF PROGRAM 4- 1 TESTING OF SPECIMENS FROM CAPSULE T 5- 1 5 5.1 Overview 5- 1 5.2 Charpy V-Notch Impact Test Results 5- 3 5.3 Tensile Test Results 5- 5 5.4 Wedge Opening Loading Tests 5- 5 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6- 1 6.1 Introduction 6- 1 6.2 Discrete Ordinates Analysis 6- 1 6.3 Neutron Dosimetry 6- 6 6.4 Transport Analysis Results 6-11 6.5 Dosimetry Results 6-22
^~1 References ii
LIST OF ILLUSTRATIONS Figure Title Page 4- 1 Arrangement of Surveillance Capsules in the R. E. Ginna Reactor Vessel (Updated Lead Factors for the Capsules Are Shown in Parentheses) 4- 3 4- 2 Schematic Diagram of Capsule T Showing Location of Specimens, Thermal Monitors, and Dosimeters 4- 4 5- 1 Charpy V-Notch Impact Data for R. E. Ginna Reactor Vessel Intermediate Shell Forging 125S255 5-15 5- 2 Charpy V-Notch Impact Data for R. E. Ginna Reactor Vessel Lower Shell Forging 125P666 5-16 5- 3 Charpy V-Notch Impact Data for R. E. Ginna Reactor Vessel Weld Metal 5-17 5- 4 Charpy V-Notch Impact Data for R. E. Ginna Reactor Vessel Weld HAZ Metal 5-18 5- 5 Charpy V-Notch Impact Data for ASTM A302B Correlation Monitor Material 5-19 5- 6 Charpy impact Specimen Fracture Surfaces for R. E. Ginna Intermediate Shell Forging 125S255 5-20 5- 7 Charpy impact Specimen Fracture Surfaces for R. E. Ginna Lower Shell Forging 125P666 5-21 5- 8 Charpy Impact Specimen Fracture Surfaces for R. E. Ginna Weld Metal 5-22 5- 9 Charpy impact Specimen Fracture Surfaces for R. E. Ginna Weld HAZ Material 5-23 5-10 Charpy impact Specimen Fracture Surfaces for R. E. Ginna ASTM Correlatior Monitor Material 5-24 5-11 Comparison of Predicted Versus Actual 41-Joule Transition Temperature Increases for R. E. Ginna Reactor Vessel Materials 5-25 5-12 Tensile Properties for R. E. Ginna Reactor Vessel Shell Forging 125S255 5-26 5-13 Tensile Properties for R. E. Ginna Reactor Vessel Shell Forging 125P666 5-27 iii
LIST OF ILLUSYHATIONS (Continued) Title Page Figure 5-14 Tensile Properties for R. E. Ginna Reactor Vessel 5-28 Weld Metal Typical Stress-Strain Curve for Tension Specimens 5-29 5-15 5-16 Fractured Tensile Specimens From R. E. Ginna Intermediate Shell Forging 125S2S5 5-30 5-17 Fractured Tensile Specimens From R. E. Ginna Lower Shell Forging 125P666 5-31 5-18 Fractured Tensile Specimens From R. E. Ginna 5-32 Pressure Vessel Weld Metal R. E. Ginna Reactor Geometry 6- 4 6- 1 Plan View of a Reactor Vessel Surveillance Capsule 6- 5 6- 2 6- 3 Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1 Mev) Within the Pressure Vessel 6-16 Surveillance Capsule Geometry 6- 4 Calculated Radial Distribution of Maximum Fast 6-17 Neutron Flux (E > 1 Mev) Within the Pressure Vessel 6- 5 Relative Axial Variation of Fa.t Neutron Flux 6-18 (E > 1.0 Mev) Within the Pressure Vessel 6- 6 Calculated Radial Distribution of Maximum Fagt Neutron Flux (E > 1 Mev) Within the Surveillance 6-19 Capsules V, R and T 6- 7 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules V und R 6-20 6- 8 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule T 6-21 6- 9 Comparison of Measured and Calculated Fast Neutron Fluence (E > 1 Mev) for Capsules V, R and T 6-34 l iv
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[' E. 4 y j.; ..* w a LIST OF TABLES .~ _.- 9.- Tably Title Page .. 4- 1 Chemistry and Heat Treatment of Material Representing '.' the Core Region Shell Forgings and Weld Metal From .
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the R. E. Ginna Reactor Vessel 4- 5 4- 2 Chemistry and Heat Treatment of Surveillance Material ,. Representing 6-Inch-Thick A3028 ASTM Correlation Monitor Material 4- 6 W 5- 1 Charpy impact Data for R. E. Ginna Reactor i..'"' Pressure Vessel Shell Forgings 5- 6 5- 2 Charpy impact Data for R. E. Ginna Reactor Pressure -~. ~ Vessel Weld Metal and HAZ Material 5- 7 A 5- 3 Charpy impact Data for the ASTM Correlation 7 .. : i; Monitor Material 5- 8 *-- 5- 4 Instrumented Charpy impact Test Results for . J. t - , R. E. Ginna Shell Forgings 5- 9 G 5- 5 Instrumented Charpy impact Test Results for ~. .' . R. E. Ginna Weld Metal and HAZ Material 5-10 ... .1 5- 6 Instrumented Charpy Impact Test Results for the '. .,-. ASTM Correlation Monitor Material 5-11 .;. 5- 7 The Effect of 288'C Irradiation to 1.75 x 10 n/cm ' .I ~ . ' ' (E > 1 Mev) on the Notch Toughness Properties of ;- ? R. E. Ginna Reactor Vessel Material 5-12 :-, '... 5- 8 Summary of R. E. Ginna Reactor Vessel Surveillance . Capsule Charpy impact Test Results 5-13 1 . 5- 9 Tensile Properties for R. E. Ginna Reactor Vessel ~~i Material Irradiated to 1.75 x 10 n/cm* 5-14 i. 6- 1 21 Group Energy Structure 6- 3 : 6- 2 Nuclear Parameters for Neutron Flux Monitors 6- 7 q7 6- 3 Calculated Fast Neutron Flux (E > 1 Mev) and Lead i. ' Factors for R. E. Ginna Surveillance Capsules 6-13 e 6- 4 Calculated Neutron Energy Spectra at the Dosimeter ' 2 -' - Block Location for R. E. Ginna Surveillance Capsules 6-14 ;
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- LIST OF TABLES (Continued)
Title Page Table 6- 5 Spectrum Averaged Reaction Cross Sections at the Dosimeter Block Location for R. E. Ginna Surveillance Capsules 6-15 6- 6 Irradiation History of R. E. Ginna Reactor Vessel Sur 'eillance Capsules 6-24 6- 7 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule V 6-28 6- 8 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule R 6-29 6- 9 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule T 6-30 6-10 Results of Fast Neutron Dosimetry for Capsules V, R and T 6-31 Results of Thermal Neutron Dosimetry for Capsule T 6-32 6-11 6-12 Summary of Neutron Dosimetry Results for Capsules V, R and T 6-33 vi
C SECTION 1
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in the third reactor vessel material surveillance capsule which was removed in the Spring or 1980 from the Rochester Gas and Electric Corporation R. E. Ginna reactor pressure vessel after 6.8 effective full power years of plant operation led to the following conclusions: a The capsule received an average fast neutron fluence (E > 1 Mev) of 1.75 x 10 n/cm' compared to a calculated fluence of 1.65 x 10 n/cm'. m Irradiation to 1.75 x 10 n/cm 2resulted in no increase in the 41-joule (30 ft Ib) transition temperature for the vessel intermediate shell forging and only a 17*C (30 F) increase in the 41-jou!e transition temperature for the vessellower shell forging when compared with unirradiated values.These increases were essentially the same as obtained on prior tests at 1.01 x 10 n/cm'and indicate the materialis very insensitive to radiation. m Irradiation to 1.75 x 10 n/cm 2resulted in a 41-joule transition temperature increase of 83*C for the submerged arc weld metal when compared with the unirradiated value. This increase is approximately the same as obtained on prior tests at .703 and 1.01 x 10n/cm*and indicates that saturation of radiation damage may be occurring. The submerged arc weld is considered to be the limiting material in the reactor vessel. Since its initial RTNDT is estimated to be - 18*C (O F) the RTNDT at a fluence of 1.75 x 10 n/cm2 is 65* C (150* F). The upper shelf energy of the weld after irradiation to 1.75 x 10 n/cm2 is 74-joule (55 ft Ib) and therefore meets the requirements of 10CFR Part 50 Appendix G. I 1-1
I I I h I e Weld HAZ material irradiated to 1.75 x 10 n/cm'showed a 56 C increase in the 41-joule transition temperature when compared with the unirradiated value. The increase was only 6'C higher than prior tests on HAZ materialirradiated to 1.01 x 10 n/cm'. m Comparisons of the 41-joule transition temperature increases for the R. E. Ginna vessel materials with Regulatory Guide 1.99 Revic. ion 1 predictions show that the increases are significantly less than predicted. m End-of-life projected fast neutron fluences for the reactor vessel based on 32 full power years of operation at 1520 MWt are as follows: Fast Neutron Fluence (n/cm2) Vessel Location Calculated Measured Inner Surface 3.96 x 10 4.03 x 10
% Thickness 2.64 x 10 2.68 x 10
% Thickness 7.78 x 10'8 7.91 x 10'8 Based on the analysis of results from the third capsule, the plant heatup and cooldown limit curves currently being used for plant operation are considered to be appropriate for use up to 21 effective full power years of operation.
