ML20010A615

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Analysis of Capsule R from Wi Public Svc Corp Kewaunee Nuclear Plant Reactor Vessel Radiation Surveillance Program.
ML20010A615
Person / Time
Site: Kewaunee Dominion icon.png
Issue date: 03/31/1981
From: Shaun Anderson, Mager T, Yanichko S
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML111660108 List:
References
WCAP-9878, NUDOCS 8108110609
Download: ML20010A615 (88)
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ANALYSIS OF CAPSULE R FROM THE WISCONSIN PUBLIC SERVICE CORPORATION KEWAUNEE NUCLEAR PLANT

.; REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM (WCAP-9878)

EPRI RESEARCH PROJECT 1021-3 2 TOPICAL REPORT March 1981
Prepared by

! WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvania 15230 l

T. R. Mager, Principal Investigator Prepared for ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue Palo Alto, California 94304 T. U. Marston, Project Manager l fD 8110609 810807 i

j p ADOCK 05000305 j PDR 1 J

1 ANALYSIS OF CAPSULE R FROM THE WISCONSIN PUBLIC SERVICE CORPORATION KEWAUNEE NUCLEAR PLANT REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM (WCAP-9878)

EPRI RESEARCH PROJECT 1021-3 TOPICAL REPORT S. E. Yanichko S. L. Anderson R.P.Shogan R. G. Lott j

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I March 1981 1

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Prepared by l

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! WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Divicion f

/ P. O. Box 355 Pit burgh, Pennsylvania 15230 l

l j T. R. Mager, PrincipalInvestigator f

f Prepared for i

ELECTRIC POWER RESEARCH INSTITUTE j

j 3412 Hillview Avenue Palo Alto, California 94304

, I T. U. Marston, Project Manager f

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LEGAL NOTICE This report was prepared by Westinghouse Electric Corporation (WESTINGHOUSE) as an account of work sponsored by the Electric 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 accu-racy, 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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TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 1-1 2 IN i?.ODUCTION 2-1 3 BACKGROUND 3-1 4 DESCRIPTION OF PROGRAM 4-1 5 TESTING OF SPECIMENS FROM CAPSULE R 5-1 5-1. Overview 5-1 5-2. Charpy V-Notch impact Test Results 5-2 5-3. Tensile Test Results 5-24 5-4. Wedge Opening Loading Tests 5-32 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 6-10 6-5. Posimetry Results 6-18 References A-1 iii m . _

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L LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in Kewaunee Reactor Vessel (Updated Lead Factors

! for the Capsules Are Shown in Parentheses) 4-2 4-2 Schematic Diagram of Capsule 3 Showing Location of Spectnens. Ther -al Monitors, and Dosimeters 4-3 5-1 Charpy V-Notch Impact Data for Kewaunee Reactor Vessel Shell Forging 122X208VA1 5-6 5-2 Charpy V-Notch impact Data for Kewaunee Reactor Vessel Shell Forging 123X167VA1 5-7 5-3 Charpy V-Notch Impact Data for Kewaunee Reactor Vessel Weld Metal 5-8 5-4 Charpy V-Notch Impact Data for Kewaunee Reactor s Vessel Weld HAZ Metal 5-9 5-5 Charpy V-Notch Impact Data for A533 Grade B Class 1 ASTM Correlation Monitor Material 5-10 5-6 Charpy impact Specimen Fracture Surfaces for Kewaunee Intermediate Shell Forging 122X208VA1 5-19 5-7 Charpy impact Specimen Fracture Surfaces for Kewaunee Lower Shell Forging 123X167VA1 5-20 5-8 Charpy impact Specimen Fracture Surfaces for Kewaunee Weld Metal 5-21 5-9 Charpy impact Specimen Fracture Surfaces for Kewaunee HAZ Material 5-22 5-10 Charpy impact Specimen Fracture Surfaces for Kewaunee ASTM Correlation Monitor Material 5-23 i

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LIST OF ILLUSTRATIONS (cont)

Figure Title Page 5-11 Comparison of Predicted Versus Actual 41-Jmule Transition Temperature increases for Kewaunee Reactor Vessel Materials 5-26 5-12 Tensile Properties for Kewaunee Reactor Vessel Shell Forging 122X208VA1 5-29 5-13 Tensile Properties for Kewaunee Reactor Vessel Shell Forging 123X167VA1 5-30 5-14 Tensile Properties for Kewaunee Reactor Vessel Weld Metal 5-31 5-15 Typical Stress-Strain Curve for Tension Specimens 5-33 5-16 Fractured Tensile Specimens From Kewaunee intermediate Shell Forging 122X208VA1 5-34 5-17 Fractured Tensile Specimens From Kewaunee Lower Shell Forging 123X167VA1 5-35 5-18 Fractured Tensile Specimens From Kewaunee WeM Metal 5-36 6-1 Kewaunee Reactor Geometry 6-2 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-4 6-3 Calculated AzimuthalDistribution of Maximum Fast Neutron Flux (E > 1 Mev) Within the Pressure Vessel Surveillance Capsule Geometry 6-11 6-4 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1 Mev) Within the Pressure Vessel 6-12 vi

LIST OF ILLUSTRATIONS (cont)

Figure Title Page 6-5 Relative AxialVariation of Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-13 6-6 Calculated RadialDistribution of Maximum Fast Neutron Flux (E > 1 Mev) Within the Surveillance Capsules 6-14 6-7 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules V and R 6-15

" 6-8 Comparison of Measured and Calculated Fast Neutron Fluence (E > 1 Mev) for Capsules V and R 6-27 I

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t LIST OF TABLES Table Title Page 4-1 Chemistry and Heat Treatment of Material Representing the Core Region Shell Forgings and Weld Metal From Kewaunee Reactor Vessel 4-5 4-2 Chemistry and Heat Treatment of Surveillance Material Representing 12-inch-Thick A533 Grade B Class 1 ASTM Correlation Monitor Material From HSST Plate 02 4-6 -

5-1 Charpy Impact Data for Kewaunee Reactor Pressure Vessel Shell Forgings 5-3 5-2 Charpy impact Data for Kewaunee Reactor Pressure Vessel Weld Metal and HAZ Material 5-4 5-3 Charpy impact Data for the ASTM Correlation Monitor Material (HSST Plate 02) 5-5 5-4 Instrumented Charpy impact Test Results for Kewaunee Shell Forgings 5-11 5-5 Instrumented Charpy impact Test Results for Kewaunee Weld Metal and HAZ Material 5-13 l

5-6 Instrumented Charpy impact Test Results for the ASTM Correlation Monitor Material 5-15 5-7 The Effect of 288 C Irradiation to 2.07 x 1019 n/cm2 (E > 1 Mev) on the Notch Toughness Properties of Kewaunee Reactor Vessel Surveillance Test Material 5-17 5-8 Summary of Kewaunee Reactor Vessel Surveillance Capsule Charpy impact Test Results 5-25 5-9 Tensile Properties for Kewaunee Reactor Vessel MaterialIrradiated to 2.07 x 1019 n/cm2 (E > 1 Mev) 5-27 I

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c LIST OF TABLES (cont)

Table Title Pago 6-1 21 Grour Energy Structure 6-5 6-2 Nuclear Parameters for Neutron Flux Monitors 6-7 6-3 Ca . fated Fast Neutron Flux (E > 1 Mev) and Lead Factors for Kewaunee Surveillance Capsules 6-16 6-4 Calculated Neutron Energy Spectra at the Center of Kewaunee Surveillance Capsules 6-17 6-5 Spectru:s ^."eraged Reaction Cross Sections at the Dosimete r Block Location for Kewaunee Surveillance Capsules 6-18 6-6 trradiation History of Kewaunta :leactor Vessel Surveillance Capsules 6-19 6-7 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule V F _2 6-8 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule R 6-23 l 6-9 Results of Fast Neutron Dosimetry for Capsules V and R 6-25 6-10 Results of Thermal Neutron Dosimetry for Capsules V and R 6-26 6-11 Summary of Fast Neutron Dosimetry Results Capsules V and R 6-28

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SECTION 1

SUMMARY

OF RESULTS The analyses of the reactor vessel material contained in the second surveillance capsule, designat-ed R from the Wisconsin Public Service Corporation Kewaunee nuclear plant reactor pressure vessel, led to the f ollowing conclusions:

a The capsule received an average fast fluence of 2.d7 x 1019 n/cm2 (E > 1 Mev) compared to a calculated value of 2.06 x 1019 n/cm2, a Based on the fluence measurements for Capsule R, the vessel 1/4 thickness fluence af ter 4.5 effective full power years of operation is 3.77 x 1018 n/cm2 compared to a calculated fluence of 3.75 x 1018 n/cm2, a Results of Charpy impact tests for the two Kewaunee capsules tested to date indi-cate that radiation damage has not satursted.

m End-of-life projected fast neutron fluences for the reactor vessel, based on 32 full-power years of operation at 1650 MW, are as follows:

Fast Neutron Fluence (n/cm2) t Vessel Location Calculated Measured Inner surface 4.34 x 1019 4.47 x 1019 1/4 thickness 2.83 x 1019 2.91 x 1019 3/4 thickness 8.53 x 1018 8.78 x 1018 l

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SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule R, the second capsule of the con-tinuing surveillance program which monitors the effects of neutron irradiation on the Wisconsin Public Service Corporation Kewaunee Nuclear Plant reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Kewaunee reactor pressure vessel materials was designed and recommended by the Westinghouse Electri : Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented in WCAP-8017.Ill The surveillance program vias planned to cover the 40-year life of the reactor pressure vessel and was based on ASTM E-185-70, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels.[2]

This report summarizes testing and the postirradiation data obtained from the second material sur-veillance capsule (Capsule R) remos ad from the Kewaunee reactor vessel and discusses the analy-sis of these data. The data rre also compared to results of the previously removed Kewaunee Capsule V, reported in W6AP-8908.[3]

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i SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vesselis the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the me-chanical properties of low allny ferritic pressure vessel steels such as A508 Class 2 (base material of the Kewaunee reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties 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 111 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 is 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 temperature (or 35-miliateral expan-

! sion if this is greater) temperature as determined from Charpy specimens oriented normal to the working direction of the material. The RTNDT of a given materialis used to index that material to a reference stress intensity factor curve (KIR curve) which appearsin Appendix G of the ASME Code.

