ML19332C153
| ML19332C153 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 08/31/1989 |
| From: | Albertin L, Shaun Anderson, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19332C149 | List: |
| References | |
| WCAP-12354, NUDOCS 8911220368 | |
| Download: ML19332C153 (103) | |
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$' ANALYSIS OF CAPSULE X FROM THE
- DUKE POWER COMPANY
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MCGUIRE UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM <g p S. E.- Yanichko S. L. Anderson l L. Albertin i N. K. Ray
-August 1989 Work performed under Shop Order No. DSMJ-106 R
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APPROVED: - T.A.Meyer,Mabager Structural Materials and Reliability Technology Prepared by Westinghouse for.the Duke Power Company [U WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230 3487s/061440 10
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-PRETACE i
- This report has been technically reviewed and verified.
1 i Reviewer . 7 Sections 1 through 5 and 7, 8 E. Terek 8. M : Section 6 E. P. Lippincott Y # i C SY e
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- TABLE OF CONTENTS !
E Section Title Page i 1 SUMMAR'Y OF RESULTS 1-1
-2 INTRODUCTION 2-l' 3 BACKGROUND- 3 . .
4 DESCRIPTION 0F PROGRAM- 4-1 5 TESTING OF1 SPECIMENS FROM CAPSULE X 5-1 s 5-1. Overview 5-1 ; 5-2. Charpy V-Notch Impact Test Results 5-3 5-3. Tension Test Results 5-4 . L 5-4. Compact Tension Tests 5-5 -) l ? 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 , l 6-1. . Introduction 6-1 6-2.- Discrete Ordinates Analysis 6-2 6-3.- Neutron Dosimetry 6-7 . 7 -SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8-1 8- REFERENCES Appendix A - Heatup and Cooldown Limit Curves for Normal Operations - l l. sears /cei4soic jjj
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F 1 i l LIST OF ILLUSTRATIONS i Figure Title Page 4-1 Arrangement of Surveillance Capsules in the 4-5 , McGuire Unit 1 Reactor Vessel 4-2 Capsule X Diagram Showing Location of Specimens. 4-6 Thermal Monitors, and Dosimeters 5-1. Charpy V-Notch Impact Data for McGuire Unit 1 5-13 ) Reactor Vessel Shell Plate B5012-1 (Transverse Orientation) 5-2 Charpy V-Notch Impact Data for McGuire Unit 1 5-14 Reactor Vessel Shell Plate B5012-1 (longitudinal Orientation) ; 5-3 Charpy V-Notch Impact Data for McGuire Unit 1 5-15 Reactor Vessel Weld Metal 5-4 Charpy V-Notch impact Dats for McGuire Unit 1 5-16 ; Reactor Vessel Weld Heat Affected Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for McGuire 5-17 Unit 1 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 5-6 Charpy Impact Specimen Fracture Surfaces for McGuire 5-18 Unit 1 Reactor Vessel Shell Plate 85012-1 (Transverse Orientation) 5-7 Charpy impact Specimen Fracture Surfaces for 5-19 McGuire Unit 1 Reactor Vessel Weld Metal , 5-8 Charpy impact Specimen Fracture Surfaces for 5-20 , McGuire Unit 1 Reactor Vessel HAZ Metal - m.i.wmo gy
p V , L LIST OF ILLUSTRATIONS (Cont)
p figure Title Page 5-9 Tensile Properties for McGuire Unit 1 Reactor. 5-21
- Vessel Shell Plate B5012-1 (longitudinal Orientation) i 5-10 Tensile Properties for McGuire Unit-1 Reactor 5-22 Vessel Shell Plate B5012-1 (Transverse Orientation) 5-11 Tensile Properties for McGuire Unit 1 Reactor 5-23 Vessel Weld Metal 5-12 Fractured Tensile Specimens for McGuire Unit 1 5-24 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 5-13 Fractured Tensile Specimens for McGuire Unit 1 5-25 Reactor Vessel Shell Plate B5012-1 (Transverse Orientation) 5-14 Fractured Tensile Specimens for McGuire Unit 1 5-26 Reactor Vessel Weld Metal 5-15 Typical Stress-Strain Curve for Tension Specimens 5-27 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-12 i r
6-2 Core Power Distributions Used in Transport Calculations 6-13 For McGuire Unit 1 e newo.u.. io y
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p LIST OF TABLES Table Title Page l 4-1 Chemical Composition of the McGuire Unit 1 4-3 Reactor Vessel Surveillance Materials 4-2 Heat Treatment of the McGuire Unit 1 4-4 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for the McGuire Unit 1 5-6 ; Reactor Vessel Shell Plate B5012-1 Irradiated at 550'F Fluence 1.38 x 10 19 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the McGuire Unit 1 5-7 Reactor Vessel Weld Metal and HAZ Metal Irradiated . at 550'F, Fluence 1.38 x 10 19 n/cm2 (E > 1.0 MeV) 5-3 Instrumented Charpy Impact Test Results for McGuire 5-8 Unit 1 Reactor Vessel Shell Plate B5012-1 5-4 Instrumented Charpy Impact Test Results for 5-9 McGuire Unit 1 Reactor Vessel Weld Metal and HAZ Matal 5-5 The Effect of 550'F Irradiation at 1.38 x 10 19 n/cm 2 5-10 > (E > 1.0 MeV) on the Notch Toughness Properties of The McGuire Unit 1 Reactor Vessel Materials 5-6 Comparison of McGuire Unit 1 Reactor Vessel Surveillance 5-11 Capsule Charpy impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions 5-7 Tensile Properties for McGuire Unit 1 Reactor Vessel 5-12 Materialirradiatedto1.38x1019 n/cm2 (E > 1.0 MeV) i i l no.wm ie y3 l
N{' LIST OF TABLES (Cont) Table Title Page 6-1 Calculated Fast Neutron Exposure Rates at the 6-14 Surveillance Capsule Center 6-2 Calculated Fast' Neutron Exposure Parameters at the 6-15 Pressure Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux 6-16 i (E>1.0 MeV) Within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-17 (E>0.1 MeV) Within the Pressure Vessel Wall 6-5 Relative Radial Distribution of Iron Displacement 6-18 Rate (dpa) Within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-19 6-7 . Irradiation History of Neutron Sensors Contained 6 in Capsule X 6-8 Measured Sensor Activities and Reaction Rates 6-23 F 6-9 Summary of Neutron Dosimetry Results 6-25 , 6-10 Comparison of Measured and FERRET Calculated 6-26 Reaction Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-27 l Capsule Center -!
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p f' LIST OF TABLES (Cont) E :Teble
; Title. Page
'6-12' Comparison of Calculated and Measured Exposure 6-28 Levels for Capsule X i
( 6-13 Neutron Exposure Projections at Key Locations on the 6-29 Pressure Vessel Clad / Base Metal Interface 14 ' Neutron Exposure Values for Use in the Generation 6-30'
'of Heatup/Cooldown Curves 6-15 Updated Lead Factors for McGuire Unit 1 Surveillance 6-31 Capsules i
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i SECTION 1 !
SUMMARY
OF RESULTS !