1-2 l
SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule T, the third capsule of the continuing surveillance program which monitors the effects of neutron irradia-tion on the Rochester Gas and Electric Corp. R. E. Ginna Nuclear Plant reactor pressure vessel materials under actual operating conditions. The surveillance program for the R. E. G:nna reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A descrip-tion of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented in WCAP-7254.[1] The surveillance program was planned to cover the minigum 40-year life of the reactor pressure vessel and was based on ASTM E-185-66," Recommended Practice for Surveillance Tests on Struc-tural Materials in Nuclear Reactors.[2] This report summarizes testing and the postirradiation data obtained from the third material surveillance capsule (Capsule T) removed from the R. E. Ginna reactor vessel in the Spring of 1980 and discusses the analysis of these data.The data are also compared to results of Capsule V,[3] which was removed in the Spring of 1971 and Capsule R.l4} which was removed in the Spring of 1974. 2-1
SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist f racture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bom-bardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as A508 Class 2 (base ma+erial of the R. E. Ginna reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile prop-erties and a decrease in ductility and toughness under certain conditions of irradiation. A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure,". Appendix G, to Section lli of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature, RTNDT-RT NDT i s defined as the greater of the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the temperature 60 F less than the 50 ft Ib temperaNre (or 35-rnil lateral expansion if this is greater) temperature as determined from Charpy specimens oriented normal to the working direction of the material.The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (K IR curve) which appears in Appendix G of the ASME Code. The K IR curve is a lower bound of dynamic, crack arrest, and static fracture tough-ness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K IR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors. l 3-1 i 1
EI I ll RT NDT and,in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties'of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the R. E. Ginna Reactor Vessel Radiation Surveillance Program,[1] in which a surveil-lance capsule is periodically removed f rom the operating nuclear reactor and the en-capsulated specimens are tested. The increase in the Charpy V-notch 50 ft Ib temperature ( ARTNDT) due to irradiation is added to the original RT NDT toadjust the RT NDT for radiation embrittlement. This adjusted RTNDT (RT NDT nitial + i ARTNDT) is used to index the material to the KIR curve and,in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
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SECTION 4 . DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the R. E. Ginna reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. Capsule T was removed after 6.8 effective full power years of plant operation. This capsule contained Charpy V-notch impact, tensile, and WOL specimens (Figure 4-2) from the intermediate and lower shell ring forgings and submerged arc weld metal representative of the core region of the reactorvesseland CharpyV-notch specimens from weld heat-affected zone (HAZ) material. The capsule also contained Charpy V-notch specimens from the 6-inch-thick ASTM correlation monitor material (A302 Grade B). The chemistry and heat treatment of the surveillance material are pre-sented in Tables 4-1 and 4-2. The chemical analyses reported in Table 4-1 were obtained from unirradiated material used in the surveillance program. In addition a chemical analysis was performed on an irradiated charpy specimen from the weld metal and is reported in Table 4-1. This analysis indicates that the copper, nickel and phosphorus content for the irradiated material is in good agreement with the unirradiated analysis. All test specimens were machined from the % thickness location of the forgings. Test specimens represent material taken at least one forging thickness from the quenched end of the forging. All base metal Charpy V-notch and tensile specimens were oriented with the longitudinal axis of the specimen parallel to the principal working direction of the forgings. The WOL test specimens were machined with the simulated crack of the specimen perpendicular to the surfaces and rolling direction of the forgings. 4-1 1
Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the welding direction. Tensile specimens were oriented with the longitudinal axis of the specimen parallel to the welding direction. Capsule T contained dosimeter wires of pure copper, nickel. and aluminum-cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dcsimeters of Np237 and U238 were contained in the capsule and located as shown in Figure 4-2. Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule and were located as shown in Figure 4-2. The two eutectic alloys and their melting points are: 2.5 Ag,97.5 Pb Melting Point 579 F (304 C) 1.75 Ag,0.75 Sn,97.5 Pb Melting Point 590* F (310 C)
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TABLE 4-1 ,4 . CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE ,- CORE REGION SHELL FORGINGS AND WELD METAL FROM THE -c R. E. GINNA REACTOR VESSEL - Chemical Analyses (percent) ~~.~' Element Forging 125P666 Forging 125S255 Weld Metal 'i C 0.19 0.18 0.075 0.06 . '. " :
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Mn 0.67 0.66 1.31 1.29 P 0.010 0.010 0.012 0.006 'd S 0.011 0.007 0.016 0.020 D~ g-Si 0.20 0.23 0.59 0.41 : Mo 0.57 0.58 0.36 0.22 1'+ Cu 0.05 0.07 0.23 0.22 Ni 0.69 0.69 0.56 0.50 ,, : 9 Cr 0.37 0.33 0.59 0.03 [~, ' Al 0.004 0.003 0.02 - N2 - - 0.015 - V 0.02 0.02 0.001 -
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-27 Sn 0.01 0.01
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- a. Analysis Performed on Charpy Specimen W26 From Capsule T .
HEAT TREATMENT {, ' FORGING Heated to 1550 F - 9 hr, Water-quenched _. 125P666 Tempered at 1220 F - 12 hr, Aircooled 1- - Stress relieved at 1100 F - 11 hr, Furnace-cooled ' Q L,
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FORGING Heated at 1550 F - 15% hr, Water-quenched .G'; 125S255 Tempered at 1220 F - 18 hr, Aircooled :' . . Stress relieved at 1100 F - 11% hr, Furnace-cooled .~ WELDMENT Stress relieved at 1100 F - 11% hr, Furnace-cooled 4.
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t TABLE 4-2 ) CHEMISTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIAL REPRESENTING 6-INCH-THICK A302B ASTM CORRELATION MONITOR MATERIAL Chemical Analysis (percent) C Mn P S Mo Si Cu Ni Cr . 0.24 1.34 0.011 0.023 0.51 0.23 0.20 0.18 0.11 HEAT TREATMENT The 6-inch-thick plate was charged into a furnace operating at 110 F heated at a maximum rate of 63 F per hour to 1650* F, held at temperature for 4 hours, and water-quenched to 300 F. The plate was then recharged into a furnace operating at 700 to 750 F, heated at a maximum rate of 63 F per hour to 1200 F, and held at that temperature for 6 hours. M 4-6
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TESTING OF SPECIMENS FROM CAPSULE T H-{ 9
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5-1. OVERVIEW F.? The postirradiation mechanical testing of the Charpy V-notch and tensile specimens 6 *- ~ ' was performed at the Westinghouse Research and Development Laboratory with , consultation by Westinghouse Nuclear Energy Systems personnel. Testing was per- f ,} formed in accordance with 10CFR50, Appendices G and H. __ J.f;
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carefully removed, inspected for identification number, and checked against the master list in WCAP-7254.Il} No discrepancies were found. . Examination of the twolow-melting 304 C(579 F)and 310 C(590 F)eutecticalloys # - indicated no melting of either type of thermal monitor. Based on this examination,the a . maximum temperature to which the test specimens were exposed was less than .Ii' 304 C (579 F). -? . The Charpy impact tests were performed on a Tinius-Olsen Model 74,358J machine. , The tup (striker) of the Charpy machine is instrumented with an Effects Technology . 7 ' model 500 instrumentation system. With this system, load-time and energy-time :~ - signals can be recorded in addition to the standard measurement of Charpy energy N (E o). From the load-time curve, the load of general yielding (PGY), the time to gen- ; } " ,. . eral yie! ding (tGY), the maximum load (PM ), and the time to maximum load (t M) can ..,' be determined. Under some test conditions, a sharp drop in load indicative of fast ' A. fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pp), and the load at which fast fracture terminated in identified as ? '. / . the arrest load (PA)- '*D .-- The energy at maximum load (EM) was determined by comparing the energy time Nrh record and the load-time record. The energy at maximum load is roughly equivalent ';. -4 p 4 5-1 s- r ,- J . ,- f 1 L.
i to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E p) is the difference between the total energy to fracture (E o) and the energy at maximum load. The yield stress (ay) is calculated from the three point bend formula. The flow stress is calculated from the average of the yield and maximum loads, also using the three point bend formula. Percent shear was determined from postfracture photographs using the ratio-of-areas method in compliance with ASTM Specification A370-74. The lateral ex-pansion was measured using a dial gage rig similar to that shown in the same specification. Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8 and E21, and MHL Procedure 7604 Revision 2. All pull rods, grips, and pins were made of inconel 718 hardened to R c45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test. Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83.
' Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air.