The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results ob-

.ained from several heats of pressure vessel steel.When a given materialis indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of tempera-ture. Allowable operating ihnits can then be determined utilizing these allowable stress intensity factors.

RTNDT and,in turn, the operating limits of nuclear power plants can be adjusted to account for the ef fects of radiation on the roactor 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 Kewaunee Reactor Vessel Radiation Surveillance Program)ll in which a surveillance capsule is periodically removed from the operating nuclear 3-1

reactor and the encapsulated specimens are tested. The increase in the Charpy V-notch 50 f t Ib temperature (AR TNDT) due to irradiation is added to the original RTNDT to adjust the RTNDTf 0f radiation embrittlement. This adjusted RTNDT (RTNDT initial + ARTNDT) is used to index the mate-rial to the KIR curve and,in turn, to set operating limits for the nuclear power plant which take into account the ef fects of irradiation on the reacto- vessel materials.

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SECTION 4 DESCRIPTION OF PROGRAM f Six surveillance capsules for monitoring the ef fects of neutron exposure on the Kewaunee reactor p; essure vessel core region material were inserted in the reactor vessel prior to initial plant startup.

Tho six capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in figurs 4-1. The vertical center of the capsules is opposite the vertica' center of the core.

Capsule R was removed af ter 4.5 effective full power years of plant operation. This capsule con-l tained Charpy V-notch impact, tensile, and WOL specimens (figure 4-2) from the intermediate and lower shell ring forgings and weld metal representative of the core region of the reactor vessel and Charpy V-notch specimens from weld heat-affected zone (HAZ) material. The capsule also con-tained Charpy V-notch specimens from the 12-inch-thick ASTM correlation monitor material (A533 Grade B Class 1). The chemistry and heat treatment of the surveillance material are present-ed in tables 4-1 and 4-2.

All test specimens were machined from the 1/4 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 direc-tion of the forgings.

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 longitudi-nal axis of the specimen parallel to the welding direction.

Capsule R contained dosimeter wires of pure iron, copper, nickel, and aluminum-cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Np237 and U238 were contained in the capsule and located as shown in figure 4-2.

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TABLE 4-1 CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE CORE REGION SHELL FORGINGS AND WELD METAL FROM KEWAUNEE REACTOR VESSEL CHEMICAL ANALYSIS (wt %)

Forging Forging Weld Element 122X208VA1 123X167VA1 Metal m

C 0.21 0.20 0.12 Si O.25 0.28 0.20 Mo 0.58 0.58 0.48 Cu O.06 0.06 0.20 Ni O.71 0.75 0.77 Mn O.69 0.79 1.37 Cr 0.40 0.35 0.090 V <0.02 <0.02 0.002 Co 0.011 0.012 0.001 Sn 0.01 0.01 0.004 Ti <0.001 <0.001 <0.001 Zr 0.001 0.001 <0.001 As 0.001 0.004 0.004 Sb <0.001 0.001 0.001 S 0.011 0.009 0.011 P 0.010 0.010 0.016 Al O.004 0.006 0.010 B <0.003 <0.003 <0.003 N2 0.006 0.010 0.012 Zn - -

<0.001 HEAT TREATMENT Intermediate Shell Heated at 1550 F for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, water-quenched Forging Heat 122X208VA1 Tempered at 1230*F for 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, air-cooled Stress-relieved at 1150'F for 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, furnace-cooled Lower Shell Forging Heated at 1550 F for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, water-quenched Heat 123X167VA1 Tempered at 1220 F for 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, air-cooled Stress-relieved at 1150*F for 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, furnace-cooled Submerged ArcWeldment Stress-relieved at 1150 F for 19-1/4 hours, furnace-cooled 4-5

TABLE 4-2 CHEMISTRY AND HEAT TREATMENT OF SURVElLLANCE MATERIAL

, REPRESENTING 12-INCH-THICK A533 GRADE B CLASS 1 ASTM CORRELATION MONITOR MATERIAL FROM HSST PLATE 02 Chemical Analysis (wt %)

l C Mn P S Si Ni Mo Cu O.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14 Heat Treatment Heated at 1675 t 25 F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - air-cooled Heated at 1600 25 F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - water-quenched Tempered at 1225 t 25 F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - furnace-cooled Stress-relieved at 1150 25*F - 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> - furnace-cooled to 600 F 4-6

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) 4-7

SECTION 5 TESTING OF SPECIMENS FROM CAPSULE R 5-1. OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was per-formed at the Westinghouse Research and Development Laboratory with consultation by Westing-house Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H.

Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully re-moved, inspected for identification number, and checked against the master list in WCAP-8017.Ill No discrepancies were found.

Exarnination of the two low-melting 304*C (579 F) and 310 C (590*F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304 C (579*F).

The Charpy impact tests wers performed on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 instrumen-tation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve, the load of general yielding (Psy), the time to general yielding (tGY), the maximum load (PM), and the time to maxi-mum load (ty) can be determined. Under some test conditions, a sharp drop in load indicative of 1 fast fracture was abserved. The load at which fast fracture was initiated is identified as the fast fracture load (Pp), and the load at which fast fracture terminated is identified as the arrest load (PAI-The energy at maximum load (EM) was determined by comparing the energy-time record and the load-time record. Tha energy at maximum load is roughly equivalent to the energy required to initi-ate a crack in the specimen. Therefore, the propagation energy for the crack (Ep ) is the difference between the total energy to fracture (ED) and the energy at maximum load.

The yield stress (vry) 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.

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Percent shear was determined from postfracture photographs using the ratio-of areas method in compliance with ASTM Specification A370-74. The lateral expansion 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 Rc 45. 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) exten-someter. The extensometer knife edges were spring-loaded to the specimen and operated through specimen f ailure. 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 thermocouple directly to the specimen, the follow-ing procedure was used to monitor specimen temperature. Chromel-alumel thermocouples were in-serted 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 tem-perature to 550'F. The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures. Experiments in-dicated that this method is accurate to plus or minus 2'F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were deter-mined 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 redction in area was computed using the final I

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 R afterirradiation to 2.07 x 1019 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.

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TABLE 5-1 CHARPY IMPACT DATA FOR KEWAUNEE REACTOR PRESSURE VESSEL SHELL FORGINGS (Irradiated to 2.07 x 1019 n/cm2)

Sample Temperature impact Energy Lateral Expansion Shear Number ('C) (* F) (J) (ft lb) (mm) (mils) (%)

l Forging 122X208VA1

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l P-23 -46 -50 11.0 8.0 0.20 8.0 0 P-16 -32 -25 17.5 13.0 0.25 10.0 0 P-18 -18 0 59.0 43.5 0.94 37.0 17 P-21 -18 0 97.5 72.0 1.50 59.0 28 P-13 -1 30 151.0 111.5 1.93 76.0 54 P-14 26 78 68.5 50.5 1.04 41.0 38 P-17 26 78 170.0 125.5 2.26 89.0 63 P-24 52 125 164.5 121.5 2.16 85.0 78 P-15 93 200 170.0 125.5 2.26 89.0 77 P-20 121 250 228.0 168.0 2.36 93.0 100 P-22 149 300 210.0 155.0 2.31 91.0 100 l

Forging 123X167VA1 S-20 -73 -100 9.5 7.0 0.15 6.0 0 S-13 -46 -50 44.5 33.0 0.66 26.0 19 i S-19 -32 -25. 74.0 54.5 1.09 43.0 17 S-21 -32 -25 77.5 57.0 1.09 43.0 21 S-15 -18 0 35.5 26.0 0.53 21.0 8 S-23 -18 0 77.5 57.0 1.17 46.0 17 f S-22 -1 30 131.5 97.0 1.80 71.0 51 S-24 10 50 103.5 76.5 1.45 57.0 38 S-14 16 60 135.5 100.0 1.85 73.0 63 S-18 26 78 221.0 163.0 2.54 100.0 100 S-16 66 150 198.5 146.5 2.44 96.0 100

S-17 93 200 203.5 150.0 2.36 93.0 100 I

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TABLE 5-2 CH AHPY IMPACT DATA FOR KEWAUNEE REACTOR PRESSURE VESSEL WELD METAL AND HAZ M ATERIAL (irradiated to 2.07 x 1019 n/cm2)

Sample Temperature impact Energy Lateral Expansion Shear Number (J) (ft Ib) (mils) (%)

(*C) l (*F) l (mm) l l

Weld Metal W-10 26 78 10.0 7.5 0.30 12.0 19 j W-15 66 150 36.0 26.5 0.53 21.0 30 W-11 93 200 42.5 31.5 0.61 24.0 30

, W- 16 99 210 55.0 40.5 1.04 41.0 50 W-12 107 225 67.0 49.5 1.02 40.0 65 W-9 121 250 106.5 78.5 1.47 58.0 95 W-14 149 300 105.0 77.5 1.63 64.0 100 W'- 13 177 350 107.0 79.0 1.78 70.0 100 HAZ Material H-10 -18 0 21.0 15.5 0.36 14.0 14 H-13 10 50 159.5 117.5 2.08 82.0 69 H-14 26 78 75.0 55.5 1.04 41.0 36 l H-11 52 125 67.0 49 5 1.04 41.0 50 H-9 66 150 137.0 101.0 2.03 80.0 76 H-16 93 200 177.0 130.5 2.44 96.0 100 H-12 149 300 218.5 161.0 2.44 96.0 100 H-15 177 350 179.5 132.5 2.49 98.0 100 5-4

TABLE 5-3 CHARPY IMPACT DATA FOR THE ASTM CORRELATION MONITOR MATERIAL (HSST Plate 02)

Sample Temperature impact Energy Lateral Expansion Shear Number (*C) ('F) (J) (f t Ib) Imm) (mils) (%)

R-13 26 78 10.0 7.5 0.13 5.0 14 R-12 66 150 30.5 22.5 0.48 19.0 25 R-9' 93 200 44.0 32.5 0.76 30.0 27 R-10 99 210 50.0 37.0 0.76 30.0 35 R-16 107 225 103.0 80.5 1.40 55.0 70 R-15 121 250 131.5 97.0 1.75 69.0 69 R-11 149 300 122.5 90.6 1.91 75.0 100 R-14 177 350 134.0 99.0 2.36 93.0 100 5-5

18566 13 TEMPERATURE ( C)

-100 -50 a 50 100 150 200 250 12 I I I I 3 I l l 100 -

g-g

@ 80 -

g g

$ 60 -

40 -

g 20 -

p e _ . .