.i The analysis of tho' reactor vessel material contained in Capsule X, the second !
surveillance capsule to be removed from the Duke Power Company McGuire Unit I reactor pressure vessel, led to the following conclusions: i o The capsule received an average fast neutron fluence (E > 1.0 MeV) l of 1.38 x 10 19 n/cm2 , ; t o Irradiation of the reactor vessel intermediate shell Plate B5012-1, to ; 1.38 x 10 19 n/cm, resulted in 30 and 50 ft-lb transition temperature [ increases of 65 and 55'F respectively, for specimens oriented normal to the major working direction (transverse orientation) and 45'F for specimens oriented parallel to the major working direction (longitudinal orientation). l Weld metal irradiated to 1.38 x 10 19 n/cm resulted in a 165 and 2 ! o 185'F increase in the 30 and 50 f t-lb transition temperature respectively. o Irradiation to 1.38 x 10 19 n/cm2 resulted in no decrease in the I average upper shelf energy of Plate B5012-1 (transverse orientation) ; and an upper shelf energy decrease of 29 ft-lbs for the weld metal. i Both materials exhibit a more than adequate shelf level for continued safe plant operation. > o Comparison of the 30 ft-lb transition temperature increases for the McGuire Unit I surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2, shows that the Plate B5012-1 material transition temperature increase was 4'F greater than predicted. This increase is bounded by the 2 sigma allowance for , shift prediction of 34'F. The weld metal showed a transition temperature increase that was 64'F less than the prediction. w.num = 11
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F l t Impact Of Test Results On Plant Life Extension i o The measured ARTNDT values are significantly lower than those values predicted at 1.38 x 10 19 n/cm2 (-22 EFPY) for the axial welds. This can provide-for less restrictive ASME, Section III, Appendix G heatup and cooldown curves for future plant life. The future surveillance capsule's test data will be required to determine what potential benefit, if any, L may be utilized for heatup and cooldown curves developed for an extended vessel life, i.e. Plant Life Extension. . o PTS margin should exist for some amount of life extension beyond the current license life of the McGuire Unit 1 based on the predicted values of RT PTS. The data reported here can imply additional PTS margin since the measured RTNDT values for the axial weld material are significantly less than the predicted values using Regulatory Guide 1.99, Revision 2 prediction methods. However, this benefit cannot be readily obtained since the PTS rule requires the use of only predicted RTNDT (i.e. RTPTS) values. an.ncm. io 1-2
g - , i p I ): SECTION 2 i INTRODUCTION J This report precents the results of the examination of Capsule X, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Duke Power Company McGuire Unit I reactor pressure vessel materials under actual operating conditions. . The surveillance program for the Duke Power Company McGuire Unit I reactor l pressure vessel materials was designed.and recommended by the Westinghouse Electric Corporhtion. A description of the surveillance program and the ' preirradiation mechanical properties of the reactor vessel materials are_ presented by Davidson and Yanichko.I13 The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based on ASTM E-185-73, " Recommended Practice for Surveillance Tests for Nuclear
- Reactor Vessels". Westinghouse Energy Systems personnel were contracted to
- aid in the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechardeal testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes testing and the postirradiation data obtained from surveillance C6psule X removed from the Duke Power Company McGuire Unit 1 reactor vessel and discusses the analysis of the data. The data are also ' compared to capsule U(2) which was removed from the teactor in 1984, no.w... io 2-1
p j 1 i e ! i 1 SECTION 3 i BACKGROUND , i 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 vessel is 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 mechanical properties of low alloy ferritic l' pressure vessel steels such as SA533 Grade B Clast,1 (base material of the McGuire Unit i 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. t A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non-ductile Failure," Appendix G to Section !!! of the ASME Boiler and Pressure Vessel l Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT}* i i f RT NDT is defined as the greater of either the drop weight nil-ductility , transition temperature (NDTT per ASTM E-208) or the temperature 60*F less than ; the 50 ft 1b (and 35-mi.1 lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction ! of the material. The RT NDT f a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel, When a given material is indexed to mr.mue 1o 31
the Kgg curve, allowable stress iitensity factors can be obtained for this material as a function.of temperature. Allowable operating limits can then be f _ determined utilizing these allowable stress intensity factors, p I RTNDT 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 McGuire Unit 1 Reactor Vessel Radiation Surveillance Program .Ill in which a surveillance capsule is
'periodica11y' removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft ib temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RT NDT for radiation embrittlement. This adjusted RTNDT (RT NDT initial + ARTNDT) is used to index the material to the Kyp 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,
- w. aim io 3-2
f' I SECTION 4 , DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the McGuire Unit I reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the neutron shield pads and the vessel wall at locations shown in Figure 4-1. lhe vertical center of the capsules is opposite , , the vertical center of the core. Capsule X (Figure 4-2) i<as removed af ter 4.33 effective full power years of plant operation. This capsule contained Charpy V-notch impact, tensile, and 1/2T - Compact Tension fracture mechanics specimens from the reactor vessel intermediate shell Plate B5012-1, submerged arc weld metal representative of the ' beltline region intermediate shell longitudinal weld seams and Charpy V-notch specimens from weld heat-affected zone (HAZ) material. All heat-affected zone specimens were obtained from within the HAZ of Plate B5012-1 of the representative weld. The chemistry and heat treatment of the surveillance material are presented in Table 4-1 and Table 4-2, respectively. 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 irradiated Charpy specimens from the intermsdiate shell Plate B5012-1 and weld metal and is reported in Table 4-1. All test specimens were machined from the 1/4 thickness location of the plate. Test specimens represent material taken at least one plate thickness from the quenched end of the plate. All base metal Charpy V-notch impact and tensile specimens were oriented with the longitudinal axis of the specimen both normal ' to (transverse orientation) and parallel to (longitudinal orientation) the mr.wm to 41
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j principal working direction of the plate. Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse I to the weld direction. Tensile specimens were oriented with the longitudinal axis of the specimens normal to the welding direction. The 1/2T Compact Tension l (CT) test specimens in Capsule X were machined such that the simulated crack in l the specimen would propagate normal and parallel to the major working direction ) for the plate specimens and parallel to the weld direction for weld specimens. l All specimens were fatigue precracked per ASTN E399-70T. ; i
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Capsule X contained dosimeter wires of pure iron, copper, nickel, and unshielded j aluminum-cobalt. In addition, cadmium-shielded dosimeters of Neptunium l (Np237)andUranium(U238) were contained in the capsule. Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule and were loceted as shown in Figure 4-2. The 4 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) The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in-Capsule X are shown in Figure 4-2. , l l i no.w. no 42 i
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i TABLE 4-1
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CHEMICAL COMPOSITION OF THE MCGUIRE UNIT 1 REACTOR VESSEL : SURVEILLANCE MATERIALS ; Plate B5012-1 Weld Metall "3 Element (Wt. %) (Wt. %) i C 0.21 - 0.10 -
-S 0.016 -
0.008 - N 0.003 - 0.008 -
;
2 Co 0.016 - 0.014 - Cu 0.087 - 0.21 0.20 Si 0.23 - 0.24 0.23 1 Mo 0.57 - 0.55 0.54 Ni 0.60 - 0.88 0.91 Mn 1.26 - 1.36 1.19 > b Cr 0.068 - 0.04 0.05 V 0.003 - 0.04 - P 0.010 - 0.011 0.010 Sn 0.007 - 0.007 - Ti 0.005 -
<0.010 -
Pb 0.001 -
<0.001 -
W <0. 001 -
<0.0100 -
2r <0.003 -
<0.001 -
As 0.008 - 0.009 - Cb < 0. 001 -
<0.010 -
I B <0.003 -
<0.001 -
l Sb <0. 001 - 0.002 - p (a) Surveillance weld specimens were made of the same weld wire and flux as ! the intermediate shell longitudinal weld seams (Tandem Weld Wire Heats l 20291 and 12008 and Linde 1092 Flux Lot 3854) l (b) Analysis performed on irradiated Charpy weld specimen DW-15 from capsule U.
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L j 1 E TABLE 4-2 j j HEAT TREATMENT OF.THE MCGUIRE UNIT 1 ! REACTOR VESSEL SURVEILLANCE MATERIALS ! i.. . !
. Material Temperature (*F) Time (hr) Coolant !
i o . .s !: Intermediate Shell 1550/1650- '4 . Water quenched- ; Plate B5012-1 - 1200/1250 4- Air cooled , f 1125/1175 40 furnace cooled- i !- i V . . t
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'l SECTION 5 ;
TESTING OF SPECIMENS FROM CAPSULE X 5-1. OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile r specimens was performed at the Westinghouse Research and Development l Laboratory with consultation by Westinghouse Nuclear Energy Systems '
; personnel. Testing was performed in accordance with 10CFR50, Appendices G and HI33, ASTM Specification E185-82 and Westinghouse Procedure MHL 8402, Revision 1 as modified by Westinghouse RMF Procedures 8102, Revision 1 and 8103, Revision 1.