Because of the difficulty in remotely attaching a thermocoupie directly to the speci-men, the following procedure was used to monitor specimen temperature. Chrornel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550 F. The upper grip was used to control the f urnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures. Experiments indicated that this method is accurate to plus or minus 2 F. 5-2
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement. 5-2. CHARPY V-NOTCH IMPACT TEST RESULTS The toughness results from Charpy V-notch impact tests performed on the various surveillance materials in Capsule T after irradiation to 1.75 x 10 n/cm2 are presented in Tables 5-1 through 5-3 and Figures 5-1 through 5-5. Instrumented Charpy impact test results for the various materials are shown in Tables 5-4 through 5-6. A summary of the surveillance test results is presented in Table 5-7. The fractured surfaces of the impact specimens are shown in Figures 5-6 through 5-10. Irradiation of Charpy V-notch specimens from theintermediate shell forging 125S255 to 1.75 x 10n/cm* resulted in no increase in either the 41-joule (30 f t Ib) or 68-joule (50 ft Ib) transition temperature and no decrease in the upper shelf energy when compared with unirradiated values as shown in Figure 5-1. Specimens from the lower shell forging 125P666 which were irradiated to 1.75 x 10 n/cm'showed a 41 and 68-joule transition temperature increase of 17 C (30 F) and an upper shelf energy decrease of 54-joule (40 ft Ib) when compared with unirradiated values as shown in Figure 5-2. A comparison of these resultswith prior results f rom irradiations at lower fluence shown in Table 5-8 indicates that the increased neutron fluence caused only a 3* C increase in transition temperature for forging 125P666 and no increase for forging 125S255. Weld metal specimens irradiated to 1.75 x 10n/cm resulted in a 41 and 68-joule transition temperature increase of 83* C and 144 C respectively and an upper shelf energy decrease of 34-joule when compared to unirradiated values as shown in Figure 5-3. A comparison of these results with prior test results as shown in Table 5-8 shows essentially no increase in the 41-joule transition temperature and a 47 C increase in the 68-joule transition temperature. 5-3
Weld HAZ specimens irradiated to 1.75 x 10 n/cm'showed considerable scatter in data as shown in Figure 5-4. Transition temperature increases of 56* C and 39'C at the 41-joule and 68-joule levels respectively were determined f rom the comparison of unirradiated ,and irradiated test results. The irradiated upper shelf energy of the HAZ material showed no decrease in energy. A comparison of these results with prior irradiation results shown in Table 5-8 indicates a 6* C increase in the 41-joule transition temperature when compared with tests at 1.01 x 10 n/cm . ASTM A302B correlation monitor materialirradiated to 1.75 x 10'* n/cm'resulted in a 41 and 68-joule transition temperature increase of 78'C and 91* C respectively when compared with unirradiated values as shown in Figure 5-5. The upper shelf energy of the monitor material decreased by 23-joule. The transition temperature increases were 20-25' C higher than previous tests at 1.01 x 10 n/cm'as shown in Table 5-8. The fracture appearance of each irradiated Charpy specimen from the various vessel materials is shown in Figures 5-6 through 5-10. Each of the vessel materials shows an increasing ductile or tougher appearance with increasing test temperatures. Figure 5-11 shows a comparison of the 41-joule transition temperature increase for the R. E. Ginna surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99 Revision 1. This comparison shows that all the reactor vessel surveillance materials exhibited 41-joule transition temperature increases lower than would be predicted. This behavior is consistent with the resuits from other reactor vessel capsules evaluated as part of this EPRI program. Based on the R. E. Ginna surveillance capsule test results to date. it can be concluded that the reactor vessel beltline region material is not as sensitive to radiation as predicted and a saturation of radiation damage may be occurring. 5-4
5-3. TENSION TEST RESULTS The results of tensile tests performed on materialfrom the two shellforgings 125S255 and 125P666 and the weld metal are shown in Table 5-9 and Figures 5-12 through 5-14 respectively. These results show that the 0.2% yield strength and other properties of the two forgings are not highly affected by irradiation to 1.75 x 10 n/cm . The increase in 0.2% yield strength of approximately 100 MPa exhibited by the weld metal indicates that the weld metal is more sensitive to radiation than are the forgings which is consistent with transition temperature increases observed for these materials. A typical stress strain curve for the tensile tests is shown in Figure 5-15. Photographs of the fractured tensile specimens for the two forgings and weld metal are shown in Figures 5-16 through 5-18, respectively. 5-4. WEDGE OPENING LOADING TESTS Wedge Opening Loading (WOL) fracture mechanics specimens which were con-tained in the surveillance capsule have been stored at the Westinghouse Research Laboratory, they will be tested and reported on later. 5-5
TABLE 5-1 CHARPY IMPACT DATA FOR R. E. GINNA REACTOR PRESSURE VESSEL SHELL FORGINGS (IRRADIATED TO 1.75 x 10" n/cm2) Sample Temperature impact Enargy Lateral Expansion Shear Number ( C) ( F) (J) (f t Ib) (mm) (mils) (%) Forging 125P666
-50 16.5 12.0 0.18 7.1 2 P26 -46
-23 -10 52.0 38.5 1.13 44.5 26 P22
-18 0 57.0 42.0 .80 31.5 22 P28
-4 25 49.0 36.0 0.91 35.8 18 P29 10 50 128.0 94.5 1.50 59.1 47 P25 24 75 164.0 121.0 2.01 79.1 74 P24 66 150 133.0 98.0 1.67 65.8 72 P27 79 175 203.5 150.0 2.14 84.3 100 P30 121 250 190.0 140.0 2.15 84.5 100 P23 177 350 186.5 137.5 1.97 77.5 100 P21 Forging 125S255 S27 -73 -100 3.5 2.5 0.07 2.8 2
-50 36.5 27.0 0.43 16.9 9 S25 -46 S28 -18 0 92.0 68.0 1.29 50.8 31
-18 0 106.0 78.0 1.35 53.2 38 S22
-4 25 95.5 70.5 1.29 50.8 34 S24 24 75 76.0 56.0 1.10 43.3 32 S23 150 139.5 103.0 1.74 68.5 70 I S21 66 79 175 153.0 113.0 2.01 79.0 70 S26 250 210.0 155.0 2.01 79.0 100 l S30 121 177 350 202.5 149.5 1.83 72.0 100 S29 l
5-6 l
TABLE 5-2 CHARPY IMPACT DATA FOR R. E. GINNA REACTOR PRESSURE VESSEL WELD METAL AND HAZ MATERIAL (IRRADIATED TO 1.75 x 10" n/cm2) Sample Temperature Impact Energy Lateral Expansion Shear Number ( C) ( F) (J) (ft Ib) (mm) (mils) '/o Weld Metal W30 -18 0 7.0 5.0 0.11 4.3 13 W27 24 75 30.0 22.0 0.36 14.2 38 W23 52 125 39.5 29.0 0.72 28.4 54 W22 66 150 48.0 35.5 0.60 23.6 73 W28 99 210 55.5 41.0 0.89 35.0 90 W21 121 250 63.0 46.5 0.77 34.5 98 W25 149 300 68.5 50.5 1.13 44.5 100 W26 177 350 71.0 52.5 0.97 38.2 100 W29 218 425 75.0 55.5 1.32 52.0 100 W24 218 425 85.5 63.0 1.43 56.3 100 HAZ Material H24 -73 -100 5.5 4.0 0.12 4.7 2 H21 -51 -60 18.5 13.5 0.16 6.3 10 H22 -46 -50 89.5 66.0 1.06 41.7 31 H28 -18 0 33.0 24.5 0.71 28.0 27 H23 10 50 45.5 33.5 0.70 27.6 46 H25 24 75 158.0 116.5 1.37 53.9 78 H29 66 150 173.5 128.0 1.58 62.2 100 H30 121 250 188.5 139.0 1.97 77.6 100 H27 177 350 116.5 86.0 1.13 44.5 100 5-7
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7 1-h i - l l TABLE 5-3 ! CHARPY IMPACT DATA FOR THE ASTM CORRELATION MONITOR MATERIAL (IRRADIATED TO 1.75 x 10" n/cm2) Sample Temperature Impact Energy Lateral Expansion Shear Number ('C) ( F) (J) (f t Ib) (mm) (mils) (%) R20 24 75 54.0 40.0 0.63 24.8 53 38 100 26.0 19.0 0.34 13.4 15 R24 66 150 35.5 26.0 0.73 28.7 41 R22 79 175 40.5 30.0 0.71 28.0 44 R23 99 210 45.5 33.5 0.64 25.2 48 R19 121 250 83.5 61.5 1.21 47.6 100 R18 R17 177 350 85.5 63.0 1.25 49.2 100 R21 218 425 80.0 59.0 1.09 43.0 100 5-8 1
TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR l R. E. GINNA SHELL FORGINGS Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress l No. ( C) (J) (kJ/m2) (kJ/m2) (kJ/m2) (N) ( s) (N) ( s) (N) (N) (MPa) (MPa) Forging 125P666 l P26 -46 16.5 203 130 74 12800 105 13400 210 13300 0 659 675 P22 -23 52.0 652 610 43 15100 105 18700 655 18500 0 773 869 l P28 -18 57.0 712 593 118 14100 150 175GO- 695 17300 0 724 813 , P29 -4 49.0 610 545 65 14000 95 18200 600 18200 0 719 828 l
$ P25 10 128.0 1602 697 905 13600 100 18600 755 14800 3100 701 830 P24 24 164.0 2051 655 1396 12400 115 16700 790 9400 2800 638 748 P27 66 133.0 1661 656 1005 13000 105 17400 760 14200 6300 668 782 P30 79 203.5 2542 648 1894 12200 100 17600 750 628 767 P23 121 190.0 2373 650 1723 11700 105 16200 800 602 719 P21 177 186.5 2330 614 1717 11800 135 16200 790 607 721 Forging 125S255 S27 -73 3.5 42 42 0 1470:0 85 14700 0 759 759 S25 -46 36.5 458 355 102 14000 115 16900 440 16900 0 718 794 S22 -18 106.0 1322 638 684 15200 100 19500 655 16700 0 784 893 S28 -18 92.0 1152 603 550 15900 105 19700 615 18300 100 820 916 S24 -4 95.5 1195 619 576 12300 120 19200 680 17700 0 631 810 S23 24 76.0 949 634 315 15000 95 19800 645 19000 200 771 894 S21 66 139.5 1746 541 1204 14200 110 19100 650 15100 8900 732 858 S26 79 153.0 1915 594 1321 13400 110 17700 680 10600 5000 691 802 S30 121 210.0 2627 617 2010 12800 105 17700 700 657 784 S29 177 202.5 2534 652 1882 13100 135 17500 750 675 788
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l t TABLE S-5 INSTRUMENTED CHARPY IMPACT RESULTS FOR R. E. GINNA WELD METAL AND HAZ MATERIAL Normalized Energies Arrest Yield Flow Test Charpy Charpy Maxi _ mum . Prop Yield Time Maximum Time to Fracture Em/A ' - Ep/A Load to Yield Load Maximum Load Load Stress Stress Sample Temp Energy Ed/A ( s) (N) (ys) (N) (N) (MPa) (MPa) No. ( C) (J) (kJ/m2) (kJ/m2) (kJ/m2) (N) We:d Metal
~
85 73 12 17000 110 17300 125 17300 0 874 884 W30 -18 7.0 214 159 13900 105 15900 285 15600 4100 716 768 W27 24 30.0 373 312 180 15800 105 18300' 350 17800 7400 811 876
,, W23 52 39.5 491-602 357 244 14800 100 -- 17600 405 17600 10400 760 832 y W22 66 48.0 17800 400 17300 10000 778 847 W28 99 55.5 695 429 265 15100 100 788 297 492 13900 100 16700 360 716 788 W21 121 63.0 856 381 475 14400 130 17000 460 741 807 W25 149 68.5 890 316 574 12900 '100 16500- 425 664 755 W26 .177 71.0 85.5 '1068 388 679 14800 130 17900 455 763 841 W24 218 75.0 941 377 564 13300 105 16700 455 686 773 W29 218 s Weld HAZ Material
-s H24 -73 5.5 68 229 181 48~ 17800 105 19200 210 18900 0 916 051 H21 -51 18.5
-46 89.6 1119 709 410 18600 115 22600 635 ~22000 0 958' 1061 H22 415 337 + 781 16900 105 '19400 360 19200 0 871 934 H28 -18 33.0 . ,
H23 10 4S.5 568 245 323 12500 135 13700 -390" ' ~13700 4800 641 672 158.0 1974 735 1239 15400 105 19600 755 14000 6300 794 901 H25 -24 , 66 173.5 2169 631 1539 13/00 95 19500 665 705 855 H29 2356 738 1617 13400 100 19400 780 691 846 - H30 121 188.5 806
'177 116.5 1458 531 927 13600 100 "17800 600 698 l H27
^%
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% g TABLE 5-6 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE ASTM CORRELATION MONITOR MATERIAL 4 Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow w Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress r
g No. ( C) (J)' (kJ/m2) (kJ/m') (kJ/m ) (N) ( s) (N) ( s) (N) (N) (MPa) (MPa) R20 24 54.0 678 418 260 17000 110 20500 425 20400 9400 876 965 R24 38 26.0 322 226 96 14800 115 17200 290 17100 1700 761 824 R22 66 35.5 441 298 143 !3800 100 17600 360 17600 2600 712 808 R23 79 40.5 508 297 212 14300 100 17500 350 17500 5800 734 818 Rt9 99 45.5 568 429 138 13600 100 17600 400 17600 12500 698 801 R18 121 83.5 1042 412 631 13800 100 18700 455 709 836 R17 177 85.5 1068 447 620 13500 100 18400 500 695 820 R21 218 80.0 1000 355 645 12200 120 16000 460 629 725
TABLE 5-7 THE EFFECT OF 288 C 1RRADIATION TO 1.75 x 10" n/Cm2 (E > 1.0 Mev) ON THE NOTCH TOUGHNESS PROPERTIES OF R. E. GINNA REACTOR VESSEL MATERIALS Transition Temperature Unirradiated Irradiated ATransition Temperature Average Energy Absorption at Full Shear 50 ft Ib 30 ft Ib 35 mils 50 ft Ib 30 ft Ib 35 mils 50 ft Ib 30 ft Ib 35 mits 68 J 41 J .9mm 68 J 41 J .9mm 68 J 41 J .9mm Unirradiated Irradiated AEnergy (*C) (*F) (*C) (' F) (*C) (*F) (*C) (*F) (*C) (*F) (*C) ('F) ('C) (* F) (*C) ('F) (*C) ('F) (J) (ft th) (J) (ft Ib) (J) (ft Ib)
?