=

80 -

Qp m -. 2 2;-

g 60 -

q -

1.5 _

l a 40 -

O 0

1.0

~

H 2 8 C (15 F) j 20 - -

0.5 0 . 0.0

'i80 -

240 160 -

! g UNIRR ADI ATED -

200 5 120 -

S O -

160 E CS y 100 -

E 80 -

O 120 2 w e 1RRADIATED AT 288 C -

3 60 0 EO (10 F) 80 40 -

. x OM n/cm2 l 0

8 C (15 F) -

40 20 -

0 - 0

-290 -100 0 100 200 300 400 500 TEMPERATURE ( F)

Figure 5-1. Charpy V-Notch Impact Data for Kewaunee Reactor Vessel Shell Forging 122x208 VA1 5-6

18566-14 TEMPERATURE ( C)

-100 -50 0 50 100 150- 200 2Le 120 100 -

2 #~k 3 g

80 -

g m - O 4 m

g; 40 -O o O g 20 b 0 S 100 @ --

q 2.5

-w 5

E 80 -

OO -

2.0 60 -

20 *go -

1.5 7

E

' 40 -

O -

1'0 ~

17 C (30 F) 20 O 9 -

0.5 0 0.0 180 240 160 - O UNIR R ADI ATED - g@ -

200

  • 2 g 120 --

h -

160 5 100 -

O b 80 - IRRADIATED AT 288 C - 120 $

E O

  1. (5500F) 2.07 x 1019n/cm2 3 $ 60 O -

so O O 11 C (20oF) 40 -

0 O e 11 C (20 F) 40 0 O I  !

O

-200 -100 0 100 200 300 400 500 TEMPER ATURE (oF)

Figure 5-2. Charpy V Notch Impact Data for Kewaunee Reactor Vessel Shell Forging 123x167 VA1 5-7

18566-15 TEMPER ATURE (OC)

-150 -100 -50 0 50 100 150 200 250 l l l l l 3l l l l 100 -

D g pFG

^

g 80 -

[

60 -

M 40 - O 8 e 1 2o -

2 O + 1 0

100

^ 2.5 C 2 2.0 5 80 -

X 60 -

9 /O -

1.5 E

2 E 1.0 -

g 1310C

$ 40 -

7 ka (2J50F) -

0.5 20 -

O e/

0 -

0.0 i 160 140 - IRRADIATED AT -

"^ '^

120 -

0 x1 n/cm2- 160

= 100 -

3 w -

120

$ 80 -

$ g-G q x j -

y 60 -

1310C 1 80 ua 40 - O (2350F)

Q 1310C -

40 20 - (235 F)

O 0 0

-300 -200 -100 0 100 200 300 400 500 TEMPERATURE (OF)

Figure 5-3. Charpy V-Notch Impact Data for Kewaunee Reactor Vessel Weld Yetal 5-8

18566-16 TEMPERATURE ( C)

--150 -100 -50 0 50 100 150 200 250 120

,  ; , , l  ;  ; i ,

100 -

O~'

2 g 80 -

3g 3 5

E 60 40

_ O o' /*

/

m O 20 -

7 2

/

0 -

100 m gzg 2.5 2 CO -

Q U

G -

2.0 E f O

60 40 O /

99 1.5 1.0 5 g

20 - -

0.5 g

0 0.0 200 180 -

UNIRRADIATED 24 O

160 -

S 200 140 -

o p ,

g 120 -

g !h -

160 y 100 -

9 'l cc Q -

120 w 80 -

d O IRRADI ATED AT 288 C 80 60 830C (150 F) S -

(550 F) 2.07x1019 40 n/cm2 9

m _

40 20 -

g , 83eC n50eF3 0 - 0

-300 -200 -100 0 100 200 300 400 500 TEMPERATURE (OF)

Figure 5-4. Charpy V-Notch Impact Data for Kewaunec Reactor Vessel Weld HAZ Metal 5-9

1 l

18566-11 1

5 t

TEMPERATURE (OC) l

-50 0 50 100 150 200 250 300 120 l3 i ,

l l l l I  ! l  !

100 -

S i g

80 60

[..  ;^

w 3 '

I 40 -

2 i 20 -

e/e* i 0 (

100 2.5 n n#

. 80 -

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2.0 60 - -

1.5 2 C 9 i

40 -

50 m/ 81 C (145 F) 1.0

~

$ 20 -

2 g/ ie -

0.5 0 .

O 0.0 140 120 NIR R ADI ATED Q IRRADIATED AT 160 O 288"C (550 F) -

@ 100 - '

$ 2.07 x 10 19n/cm2 g_O

~

80 - gO e

/

120

~

o - -

$ _ 72 C (130 F) z W

s 40 -

2 78 C (1400F) -

40 20 - 9/

0 0

-100 0 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-5. Charpy V-Notch Impact Data for A533 Grade R Class 1 ASTM Correlation Monitor Material 5-10

9 l;

(N INST @

Normalized Eni Test Charpy Charpy Init Sample Temp Energy Ed'% Em/A Number ( C) (J) (kJ, ..,2) (kJ/m2)

P23 -46 11.0 136 68 P16 -32 17.5 220 101 P18 -18 59.0 737 675 P21 -18 97.5 1220 651 P13 -1 151.0 1890 737 P14 26 68 5 856 628 P17 26 170.0 2127 768 P24 52 164.5 2059 783 P15 93 170.0 2127 698' P20 121 228.0 2847 698 P22 149 210.0 2627 651 S2O -73 9.5 119 84 S13 -46 44.5 559 470 S19 -32 74.0 924 667 S21 -32 77.5 966 683 S15 -18 35.5 441 S23 -18 77.5 966 698 S22 -1 131.5 1644 714 S24 10 103.5 1297 760 S14 16 135.5 1695 698 S18 26 221.0 2762 729 S16 66 198.5 2483 698 S17 93 203.5 2542 651 s

1

\

o

t

. I e

! . j TABLE 5-4 3MENTED CHARPY IMPACT TEST RESULTS FOR KEWAUNEE SHELL FORGINGS Eies Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Ep/A Load to Yield Load Maximum Load Load Stress Stress (kJ/m2) (N) ( s) (N) { s) (N) (N) (MPa) (MPa)

Forging 122X208VA1 68 17300 130 17300 0 119 14900 90 16700 200 16700 0 767 812 62 15100 100 19300 670 19300 0 778 887 569 14900 110 19100 680 17800 0 767 875 1153 14900 130 19600 730 14900 4400 767 887 228 12900 115 18700 690 18700 3300 664 812 1359 13600 140 18000 800 11100 3300 698 812 1276 13300 140 18700 850 12900 6000 687 824 1429 13100 115 18200 790 10900 6400 675 807 2149 12200 130 17600 840 629 767 1975 10700 90 16500 800 549 698 Forging 123X167VA1 34 17800 90 17800 0 90 16000 130 19800 500 19800 0 824 921 257 15600 100 20200 660 19800 0 801 921 283 14900 130 20200 670 19600 . 0 767 904 268 15300 110 19800 670 19600 0 790 904 930 14900 130 19300 700 15300 0 767 881 537 14700 110 20000 730 18900 0 755 893 997 14000 110 19100 700 13800 0 721 853 2033 14500 120 20000 750 744 887 1785 12900 120 19100 730 664 824 1891 12900 130 17300 807 664 778 r

I

. 5-11

t i

INSTRUMENT Normalized Energie:

Test Charpy Charpy init I Sample Temp Energy Ed/A Em/A I Number ('C) (J) (kJ/m2) (kJ/m2) (3 W10 26 10.0 127 68 W15 66 36.0 449 381 W11 93 42.5 534 365 W16 99 55.0 686 462 W12 107 67.0 839 397 W9 121 106.5 1330 486 W14 149 105.0 1313 429 W13 177 107.0 1339 502 l l

H10 -18 21.0 263 198 H13 10 159.5 1991 737 H14 26 75.0 941 557 H11 52 67.0 839 565 H9 66 137.0 1712 659 H16 93 177.0 2212 745 H12 149

}

218.5 2729 659 q HIS 177 179.5 2246 644 f l

l t

i 1

TABLE 5-5 ED CHARPY IMPACT TEST RESULTS FOR KEWAUNEE WELD METAL AND HAZ MATERIAL Dr:p Yield Time Maximum Time to Fracture Arrest Yield Flow

/A Load to Yield Load Maximum Load Load Stress Stress 2 ( s) ( s) (N) (MPa) (MPa)

/rm) (N) (N) (N)

Weld Metal 60 16000 90 16000 0 68 14200 120 18000 400 18000 6200 732 830 169 15100 110 18200 410 18200 7600 778 858

$25 14700 100 18500 480 18500 10700 755 853 2 14700 110 18900 430 18900 14200 755 864 45 13300 140 1820( 550 687 P'2 04 12900 130 17600 510 664 784 837 13800 130 17800 530 709 812 HAZ Material 65 14500 117 17600 265 17600 0 744 824 255 13800 100 19800 730 14500 10200 709 864 384 13800 120 20500 640 20000 6200 709 881 374 15600 130 18900 610 17800 6700 801 887 a52 14200 130 18700 730 13800 11100 732 847 467 15100 112 20000 780 778 904 d69 12900 150 17300 780 664 778 G02 13100 180 17100 760 675 778 5-13 g

o I*

INS Normalized Test Charpy Charpy init Sample Temp Energy Ed/A Emf Number (*C) (J) (kJ/m2) (kJ/ni l

l R13 26 10.0 127 84 R12 66 30.5 381 316 R9 93 44.0 551 37f R10 99 50.0 627 49d R16 107 109.0 1364 62%