1 Upon receipt of the capsule at the laboratory, the specimens and spacer blocks . were carefully removed, inspected for identification number, and checked l against the master list in WCAP-9195.I13 No discrepancies were found. : Examination of the two low-melting 304'C (579'F) and 310*C (590'F) eutectic alloys indiceted no melting of either type of thermal monitor. Br. sed on this examination, the maximum temperature to which the test specimens were exposed was less than 304*C (579'F). The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358] machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology - model 500 instrumentation system. With this systen, 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 (Pgy), the time to general yielding (tGY), the maximum load (PM ), and the-time to maximum load (tM) can be determined. Under some test l ust,misse to 5-1 l
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. conditions, a sharp drop in load indicative of fast 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 is identified as the arrest load (PA)* , The energy at maximum load (E g ) was determined by comparing the energy-time , record and the' load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. l Therefore, the propagation energy for the crack (E p
) is the difference between the total energy to fracture (E )Dand the energy at maximum load.
1-The yield stress (oy) is calculated from-the three point bend formula. The flow stress is calculated from the average of the yield and maximum loads, also usino the three point bend formula. Percentage shear was determined from postfracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77. The , lateral expansion was measured using a dial gage rig similar to that shown in the same specification. Tension tests were performed on a 20,000 pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-83 and E21-79, and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc45. 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. 4 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 E83-67. Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch not zone. All tests were conducted in air. mwmm io 5-2
i Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature. Chromel-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 (288'C). 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 indicated that this method is accurate to plus or minus 2'F.
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 results of Charpy V-notch impact tests performed on the various materials 2 contained-in Capsule X irradiated to approximately 1.38 x 10 19 n/cm at 550'F are presented in Tables 5-1 through 5-4 and Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the Capsule X material are shown in Table 5-5. Irradiation of the vessel intermediate shell Plate B5012-1 material (transverse orientation) specimens to 1.38 x 10 19 n/cm2 (Figure 5-1) resulted in a 30 and 50 ft-lb transition temperature increase of 65 and 55'F respectively, and an upper shelf energy increase of I f t-lb when compared to the unirradiated data. Irradiation of the vessel intermediate shell Plate B5012-1 material (longitudinal orientation) specimens to 1.38 x 10 I9 n/cm2 (Figure 5-2) resulted in a 30 and 50 ft-1b transition temperature increase of 45'F and an
- w. noun so 53
I' I i r i upper shelf energy decrease of 7 ft-lb when compared to the unirradiated data.[1] Weld metal irradiated to 1.38 x 1019 n/cm2(Figure 5-3)resultedina30 and 50 ft-lb transition temperature increcse of 165 and 185'F respectively and an upper shelf energy decrease of 29 ft-lb. Weld HAZ metal irradiated to 1.38 x 10 19 n/cm2 (Figure 5-4) resulted in a 30 and 50 f t-lb transition temperature increase of 115 and 120'F respectively and an upper shelf energy decrease of 22 ft-lb. , The fracture appearance of each irradiated Charpy specimen from the various
- materials is shown in Figures 5-5 through 5-8 and show an increasing ductile or tougher appearance with increasing test temperature.
Table 5-6 shows a comparison of the 30 f t-lb transition temperature (ARTNDT) increases for the various McGuire Unit I surveillance materials ; with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2.I43 This comparison shows that the transition temperature increase resulting from irradiation to 1.38 x 10 19 n/cm2 is less than predicted by the Guide for Plate B5012-1 longitudinal specimens but 4'F higher than predicted for transverse specimens. The weld metal transition temperature increase resulting from 1.38 x 10 19 n/cm2 is less than the Guide prediction. ; 5-3. TENSION TEST RESULTS The results of tension tests performed on Plate B5012-1 (transverse and longitudinal orientation) and weld metal irradiated to 1.38 x 10 19 n/cm2 are shown in Table 5-7 and Figures 5-9, 5-10 and 5-11, respectively. These results show that irradiation produced a 10 to 15 Ksi increase in 0.2 percent yield strength for Plate B5012-1 and 18 to 25 Ksi increase for the weld metal. Fractured tension specimens for each of the materials are shown in Figures 5-12, 5-13 and 5-14. A typical stress-strain curve for the tension specimens is shown in Figure 5-15. wo.e.. io 5-4
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- 5-4. COMPACT TENSION TESTS e
Per the surveillance capsule testing contract with the Duke Power Company.. 1/2T - Compact Tension Fracture Mechanics specimens will not be tested and , will be stored at the Hot Cell at the Westinghouse R&D Center. l
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t 6 u n. = i. io 5-5
;
j TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE MCGU1RE UNIT 1 REACTOR VESSEL SHELL PLATE B5012-1 ) 19 IRRADIATED AT 550'F, FLUENCE 1.38 x 10 n/cm2 (E > 1.0 MeV) l 5 Temperature Impact Energy Lateral Expansion Shear f Sample No. 1*f.). G (ft-lb) M (alls) 13R). A ; Lonnitudinal Orientation DL50 0 (-18 8.0 ( 11.0) 7.0 (0.18) 5 ; DL55 25 (- 4 18.0 ( 24.5) 13.0 (0.33) 10 . DL51 40 ( 4 25.0 ( 34.0) 21.0 (0.53 10 ; DL47 50 10) 35.0 (47.5) 26.0 (0.66 15
)
DL53 50 10) 39.0 ( 53.0) 29.0 0,74 20 DL60 74 23) 44.0 ( 59.5) 33.0 0.84) 30 DL59 74 23) 55.0 ( 74.5) 36.0 0.91) 30 - DL57 100 62.0 84.0) 84.0 1.17 40 DL54 125 73.0 99.0) 56.0 (1.42 60 , DL46 150 85.0 115.0 68.0 (1.73 75 ' DL58 200 105.0 (142.5 72.0 (1.83) 85 DL49 250 (121) 136.0 (184.5 80.0 (2.03) 100 DL52 300 (149) 121.0 (164.0 70.0 (2.01 100 DL48 350 (177) 134.0 (181.5 90.0 (2.29 100 DL5$ 400 (204) 140.0 (190.0 79.0 (2.01 100 Transverse 1rientatLon DT58 0 -18 13.0 ( a7.5 :.2.0 (0.30) 10 DT54 25 - 19.0 ( 26.0 18.0 0.46) 15 , DT57 50 19.0 26.0 19.0 0.48 20 DT52 60 15.0 20.5 18.0 0.46 20 i DT53 74 49.0 66.5 37.0 (0.94 35 l DT50 74 35.0 47.5 26.0 (0.66 25 t DT60 100 56.0 76.0 48.0 (1.17 45 DT49 100 45.0 61.0 38.0 (0.97 40 DT55 125 49.0 66.5 42.0 (1.07 50 . l DT59 150 62.0 84.0 50.0 1.27 60 - l DT47 200 94.0 127.5 69.0 1.75 100 L DT56 250 1 103.0 (139.5 74.0 1.88 100 l DT48 300 1 113.0 (153.0 81.0 (2.06 100 DT51 350 177) 93.0 (126.0 76.0 (1.93) 100 DT46 400 (204) 100.0 (148.0 76.0 (1.93) 100 uswuun io 5-6
-- - . _ . .. -. - -. - - - - - . ~ _ _ - _ - .