a teeterial
" Forging 125P666 15 40 25 -9 15 10 -15 5 17 30 17 30 17 30 i248 183 194 143 54 40 Forging 125S255 -11 12 25 -21 11 12 25 -21 -5 0 0 0 0 0 0 190 140 206 152 +16{a.] +12[a.]
'Neld Metal -4 25 25 20 140 285 52 125 104 220 144 260 83 150 133 240 108 80 74 55 34 25 HAZ Metal -23i-10 75 30 16 60 -4 25 10 50 39 70 56 100 44 80 122 90 160 118 +38{a.] +28 {a.]
Corre-lation Monitor 22 72 4 40 10 50 113 235 82 180 99 210 91 163 78 140 89 160 106 78 , 83 61 23 17
- a. Shelf Energy increase
TABLE 5-8
SUMMARY
OF R. E. GINNA REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS 68 J 41 J 50 f t Ib 30 ft Ib Decrease in Trans. Ternp Trans. Temp Upper Shelf Fluence increase (b) increase (b) Energy (b) Material 10" n/cm2 ( C) ( C) _( JF) (* F) (J) (f t Ib) Forging 125P666 0 703 14 25 14 25 31 23 1.01 14 25 14 25 18 13 1.75 17 30 17 30 54 40 Forging 125S255 0.703 0 0 0 0 0 0 1.01 0 0 0 0 0 0 1.75 0 0 0 0 +16[a.] +12[a.] Weld Metal 0.703 97 175 78 140 40 29.5, 1.01 97 175 92 165 41 30 1.75 144 260 83 150 34 25 HAZ Metal 0.703 28 50 0 0 +54[a.] +40[a.] 1.01 47 85 50 90 20 15 1.75 39 70 56 100 +38I *'l +28 I 'l Correlation 0.703 66 118 50 90 18 13 Monitor 1.01 71 128 53 95 25 18.5 1.75 91 163 _7_8 140 23 17
- a. Shelf Energy increase
- b. Change resulting from a comparison of Unirradiated versus Irradiated Data 5-13
TABLE 5-9 TENSILE PROPERTIES FOR R.E. GINNA REACTOR VESSEL MATERIAL IRRADIATED TO 1.75 x 10" n/cm2 Uniform Total Reduction Test Yield Ultimate Fracture Fracture Fracture in Area Load Stress Strength Elongation Elongation Sample Temp Strength Strength (%) (%) No. Material C ( F) MPa (ksi) MPa (ksi) N(kip) MPa (ksi) MPa (ksi) (%) 12900 1180 407 9.0 23.3 65 S20 Forging 26 544 667 125S255 (78) (78.9) (96.8) (2.90) (170.9) (59.1) 1350 449 7.8 19.4 67 Forging 93 562 678 14200 S19 125S255 (200) (81.5) (98.3) (3.20) (196.5) (65.2) , o, 14700 1190 464 7.4 18.2 61 S21 Forging 260 534 667 k 125S255 (500) (77.4) (96.8) (3.30) (172.7) (67.2) 12500 1250 393 11.3 25.4 69 P21 Forging 26 478 625 125P666 (78) (69.3) (90.7) (2.80) (181.9) (57.0) 11100 1300 351 9.8 23.3 73 P20 Forging 93 456 576 125P666 (200) (66.2) (83.5) (2.50) (188.3) (50.9) 11800 1020 372 8.6 20.4 64 P19 Forging 260 428 583 125P666 (500) (62.1) (84.5) (2.65) (148.0) (54.0) 16900 1320 534 11.3 22.5 60 W7 Weld 26 660 758 Metal (78) (95.7) (110) (3.80) (191.4) (77.4) 1300 506 10.4 21.0 61 W9 Weld 93 625 699 16000
- Metal (200) (90.7) (101.4) (3.60) (188.3) (73.3) 16300 1230 516 9.8 19.5 58 W8 Weld 260 558 671 Metal (500) (81.0) (97.3) (3.68) (178.3) (74.9)
('C)
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FIGURE 5-1. CHARPY V-NOTCH IMPACT DATA FOR R. E. GINNA REACTOR VESSEL INTERMEDIATE SHELL FORGING 125S255 5-15 1
('C)
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FIGURE 5-2. CHARPY V-NOTCH IMPACT DATA FOR R. E. GINNA REACTOR VESSEL LOWER SHELL FORGING 125P666 5-16
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FIGURE 5-3. CHARPY V-NOTCH IMPACT DATA FOR R. E. GINNA REACTOR VESSEL WELD METAL 5-17
(* C)
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FIGURE 5-4. CHARPY V-NOTCH IMPACT DATA FOR R. E. GINNA REACTOR VESSEL WELD HAZ METAL 5-18
('C)
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FIGURE 5-5. CHARPY V-NOTCH IMPACT DATA FOR ASTM A302B CORRELATION MON: TOR MATERIAL 5-19
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600
- 300 500 -
400 -
- 200
_ 300 Y WELD METAL PREDICTION CORRELATION .x 200 - MONITOR PREDICTION y WELD METAL - 100 ? e /
- E -eo 5 -
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- 30 -
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10 (FORGING 125S255 SHOWED NO INCREASE AT ALL THREE FLUENCES) 10 I ! I I I III I I I I I I II 10 '" 2 4 6 8 10 " 2 4 6 8 10" 2 FLUENCE (n/cm ) FIGURE 5-11. COMPARISON OF PREDICTED VERSUS ACTUAL 41-JOULE TRANSITION TEMPERATURE INCREASES FOR R. E. GINNA REACTOR VESSEL MATERIALS (PREDICTED INCREASES BASED ON METHODS IDENTIFIED IN REGULATORY GUIDE 1.99 REVISION 1) 5-25
(*C) 50 100 150 200 250 300 0 120 i i e i s # - 800 110 - O TENSILE STRENGTH - 700 100 O A- -A A .A g O - 600 b o E g 0.20o YlELD STRENGTH
$ 80 - --g O
N O l o ]500 70 - g o o l 60 - M400 l 1 50 CODE OPEN POINTS - UNIRRADIATED CLOSED POINTS - IRR ADIATED AT 1.75 x 10 n/cm'
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70 - o_ g o o e-60 - REDUCTION IN AREA
,] 50 -
Y 3 40 - g 30 - 9 0 ; 20 f TOTAL ELONGATION 10 0 300 400 500 600 O 100 200 TEMPERATURE (* F) FIGURE 5-12. TENSILE PROPERTIES FOR R. E. GINNA REACTOR VESSEL SHELL FORGING 125S255 l 5-26
('C) 0 50 100 150 200 250 300 110 i a i i i i i
- 700 100 -
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%% g - g
- 20 -
TOTAL ELONGATION 10 - 0 O 100 200 300 400 500 600 TEMPERATURE (* F) FIGURE 5-13. TENSILE PROPERTIES FOR R. E. GINNA REACTOR VESSEL SHELL FORGING 125P666 5-27
(' C) 100 150 200 250 300 0 50 , 120 , , , i i 110 - A' 100 - g e TENSILE STRENGTH , g _ - 600 _ y 3% ' ~ m
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.i 40 -
I O g 30 - 2 20 - k e -- 8 . l TOTAL ELONGATION 10 - 0 400 500 600 O 100 200 300 TEMPERATURE (* F) FIGURE 5-14. TENSILE PROPERTIES FOR R. E. GINNA REACTOR VESSEL WELD METAL 5-28
100 90 - 600 80 - 500 70 - g* '-
- 400 6 =
50 - E - 300
- 40 -
g 30 -
- 200 20 -
SPEClMEN P19
- 100 10 - ,
O I ' ' ' ' ' ' ' O O .025 .050 .075 .100 .125 .150 .175 .200 .225 STRAlrJ (in/in) FIGURE 5-15. TYPICAL STRESS-STRAIN CURVE FOR TENSION SPECIMENS
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M- g%=-2,v <g. . . -. ,j ( > S-21 260*C t t I r l FIGURE 5-16. FRACTURED TENSILE SPECIMENS FROM R. E. GINNA PRESSURE VESSEL INTERMEDIATE SHELL FORGING 125S255 t 5-30 l l _ . , ~ . - . --,--.--------,---..,7 _.-.- -- - - - _ . - . - m -
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SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1. INTR 3 DUCTION This section presents the dosimetry results of the examination of capsule T and an updated evaluation of dosimetry from capsules V and R.The three capsules are a part of the continuing program which monitors the effects of neutron irradiation on the R. E. Ginna reactor vessel materials under actual operating conditions. Knowledge of the neutron environment within the pressure vessel-surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second,in relating the changes observed !n the test specimens to the present and future condition of the reactor pressure vessel, a relationship between the environment at various positions within the reactor vessel and that experienced by the test specimens must be estab-lished. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monito!s contained in each of the surveillance capsules. The latter information, on the other hand, is derived solely from analysis. This section descnbes a discrete ordinates Sn transport analysis performed for the R. E. Ginna reactor to determine the fast neutron (E > 1.0 Mev) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance cap-sules: and, in turn, to develop lead f actors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. 6-2. DISCRETE ORDINATES ANALYSIS A plan view of the R. E. Ginna reactor geometry at the core midplane is shown in Figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a 0 -45 sector is 6-1