R15 121 131.5 1644 69%

R11 149 122.5 1534 58 R14 177 134.0 1678 58 q l

4

\

~ a TABLE 5-6 jRUMENTED CHARPY IMPACT TEST RESULTS FOR THE ASTM CORRELATION MONITOR MATERIAL '

}nergias Prep Yield Time Maximum Time to Fracture Arrest Yield Flo w Ep/A Load to Yield L.oad Maximum Load Load Stress Stress

3) (kJ/m2) (N) ( s) (N) (p s) (N) (N) (MPa) (MPa) 43 14200 90 14200 0 65 14700 130 16900 340 16900 3300 755 812 178 14000 110 18200 440 17800 6000 721 830 134 13300 120 18700 540 18700 8900 687 824 744 13600 130 18200 640 17300 16200 698 818 946 13800 130 19600 807 16900 11100 709 858 953 13300 150 18200 660 687 812 1089 12900 115 17800 630 664 790 I

l l

i l

5-15 l

e i

(

THE EFFECT @

NOTCH TOUGHNESS PRO Transition Tempera Unirradisted j 50 f t Ib 30 ft Ib 35 mits 50 if 68J 41 J .9 m m 63, Material *C (*F) *C (*F)

  • C ( F) *C (*

122X208VA1 -9 -31 -9 -

(15) (-25) (- 15) (2' 123X167VA1 -31 -46 -43 -2

(-25) (-50) (-45) (-!

Weld Metal -23 -46 -37 10'

(- 10) (-50) (-35) (22!

HAZ Metal -57 -82 7'3 2' (122X208VA1) (-70) (- 115) (-100) (81 Correlation 27 7 16 9 (80) (45) (60) (214

(

l

P TABLE 5-7 2

5 288*C IRRADIATION TO 2.07 x 10 19 n/cm .(E > 1 Mev) ON THE PERTIES OF KEWAUNEE REACTOR VESSEL SURVEILLANCE TEST MATERIAL

+

tura Average Energy Absorption Irr:disted .1 Transition Temperature at Full Shear 35 mils i Ib 39 f t Ib 50 f t Ib 30 f t Ib 35 mils 1 41 J .9mm 68J 41 J .9 m m Unirradiated Irradiated .i Energy F) *C ('F) *C(*F) 'C ('F) *C(*F) 'C ('F) J (f t Ib) J (f t Ib) J (f t Ib) l -23 -18 6 8 8 217 217 0 i) (-10) (0) (10) (15) (15) (160) (160) (0)

) -34 -26 11 11 17 213 207 6 i) (-30) (-15) (20) (20) (30) (157) (153) (4) 85 93 131 131 131 171 106 65 i) (185) (200) , (235) (235) (235) (126) (78) (48)

' 2 16 83 83 89 244 191 53

)) (35) (60) (150) (150) (100) (180) (141) (39)

) 85 96 72 78 81 167 129 38

)) (185) (205) (130) (140) (145) (123) (95) (28) l l

l

\

i i

5-17 i

18566-1 I

y _. ,.

y .. _ 3

' tr+=*741:. > \

P-23 P-16 P-18 P-21 F 13 P-14 l P-17 P-24 P-15 P-20 P-22 l

l l

Figure 5-6. Charpy impact Specimen Fracture Surfaces for Kewaunee Intermediate Shell Forging 122x208 VA1

.l 5-19

18566-2 i

l

,, .~ '~ -~ ~ ~

a  :

S-20 S-13 S-19 S-21 S-15 S-23 l

I 1

S-22 S-24 S-14 S-18 S-16 S-17 l

Figurt 5-7. Charpy impact Specimen Fracture Surfaces for l Kewaunee Lower Shell Forging 123x167 VA1 l

l 5-20 x _ _ _ _ - - . _ . _ _ _ _ _ _ _ - _ _--- _ - . - _ . . . _ _ . - - _ _ _ _ . _ _ - _ . .

18566-3 W-10 W-15 W-11 W-16

~ ~~~

/

n. .- , y

.: .G.

W-12 W-9 W-14 W-13 Figure 5-8. Charpy In. pact Specimen Fracture Surfaces for Kewaunee Weld Metal 5 21

/

g 18566-4 l pla ,

^~

v t_~ ., ^q , " -' -

~:s - -

H-10 H-13 H-14 H-11 k> 'N Y  ?

.n I, "?

f

! H-9 H-16 H-12 H-15 l

l Figure 5-9. Charpy impact Specimen Fracture Surfaces for Kewaunee HAZ Material 5-22 s 1

I

' r A

!! 18566-5 i.

i i

7 a .4,'-

f' .

t.

. w, i i

% $. i

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R 13 R-12 R-9 R-10 i

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k.. g - .._.

irg

( 45 __

R-16 R-15 R-11 R-14 t

I Figure 5-10. Charpy impact Specimen Fracture Surfaces for Kewaunee ASTM Correlation Monitor Material 5-23

trradiation of material from intermediate shell forging 122X208VA1 to 2.07 x 1019 n/cm2 re- ,

suited in 68-joule (50 f t Ib) and 41-joule (30 f t Ib) transition temperature increases of 6 C (10 F) f and 8*C (15 F), respectively, as noted in figure 5-1. As shown in figure 6-2, the 68-joule and 41-joule transition temperature increase for material from lower shell forging 123X167VA1 was 11 C (20*F). No decrease in upper shelf energy appears to have occurred for forging 122X208VA1 and only a 6-joule (4 f t Ib) decrease was exhibited by forging 123X167VA1.

Irradiation to 2.07 x 1019 n/cm2 resulted in a 68-joule and 41-joule transition temperature in-crease of 131*C (235 F) for the weld metal (figure 5-3). The upper shelf energy of the weld metal decreased by 65 joules (48 f t Ib) to 106 joules (78 ft Ib) which is sufficiently high enough to allow q I

for continued saf e operation of the plant. Weld HAZ material shown in figure 5-4 exhibited a 68-joule and 41-joule transition temperature increase of 83*C (150 F) afterirradiation to 2.07 x 1019 n/cm2 and a decrease in upper shelf energy of 53 joules (39 f t Ib).

ASTM correlation monitor material from HSST Plate 02 exhibited a 68-joule transition temperature increase of 72*C (130*F) and a 41-joule temperature increase of 78 C (140*F) as shown in figure 5-5. Upper shelf energy of the correlation monitor material decreased by 38 joules (28 ft l'ui.

A su.nmary of the Charpy impact tests performed on the two capsules removed from the Kewau-nee reactor to date is shown in table 5-8. These results show that the two low copper (0.06 per-cent Cu) shell forgings are very insensitive to irradiation up to 2.07 x 10 19 n/cm 2. A comparison of the 41-joule transition temperature increases with the predicted increases based on the U.S. Nucle-ar Regulatory Commission Regulatory Guide 1.99 Revision 1I4l is shown in figure 5-10. This com-parison shows that the Guide overpredicts the 41-joule transition temperature increase for the forgings by approximately 28 C (50 F).

The results o'f Charpy impact tests performed to date on the weld metal (0.20 percent Cu) indicate, as shown in table 5-8, that irradiation to 2.07 x 1019 n/cm2 resulted in considerable additional in-crease in transition temperatures when compared with irradiation t :sts at 5.99 x 1018 n/cm2. A companson of the 41-joule transition temperature increases with the Guide predictions, as shown in figure 5-11, tends to indicate that the Guide will overpredict the transition temperature increase at fluences greater than 2 x 1019 n/cm2 5-3. TENSlLE TEST RESULTS The results of tensile tests performed on material from shell forgings 122X208VA1 and 123X167VA1 and the weld metal are shown in table 5-9 and figures 5-12 through 5-14, respec-tively. These results show that the 0.2 percent yield strength and other properties of the two forg-ings are only slightly af fected by irradiation to 2.07 x 1019 n/cm2. The large increase in 0.2 per-cent yield strength exhibited by the weld metal as shown in figure 5-14 indicates, as did the Charpy impact tests, that the weld metalis very sensitive to radiation at 2.07 x 10 19 n/cm2. A typical 5-24

TABLE 5-8

SUMMARY

OF KEWAUNEE REACTOR VESSEL SURVEILLANCE CAPSULE CH ARPY IMPACT TEST RESULTS 68 J 41 J 50 ft Ib 30 ft Ib Decrease in Trans. Temp Trans. Temp Upper Shelf Fluence increase increase Energy [a]

Material 1018 n/cm2 (*C) (*F) ( C) (*F) (J) (ft Ib)

Forging 5.99 0 0 0 0 +27 +20 122X208VA1 20.70 6 10 8 15 0 0 Forging 5.99 0 0 0 0 +28 +21 123X167VA1 20.70 11 20 11 20 6 4 Weld Metal 5.99 108 195 97 175 60 44 20.70 131 235 131 235 65 48 Weld HAZ Metal 5.99 44 80 44 80 47.5 35 20.70 83 150 83 150 53 39 Correlation Monitor 5.99 53 95 53 15 19 14 20.70 72 130 78 140 38 28 a Values preceded by a plus sign indicate an increase in upper shelf energy.