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE MCGUIRE UNIT 1 l REACTOR VESSEL WELD METAL AND HAZ METAL IRRADIATED AT 550*F ! 19 FLUENCE 1.38 x 10 n/cm2 (E > 1.0 MeV) [ Temperature Impact Energy Lateral Expansion Shear : Sample No. G G (ft-lb) 2), (mils) 133), .J)),,,, ; i Welcl Metal DW21 74 23) 14.0 i 19.0) 19.0 (0.48 15 DW49 100 38) 15.0 20.5) 14.0 (0.36 20 : DW53 125 52) 17.0 23.0) 20.0 (0.51 20 ; DW56 150 66) 29.0 39.5) 28.0 (0.71 35 ! DW59 150 ( 66 MACHINE MAIJUNCTION - - DW50 150 (66 33.0 (44.5 28.0 (0.71) 40-DW58 175 ( 79 38.0 51.5 27.0 .(0.89) 50 DW46 175 ( 79 39.0 53.0 34.0 (0.86 65 . DW60 200 ( 93 39.0 53.0 '36.0 (0.91 65 i DW57 210 (99 43.0 ( 58.5 34.0 (0.86 75 DW52 225 107 83.0 (112.5)' 57.0 (1.45 100 DW48 225 107) MACHINE MAIJUNCTION - - DW47 300 149) 87.0 (118.0 63.0 1.60 100 DW55 350 177) 83.0 112.5 67.0 1.70 100 DW54 400 (204) 79.0 107.0 64.0 1.63 100 RAZ Metal DB46 - 25 -3 9.0 12.0 9.0 (0.23 10 DR54 25 - 15.0 20.5 11.0 (0.28 15 DB53 40 26.0 35.5 17.0 (0.43 20 DE58 50 28.0 38.0 24.0 (0.61 30 DB53 50 34.0 46.0 24.0 (0.61 30 DR55 74 33.0 44.5 29.0 (0.74 40 DE50 74 54.0 73.0 34.0 0.86 50 : DB48 100 32.0 43.5 23.0 0.61 35 l L DR56 125 53.0 72.0 45.0 1.14 50 D547 150 77.0 104.5 58.0 1.47 75 DB49 200 59.0 80.0 52.0 (1.32 60 l DE51 250 92.0 (124.5 71.0 (1.80 100 DR59 300 100.0 148.0 74.0 (1.88) 100 D857 350 95.0 129.0) 75.0 (1.88) 100 DB60 400 (204 88.0 119.5) 66.0 (1.68) 100 l w.menee no 57
TABLE 5-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR MCGUIRE UNIT 1 l l REACTOR VESSEL SHELL PLATE B5012-1 l l Normalised Emerates Test Charpy Charpy Maximes Prop Yield Time Naminum Time to Fracture Arrest Yield Flow- i Sample Temp Energy Ed/A Ee/A 2 Ep/A Load to Yield Load Maximum Load Load Stress Stress Number ,[* 1 (f t-1b) (ft-lb/in 1 (hips) _ (msec) (kips) (ssec) (hips) [ hips) (ksi) (ksi) Lommitudinal Orientation DL50 0 8.0 64 33 31 3.50 25 3.90 250 3.90 -
-1 64 DL55 25 18.0 145 60 85 1.50 50 3.90 405 3.85 0.38 50 89 ,
DL51 40 25.0 201 149 53 3.25 0 4.15 580 4.15 - 0 69 > DL47 50 35.0 282 171 111 3.10 20 3.95 774 3.85 - G 65 l DL53 50 39.0 314 252 62 3.05 85 4.95 530 4.55 0.15 101 126 DL60 74 44.0 354 224 131 3.10 120 4.35 530 4.35 0.00 102 123 ' DL59 74 55.0 443 311 132 2.20 50 4.50 055 4.40 0.85 73 111 DL57 100 62.0 499 269 230 3.00 15 4.10 1065 4.05 0.90 0 68 T DL54 125 73.0 588 364 224 3.60 40 4.40 945 4.10 2.00 36 91
- 85.0 3.35 4.30 DL46 150 684 271 414 110 605 3.75 2.20 108 125 :
i DLS8 200 105.0 845 282 563 2.85 25 4.10 740 3.10 2.15 -1 67 , DL49 250 136.0 1095 319 776 2.55 40 4.10 785 - - 50 93
DL52 300 121.0 974 299 675 2.85 100 3.95 720 - - 93 112 DL48 350 134.0 1079 253 826 2.40 150 3.60 720 - - 79 99 < DL56 400 140.0 1127 297 830 2/65 95 3.80 740 -- - 88 107 Transverse Orientation DT58 0 13.0 105 70 34 3.55 10 4.00 320 3.95 - 0 66 DT54 25 19.0 153 108 45 3.20 25 3.65 570 3.65 -
-1 80 DT57 50 19.0 153 95 58 3.00 15 3.50 805 3.50 0.35 0 58 i Irr52 80 15.0 121 40 81 3.00 10 3.45 450 3.15 0.55 -1 57 DT50 74 35.0 282 180 101 3.20 55 4.30 580 4.25 0.25 0 71 ,
49.0 3.15 4.35 4.35 i DT53 74 395 216 179 55 470 1.15 104 124 DT49 100 45.0 362 209 154 3.00 20 4.30 550 4.10 1.15 41 90 DT60 100 56.0 451 293 158 3.15 40 4.25 905 4.20 1.15 0 70 DT55 125 49.0 395 247 148 2.76 30 4.10 740 4.05 1.35 -1 67 DT59 150 62.0 500 COMPtT11R NALFUNCTION - - - - - - - DT47 200 94.0 757 277 480 0.00 25 4.10 980 - - 0 67 ; DT56 250 103.0 829 259 571 2.65 60 3.95 630 - - 87 109 DT48 300 113.0 910 254 656 2.50 65 3.90 625 - - 82 106 DT51 350 93.0 749 224 525 0.00 10 3.80 870 - - 0 63 DT46 400 109.0 878 244 634 2.20 25 3.85 605 - - 73 100
- wveesm se i
TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR MCGUIRE IMIT 1 REACTOR VESSEL WELD METAL AND HAZ METAL 3 Moraalised Emeraies l Test Charpy Charpy Maximum Prop Tield Time Masimum Time to Fracture Arrest Yield Flow 4 Sample Temp Energy Ed/A Em/A 2 Ep/A Load to Yield Load Maxieue I.ced Lead Stress Ptress
- Number ,(*F1 fit-lb) fft-lb/in ) [ kips) (seec) (hips) fasec) (kips) [hipe] (ksi) (ksi) i Weld Metal i DW51 74 14.0 113 44 69 2.60 55 3.80 130 3.80 0.25 86 106 i DW49 100 15.0 121 61 60 3.40 20 3.60 795 3.60 - 0.35 0 59 DW53 125 17.0 137 68 69 3.00 10 3.40 700 3.35 0.70 -I 56 DW56 150 29.0 234 142 92 3.30 5 3.80 560 3.80 1.40 0 62 /
- DW59 150 MACEI M MALFUNCTION - - - - - - - - - -
DW50 150 33.0 266 180 86 3.50 10 4.10 665 4.10 1.30 -1 68 DW58 175 38.0 306 211 95 3.00 15 4.06 755 4.06 1.70 0 67 4 DW46 175 39.0 314 172 142 3.40 20 4.06 855 4.06 2.40 0 67 DW60 200 39.0 314 177 137 3.25 15 4.00 915 3.85 1.70 0 66 m DW57 210 43.0 346 217 129 3.10 0 4.10 700 4.10 2.30 0 67 to DW52 225 83.0 666 229 440 3.2 170 4.25 570 - - 106 123 DW48 225 MACRI E MALFUNCTION - - - - - - - - - - DW47 300 87.0 701 244 457 0.00 5 4.05 870 - - 0 67 DW55 350 83.0 668 237 432 -7.45 20 3.95 840 - -
-1 65 j DW54 400 79.0 636 221 415 -7.45 5 3.95 855 - -
-1 65 EAX Metal l DH46 - 25 9.0 72 51 21 3.40 65 3.75 145 4.05 0.20 88 106 DH54 25 15.0 121 40 80 3.70 5- 3.50 435 3.50 0.85 3 58
! DH52 40 26.0 209 119 91 3.60 5 4.20 460 4.20 0.95 0 69 DH58 50 28.0 225 118 107 3.60 50 4.20 335 4.20 1.75 50 94 I DH53 50 34.0 274 144 130 3.30 10 3.95 675 3.90 1.30 -1 65 DB55 74 33.0 266 117 149 3.60 5 4.15 545 4.15 2.85 0 68 DH50 74 54.0 435 225 209 3.40 45 4.50 460 4.40 2.50 112 130 DH48 100 32.0 258 192 66 3.40 85 4.25 425 4.25 1.06 112 127 DH56 125 53.0 427 236 191 3.50 10 4.45 740 4.35 3.95 -1 73 DH47 150 77.0 620 287 333 3.00 5 4.20 750 4.10 3.75 0 69 i DH49 200 59.0 475 293 182 3.25 10 4.10 1215 3.95 1.95 0 68 DH51 250 92.0 741 287 454 3.15 120 4.10 705 0.25 0.25 76 106 DH59 300 109.0 878 313 565 3.00 5 4.00 1175 - - 0 66 ' DH57 350 95.0 765 266 500 2.50 50 4.05 620 - - 83 109 DH60 400 88.0 709 199 510 3.10 15 3.08 800 - -
-1 62
399Fs/58f 499 TO I
,, .. ,_ _ . ~ _ . . .. _ . . . . . . . _ . . .-_ , - - , -. ... .-._.._ ..-._.,_ _-_ ._ _ -..