depicted. Six irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance program. Three capsules are located symmetrically at 13 ,23*, arid 33' from the cardinal axis as shown in Figure 6-1. A plan view of a single surveillance capsule attached to the thermal shield is shown in Figure 6-2. The stainless steel specimen container is 1-inch square and approxi-mately 63 inches in height. The containers are positioned axially such that the specimens are centered on the core midolane, thus spanning the central 5.25 feet of the 12-foot-high reactor core. From a neutronic standpoint, the surveillance capsule structures are significant. In fact, as will be shown later, they have a marked impact on the distributions of neutron flux and energy spectra in the water annulus between the thermal shield and the reactor vessel. Thus, in order to properly ascertain the neutron environment at the test specimen locations. the capsules themselves must be included in the analytical model. Use of at least a two-dirnensional computation, is, therefore, mandatory. In the analysis of the neutron environment within the R. E. Ginna reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the DOT [5] two-dimensional discrete ordinates code. The radial and azimuthal distribu-tions were obtained from an R. O computation wherein the geometry shown in Figures 6-1 and 6-2 was described in the analytical model. In addition to the R. O com-putation, a second calculation in R, Z geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest. In the R. Z analysis the reactor core was treated as an equivalent volume cylinder and, of course, the surveillance capsules were not included in the model. Both the R, O and the R, Z analyses employed 21 neutron energy groups,an S8 ngu-lar quadrature, and a P 1 cross-section expansion. The cross sections were gener-ated via the Westinghouse G AMB1T [6] code system with broad group processing by the APPROPOS[7] and ANISN[8] codes. The energy group structure used in the analysis is listed in Table 6-1. A key input parameter in the analysis of the integrated f ast neutron exposure of the reactor vessel is the core power distribution. For the analysis, power distributions representative of time-averaged conditions derived from statistical studies of long-6-2
_ .-. - ~ _ _ - . . ...- . . _. . - . . -- 1 1 . TABLE 6-1 21 GROUP ENERGY STRUCTURE 4 Group Lower Energy (Mev) 1 7.79* 2 6.07 3 4.72 4 3.68 . 5 2.87 6 2.23 7 1.74 8 1.35 9 1.05 10 0.821
! 11 0.388 I 12 0.111 ! 13 4.09 x 10 ! 14 1.50 x 10
15 5.53 x 10~ 16 5.83 x 10 17 7.89 x 10-5 18 1.07 x 10-5 j 19 1.86 x 10-8 20 3.00 x 10-7 21 0.0 Upper energy of group 1 is 10.0 Mev d 6-3 l e
=w-- ---y= -,x- --w ew- *-- +'%--w-,*---,w r- --1-- -wm + - - - - - - - = - - - - +-&--yn a -
yy=r,v-w-v -------m-- V y -
+'N 'T
PRESSURE VESSEL SURVEILLANCE CAPSULE 0' 13* ! CAPSULES V, R) [/ ! 23' (CAPSULES T. P) THERMAL 33' (CAPSULES S, N)
/
SHIELD / /
/
j Y // , I '
' 45*
WI/III'/////i ,/ / , in, , , , . nii, '. , , ,,y, t ,
)
l l usu,pyi,u,pu,- l /
~
/ / / in,. u,g .,
/ / /
/ / /
/
// / REACTOR CORE l/ /
/// ~
l /// l I FIGURE 6-1. R. E. GINNA REACTOR GEOMETRY 6-4 n
I ( 13', 23* , 33* ) I (12' , 22* , 32' ) gefg e jpHARPY l / s
}
THERMAL SHIELD FIGURE 6-2. PLAN VIEW OF A REACTOR VESSEL SURVEILLANCE CAPSULE 6-5 l l _ . - _ . _ l
s p. i A p (k f; ( /. 1, A- . i; W . y ed.Theseinput distri- a term operation lof Westinghouse two-loop plants were emplo' butions idclude rod-by-rod spatial variations for all peripheral fuel assemblies.
,t
;n it should be no'ed that this particular power distribution is in' tended to produce accurate end-of-life neutron exposure levels for the pressure vessel. As such, the calculatio. is iribeed representative of an average neutron flux andsmall (= 15-20%)
3 ~ deviation 2'frorp !::ycie 'o cycle are to be expected. c* , Having the Tsulg 'of the R, O and R, Z calculations, three-dimensional variations of neutron flux may be approximate'd by assuming that the following relation holds for the applicabl,e regions ol'tne reacto,r. 4 4
)I 6-1 (R,Z,0.E g ) = $(R.O.Eg ) F(Z.Eg) where: ,; $(R,Z.0,Ej) xneutron flux at point R.Z.O within energy group g k
( (R.O.E g )g i neutron flux at point R.O within energy group g obtained from the R.O calculation
/.
F(2!.Eg) = telative axial distribution of neutron flux within energy group g obtained from the R.Z calculation
~
N 6-3. NEUTEON DOSIMETRY The passive rieutron flux monitors included in the R. E. Ginna surveillance program are listed in Table'6-2. Ths first five reactions in Table 672 are used as fast neutron monitors to felatineutron fluence (E. 1.0 Mev) to measured materials properties changes. To o'rbperly account for burnout of the product isotope generated by fast neutron reactions, it is necessary to also determine th"e magnitude of the thermal neutron flu f at the monitor location. Therefore. barr and cadmium-covered co , aluminum monitors were also included. .I The relative locations of the various monitors within the surveillance capsules are shown in Figure 4-2. The nickel, copper,and cobalt-aluminum monitors,in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The iron monitors are obtained by drilling samples from selected Charpy test specimens. r The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule. b
)
fi l 6-6 f
=;
9 TABLE 6-2 4 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Target , Fission Reaction Weight Response ,. Product Yield Monitor Material of Interest Fraction Range . Half-Life (%) , _ ~ _ _ _ _ . ._c ,. , Copper Cu(n.a )Co* 0.6917 E 4.7 Mev 5.2'7 years Fe**(n. p) M n'd 0.0585 E :> 1.0 Mev 314 days { Iron Nickel N "(n.p)Co 68 0.6777 F 1.0 Mev 71.4 days Uranium-238' U2'8(n.f)Cs"' 1.0 E 0.4 Mev 30.2 years 6.3 , Neptunium-237' Np?"(n,f)Cs'3' 1.0 E 0.08 Mev 30.2 years 6.5 Cobalt-Aluminum
- Co"(n,y)Co* 0.0015 0.4 eV < 0.015 Mev 5.27 years Cobalt-Aluminum Co"(n,y)Co" 0.0015 E < 0.0015 Mev 5.27 years
- Denotes that monitor is cadmium shielded I
The use of passive monitors such as those listed in Table 6-2 does not yield a direct measure of the energy-dependent flux level at the point of interest. Rather, the activa-tion or fission process is a measure of the integrated effect that the time-and energv-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of 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: a The operating history of the reactor a The energy response of the monitor a The neutron energy spectrum at the monitor location a The physical characteristics of the monitor The analysis of the passive monitors and subsequent derivation of the average neutron flux requires completion of two procedures. First. the disintegration rate of product isotope per unit mass of monitor must be determined. Second,in order to define a suitable spectrum averaged reaction cross section. the neutron energy spectrum at the monitor location must be calculated. The specific activity of each of the monitors is determined using established ASTM procedures.[9.10.11.12.13] Following sample preparation, the activity of each monitor is determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing. the uncertainty in counting. and the acceptable error in detector calibration. For the samples removed from R. E. Ginna the overall 2o deviation in the measured data is determined to be : 10 percent. The neutron energy spectra are determined analytically using the method described in Section 6-1. l Having the measured activity of the monitors and the neutron energy spectra at the locations of interest, the calculation of the neutron flux proceeds as follows. I The reaction product activity in the monitor is expressed as f l
- N ~
d 6-2 R= fY i g(E)&(E) (1 - e b) e 1 6-8
where: R = induced product activity No = Avagadro's number A = atomic weight of the target isotope f j = weight fraction of the target isotope in the target material Y = number of product atoms produced per reaction u(E) = energy-dependent reaction cross section (E) = energy-dependent neutron flux at the monitor location with the reactor at full power Pj = average core power level during irradiation period j Pmax = maximum or reference core power level A = decay constant of the product isotope tj = lengtt; of irradiation period j td = decay time following irradiation period j Since neutron flux distributions are calculated using multigroup transport methods and, further, since the prime interest is in the fast neutron flux above 1.0 Mev, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-2) is replaced by the following relation. a(E) $(E)dE = F &(E > 1.0 Mev)