5-25

- _ .. . . .. i

5 4 -

200 C

- 3 -

WELD METAL E

6 2 -

100 _

cc WELD METAL S PREDICTION o[

a.

w cr E 10 2 i

w 50 3 9 -

H

& g -

40 y m

y 7 -

w I 6 -

- 30 c.

k -H E cn g; 5 -

N k

tr 4 -

BASE METAL PREDICTION- 20 H 3 -

i uj FORGING 122X208 val l 8 2 -

10 FORGING 123X167 val x

101 I I I I I Ill! 'l I I I. I I I IIIlI 8 10 18 8 10 19 4 6 8 10 20 10 17 2 4 6 2 4 6 2 FLUENCE (n/cm2) 8 8

Figure 5-11. Comparison of Predicted Versus Actual 41-Joule Transition Temperature $

Increases for Kewaunee Reactor Vessel Materials

{

(

l

.8 Test i Sample Temp @

Number Material ( C)

P-6 122X208VA1 10 P-5 122X208VA1 66 P-7 122X208VA1 121 P-8 122X208VA1 288 S-6 123X167VA1 -4 S-5 123X167VA1 24 S-4 123X167VA1 288 W-3 Weld 121 W-4 W eld 288 t

9

?

i TABLE 5-9

' ENSILE PROPERTIES FOR KEWAUNEE REACTOR VESSEL MATERIAL IRRADIATED TO 2.07 x 1019 n/cm2 (E > 1 Mev)

V; Yi:Id

)ffnt Ultimate Fracture Fracture Fracture Uniform Total Reduction Ir;ngth Strength Load Stress Strength Elongation Elongation in Area MPa) (M Pa) (N) (M Pa) (M Pa) (%) (%) (%)

467 629 12,700 1480 400 12.0 25.2 73 464 597 11,700 1360 369 11.1 24.8 73 832 572 11,500 1240 362 10.5 22.8 71 414 586 11,800 1190 372 10.7 22.0 69 427 674 12,900 1400 407 11.0 24.9 71 899 646 12,200 1550 386 11.3 24.9 75 835 607 12,000 1210 379 9.8 20.5 69 174 758 16,900 1300 534 12.0 22.5 59 604 716 17,700 1210 558 12.0 19.2 54 5-27

18566-17 TEMPER ATURE (OC) 0 50 100 150 200 250 300 350 100 700 l I l l I I I I 90 A A- ._ e . -

600 80 C A

3 TENSILE STRENGTH - 500 70 -

O ~

m wJ -_ _ _

60 - _

4UU O -Q 50 -

0.2% YlELD STRENGTH 300 LEGEND:

OPEN POINTS - UNIRRADIATED CLOSED POINTS - IRRADI ATED AT 2.07 x 10 1U n/cm 2 80 70 -

~~+ ------g 9 60 -

REDUCTION IN AREA 2

- 50 -

2 40 -

D 30 -

TOTAL ELONGATION U

Ob" ~ Q 20 -

A 10 -

~M - -M - ~ ~ - - -A UNIFORM ELONGATION 0

O 100 200 300 400 500 600 700 TEMPERATURE ( F)

Figure 5-12. Tensile Properties for Kewaunee Reactor Vessel Shell Forging 122x208 VA1 5-29

18566-19 l TEMPER ATURE (OC) 0 50 100 150 200 250 300 350 110 I I I I i i I l 100 700

-A NA 90 -

='= A j -

600

^ ~

$ 80 -

ON s TENSILE STRENGTH _ SM ag

  • 70 p%

_ - . a M 60 -

2 O -

400 2

50 -

0.2% YJELD STRENGTH aoe LEGEND:

OPEN POINTS - UNIRRADI ATED 19 2 CLOSED POINTS -- IRRADI ATED AT 2.07 x 10 n/cm 80

-O M- n i, 2 70 V -

-W- *=-~g Q 60 -

' ^ ^

s~ 50 -

D E 40 -

U 3 30 -

TOTAL ELONGATION O g A q )

n m 20 -

' -- 34 - -9 A"

A' =g- -- -- -

10 A _ _ ~ - - A l l ' UttlFORM ELONGATI,0N O

O 100 20G 300 400 500 600 700 TEMI'ERATUME ( F)

)

Figure 5-13. Tensile Properties for Kewaunee Reactor Vessel Shell Forning 123x167 VA1 )

l i

5-30 l I

l 18566-10 TEMPERATURE ( C) 0 50 100 150 200 250 300 350 l l l l l l l l-l800 110 -

A--  %

'-- -A _

700 0- rTENSILE STRENGTH 90 -

T*" -

  • 9 -

600 3  %

80 - -

A P M o-500 cc 70 -

0.2% YlELD STRENGTH 60 -

O -

400 2

2 _

300 LEGEND:

OPEN POINTS - UNIRRADI ATED 19 2 CLOSED POINTS -IRRADIATED AT 2.07 x 10 n/cm 80 70 -

O &

h 60 -

g._ _~ ~ ~ REDUCTION IN AREA y ~e

50 -

] 40 -

H TOTAL ELONGATION S 30 20 W en b

-v --- _g 9 0 -

4 Au A4 UNIFORM ELONGATION I I I I O

O 100 200 300 400 500 600 700 TEMPER ATURE (OF)

Figure 5-14. Tensile Properties for Kewaunee Reactor Vessel Weld Metal 5-31

stress strain curve for the tensile tests is shown in fi;;9re 5.15. Photographs of the fractured tensile specimens for the two forgings and the weld metal are shown in figures 5-16 though 5-18, respectively.

5-4. WEDGE OPENING LOADING TESTS Wedge Opening Loading (WOL) fracture mechanics specimens which were contained in the sur-veillance capsule have been stored temporarily at the Westinghouse Research Laboratory on the recommendation of the U.S. Nuclear Regulatory Commission; tney will be tested and reported on later.

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4 I

..r

, t s:.c  :-

~-

_g VIM,.: ?4 g .g -- ,w~d%g. g - y , ..AI.. , aW M ,y% '-

4 l

a,o,a ~ ~ - , _

-em-w-m 4,

, ~ . -. . -

' P-7 121 C i

1 . . _ . , . ,

^

, h g,,,,

' .t

  • y % , ,

w a w: a , - -

j i

i

> A g y4- ~ g ll~ ?' ;;,t;q n g 2{4'f1  : .

-d ~ .+ ,+ x , .n n.n m .- . -. -'.~'~ ,. 3 u . + . LL ,, n . , s P-8 288 C Figure 5-16. Fractured Tensile Specimens From Kewaunee intermediate Shell Forging 122x208 VA1 i

5-34

18566-8 i

.pTyp y

- g-rvn , p,,gmr-

  1. p . . - pgM, x 1 r ,*,*

~

p.1-s.

j y : '+ ,n ; ,

.)-.,;,,.-

e

__a-.....~_; s S-6 -4 C n,- ,,

m .:g7+

.p a.4o v, .:r~crr--e

.. .. >w . ~ o, .

I f . ne,; */

~ py _ - r 8

, .::2 ,, .

5 cgy, l

,4 . ; ,;

44 ap .

.' z. ,G ' m,.

-,m hw h .__ . d.;

S-5 24 C l

l l

y. ,

. . . , - , %ey ,

c

, . .re y, . ~.

. >.. r. s .s.* * .

e 'r. '*o

-+.4,.

,y .

_,.v y .;5.T' Q$kh.b!l$ Ulb ,,ib 4:. .,' _st .

% *Je .w r, a b . p e<%;w,.mn--

l 1 ~:~.....,.,..n.-, ._ ,

4 l

I S-4 288 C 1

l 9

Figure 5-17. Fractured Tensile Specimens From Kewaunee Lower Shell Fcrging 123x167 VA1 5-35 l l

l l

18566-S 1

7; m 7y ,- -m 7 7

.w. >

9 .

,e . v 4

Y i

.. . < w% .

. ;:qff.::Qi .;W;6.:e\h;&,  ?.q <,~, . ;<. ~

.. .. e 'r s R

% ,,. ,y- 4Jm > : +,,u,e. -,w~mv m.-4 3,

  • .'s<4, mm .s o . .w .v s .. . Q S s , ,y

,j W-3 121 C

- - _ . ~

t e ...

3

. ;,.' we m . .c, - - &<-

9'e

..w, -- ' .m? 14 . , >

w y;e , ., a ,; ~~~ mew w - w ~ r . , ,

-; - . . , . =

W-4 288 C l l

Figure 5-18. Fractured Tensile Specimens From Kewaunee Weld Metal 5-36

SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1. INTRODUCTION 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 in the test specimens to the present and future condition of the reactor vessel, a relationship between the environment at various positions within the reactor vessel and that experienced by the test specimens must be established. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements ob-tained with passive neutron flux monitors contained in each of the surveillance capsules. The latter requirement is derived solely from analysis.

This section describes a discrete ordinates Sn transport analysis performed for the Kewaunee reac-tor 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 capsules, and,in turn, to develop lead factors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averaged reaction cross sections derived from this calculation, the analysis of the neutron dosimetry contained in Capsule R is discussed, and updated evaluations of dosimetry from Capsule V are presented.

6-2. DISCRETE ORDINATES ANALYSIS A plan view of the Kewaunee reactor geometry at the core midplane is shown in figure 6-1. Be-cause the reactor exhibits 1/8th core symmetry, only a zero- to 45-degree sector is depicted. Six irradiation cepsules attached to the thermal shield are included in the design to constitute the reac-tor vessel surveillance program. Two capsules are located symmetrically at 13,23, and 33 degrees from the cardinal axis as shown in figure 6-1.

6-1

-- d

18,318-20 PRESSURE VESSEL SURVEILLANCE CAPSULE l 13 (CAPSULES V,R)

/ 23 (CAPSULES T,P)

THERMAL 33 (CAPSULES S,N)

SHIELD 9

'//M///

W/////// 4go fH,HH,,HHHH/ 4

/

/

/

/ /

/

/

l /

. / ,/

/

f I / l/ REACTOR CORE

//

[/

ly!/

Figure 6-1. Kewaunee Reactor Geometry 6-2 i

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 approximately 63 inches in height. The containers are positioned axially such that the specimens are centered on the core midplan 3, 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, they have a marked impact on the distributions of neutron flux and energy spectra in the water annulus be-tween the thermal shield and the iactor vessel. Thus, to properly ascertain the neutron environ-ment at the test specimen locations, the capsules themselves must be included in the analytical model. Use of at least a two-dimensional computation is, therefore, mandatory.

In the analysis of the neutron environment within the Kewaunee 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 distributions were obtained from an R,() computation wherein the geometry shown in figures 6-1 and 6-2 was described in the analytical model. in addi-tion to the R,9 computation, 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 the surveillance capsules were not included in the model.