TABLE 5-5
-THE EFFECT OF 550*F IRRADIATION AT 1.38 x 100 n/cm2 (E > 1.0 MeV) 4 ON THE NOTCH TOUGHNESS PROPERTIES OF THE MCGUIRE UNIT 1 REACTOR VESSEL MATERIALS Average Average 35 att Average Average Energy Absorptfon 30 ft-1b Temp (*F) Lateral E xpans ton Temp (
- F ) 50 f t -10 Temp *
- F ) at FU1.4 Shese (ft-tD) ,
Material Unteradtated Irradiated AT Unteradiated Irradiated AT Untrradiated Irradiated AT Untrraatated Irradtated A(ft-10) Plate 85082-1 5 50 45 35 75 40 35 80 45 140 133 -7 (LongltudInaI) Plate B5012-1 0 65 65 50 95 45 75 130 55 101 102 +1
, (Transverse) b Weld Metal -5 160 165 0 190 190 20 205 185 112 83 -29 HAZ metal -50 65 115 -15 100 115 -5 115 120 118 96 .-22 4
4 3887s/081489:10 ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . . , ~_ . _..._. _ _- _ _ _ _ . . _ . -
TABLE 5-6 COMPARISDN OF MCGUIRE UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS ARTNDT ( F) USE DECREASE (%) Fluence R.G 1.99 R.G 1.99 19 2 Pred. Material Capsule 10 n/cm Meas. Pred. Meas. Plate B5012-1 U 0.414 45 42 5 15 (Longitudinal) X 1.38 45 61 5 20 Plate B5012-1 0 0.414 50 42 1 15 (Transverse) X 1.38 65 61 0 20 Weld Metal U 0.414 160 159 33 28 X 1,38 165 229 26 37 6 msnmi .. io 5-11 L a
L TABLE 5-7 TENSILE PROPERTIES FOR MCGUIRE UNIT 1 REACTOR VESSEL MATERIAL IRRADIATED TO 1.38 x 10 I9 n/cm2 (E > 1.0 MeV) i Teet 0.25 Yield Ultimate Fracture Fracture Fracture Uniform Total Beductice Sample Temp. Strength Strength Load Stress Strength Blongation Blongation in Ares l Material Number ,(*F1 (ksi) (ksi) (hip) (ksil (ksi) (5) (5) (5) i i Plate DL11 74 78.9 98.9 3.10 175.4 63.2 10.5 25.2 64 l B6012-1 DL12 200 75.4 94.7 3.06 178.5- 62.1 9.8 22.2 65 m (Long. DL10 560 69.8 91.7 3.10 208.8 63.2 9.8 21.2 57 y Orient.) - t Plate D711 74 76.4 96.8 3.45 171.6 70.3 12.0 24.6 64
- B5012-1 M12 200 73.3 91.5 3.10 180.2 63.2 10.5 23.1 64 !
(Transv. D710 560 67.0 90.7 3.65 132.2 74.4 9.0 12.9 80 Oriest.) Weld N11 175 84.0 97.8 3.45 216.3 70.3 12.0 24.1 88 i W12 225 78.4 89.6 3.35 189.6 68.2 10.5 20.9 64 l N10 550 77.4 94.3 3.30 174.9 67.2 9.0 19.1 62 i l 1 i i I i $ 3nervoe1*e9 to
. _~ - .,~,,.a . , , - .
. . ._= - - _ _ _ . - . _ _ _ _
l'
-l p.- Curie 757276 A
('C) .i
-150 -100 -50' O 50 100 150 200 250 I- I I I I I 1 3 I 3
o
=
100 - J= = = = - 0 o
?, 80 .
- ;
_ g 60 -- m
.e - @ -
2 - E _ 2 0 ' ' ' ' ' ' ' 100 T- 2.5 _ i i i i i i , , ; E 80 2.0 E _ i
-2 - * -
L 51
.. I# -
o = 45.p - LO ' { . 20.0 [
' ' ' ' i
- 0. 5 0-2% , , , , , ,
180 - 24 12 -
., ,. 14 - 200 a
Q120 2 . 160
~ 100 - Unirradiated -
i F
= 80 -
o 13 0 3 g _ Irradiated ( 550'F) - 80 e _ g 55*F l'. 38 x10 n/cm 19 2 20 0 i i
.v A i i i i i i 0
-200 -100 0 100 200 300- 40 500 Temperature (*F)
FIGURE 5-1 CHARPY V-NOTCH IMPACT DATA FOR MCGUIRE UNIT 1 REAC10R VESSEL SHELL PLATE B5012-1 (TRANSVERSE ORIENTATION) nei.ameno 5-13 1
i ., curve 757277-A s (
- C) .
-150 -100 -50 0 50 100 150 200 250 1 22'3 ' '
100 - v-v - : : : - 1
- n -
g2 - 5@ - m - 2 2 1 0 ' ' ' ' ' ' ' 100 i , , , 2.5 5%E
- o . . , -
20
-2 --
L5 e fe -
=
4o.7
- 1. 0 "
Y$~i 2E i i i i i i -I' i- i i i i i i i i 180 - 24 g 160 - le -
- p. .
2W g120 - Unirradiated * * - 160 (g
~ 100 -
_ Irradiated ( 550*F) - 120 3 6 g o 1. 38 x 10 19 n/cm 2 L
- _ g M -
o)6[- 45'F 45'F - 20 0 i i [ i i i i i 0
- 200 -100 0 100 200 300 40 500 Temperature (*F)
FIGURE 5-2 CHARPY V-NOTCH IMPACT DATA FOR MCGUIRE UNIT 1 REACTOR VESSEL SHELL i PLATE B5012-1 (LONGITUDINAL ORIENTATION) u n . m u n ic 5-14
l- , n Curva 757 278-A ( ' C) ,
-150 -100 -50 0 50 100 150 200 250 I I I I I I I I i 3
100 - e / _ _ _ o f 80 -
) 60 -- .o -
Si # - 20 - 2 0 ' - ' ' ' i i I I I I I I I I I l
~
c a l Eg -
= -
10 i E * ^ ' L - 60 - o r -
- 1. 5 ^
8@ - o 190*F _ 1.0 4'" ,
/ 7, , , ,- t' .