-E 6-9
l where: N a(E) $ (E)dE 9 #9
- o G1 O p x. N s 10 Mew 0 GG 9 o y,,,
Thus, equation (6-2) is rewritten N No P -At -At d 1 R= f, y o (E 1.0 Mev) (1-e 1) e A P max g1 or, solving for the neutron flux. R (E 1.0 Mev) N N P -At -At d I f, y o (1-e I) e 6-3 l The total fluence above 1.0 Mev is then given by P I
$ (E '- 1.0 Mev) = $ (E 1.0 Mev){ max t; 6-4 I
6-10 i
where: N P-P max tj = total effective fu!! power seconds of reactor operation
, up to the time of capsule removal An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered Co"(n,6)Co* data by means of cadmium ratios and the use of a 37-barn 2200 m/sec cross section. Thus, iD-1 (
pTh R bare VD \ 6-5 N No Pj
-- f, yo (1-e .A tj) e-A td A P max 11 where:
are D is defined as R Cd covered 6-4. TRANSPORT ANALYSIS RESULTS Results of the Sn transport calculations for the R. E. Ginna reactoraresummarized in Figures 6-3 through 6-8 and in Tables 6-3 through 6-5. In Figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius, 'a thickness location, and S thickness location are presented as a function of azimuthal angle. The influence of the surveillance capsules on the fast neutron flux distribution is clearly evident. In Figure 6-4, the radial distribution of maximum fast neutron flux (E > 1.0 Mev) through the thickness of the reactor pressure vessel is shown. The relative axial variation of neutron flux within the vessel is given in Figure 6-5. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in Figures 6-3 or 6-4 by the appropriate values from Figure 6-5. 6-11
in Figure 6-6 the radial variations of fast neutron flux within surveillance capsules V, R, and T are presented. These data,in conjunction with the maximum vessel flux,are used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E > 1.0 Mev) at the dosimeter block location (capsule center) to the maximum f ast neutron flux at the pressure vesselinner radius. Updated lead factors for all of the R. E. Ginna surveillance capsules are listed in Table 6-3. Since the neutron flux monitors contained within the surveillance capsules are not all located at the same radial location, the measured disintegration rates are analytically adjusted for the gradients that exist within the capsules so that flux and fluence levels may be derived on a common basis at a common location. This point of companson was chosen to be the capsule c9nter. Analytically determined reaction rate gradients for use in the adjustment procedures are shown in Figures 6-7 and 6-8 for Capsules V, R, and T. All of tne applicable fast neutron reactions are included. In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required.The neutron energy spectrum calculated to exist at the center of each of the R. E. Ginna surveil-lance capsules is given in Table 6-4. The associated spectrum-averaged cross sections for each of the five fast neutron reactions are given in Table 6-5. 6-12
TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MEV) AND LEAD FACTORS FOR R. E. GINNA SURVEILLANCE CAPSULES Capsule Azimuthal $ (E > 1.0 Mev) Lead identification Location (n/cm'-sec) Factor V 13 1.33 x 10" 3.37 R 13" 1.33 x 10 3.37 T 23' 7.66 x 10'" 1.94 P 23 7.66 x 10"> 1.94 S 33' 7.06 x 10'" 1.79 N 33 7.06 x 10's 1.79 ti-13
TABLE 6-4 CALCULATED NEUTRON ENERGY SPECTRA AT THE DOSIMETER BLOCK LOCATION FOR R. E. GINNA SURVEILLANCE CAPSULES Group Neutron Flux (n/cm2-sec) No. Capsules V & R l Capsules T & P Capsules S & N 1 7.52 x 10' 5.51 x 10' 4.84 x 10' 2 2.47 x 10' 1.83 x 10+ 1.61 x 10' 3 4.08 x 10" 2.83 x 10' 2.51 x 10' 4 4.58 x 10' 2.93 x 10' 2.65 x 10* 5 7.97 x 10" 4.78 x 10' 4.37 x 10* 6 1.56 x 10' 9.29 x 10' 8.52 x 10) 7 2.26 x 10"' 1.30 x 10' 1.20 x 10'3 8 I 3.25 x 10' 1.81 x 10' 1.68 x 10'c 9 4.30 x 10' ' 2.33 x 10' 2.16 x 10s 10 , 4.64 x 10' 2.46 x 10' 2.28 x 10" 11 j 1.54 x 10" 7.97 x 10 7.39 x 10"' ( 12 i 1.94 x 10' 9.68 x 10' ' 8.98 x 10 13 8.67 x 10'o 4.28 x 19'" 3.99 x 10'o 14 6.54 x 10'3 3.24 x 10'" 3.02 x 10'" ! 15 5.22 x 10'" 2.58 x 1C 2.41 x 10' ' 16 1.21 x 10' 5.90 x 10'" 5.51 x 10"' l 17 9.52 x 10"> 4.66 x 10"' 4.25 x 10'; 18 9.75 x 10'o 4.73 x 10'" 4.43 x 10w 19 7.74 x 10'o 3.76 x 10'" 3.52 x 10'o 20 8.59 x 10") 4.16 x 10' 3.89 x 10" l 21 2.73 x 10" 1.39 x 10" 1.25 x 10" l l l 6-14 1
TABLE 6-5 SPECTRUM AVERAGED REACTION CROSS SECTIONS AT THE DOSIMETER BLOCK LOCATION FOR R. E. GINNA SURVEILLANCE CAPSULES F (barns) Reaction Capsules V & R Capsules T & P Capsules S & N I Fe(n. p) M n O.0595 0.0683 0.0666 N "(n.p)Co
- 0.0811 0.0912 0.0893 o
Cu">(n cc)Co 0.000404 0.000517 0.000494 U2*(n,f)F.P. 0.333 0.345 0.344 Np237(n f)F.P. 2.93 l 2.80 2.82 a(E) (E)dE
, = ,
fJ (E)dE i c, .. 6-15 l
10'; 5 - 2 - 10" U - h SURVEILLANCE 5 5 - CAPSULES Z 2 - PRESSURE VESSEL IR 10 14T LOCATION 5 - l 3 4T LOCATION 2 - I l 10- 60 70 20 30 40 50 O 10 AZlMUTHAL ANGLE (DEGREES) FIGURE 6-3. CALCULATED AZIMUTHAL DISTRIBUTION OF MAXIMUM FAST NEUTRON FLUX (E > 1.0 MEV) WITHIN THE PRES VESSEL - SURVEILLANCE CAPSULE GEOMETRY 6-16
10" _ 167 64 5 - -
- 171.77 I"
175.90 2 -
^
g 174T E 180 02 5 10 g - 1 2T z O - E - 18 . '.5 35 Z 3/4T OR 2 - HO 2 3 PRESSURE VESSEL $ 10' 160 162 164 166 168 170 172 174 176 178 180 18? 184 186 188 RADIUS (cm) FIGURE 6-4. CALCULATED RADIAL DISTRIBUTION OF MAXIMUM FAST NEUTRON FLUX (E > 1.0 MEV) WITHIN THE PRESSURE VESSEL 6-17
10
- _
5 - 2 - 10-'
~
X 3 s - 3 5 - E o - Z y 2 - P 5 E 10 s - CORE MIDPLANE f 2 - _ TO VESSEL
' CLOSURE HEAD 3 ; I I I i 30
-300 -200 -100 0 100 200 300 DISTANCE FROM CORE MIDPLANE (cm)
FIGURE 6-5. RELATIVE AXIAL VARIATION OF FAST NEUTRON FLUX l (E > 1.0 MEV) WITHIN THE PRESSURE VESSEL 6-18
10 5 _ 157.33 158 10 U2 - y - 158 33 b
,E_.
10 CAPSULE p _ V&R h FRONT - g - MONITORS
$5 -
DOSIMETER - BLOCK CAPSULE T RE AR - MONITORS
~
2 THERMAL CAN CAN SHIELD H2O v g TEST SPECIMENS j HO 2 10'o I I- ! I I C 157 158 159 160 161 155 156 R ADIUS (cm) FIGURE 6-6. CALCULATED RADIAL DISTRIBUTION OF MAXIMUM FAST NEUTRON FLUX (E > 1.0 MEV) WITHIN SURVEILLANCE CAPSULES V, R, AND T t 6-19
10' _ 157;33 158.10
- l } 158.33
~ Ni"(N,P)Co*
~
l
- 1l Np*(N.F)Cs' p _
10 _
~
3l cn ws l
>5 - l {
s U ~(N.F)Cs' j g
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$ Fe"(N.P)Mn
- E k Y C 2 N C' R c 35
< 2 m+
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g 5g
- o y7 E se
- Di <
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i l ' 1 l Cu tN..t)Co' 2 - CAN CAN THERMAL SHIELD f H2O ? $ TEST SPECIMENS l q H2O 10' 159 160 161 156 157 158 155 R ADIUS (cm) FIGURE 6-7. CALCULATED VARIATION OF FAST NEUTRON FLUX MONITOR SATURATED ACTIVITY WITHIN CAPSULES V AND R l 6-20
2 - l 157.33 i 158.10 l 158.33 2 Ni>8(N P)Co'*
- l
. 5 - l1
~
N p> "(N,F)C s 2 - O D cn b 10 -
- II '
5 _ U2e(N F)Cs P 1 O 5 - I i
< I 0
w N - Fe(N.P)Mn ' c $ Y O O Om c F O O 2 - Q E dt m O 2 z z w O E w 2 O E 2E OE C 5 ll Cu*'(N,a)Co'" 2 - THERMAL CAN CAN SHIELD gHO; 2 , EST SPECIMENS $ ; HO 2 j ; - l l 5 $ I I I ' ' 10' 155 156 157 158 159 160 161 H A01US (cm) FIGURE 6-8. CALCULATED VARIATION OF FAST NEUTRON FLUX
- MONITOR SATURATED ACTIVITY WITHIN CAPSULE T 6-21
~