Both the R,# and the R,Z analyses employed 21 neutron energy groups, an S8 angular quadrature, arH a P1cross-section expansion. The cross sections were generated via the Westinghouse GAMBIT [6] code system with broad group processing by the APPROPOSI71 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 fast neutron exposure of the reactor vessel is the core power distribution. For this analysis, power distributions representative of time-averaged conditions derived from statistical studies of long-term operation of Westinghouse two-loop plants w% 7 employed. These input distributions include rod-by-rod spatial variations for all l peripheral fuel assemblies.

I 11 should be noted that this particular power distribution is intended to produce accurate end-of-life j neutron exposure levels for the pressure vessel. As such, the calculation is indeed representative of I an average neutron flux and small (plus or minus i o to 20 percent) deviations from cycle to cycle are to be expected.

Having the results of the R,# and R,Z calculations, three-dimensional variations of neutron flux may be approximated by assuming that the following relation holds for the applicable regions of the reactor.

4(R,Z,#,Eg) = 4(R,0,Eg) F(Z,Eg) (6-1) 6-3 i _ .

l 18,318-21 3 (13 , 23 , 33 )

(12 , 22 , 32 ) [ CHA ppC EN

'/

/j #

/

/

// /i /// //)

THERMAL SHIELD Figure 6-2. Plan View of a Reactor Vessel Surveillance Capsule 6-4

TABLE 6-1 21 GROUP ENERGY STRUCTURE Group Lower Energy (Mev) 1 7.79[a]

2 6.07 3 4.72 4 3.68 5 2.87 6 2.23 7 1.74 2 1.35 9 1.05 10 0.821 11 0.388 12 0.111 13 4.09 x 10-2 ~

14 1.50 x 10-2 15 5.53 x'10-3 16 5.83 x 10-4 17 7.89 x 10-5 18 1.07 x 10-5 19 1.86 x 10-6 20 3.00 x 10-7 21 0.0

a. Upper energy of group 1 is 10.0 Mev 6-5

where 4(R,Z,0,Eg) = neutron flux at point R,Z,# within energy group g 4(R,#,Eg) = neutron flux at point R,9 within energy group g obtained from the R,0 calculation I

l F(Z,Eg) = relative axial distribution of neutron flux within energy group g obtained from the R,Z calculation 6-3. NEUTRON DOSIMETRY The passive neutron flux monitors included in the Kewaunee surveillance program are listed in table 6-2. The first five reactbas in table 6-2 are used as fast neutron monitors to relate neutron fluence (E > 1.0 Mev) to rueasure materials properties changes. To properly account for burnout of the product isotope generated by fast neutron reactions, it is necessary to also determine the magni-tude of the thermal neutron flux at the monitor location. Therefore, bare and cadmium-covered cobalt-aluminum monitors were also included.

  • The relative locations of the various monitors within the surveillance capsules are showa in figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are piaced in holes united in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission rr ,nitors are accommodated within the dosimmer block located near the center of the capsule.

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 activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron flux has on 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 mo,otor The analysis of the passive monitors and subsequent derivation of the average neutron flux re-quires completion of two plodres. 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 neutrun energy spectrum at the monitor location must be calculated.

6-6

TABLE 6-2 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Target Fission Monitor Reaction Weight Product Yield Material ofInterest Fraction Response Range Half-Life (%)

Copper Cu63(n.a)Co60 0.6917 E> 4.7 Mev 5.27 years Iron Fe54(n.p)Mn54 0.0585 E > 1.0 Mev 314 days Nickel N,58(n p)Co68 0.6777 E > 1.0 Mev 71.4 days Uranium-238(al U238(n f)Cs137 1.0 E > 0.4 Mev 30.2 years 6.3 Neptunium-237[a] Np237(n,f)Cs137 1.0 E > 0.08 Mev 30.2 years 6.5

~

Cobalt-Aluminum!al CoS9(n,y)Co60 0.0015 0.4eV > 0.015 Mev 5.27 years Cobalt-Aluminum CoS9(n,y)CoE0 0.0015 E > 0.0015 Mev 5.27 years a Denotes that momtM s$ Cadmium shielded 6-7

The specific activity of each of the monitors is determined using established ASTM proce-dures.19,10,11,12,13] Following sample preparation, the activity of each monitor is determined by means of a lithium-drif ted 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 Kewaunee, the overall 2er de,iation in the measured data is determined to be plus or minus 10 percent. The neu-tron energy Opectra are determined analytically using the method described in paragraph 6-1.

(

Having the measured activity of the monitors and the neutron energy spectra at the locations of interesi, the calculation of the neutron flux proceeds as follows.

The reaction product activity in the monitor is expressed as N p.

R= fY er(E)4(E) (1 - e- Atj) ,- Atd (6-2)

A E Pmax where R = induced product activity No = Avogadro's number A = atomic weight of the target isotope i = weight fraction of the target isotope in the target material Y = number of product atoms produced per reaction er(E) = energy-dependent reaction cross section 4(E) = energy-dependent neutron flux at the monitor location with the reactor at full pcwer Pj = average core power level during irradiation period j P max = maximum or reference core power level A = decay constant of the product isotope tj = length of irradiation period j td = decay time following irradiation period j Because neutron flux distributions are calculated using multigroup transport methods and, further, because 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.

er(E) 4(E)dE = frd(E > 1.0 Mev)  ;

6-8

where ao N o(E) 4(E)dE {o4 G=1

=

o=

N

.0 Mw 4(EldE [G 1.0 Mev Thus, equation (6-2) is rewritten o P i d R= f i y a 4 (E > 1.0 Mev) (1-e-At-)

1 e-At P max A j= 1 or, solving for the neutron flux, R

4 (E > 1.0 Mev) = N P-(6-3)

N 1 -At. -Atd o f; y a-. { p max (1-e Il e 7

A = 1 The total fluence above 1.0 Mev is then given by Pj (D (E > 1.0 Mev) = 4 (E > 1.0 Mev) tj (6-4) j=1 Pmax where N

{ max t; = tota! effective full power seconds of reactor operation up to the time of capsule removal j=1 An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bars and cadmium-covered CoS9(n, y)Co60 data by means of cadmium ratios and the use of a 7-barn 2200 m/s cross section. Thus, 6-9

hD-1J R

bare [ Dh DTh " N l N p, A

fya i T u y I

(1.e j) e4td j=1 max where D is defined as Rbare/ rcd covered-6-4. TRANSPORT ANALYSIS RESULTS Results of the Sntransport calculations for the Kewaunee reactor are summarized in figures 6-3 through 6-7 and in tables 6-3 through 6-5. In figure 6-3, the calculated r.1aximum neutron flux levels at the surveillance capsule center line, pressure vesselinner radius,1/4 thickness location, and 3/4 thickness location are presented as a function of azimuthal angle.The influence of the sur-veillance capsules on the f ast 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 pres-sure vesselis 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.

In figure 6-6 the radial variations of f ast neutron flux within each of the surveillance capsules are presented. These data,in conjunction with the maximum vessel flux, are used to developlead fac-tors 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 fast neutron flux at the pressure vesselinner radius. Updated lead factors for all of the Kewaunee surveillance capsules are listed in table 6-3. The lead f. : ors presented in table 6-3 differ from those previously pub-lished in references 1 and 3 because of an improved neutron transport methodology which takes into account neutron flux perturbations introduced by the surveillance capsules and their associat-ed structure. Additionalinformation on this perturbation effect is given in reference 14.

Because the neutron flux monitors contained within the surveillanca capsules are not alllocated at the same radial location, the measured disintegration rates are analytically adjusted for the gra-dients 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 comparison was chosen to be the capsule center. Ana-lytically determined reaction rate gradients for use in the adjustment procedures are shown in figurst 6-7 for Capsules V and R. All of the applicable fast neutron reactior's are included.

To derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum ..veraged reaction cross sections are required. The neutron energy spectrum calculated to exist at the center cf each of the Kewaunee nurveillance capsules is given in table 6-4. The asso-ciated sp ctrum-averaged cross sections for each of the five fast neutron reactions are given in table 6-5.

6-10

18566-22 10 12 PW s

8 -

l _

6 -

4 2 -

10 11 -

  • G c -

8 -

"ao 6 -

._ SURVEILLANCE 5 4 _

CAPSULES X

3 u.

2

.O 2 -

3 PRESSURE VESSEL z IR T 1/4T LOCATION 8 -

6 -

4 -

3/4T LOCATION 2 -

10 9 0 10 20 30 40 50 60 70 AZlMUTHAL ANGLE (deg)

Figure 6-3. Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Wit i the Pressure Vessel Surveillance Capsule Geometry 6-11

18566 12 10 11 8

6 - 167.64 l

4 -

171.77 I I

~

- IR I 175.90 N 2 -

1/4 T l E

i i N 1/2 T 3 10 10 -

l d 8

~

z - I cc 6 -

3/4 T 184.15 l

y 4 -

_ i OR 2

4--- H2 O

  • g: PRESSURE VESSEL rs 9  !

10 160 162 164 166 166 70 172 174 176 178 180 182 184 186 188 R ADIUS (cm)

Figure 6-4. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vesi.cl 6-12

18.318 24 10 0 5 -

2 -

10-1 x -

3 o.

z 5 -

E ua 2 2 -

H 5

e 10-2 5 -

CORE MIDPLANE 2 -

rTO VESSEL CLOSURE HEAD 10-3  !  !  ! l

-300 -200 -100 0 100 200 300 DISTANCE FROM CORE MIDPLANE (cm)

Figure 6-5. Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-13 A

18566-21 10 12 8 -

6 -

4 -

E Si h 2 _ CAPSULES N V, R E l X

11 3

u.