200 , , , , , , , , , 180 - 2@ 160 - 200 _ l@
.ru
]120 -
160'
~ 100 - Unirradiated o -
h 80 - *
* ;
120 C e o . 60 - o185'F Irradiated ( 550*F)~
/165'F_.[# 19 2 o 1. 38 x10 n/cm - @
20 0 i 2 o i 8 i i i i i 0
- 200 -100 0 100 200 300 40 500 ,
l- Temperature (*F) l l FIGURE 5-3 CHARPY V-NOTCH IMPACT DATA FOR MCGUIRE UNIT 1 REACTOR VESSEL WELD l METAL mwaunao s.15 t
. . . - . . . . . , . _ , , . - . . . - . - , , . . . . , . . . - _ _ . _ . . , _ . , , , - - . . . . . . - . - , .,wr _ .=,. - ,
Curve 757279-A y -(oC)
-150 -100 ~50 0- 50 100 150 200 250 I i i ' ' 3 ' i 2 12 100 -
oJ d- : : : - t 80 - o , _ gM - o o
- di @ _
2 .- E -
' ' I i 0 i i i I# 1 i i i 2.5-
_ _i i i i , EMe o ^ yo
^
~, e., e 10
-M -
o o L5 e I'# - o 1.03 I$~i I , , , o 200 i i i i i i i l 1 180 - 24 160 -
- _ 14 a
- 200
- O o g120 -
Unirradiated - l@ o *
~ 100 -
o _ _ 80 - o e - 120 C l w L M - 80 120'F 40 - og 7 Irradiated ( 550*F)
~
19 20 - 1.38 x 10 n/cm 0 i i i i i i i 0
- 200 -100 0 100 200 300 40 500 Temperature (*F)
L FIGURE 5-4 CHARPY V-NOTCH IMPACT DATA FOR MCGUIRE UNIT 1 REACTOR VESSEL WELD HEAT AFFECTED ZONE METAL l l' mer. m m eio 5-16 l l l
l l
?W% q -- ; sen g ,: ,,
?
( 5 (4,- l t l l
- "] J f Skf l 'k ,
g= -LE: . , _ .: i DL50 DL55 DL51 DL47 DL53
. I4 4 g l vi[ j'h EID N-(hb Jn g- - .J., #1R DL60 DL59 DL57 DL54 DL46 "g,s+4 1 EDL58
? te DL49
{. DL52 A
~
DL48 DL50 FIGURE 5-5 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR MCGUIRE UNIT 1 REACTOR VESSEL SHELL PLATE B5012-1 (LONGITUDINAL ORIENTATION) u s ** "' ' 5-17 RM-19840
g, se ,-g;p g-c ,
, ,w3-7,yg .m DT58 DT54 DT67 DT52 DT53 h i* .' fI ie
'4 ,A ;
- f. 3
'p b . ),t il, .3. }l \p ; -- - ~a +.,, .
DT50 DT60 DT49- 'DT55 ::DT59 DT47 DT56 DT48 DT51 DT46 FIGURE 5-6 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR MCGUIRE UNIT 1 REACTOR VESSEL SHELL PLATE B5012-1 (TRANSVERSE ORIENTATION) 3887s/061449 10 h.}g RM-19841
i .
- j. !
L i~ i k __--w .e . , , - -. . . , --.,, , . , .; , . - . . . ,
. g. ;
m _ . g? :DW61 DW49' . L. DW58 - DW58?' i < v [ ' t Wf .; #tA g h - l C[f i.
. is '
DW50 F['k$i-:
; DW587 1
-~DW46h
_' 2.
'DW80-' DW57
(
', 1 r *'
)
DW52.
. DW47 i DW55 :DW54 FIGURE 5-7 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR MCGUIRE UNIT 1 REACTOR VESSEL WELD METAL unwo uano 5-19 -
RM-19842
., , . . - . . _ _ _ . ~ . _ . . . _ , _ . . . . . . . . . . _ _ . _ . . . _ . . _ _ _ - . _ _ _ . _ _ _ _ . _ . _ . . . _ . . - _ _ _ . . . . . . _ . _ _ _ . _ _ . . _ . _ _ . , _
l 1 , 1 L ,', ! :
.- . - .. s; '"~" .: "~ -
-"e
- DH46 DB54 DH52 DH68 DH53 DH55 DH50 DH48 DH56 DH47 DH49 DH51 DH59 DH57 DH60 l l
FIGURE 5-8 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR MCGUIRE UNIT 1 REACTOR VESSEL WELD HAZ METAL 3m. muse ' 5-20 RM-19843
. _ , . - . _ _ _ . _ _ . . . .. ... ~ ._. .. .._ _ _ _ ._ _ . _ _ . _ _ . . _ - _ . _ . _ _ _ . . . _
;
i Curve 757281-A I
'O
- 50 0 50 100 150 200 250 300-120 .
i i i i i i i i 110 - _ g 100 - - 700 f
~
I ltimate T, ensile Strength 600g
-g -
.E '
2:* 70
- - 500 - i M -
k 0. 2 5 Yield Strength O e - m 50
- i 40 e i i i i -
300 Code : Open Points - Unirradiated Closed Points - Irradiated 1,38 x 1019 n/cm 2 E I I I I I I I I 70 - e _
, Reduction in Area -
60 -
-g -
=M -
]30 -
m 3 Total Elongation A m - 10 - E - 0 9 ~ L 0 i i i , Uniform Elongat'on , L -100 0 100 200 300 400 500 600 Temperature (*F) FIGURE 5-9 TENSILE PROPERTIES FOR MCGUIRE UNIT 1 REACTOR VESSEL SHELL PLATE B5012-1 (LONGITUDINAL ORIENTATION) l l wo.was in 5-21
Curve 757282-A e oC
- 50 0. 50 100 150 200 250 300 120 , , i. , , , , ,_
110 - I 100 - e Ultimate Tensile Strength- 700 i
= Q ll l
.g Q g. j- -
600g E _ l
-s -
'2 b 500 -
m 70 Qll
= i O e -
m 50 - 2 % Yleid Strength 40 1 i i i i i i- 300-Code : Open Points - Unirradiated 19 2 Closed Points -Irradiated 1.38x-10 n/cm l 80 i i i i i i , , 70 - 2% , Reduction in Area _
- : O
@ ~
= -
k 50 - -
$ 40 - -
]c 30 - , Total Elongation _
= .
e ^ y 10 - S : 0 0 - 0 i i i i UnWprm EloggaGon,
-100 0 100 200 300 400 500 600 Temperature ( *F)
FIGURE 5-10 TEliSILE PROPERTIES FOR MCGUIRE UNIT 1 REACTOR VESSEL SHELL PLATE p B5012-1 (TRANSVERSE ORIENTATION)
- . 1 l.
. mawosiaan to 5-22 i l
t .- ' t . curve 757280-A
'C.
- 50 0 50 100 150 200 gg i 250 300 .
i i i i i ' ' - 800 - _ lM - - 7@ 3 90 - J 2
% ltimate Tensile Strength 600 E t 80 2
' E 2:" 70
-; -
500 - g _ Q 0,2 % Yield Strength e c - 400 50 40 i i i i ! i i- 300 Code : Open Points - Unirradiated 19 Closed Points - Irradiated 1,38 x 10 n/cm 2 80 i i i , i i , , 70 - c _
, Reduction in Area -
_s - 1 _ f 2 -
;
k40 -
]30 -
g -
^
Total Elongation
^
l 20 - j 10 - + :
"2 0 i i i , Uniform Elon,gation ,
-100 0 100 200 300 400 500 600 Temperature (*F)
FIGURE 5-11 TENSILE PROPERTIES FOR MCGUIRE UNIT 1 REACTOR VESSEL WELD METAL l nwceino io 5-23
-- t y y -
r . . - + - ,,,,,-..,.,._,w.
g -. l 1 . 'tl l t .j; 6 -i.7; -e
' 34 yppig .s ; ;F1 ut es
'c.;. ;. yg 1.' .I' 100Uf$ .. # i
.2- 3' 4 5 t'.7i.84 Dis '
s.i -J
'Y'hYkhh hkh
\
- w. -
c g.gy g g g Nyg l
! 9MQ n&iu
"^
l l Specimen DL11 74*F F Mi5h[ot4 4.'30 r HS
'a 3
i i 2 - r 1300TH$ e 3
' a
,c , .2 3 4 t v ea ,
.. ,4 li,IJ,[Ilh thlAldth,I
# j.ddh il r)g .
1
., %g%. +-+ g y M .sW; ~~, ,,,, -
t
'$.'?