6-5. DOSIMETRY RESULTS The irradsation history of the R. E.Ginna reactor is given in Table 6 6. Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsules V, R, and T are listed in Tables 6-7,6-8, and 6-9, respectively.The data are presented as measured at the actual monitor locations as well as adjusted to the capsule center. The measured results for Capsules V, R and T were obtained by West-inghouse. All adjustments to the capsule center were based on the data presented in Figures 6-7 and 6-8. The fast neutron (E > 1.0 Mev) flux and fluence levels derived for Capsules V, R and T are presented in Table 6-10. The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in Table 6-11. Due to the relatively low thermal neutron flux at the capsule locations, no burnup correction was made to any of the measured activities. The maximum error introduced by this assumption is estimated se to be less than 1 percent for the Nise(n,p)Co reaction and even less significant forall of the other fast neutron reactions. Using the iron data presented in Table 6-10, along with the lead factors given in Table 6-3, ths fast neutron fluence (E > 1.0 Mev) for Capsules V. R, and T as well as for the reactor vessel inner diameter are summarized in Table 6-12 and Figure 6-9. The agreement between calculation and measurement is excellent, with measured fluence levels of 7.03 x 10's,1.01 x 105, and 1.75 x 10'S compared to calculated values of 6.88 x 10's,1.07 x 10, and 1.75 x 10 n/cm2 for Capsules V, R, and T, respectively. It should be noted that the fluence levels for Capsules V and R differ from those previously reported. This difference is the result of an upde".ed '; valuation of dosimetry from Capsules V and R using the latest analytical ,echniques. The graphical representation in Figure 6-9 indicates the accmocy of the transport analysis for R. E. Ginna and supports the use of the analytically determined fluence trend curve for predicting vessel toughness at times in the future. Projecting to end-of-life, a summary of peak fast neutron exposure of the R. E.Ginna reactorasderived from both calculation and measurement may be made as follows. Fast Neutron Fluence (n/cm') Surface 1/4 T 3/4 T 4.08 x 10'S 2.72 x 10 O.02 x 10'8 Capsule V Capsule R 3.75 x 10" 2.50 x 10 7.37 x 10'8 Capsule T 4 25 x 10 2.83 x 10'S 8.34 x 10'8 Average measurement 4 03 x 10* 2.68 x 10* 7.91 x 10'8 , 3.96 x 10 2.64 x 10" 7.78 x 10'8 Calculation These data are based on 32 full-power years of operation at 1520 MWt. 6-22
Based on the capsule withdrawal schedule identified in ASTM E185-79 it is recom-mended that the remaining capsules be removed from the reactorin accordanc,e with the following schedule. Capsule Capsule Vessel Lead Removal Fluence Identity Location Factor Time (n/cm2) P 247 1.94 17EFPY 4.10 x 10 S 57 1.79 STANDBY N 237 1.79 STANDBY [a.) Approximate fluence at inner surface location at End-of-Life 6-23
TABLE 6-6 1RRADIATION HISTORY OF R. E. GINNA REACTOR VESSEL SURVEILLANCE CAPSULES Irradiation Decay Pj P max Time Time Month (MW) (MW) Pj/ Pmax [a] (days) (days) 7/70 479 1520 .3156 273 3914 8/70 1288 1520 .8487 31 3883 9/70 1231 1520 .8107 30 3853 10/70 666 1520 4385 31 3822 11/70 1187 1520 .7822 30 3792 12/70 1163 1520 .7663 31 3761 1/71 1136 1520 .7483 31 3730 2/71 1446 1520 .9523 28 3702 3/71 37 1523 .0246 31 3671 4/71 0 30 3641 5/71 452 1520 1520 l .2976 0 0000 31 3610 3580 6/71 1173 1520 i .7730 30 7/71 1141 1520 l .7516 31 3549 8/71 1259 1520 ; .8295 31 3518 9/71 1289 1520 .8489 30 3488 10'71 1307 1520 .8609 31 3457 11/71 1349 1520 .8890 30 3427 12471 1286 1520 .8470 31 3396 1/72 1313 1520 .8653 31 3365 2/72 1401 1520 .9232 29 3336 3/72 1016 1520 6692 31 3305 4/72 929 1520 .6122 30 3275 5/72 0 1520 0.0000 31 3244 6'72 11 1520 .0072 30 3214 7/72 1123 1520 7400 31 3183l 8/72 1233 1520 .8121 31 3152 9/72 1292 1520 .8511 20 l 3122 10'72 453 1520 .2982 31 3091 11/72 542 .3568 30 3061 1520 l j 3030 12/72 1447 1520 .9531 31 1/73 1217 1520 .8017 31 2999 2/73 1296 1520 .3539 28 2971 3/73 1295 1520 l .8529 31 2940l - [a] = Average Core Power Level _PJ P MAX Maximum Reference Core Power Level 6-24
TABLE 6-6 CONTINUED 1RRADIATION HISTORY OF R. E. GINNA REACTOR VESSEL SURVEILLANCE CAPSULES Irradiation Decay Pj P max [a] Time Time Month (MW) (MW) Pjf Pmax (days) (days) 4/73 1298 1520 .8552 30 2910 5/73 1301 1520 .8573 31 2879 6/73 1287 1520 .8478 30 2849 7/73 932 1520 .6140 31 2818 8/73 1299 1520 .8557 31 2787 9/73 1351 1520 .8897 30 2757 10/73 1124 1520 .7405 31 2726 11/73 1391 1520 .9163 30 2696 12/73 1268 1520 .8354 31 2665 1/74 4 1520 .0026 31 2634 2/74 0 1520 0.0000 28 2606 3/74 0 1520 0.0000 31 2575 4/74 118 1520 .0776 30 2545 5/74 1027 1520 .6765 31 2514 6/74 830 1520 .5468 30 2484 7/74 904 1520 .6353 31 2453 8/74 1208 1520 .7957 31 2422 9/74 1400 1520 .9224 30 2392 10/74 1428 1520 .9406 31 2361 11/74 777 1520 .5117 30 2331 12/74 1486 1520 .9791 31 2300 1/75 1518 1520 1.0000 31 2269 2/75 1518 1520 1.0000 28 2241 3/75 442 1520 .2911 31 2210 4/75 0 1520 0.0000 30 2180 5/75 190 1520 .1248 31 2149 6/75 858 1520 .5653 30 2119 7/75 1462 1520 .9633 31 2088 8/75 1505 1520 .9912 31 2057 9/75 1518 1520 1.0000 30 2027 10/75 1431 1520 .9428 31 1996 11/75 1518 1520 1.0000 30 1966 12/75 1420 1520 .9354 31 1935 (a) Pj _ Average Core Power Level P MAX Maximum Reference Core Power Level 6-25
TABLE 6-6 CONTINUED IRRADIATION HISTORY OF R. E. GINNA REACTOR VESSEL SURVEILLANCE CAPSU!.ES Irradiation Decay Pj Pmax [a] Time Time Month (MW) (MW) PjfP max (days) (days) 1/76 721 1520 .4748 31 1904 2/76 0 1520 0.0000 29 1875 3/76 0 1520 0.0000 31 1844 4/76 355 1520 .2340 30 1814 5/76 906 1520 .5970 31 1783 6/76 1518 1520 1.0000 30 1753 7/76 1478 1520 .9737 31 1722 8/76 200 1520 .1316 31 1691 9/76 910 1520 .5996 30 1661 10/76 396 1520 .2610 1 31 1630 11/76 1345 1520 .8857 30 1600 12/76 1260 1520 .8300 31 1569 1/77 1324 1520 .8724 31 1538 2/77 1339 1520 .8819 28 1510 3'77 1334 1520 .8788 31 1479 4/77 665 1520 .4379 30 1449 5/77 228 1520 .1499 31 1418 6 77 1335 1520 .8797 30 1388 7/77 955 1520 .6293 31 1357 8/77 1154 1520 .7604 31 1326 9/77 1327 1520 8740 30 1296 10/77 1331 1520 .8766 31 1265 11/77 1159 1520 .7638 30 1235 12/77 1296 1520 8539 31 1204 1/78 1042 1520 6866 31 1173 2/78 1170 1520 .7706 28 1145 l l 3/78 1038 1520 .6839 31 1114 4/78 0 1520 0.0000 30 1084 5478 324 1520 .2135 31 1053 6:78 1518 1520 1.0000 30 1023 7/78 1512 1520 9958 31 992 8/78 1464 1520 .9643 31 961 9'78 1518 1520 1.0000 30 931 10/78 1518 1520 1.0000 31 900 [a] Pj _ Average Core Power Level P MAX Maximum Reference Core Power Level 6-26
TABLE 6-6 CONTINUED
.. l IRRADIATION HISTORY OF R. E. GINNA l REACTOR VESSEL SURVEILLANCE CAPSULES l l
1 Irradiation Decay Pj Pmax [a] Time Time Month (MW) (MW) Pj/ Pmax (days) (days) 11/78 1518 1520 1.0000 30 870 12/78 1475 1520 .9714 31 839 1/79 1518 1520 1.0000 31 808 2/79 443 1520 .2917 28 780 3/79 0 1520 0.0000 31 749 4/79 1224 1520 .8060 30 719 5/79 1518 1520 1.0000 31 688 6/79 1518 1520 1.0000 30 658 7/79 294 1520 .1937 31 627 8/79 1286 1520 .8473 31 596 9/79 1518 1520 1.0000 30 566 10/79 1487 1520 .9797 31 535 11/79 1483 1520 .9771 30 505 12/79 657 1520 .4331 31 474 1/80 1518 1520 1.0000 _ 31 443 2/80 1518 1520 1.0000 29 414 3/80 1243 1520 .8191 31 383 [a] Pj = Average Core Power Level P MAX Maximum Reference Core Power Level 6-27
TABLE 6-7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE V Saturated Activity Adjusted Saturated Activity Radial (dis [ dis /s) Reaction and Location 9 /sj
/ \ 9 /
Axial Location (cm) Capsule V Calculated Capsule V Calculated Fe"(n.p)Mn" W-1 157.33 6.27 x 10" 6.11 x 10'- 5.34 x 106 R-1 157.33 6.52 x 10 - 6.11 x 10 - 5.55 x 10' S-6 157.33 5.53 x 10'- 6.11 x 10- 4.71 x 10 P-7 158.33 6.52 x 10- 6.11 x 10 5.55 x 104 W-2 158.33 5.05 x 10" 4.95 x 10 5.30 x 10' R-3 158.33 4.83 x 10" 4.95 x 10' 5.07 x 10' S-8 158.33 5.00 x 10' 4.95 x 10- 5.25 x 10-P-9 158.33 5.20 x 10' 4.95 x 10' 5.46 x 10+ Average 5.28 x 10' 5.20 x 10' Cu"( N ,a ) Co' ' Top 158.33 3.86 x 10' 3.39 x 10 - 4.04 x 10' Mid-top 158.33 3.53 x 10' 3.39 x 10 3.70 x 105 Mid-bottom 158.33 3.92 x 10' 3.39 x 10 4.10 x 10' Bottom 158.33 4.25 x 10' 3.39 x 10 4.45 x 10 - Average 4.07 x 10' 3.55 x 10 Ni^(n.p)Co " Middle 158.33 8.21 x 10 7.30 x 10 8.66 x 10- 7.70 x 10 U'(n.f)Cs' Middle 158.10 6.13 x 10" 7.10 x 10" 6.13 x 10- 7.10 x 10- _ 6-28