10 T CAPSULES T, P 8 -

5 -

i

% 6 -

1 3 uj oc CAPSULES 2 4 -

gg S, N mz OU CAN CAN 2

THERNIAL

+- SHIE LD HO2 h TEST SPECIMENS =I +HO+2 154 155 156 157 158 159 160 161 R ADIUS (cm)

Figure 6-6. Calculated Radial Distribution of Maximum Fast Neutron F!ux (E > 1.0 Mev) Within the Surveillance Capsules 6-14

18566-20 109 _

8

~_

6 -

4 -

2 - NiS8 (n, p) CoS8 10 8 __

8 -

6 -

@ 4 -

o Np237 (n, f) Cs137 2 -

U238 (n, f) Cs137 h f

107 -  %

4 8 2 6 -

4 -

k Fe54 (n, p) Mn54 2 - b o

10 6 _ g 8

} Cu64 (n, ex ) Co60 6 -

4 -

HO 2 2 THERMAL

+- SHIELD TEST SPECIMENSq + H2O ->

I I I I  ! h 10 5 154 155 156 157 158 159 160 161 R ADIUS (cm)

Figure 6-7. Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules V and R I l

6-15 i

_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _______________.i__________________________ . . _ -

TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 Mov) AND LEAD FACTORS FOR KEWAUNEE SURVEILLANCE CAPSULES Capsule Azimuthal 4 (E > 1.0 Mov) Lead Identification Location (n/cm2 sec) Factor V 13* 1.45 x 1011 3.37 R 13* 1.45 x 1011 3.37 T 23' 8.33 x 1011 1.94 P 23* 8.33 x 1010 1,94 S 33' 7.67 x 1010 1,79 N 33' 7.67 x 1010 1,79 t

6-16

TABLE 6-4 CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF KEWAUNEE SURVEILLANCE CAPSULES Group Neutron Flux (n/cm2-sec)

No. Capsules V and R Capsules T and P Capsules S and N 1 8.17 x 108 5.99 x 108 5.26 x 108 2 2.68 x 109 1.99 x 109 1.75 x 109 3 4.43 x 109 3.08 x 109 2.73 x 109 4 4.98 x 109 3.18 x 109 2.83 x 109 5 8.66 x 109 5.20 x 109 4.75 x 109 6 1.70 x 1010 1.01 x 1010 9.26 x 109 7 2.46 x 1010 1.41 x 1310 1.30 x 1010 8 3.53 x 1010 1.97 x 1010 1.83 x 1010 9 4.07 x 1010 2.53 x 1010 2.35 x 1010 10 5.04 x 1010 2.67 x 1010 2.48 x 1010 11 1.67-:1011 8.66 x 1010 8.03 x 1010 12 2.11 x 1011 1.05 x 1011 9.76 x 1010 13 9.42 x 1010 4.65 x 1010 4.34 x 1010 14 7.11 x 1010 3.52 x 1010 3.28 x 1010 15 5.67 x 1010 2.80 x 1010 2.62 x 1010 16 1.32 x 1011 6.41 x 1010 5.99 x 1010 17 1.03 x 1011 5.07 x 1010 4.73 x 1010 18 1.06 x 1011 5.14 x 1010 4.82 x 1010 19 8 41 x 1010 4 09 x 1010 3.83 x 1010 20 9.34 x 1010 4.52 x 1010 4.23 x 1010 21 2.97 x 1011 1.51 x 1011 1.36 x 1011 6-17

TABLE 6-5 SPECTRUM AVERAGED REACTION CROSS SECTIONS AT THE DOSIMETER BLOCK LOCATION FOR KEWAUNEE SURVEILLANCE CAPSULES r7 (barns)

Reaction Capsules V and R Capsules T and P Capsules S and N Fe54(n,p)Mn54 0.0595 0.0683 0.0666 N,58(n.p)CoS8 0.0811 0.0912 0.0893 Cu63(n.a)Co60 0.000404 0.00051 ( 0.000494 U238(n.f)F.P. 0.333 0.345 0.344 Np237(n f)F.P. 2.93 2.80 2.82

<r(E)6(E)dE o

<7 =

l 6(E)dE 1 Mev 6-5. DOSIMETRY RESULTS The irradiation history of the Kewaunee reactoris given in table 6-6. Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsules V and R are listed in tables 6-7 and 6-8, respectively. The data are presented .a measured at the actual monitor locations as well as adjusted to the capsule center. All adjustmu.1ts to the capsule center were based on the data I

presented in figure 6-7.

As noted in reference 3, reaction rate ratios obtained from fast neutron dosimetry indicated that Capsule V had been rotated 180 degrees from the configuration shown in figure 6-2. Data from Capsule R, however, indicate that this capsule was inserted in the reactor with the orientation shown in figure 6-2. All adjustments to the capsule center were based on the assumption tha' Capsule V had been rotated while Capsule R had not.

6-18

TABLE 6-6 IRRADIATION HISTORY OF KEWAUNEE REACTOR VESSEL SURVEILLANCE CAPSULES Irradiation Decay Pmax Time Time PJ Month (MW) (MW) Pj/P max (days) (days) 4/74 267 1650 .162 17 2364 5/74 790 1650 .479 31 2333 6/74 1201 1650 .728 30 2303 7/74 1160 1650 .703 31 2272 8/74 1582 1650 .959 31 2241 9/74 909 1650 .551 30 2211 10/74 419 1650 .254 31 2180 11/74 1018 1650 .617 30 2150 12/74 1165 1650 .706 31 2119 1/75 1008 1650 .611 31 2088 2/75 1350 1650 .818 28 2060 3/75 1432 1650 .868 31 2029 4/75 1107 1650 .671 30 1999 5/75 1107 1650 .671 31 1968 6/75 1021 1650 .619 30 1938 7/75 1091 1650 .661 31 1907 8/75 1523 1650 .923 31 1876 9/75 794 1650 .481 30 1846 10/75 1190 1650 .721 31 1815 11/75 1087 1650 .659 30 1785 12/75 1493 1650 .905 31 1754 1/76 1411 1650 .855 31 1723 2/76 1530 1650 .927 13 1710 Capsule V Removed 2/76- 0 1650 .000 47 1663 3/76 4/76 487 1650 .295 30 1633 5/76 904 1650 .548 31 1602 6/76 1591 1650 .964 30 1572 7/76 1543 1650 .935 31 1541 8/76 1558 1650 .944 31 1510 9/7'6 1437 1650 .871 30 1480 10/76 1576 1650 .955 31 1449 11/76 1546 1650 .937 30 1419 12/76 1551 1650 .940 31 1388 6-19

TABLE 6 6 (cont)

IRRADIATION HISTORY OF KEWAUNEE REACTOR VESSEL SURVEILLANCE CAPSULES trradiation Decay PJ P max Time Time Month (MW) (MW) Pj/Pmax (days) (days) 1/77 784 1650 .475 31 1357 2/77 0 1650 .000 28 1329 3/77 244 1650 .148 31 1298 4/77 1379 1650 .836 30 1268 5/77 1584 1650 .960 31 1237 6/77 1587 1650 .962 30 1207 7/77 1592 1650 .965 31 1176 8/77 1389 1650 .842 31 1145 9/77 1610 1650 .976 30 1115 10/77 1599 1650 .969 31 1084 11/77 1605 1650 .973 30 1054 12/77 1571 1650 .952 31 1023 1/78 1596 1650 .967 31 992 2/78 1582 1650 .959 28 964 3/78 1564 1650 .948 31 933 4/78 1092 1650 .662 30 903 5/78 85 1650 .0518 31 872 6/78 1373 1650 .832 30 842 7/78 1485 1650 .900 31 811 8/78 1539 1650 .933 31 780 9/78 1526 1650 .925 30 750 10/78 1577 1650 .956 31 719 11/78 1528 1650 .926 30 689 12/78 1569 1650 .951 31 658 1/79 1577 1650 .956 31 627 2/79 1505 1650 .912 28 599 3/79 1472 1650 .892 31 568 4/79 1538 1650 .932 30 538 5/79 1261 1650 .764 31 507 6/79- 0 1650 .000 61 446 7/79 8/79 967 1650 .586 31 415 9/79 1538 1650 .932 30 .385 10/79 1566 1650 .949 31 354 11/79 1599 1650 .969 30 324 12/79 1587 1650 .962 31 293 6-20

TABLE 6 6 (cont)

IRRADIATION HISTORY OF KEWAUNEE REACTOR VESSEL SURVEILLANCE CAPSULES Irradiation Decay P-j Pmax Time Time Month (MW) (MW) Pj/Pmax (days) (days) 1/80 939 1650 .569 31 262 2/80 1647 1650 .998 29 234 3/80 1615 1650 .979 31 203 4/80 1591 1650 .964 30 173 5/80 1417 1650 .859 10 163 6-21

TABLE 6-7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE V Radial Saturated Activity Adjusted Saturated Activity Reaction and Location (dis /s) (dis /s)

Axial Location tem) Capsule V Calculated Capsule V Calculated Fe54(n.p)Mn54 Top 158 33 5.22 x 106 5.38 x 106 5.48 x 106 Top-Middle 158.33 4.76 x 106 5.38 x 106 5.00 x 106 Middle 158.33 4.98 x 106 5.38 x 106 5.23 x 106 Bottom-Middle 158.33 4.93 x 106 5.38 x 106 5.18 x 106 Bottom 158.33 5.36 x 106 5.38 x 106 5.63 x 106 Averaga 5.30 x 106 5.65x106 cub 3(n.a)Co60 Top-Mddle 157.33 5.00 x 105 4.57 x 105 4.22 x 105 Bottom-Mddle 157.33 5.68 x 105 4.57 x 105 4.80 x 105 Average 4.51 x 105 3.86 x 105 th58(n.p)CoS8 Middle 157 33 8.55 x 107 9.84 x 107 7.27 x 107 8.37 x 107 Np237(n.fiCs137 Middle 158.10 7.84 x 107 7.01 x 107 7.84 x 107 7.01 x 107 U238(n.f)Cs137 Middle 158.10 8.89 x 106 7.72 x 106 8.89 x 106 7.72 x 106 6-22

l TABLE 6-8 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE R Radial Saturated Activity Adjusted Saturated Activity Reaction and Location (dis /s) (dis /s)

Axial Location (cm) Capsule R Calculated Capsule R Calculated Fe54(n.p)Mn54 Top 157.87 4.05 x 106 5.92 x 106 3.87 x 106 Top-Middle 157.87 3.76 x 106 5.92 x 106 3.59 x 106 Middle 157.87 3.80 x 106 5.92 x 106 3.63 x 106 80ttom 157.87 4.17 x 106 5.92 x 106 3.98 x 106 Average 3.77 x 106 5.65 x 106 Cu63(n,#)Co60 158.87 4.20 x 105 3.28 x 105 4.94 x 105 3.86 x 105 NiS8(n.p)CoS8 Middle 158 87 7.23 x 107 7.12 x 107 8.50 x 107 8.37 x 107 Np237(n.f)Cs137 Middle 158.10 not 7.01 x 107 not 7.01 x 107 determined - determined U238(n.f)Cs137 Middle 158.10 8.33 x 106 7.72 x 106 8.33 x 106 7.72 x 106 6-23

The fast neutron (E > 1.0 Mev) flux and fluence levals derived for Capsules V and R are presented in table 6-9. The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in table 6-10. Due to the relatively low thermal neutron flux at the capsule locations, no burnup cor-rection was made to any of the measured activities. The maximum error introduced by this as-sumption is estimated to be less than 1 percent for the NiS8 (n.p) CoS8 reaction and even less sig-nificant for all of the other fast neutron reactions.