;
.ft *1 '"
s Specimen DL12 200*F 1 .. .> # 9 q. y
)j g
, r.
% , 8 9 l1
\.1 -
w .. M.didgbj
,. ,wn; a.
1 1 l [ 1 d'
. .u:
I Specimen DL10 550*F l FIGURE 5-12 FRACTURED TENSILE SPECIMENS FOR MCGUIRE UNIT 1 REACTOR VESSEL l i SHELL PLATE B5012-1 (LONGITUDINAL ORIENTATION) 1 1 3887s/081489 10 5-N RM-19844 l
N n,;y v , a . g vy y =r AT (9(. &;U:e91g;yl;3rpQpg;.sg 1 1,3 scss.c'ste::&qEIl
#:eal 6 p %q,.m.sa
*gykgf";;;c-10tec A.L .M; ' 4- 6. -
h!,v I
. ,!ikh+ hfh)NNkNI$[Nyg$
, , ie .s.mvja hjk s
- w;)K.
.g.,Jfdf%ff%gf4
' ,, ?
.., :-C;.s.C u: **:L-6 2:. 9 y' Specimen DT11 74*;y c >
s 4 <, 6 7 'A ' ,3 -l-l l, J1 .
- s. '
y-n , ,.; 1
.x.L,'y 3 ., l <> 1 ' e :: n ,. ,
+
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l t-Specimen DTIO 550*F l 1 1 FIGURE 5-13 FRACTURED TENSILE SPECIMENS FOR MCGUIRE UNIT 1 REACTOR VESSEL SHELL PLATE B5012-1 (TRANSVERSE ORIENTATION) sw.ie ,.. ,o 9_g5 l
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t
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t l 120 , , , , 100 -
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. sp.ci.. otto r 20 -
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. -- _ ; . - - . _ ___. .,_ ...,,. .
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f
-SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY
6.1 INTRODUCTION
-Knowledge'of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future-condition of the reactor vessel, a relationship must be established between the neutron environment.at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. in recent years, however, it has been suggested that' an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate ovaluation of damage gradients through the pressure vessel wall. Because of this potential shif t away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853
" Analysis and Interpretation of Light Water Reactor Surveillance Results,"
recommends reporting displacements per iron atom (dpa) along with fluence { 1. mar,m use,o i 6-1
g (E-> 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function-to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Forritic Steels l in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99, " Radiation Embrittlement of Reacter Vessel Materials." This section provides the results of the neutron dosimetry evaluations performed in conjunction with tne analysis of test specimens contained in surveillance capsule X. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron- fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided. 6.2 DISCRETE ORDINATES ANALYSIS , A plan view of the reactor geometry at the core midplane is shown in Figure
~
4-1. Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 56 , 58.5', 124', 236*, 238.5*, I- and 304' relative to the core cardinal area as shown in Figure 4-1. A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure'5-1. The stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core. From a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on both the distribution of neutron l l 1 ummim io 6-2 l I
i l' i
' flux and the neutron energy spectrum'in the water annulus between the neutron pad and the reactor vessel. -In order to properly determine the neutron environment at the test specimen locations, the capsules themselves must be included.in the analytical model.-
In performing the fast neutron exposure evaluations for the surveillance capsules and reactor' vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters (v(E > 1.0 Mev,) ((E > 0.1 Mev), and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/e(E > 1.0 MeV), within the pressure vessel geometry. The relative radial gradient information was required to permit the projection cof measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations. The second set of-calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These-importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the a locations of interest for the first 5 cycles of b radiation; and established the means to perform similar predictions and dosimetry evaluations for cil subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only l- spatial variations of fission rates within the reactor core; but, also l accounted for the effects of varying neutron yield per fission and fission l spectrum introduced by the build-up of plutonium as the burnup of individual l fuel assemblies increased, l i 3687s/102469 10 63 l
1 l The' absolute cycle specific data from the adjoint evaluations together with I relative neutron energy spectra and radial distribution'information from the forward calculation'provided the means to:
- 1. Evaluate neutron dosimetry obtained from surveillance capsule locations.
- 2. Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.
- 3. Enable a oirect comparison of analytical prediction with measurement.
- 4. Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
The' forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R, e geometry using the DOT two-dimensional discrete ordinates code (5) and the SAILOR cross-section library [6). The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for
. light water reactor applications. In these analyses anisotopic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an S8 rder of angular quadrature.
The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2e uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2a level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.
=7.w ma o 6-4
.All adjoint analyses'were also carried out using an S rder of angular 8 ~ quadrature and'the P3 cross-section approximation from the SAILOR library. ,
Adjoint source locations were chosen at several azimuthal locations clong the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R, o geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, , (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as: R (r, 0) = /r #0 E
# 1(r, 0, E) S (r, O. E) r dr de dE where: R(r,0) =
, (E > 1.0 MeV) at radius r and azimuthal angle 0 I.(r,0,E) = Adjoint importance function at radius, r, azimuthal anglo e, and neutron source energy E.
S-(r,0,E) = Neutron source strength at core location r, 0 and energy E. Although the adjoint importance functions used in the McGuire Unit I analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the' magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/4 (E > 1.0 MeV) is insensitive to changing core source distributions. In the application of these adjoint important functions to the McGuire Unit I reactor, therefore, calculation of the iron displacement rates (dpa) and the i neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/9 (E > 1.0 MeV) and v (E > 0.1 MeV)/d (E > 1.0 MeV) ratios from the l l forward analysis in conjunction with the cycle specific e (E > 1.0 MeV)
'solutione from the individual adjoint evaluations.
un,mv,eo sa g.5
r~rt. r The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design report for the first five operating cycle of McGuire Unit 1 [7 thru 11). The relative power levels in fuel assemblies that are significant contributors to the neutron exposure of the pressure vessel and surveillance capsules are sunmarized in Figure 6-2. . For comparison purposes, the core power distribution (design basis) used in the reference forward calculation is also illustrated in Figurc 6-2. Selected results from the neutron transport analyses performed for the McGuire-Unit I reactor are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall. In Table 6-1, the calculated expesure parameters (# (E > 1.0 MeV), e (E > 0.1 MeV), and dpa) are given at the geometric center of the two surveillance capsule positions for both the design basis and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison.of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar data is given in Table 6-2 for the pressure. vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the cycle 1 through 5 plant specific power distributions, it is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself. Radial gradient information f or neutron flux (E > 1.0 MeV), neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport ' calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5. 3447s/061489 10 g.g
- For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' l
aximuth is given by:
=-#(220.27,45')F(225.75,45')
'1/4T(45')
4 where.- 9 1/4T(45') = Projected neutron flux at the 1/4T position on the 45' azimuth e (220.27, 45') = Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth. F(225.75,45') = Relative radial distribution function from Table 6-3. Similar expressions apply for exposure parameters in terms of f(E > 0.1 MeV) and dpa/sec. 6.3 NEUTRON 00SIMETRY , The passive neutron sensors included in the McGuire Unit 1 surveillance q program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the' evaluation of the neutron energy spectrum within the capsule and the j subsequent determination of the various exposure parameters of interest j {e (E > 1.0 Mev), 4 (E > 0.1 MeV), dpa). l The relative locations of the neutron sensors within the capsules are shown in Figure 4-E. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors L L were accommodated within the dosimeter block located near the center of the capsule. The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest. usweeue io 6-7
r Rather, the activation or fission process is a measure of the integrated effect that the time and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived i from the activation measurements only if the irradiation parameters are well l known. In particular, the following variables are of interest: o The specific activity of each monitor. o The operating history of the reactor, o The energy response of the monitor. o The neutron energy spectrum at the monitor location, o The physical characteristics of the monitor. < The specific activity of each of the neutron monitors was determined using established ' ASTM procedures (12 through 25). Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. Tne irradiation history of the McGuire Unit I reactor.during cycles 1 through 5 was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable period. The irradiation history applicable to capsule X is given in Table 6 7. Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7. Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code (26). The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum et the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra. mweeme io 6-8 i
p 1 In the FERRET evaluations, a' log normal least-squares algorithm weights both a
-.the a priori values and the measured data in accordance with the assigned i uncertainties and correlations. In general, the measured values f are linearly'related to the flux 4 by some response matrix A: ,
f i(s,a) = I A (s)- ,g(a) g ig t S . where i indexes the measured values belonging to a single data set s, g ;
- designates'the~ energy group and a delineates spectra that may be simultaneously adjusted. For example, R I
$=g ogg #g relates a set of measured reaction rates R$ to a single spectrum pg_by the multigroup cross section ogg . (In this case, FERRET also adjusts the l
' cross-sections.) The log normal approach automatically accounts for the !