TABLE 6-8 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE R i Saturated Activity Adjusted Saturated Activity Radial dis /s Reaction and Location (dis 9 //s} 9 Axial Location (cm) Capsule R Calculated Capsule R Calculated Fe"(n.p)M n" W-13 157.33 5.82 x 10 - 6.11 x 10" 4.95 x 10-R-14 157.33 5.53 x 10" 6.11 x 106 4.71 x 104 P-18 157.33 5.76 x 10- 6.11 x 10" 4.90 x 10-W-14 158.33 4.37 x 10' 4.95 x 10" 4.59 x 10' R-15 158.33 4.55 x 10 4.95 x 10" 4.78 x 10' P-19 158.33 4.96 x 10" 4.95 x 10" 5.21 x 10-Average 4.86 x 10' 5.20 x 10' C u '( N ,a) Co* Top 158.33 4.18 x 10- 3.39 x 10- 4.38 x 10' Mid-top 158.33 3.75 x 10' 3.39 x 10' 3.93 x 10' Mid-bottom 158.33 4.46 x 10 3.39 x 10'> 4.67 x 10' Bottom 158.33 4.46 x 10 3.39 x 10 > 4.67 x 105 Average 4.41 x 105 3.55 x 10-NiSa(n,p)Co68 Middle 158.33 5.93 x 10' 7.30 x 107 6.26 x 107 7.70 x 10' Np"7(n.f)Cs'37 Middle 158.10 7.23 x 10 6.45 x 10 7.23 x 10' 6.45 x 10' U"8(n.f)Cs'3' Middle 158.10 7.35 x 10'> 7.10 x 10" 7.35 x 106 7.10 x 10'> 6-29
TABLE 6-9 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVIT!ES FOR CAPSULE T Saturated Activity Adjusted Saturated Activity Radial dis /s} Reaction and Location (dis 9 /s} d 9 4 Capsule T Calculated Capsule T Calculated Axial Location (cm) Fe5'(n,p)Mn5d 157.33 3.40 x 106 3.29 x 106 3.28 x 106 S-22 157.33 3.79 x 106 3.29 x 106 3.66 x 106 P-28 157.33 3.88 x 106 3.29 x 106 3.75 x 106 W-21 158.33 3.01 x 10' 4.05 x 106 3.59 x 106 S-23 158.33 3.08 x 10o 4.05 x 10 3.66 x 10+ P-29 158.33 3.29 x 106 4.05 x 10' 3.91 x 106 W-22 3.64 x 106 3.41 x 10+ Average Cu"(N A)Cow 158.33 3.59 x 10' 2.52 x 10 3.39 x 10' Top 158.33 3.14 x 10' 2.52 x 10' 2.97 x 10' Mid-top 158.33 3.73 x 10' 2.52 x 10' 3.52 x 10' Mid-bottom 158.33 3.91 x 10' 2.52 x 10 - 3.69 x 105 Bottom 3.39 x 105 2.67 x 10' Average Ni$8(n.p)Co'8 4.16 x 10" 4.70 x 10 3.95 x 10' 4.95 x 10' Middle 158.33 Npm(n.f)CS'37 158.10 4.44 x 10' 3.40 x 10' 4.44 x 10~ 3/0 x 10' Middle U '"(n.f ) Cs'" 4.23 x 10" 158.10 5.41 x 10' 4.23 x 10' 5.41 x 10 Middle 6-30
TABLE 6-10 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULES V, R, AND T Adjusted Saturated Activity fdis/s) $(E > 1.0 Mev) (E > 1.0 Mev) ( 9 / (n/cm2-sec) (n/cm2) Capsule Reaction Measured Calculated Measured Calculated Measured Calculated V Fe5'(n p)Mn5' 5.28 x 105 5.20 x 105 1.36 x 10" 1.33 x 10" 7.03 x 10 6.88 x 10 CuS3(n.a)CoSo 4.07 x 105 3.55 x 105 1.52 x 10" 7.86 x 10 m Ni58(n.p)Co58 8.66 x 107 7.70 x 107 1.51 x 10" 7.81 x 10's $ Np23'(n f)Cs U23'(n,f)Cs'2' 6.13 x 105 7.10 x 105 1.15 x 10" 5.94 x 10 R Fe5'(n,p)Mn5* 4.86 x 105 5.20 x 108 1.25 x 10" 1.33 x 10" 1.01 x 10 1.07 x 10 CuS3(n a)Co5o 4.41 x 105 3.55 x 105 1.65 x 10" 1.33 x 10 NiS8(n.p)Co58 6.26 x 10' 7.70 x 10' 1.10 x 10" 8.87 x 10 Np237(n,f)Cs'3' 7.23 x 10' 6.45 x 10' 1.49 x 10" 1.20 x 10 U23a(n.f)Cs'37 7.35 x 108 7.10 x 108 1.38 x 10" 1.11 x 10" T Fe5d(n p)Mn54 3.64 x 108 3.41.x 108 8.16 x 10'o 7.66 x 10'o 1.75 x 10 1.65 x 10 Cu"3(n.a)Co** 3.39 x 105 2.67 x 105 9.92 x 10'o 2.13 x 10 Ni58(n.p)CoS8 3.95 x 107 4.95 x 10' 6.16 x 10'o 1.32 x 10 Np2 '(n,f)Cs' 4.44 x 107 3.40 x 10' 9.59 x 10'o 2.07 x 10 U23e(n,f)Cs'3' 5.40 x 108 4.23 x 108 9.83 x 10'o 2.11 x 10
~
4
- i l-TABLE 6-11 -
e RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE T-i _< q
~ m. -
j l
~~~I -
Saturated Activity (dis 9 /s Th N Axial Location Bare Cd-covered (n/cm2-sec) Top 3.17 x 10' 1.21' x 10 7.77 x'^1.0' Middle top 3.06 x 10- 1.13 x 10' 7.65 x 1C ,; Middle 3.03 x 10 1.16 x 10 7.41 x 10'^ Bottom middle 3.27 x 10' 71.26 x 10' 7.97 x 10"' 's 1. Bottom 3.07 x 10 1.20 x 10 7 41 x 10'" Average _ ______, 7.64 x 10 ' J g
% y
~
y 2; N ,
sg
?
~
4 -
\
k I
,)
s .*% m
$ =
6-32 . _ _ _ _ _ _ _ . . _ _ w a
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e y t ,?b - i (, '
; _ . /'
,y t
TABLE 6-12 <
SUMMARY
OF NEUTRON DOSil Y RESULTS FOR CAPSULES V, R, AND T O I 6 Irradiation:
- s Vessel Calculated 9, Time &(E >1.0 Mev) &(E > 1.0 Mev)' Lead Floence r i Vessel Fluence
/ -
Capsule (EFPS)- (n/cm'-sec) (n/cm2) Factor ' (n/cm2) (n/cm2) V '5.17 x 10' l 1.3'6 x-10" 7.03 x 10's 3.37 2.08 x 10'8 - 2.04 x 10ie R 8.06 x 10' 1.25 x 10 1.01 x 10'S , 3.37 3.00 x 10" 3.18 x 10is T 2.15 x 108 8.16 x 10'o 1.75 x 10 ' 1.94 9.04 x 10'8 8.51 x 10'8
I 1&'
- - - - O CAPSULE V DjtTA O CAPSULE,Y DATA 5
- A CAPSULE.R D*TA
- SN C'ALLULATION P6 10^
r I"- ~ 13' CAPSULES (V, R) b 5 N E - e Z \ O f2 a b 23' CAPSULES (T) y 10 " - VESSEL INNER R ADIUS 2 5
~
2 10
! ! ! ! ! I! I I l l l l l l ll 5 10 20 50 100 1 2 OPERATING TIME (EFPY)
FIGURE 6-9. COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUENCE (E > 1.0 MEV) FOR CAPSULES V, R, AND T t 6-34
REFERENCES
- 1. Yanichko, S. E.," Rochester Gas and Electric Robert E. Ginna Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-7254, May 1969.
- 2. ASTM Designation E185-66, " Surveillance Tests on Structural Materials in Nuclear Reactors" in " ASTM Standards (1967), Part 31, Physical and Mechanical Testing of Metals - Metallography. Nondestructive Testing, Fatigue. Effect of Temperature / pp. 638-642. Am. Soc. for Testing and Materials Philadelphia, Pa.,1967.
- 3. Mager, T. R., et al," Analysis of Capsule V from the Rochester Gas and Electric R. E. Ginna Unit No.1 Reactor Vessel Radi uion Surveillance Program," Westing-house Nuclear Energy Systems - FP-R A-1 (April 1.1973).
- 4. Yanichko, S. E., et al. " Analysis of Capsule R from the Rochester Gas and Electric Corporation R. E. Ginna Unit No.1 Reactor Vessel Radiation Surv< ilance Program," WCAP-8421. Nosember,1974.
- 5. Soltesz, R. G., R. K. Disney, J. Jedruch. and S. L. Zeigler, " Nuclear Rocket Shielding Methods. Modification, Updating and input Data Preparation -
Volume 5 - Two-Dimensional, Discrete Ordinates Transport Technique " WANL-PR-(LL)-034. Vol. 5. August 1970.
- 6. Collier, G., G. Gibson. L. L. Moran, R. K. Disney, and R. S. Kaiser, "Second Version of the GAMB1T Code, WANL-TME-1969, November 1969.
- 7. Soltesz, R. G., R K. Disney, S. L. Zeigler, " Nuclear Rocket Shielding Methods.
Modification, Updating and input Data Preparation - Volume 3, Cross-Section Generation and Data Processing Techniques." WANL-PR-(LL)-034, August 1970.
- 8. Soltesz, R. G. and R. K. Disney," Nuclear Rocket Shielding Methods, Modifica-tion, Updating and input Data Pre,paration - Volume 4 - One-Dimensional Discrete Ordinates Transport Technique," WANL-PR-(LL)-034, August 1970.
A-1 l L._ . _ _ _ __ _ _ _
w
~
- 9. ASTM Designation E261-70," Standard Method for Measuring Neutron Flux b'y Radioactivation Techniques," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 745-755, American Society for Testing and Materials, Phila-delphia, Pa.,1975.
10,~ ASTM Designation E262-70," Standard Method for Measuring Thermal Neutron , Flux by Radioactivation Techniques " in ASTM Standards (1975), Part 45. Nuclear Standards, pp. 756-763, American Society for Testing and Materials, Philadelphia, Pa.,1975.
- 11. ASTM Designation E263-70. IStandard Method for Measunng Fast-Neutron Flux by Radioactivation of Iron." in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 764-769, American Society for Testing and Materials. Phila-delphia, Pa.,1975.
- 12. ASTM Designation E481-73T, "Tentatc. Method of Measuring Neutron-Flux Density by Radioactivation of Cobaltand Silver,"in ASTM Standards (1975), Part
- 45. Nuclear Standards. pp. 637-894, American Society f or Testing and Materials.
Philadelphia. Pa.,1975.
- 13. ASTM Designation E264-70. " Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel,"in ASTM Standards (1975), Part 45. Nuclear Standards, pp. 770-774, Amencan Society for Testing and Materials. Phila-dolphia, Pa.1975.
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