Using the average measured data from all of the fast neutron dosimetry presented in table 6-9 along with the lead factors given in table 6-3, the fast neutron fluences (E > 1.0 Mev) for Capsules V and R as well as for the reactor vesselinner diameter are summarized in table 6-11 and figure 6-8. The agreement between calculation and measurement is excellent, with measured fluence levels of 5.99 x 1018 and 2.07 x 1019 n/cm2 compared to calculated values of 5.71 x 1018 and I

2.06 x 1019 n/cm2 for Capsuler V and R, respectively. Further, t'he graphical representation in figure 6 8 indicates the accurac) uf the transport analysis for Kewaunee and supports the use of the analytical!y determined fluence trend curve for predicting vessel toughness at times in the future.

i it should be noted that Westinghouse normally uses Fe54 (n.p) Mn54 measured resub to quote capsule exposures. However, the iron data obtained from Capsule R are inconsistent with the re- j maining dosimetry from Capsule R as well as with iron data from Capsule V. From table 6-11,it is  !

seen that the iron data from Capsule V agree with the overall average within 12 percent while for Capsule R the iron data are low by 61 percent. The reason for this discrepancy is not known. How-ever,it is believed that the average measured fluence is a betterindicator of Capsule R exposure than are the Fe54 (n.p) Mn54 data.

Projecting to end-of-life, a summary of peak f ast neutron exposure of the Kewaunee reactor as derived from both calculation and measurement may be made as follows.

Fact Neutron Fluence (n/cm2}

Surface 1/4 T 3/4 T Capsule V 4.56 x 1019 2.97 x 1019 8.96 x 1018 Capsule R 4.37 x 10.19 2.85 x 1019 8.59 x 1018 Avg Measurement 4.47 x 1019 2.91 x 1019 8.78 x 1018 Calculation 4.34 x 1019 2.83 x 1019 8.53 x 1018 These data are based on 32 full power years of operation at 1650 MWt.

6-24

l l

TABLE 6-9 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULES V AND R.

Adjusted Saturated Activity 4 (E > 1.0 Mev) 4 (E > 1.0 Mov)

(dis /s) (n/cm2-sec) (n/cm2)

Measured Calculated Measured Calculated Measured Calculated Capsule Reaction l

5.65 x 106 1.36 x 1011 1.45 x 1011 5.36 x 1018 5.71 x 1018 V Fe54(n.p)Mn54 5.30 x 106 -

4.51 x 105 3.86 x 105 1.69 x 1011 6.66 x 1018 Cu63(n.a)Co60

. N,58(n.p)CoS8 7.27 x 107 8.37 x 107 1.26 x 1011 4.96 x 1018 Np237(n,f)Cs137 7.84 x '.e7 7.01 x 107 1.62 x 1011 6.38 x 1018 U238(n,f)Cs137 8.89.*106 7,72 :1o6 1.67 x 1011 6.58 x 1018 l

$2

  • Fe64(n.p)Mn54 3.77 x 106 0 65 x 106 9.68 x 10 1,0 1.45 x 1011 1.37 x 1019 2.06 x 1019 R

4.94 x 105 3.86 x 106 1.86 x 1011 2.64 x 1019 l Cu63(n a)Co60 NiS8(n.p)CoS8 8 50 x 107 8.37 x 10 7 1.47 x 1011 2.09 x 1019

' Np237(n.f)Cs137 not 7.01 x 10 7 not -

I determined determined U238(n,f)Cs137 8.33 x 106 7.72 x 106 1.56 x 1011 2.22 x 1019 M

TABLE 6-10 RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULES V AND R Axial Saturated Activity (dis /s) Oth Capsule Location Bare Cd Covered (n/cm2-sec)

V top 1.65 x 108 7.33 r 107 1.62 x 1011 bottom not determined 7.40 x 107 not determined R top 1.35 x 108 e,70x107 1.20 x 1011 bottom 1.42 x 108 6.01 x 107 1.44 x 1011 6-26

18566 18 l

r 10 21 ,

LEGEND:

O CAPSULE V DATA 6 -

O CAPSULE R LiATA SNCALCULATION 4 -

2 130 CAPSULES (V, R)

@ 10 20 _

E y 8 -

S ~

w 6 -

- , VESSEL E

4 INNER

$a RADIUS

u. _

z o

5 2 -

8 z

10 19 -

8 -

6 -

4 -

l i

1 2

l 10 18 I I i 1 1 1 III I I I I I I11 10 0 2 4 6 8 10 1 2 4 6 8 10 2 OPERAliNG TIME (EFPY) i Figure 6-8. Comparison of Measured and Calculated Fast Neutron Fluence (E > 1.0 Mev) for Capsules V and R 6-27

i TA BLE 6-11

SUMMARY

OF FAST NEUTRON DOSIMETRY RESULTS FOR CAPSULES V AND R 1rr'a diation Calculated -

Time 4 (E > 1.0 Mev) 6 (E > 0.0 Mev) Lead Vessel Fluence Vessel Fluence Capsule (EFPS) (n/cm2-sec) (n/cm2) Factor (n/cm2) (n/cm2)

Based .,i1 the Fe54 (n.p) Mn54 Reaction V 3.94 x 107 1.36 x 1011 5.36 x 1018 3.37 1.59 x 1018 1.69 x 1018 R 1.42 x 108 9.68 x 1010 1.37 x 1019 3.37 4.54 x 1018 6.11 x 1018 Based on the Average of All Fast Neutron Reactions V 3.94 x 107 1.52 x 1011 5.99x1018 3.37 1.78 x 1018 1.69 x 1018 R 1.42 x 108 1.46 x 1011 2.07 x 1019 3.37 6.14 x 1018 6.11 x 1018 i

6-28

Based on the fluence measurements for Capsule R, the vessel II4 thickness fluence after 4.5 effec-tive full power years of operations is 3.77 x 1018 n/cm2 compared to a calculated fluence of 3.75 x1018 n/cm2, i

Based on the new capsule to vesselinner walllead factors identified in table 6-3 and the new capsule withdrawal schedule identified in ASTM E185-79,it is recommended that future capsules be removed from the reactcr in accordance with the following schedule:

Capsule Vessel Lead Capsule Fluence identity Location Factor Removal Time (n/cm2)

V 77* 3.37 1.25 EFPY (removed) 5.99 x 1018 R 257' 3.37 4.50 EFPY (removed) 2.07 x 1019 T 67* 1.94 11.00 EFPY 2.89 x 1019 [a]

P 247 1.94 16.00 EFPY 4.21 x 1019 [b]

S 57

  • 79

. 32.00 EFPY 8.42 x 1019 M 237 1.79 Standby -

I a Represents approximate Eol fluence at the vessel 1/4 thickness tt Represents approximate Eol fluence at the vessel inside surf ace I

6-29

REFERENCES l

1. Yanichko, S. E., D. J. Lege, arid G. C. Zula, " Wisconsin Public Service Corp. Kewaunee Nuclear Power Plant Reactor Vessel Radiation Surveillance Program" WCAP-8107, April 1973.
2. ASTM Designation E185-70 " Surveillance Tests for Nuclear Reactor Vessels"in ASTM Stan-dard (1971) Part 31, American Society for Testing and Materials, Philadelphia, Pa.,1971.

i

3. Yanichko, S. E., S. L. Anderson, and K. V. Scott, " Analysis of Capsule V from the Wisconsin Public Ser. ice Corporation Kewaunee Nuclear Plant Reactor Vessel Radiation Surveillance Program," WCAP-8908, January 1977.
4. Regulatory Guide 1.99, Revision 1,"Ef fects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, April 1977.
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-Dimensiona). Dhcrete 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 GAMBIT Code, WANL-TME-1969, November 1969.
7. .Soltasz, 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 Pro-cessing Techniques," WANL-PR-(LL)-034, August 1970/
8. Soltesz, R. G. and R. K. Disney, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation - Volume 4 - One-Dimensional Discrete Ordinates Transport Technique,"WANL-PR-(LL)-034, August 1970.
9. ASTM Designation E261-70, Standard Method for Measuring Neutron Flux by Radioactivation Techniques," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 745-755, American Society for Testing and Materials, Philadelphia, Pa.,1975.
10. ASTM Designation E262-70, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Teci-iques," 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," Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Iron," in ASTM Standards (1975), Part 45, Nuclear Standards, pp.

764-769, American Society for Testing and Materials, Philadelphia, Pa.,1975.

A-1

s s

12. ASTM Designation E481-73T," Tentative Method of Measuring Neutron-Flux Density by  ;

Radioactivation of Cobalt and Silver l* in ASTM Standards (1975), Part 45, Nuclear Standards, l pp. 887-894, American Society for 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.

j 770-774, American Society for Testing and Materials, Philadelphia, Pa.,1975.

i

14. Anderson, S. L.," Characterization of the Neutron Environment for Commercial LWR Pressure Vessel Surveillance Programs"in Proceedings of the Second ASTM-EUROTOM Symposium on Reactor Dosimetry, NUREG/CP-0004, October 1977, Vol. 3, pp.1093-1107.

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