physical constraint of positive fluxes, even with the large assigned uncertainties. In the FERRET < analysis of the dosimetry data, the continuous quantities (i.e., fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code [27). This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation.in regions where j; group boundaries do not coincide. The 620 point spectrum was then easily collapsed'to the group scheme used in FERRET. The cross-sections were also collapsed into the 53 energy group structure using SAND II with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/B-V dosimetry file. Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.
= >. e u to 6-9
i y e for each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight. In some cases, as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form.is used: M gg,=Rh+R g R,Pg gg, where NR specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional uncertainties Rg specify additional random uncertainties for group g that are correlated with a correlation matrix: Pgg, = (1 - 0) 6gg. + 0 exp ( _,(g g')2) 2r
-_The first term specifies purely random uncertainties while the second term describes short-range correlations over a range r (e specifies the strength of the latter term.)
For the a priori calculated fluxes, a short-range correlation of r = 6 groups was used. This choice implies that neighboring groups are strongly correlated when e is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R. E. Maerker (28]. Maerker's results are closely duplicated when r = 6. For the integral reaction rate covariances, simple normalization and random uncertainties'were combined as deduced from experimental uncertainties. Results of the FERRET evaluation cf the capsule X dosimetry are given in Table 6-9 The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 1,38 x 10 19 n/cm2 (E > 1.0 MeV) with an associated uncertainty of + 8%. Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit of the adjusted spectrum are provided in Table 6-10. In general, l excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure. l l w,**" 6-10 1 1
q A summary of the measured and calculated neutron exposure of capsulo X is i presented in Table 6-12. The agreement between calculation and measurement falls within + 2-17% for allLexposure parameters listed. The calculated fast neutron exposure (f (E > 1.0 MeV), 6 (E > 0.1 MeV), dpa) values agreed with the measurements to.within 6-12% whereas, the thermal neutron fluence calculated for the exposure period was less than the measured value by 17 percent. Neutron exposure projections at key locations on the. pressure vessel inner radius are given in Table'6-13. Along with the current (4.33 EFPY) exposure derived from the capsule X measurements, projections are also provided for'an exposure period of 16 EFPY and to end of vessel design'1ife (32 EFPY). The time averaged exposure rates for the first 4.33 EFPY of operation were used to perform projections beyond the end of cycle 1 through 5 exposure period. In the calculation of exposure gradients for use in the development of heatup
.and cooldown curves for the McGuire Unit 1 reactor coolant system, exposure
. projections to 16.EFPY and 32 EFPY were employed. Data based on both a fluence (E > 1.0 MeV)' slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14. In order-to access RT vs.
NDT fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations t' 1/4T = + (Surface) { dpa (1/4T) ) dpa (Surface) t' 3/4T=+(Surface){ dpa (3/4T) ) dpa (Surface) Using this approach results in the dpa equivalent fluence values listed in Table 6-14. . In Table 6-15 updated lead factors are listed for each of the McGuire Unit 1 surveillance capsules. These data may be ustd as a guide in establishing future withdrawal schedules for the remaining capsules. l l meweeuulo 6-11 E _.
j'!
, ?
- 1 W e,...
i l 4
)
CHARPY SPECIMEN ,
/ / l
- /,1 ,, J /,/,, /
////// / //// ,
SNNNNNNNNNNNNNNNNNNN
\ NEUTRON PAD \
A\\\\\\\N\\\\\\\\\\\ Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule mi.mu io 6-12
v , 4 , 1 1.01 1.04 0.96 0.77 DESIGN BASI 5 0.72 0.75 0.67 0.55 CYCLE ! 1.03 1.05 0.97 0.77 CYCLE 2-0.77 0.77 0.67 0.51 CYCLE 3 0.84 0.71 0.01 0.50 CYCLE 4 0.87 0.72 0.77 0.39 CYCLE 5 1.02 1.10 1.00 1.05 1.10 0.71 0.99 1.05 0.95- 0.97 9.80 0.49 0.93 1.30 0.93 i.25 1.07 0.69 0.92 1.22 0.93 1.15 0.75 0.45 0.91 1.20 1.00 1.12 0.69 0.41 1.03 1.19 1.06 1.10 0.69 0.35 1.05 0.87 0.87 1.07 1.00 1.05 ; 1.14 1.10 1.12 1.05 0.98 0.95 'i' 0.79 0.E' O.91 0.97 1.14 0.80 1.17 1.00 1.31 1.03 1.21 0.7) i 1.09 1.01 1.27 1.01 1.19 0.83 I 1.13 0.98 1.18 1.13 1.18 0.78 l 1
-l 1.09 1.06 0.09 1.10 1.04 1.15 1.18 1.13
.f 1.13 1.18 1 0.93 0.92 0.92 1.11 1.15 I 0.93 1.33 1.02 1.33 0.91 !
1.15 1.29 0.86 1.10 1.10 ! 0.92 1,29 0.97 1.30 0.97 ! 0.90 1.04 1.12 0.92 i 1.19 1.15 1.19 1.14 I 0.90 0.94 1.15 1.11 > 7 1.28 '1.00 1.33 1.01 ; 1.09 1.11 1.26 1.13 i 1.10 1.13 1.31 0.95 j i L Figure 6-2. Core Power Distributions Used in Transport Calculations for j l McGuire Unit 1 { l 0
- u. .. io.
6-13 li L 1 1
~' 4 7 TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER IRRADIATION $ (E > 1.0 Mev) f(E > 1.0 Mev) dpa/sec 2 2 IIME (n/cm -sec) (n/cm -sec) CYCLE (EFPS) 31.5* 34.0' 31.5' 34.0' 31.5* 34.0' 1.11 x 10 11 1.29 x 10 11 2.21 x 10 -10 2.62 x 10 -10 II DESIGN BASIS 4.88 x 10 II 5.93 x 10 CYCLE 1 3.53 x 10 7 8.18 x 10 10 9.32 x 10 10 3.60 x 10 II 4.28 x 10 II 1.63 x 10 -10 1.89 x 10-10 CYCLE 2 2.32 x 10 7 1.03 x 10 I1 1.16 x 10 11 4.53 x' 10 II 5.33 x 10 II 2.05 x 10 -10 2.36 x 10 -10 3.95 x 10 I1 1.52 x 10 -10 1.74 x 10-10 7 CYCLE 3 2.49 x 10 7.67 x 10 10 8.59 x 10 10 3.35 x 10 11 7.23 x 10 10 8.03 x 10 10 1.44 x 10 -10 1.63 x 10 -10 7 CYCLE 4 2.59 x 10 3.17 x 10 II 3.69 x 10 II CYCLE 5 2.73 x 10 7 6.70 x 10 10 7.40 x 10 10 2.93 x 10 II 3.40 x 10 II 1.33 x 10 -10 1.50 x 10 -10 3907sM01489 TO
b TABLE 6-2 I i CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE L l e (E > 1.0 Mev) ( n/cm2 .3,c) h., o* 15' 30* ,4 5 ' DESIGN BASIS 1.45 x 10 10 2.21 x 10 10 1.69 x 10 10 2.44 x 10 10 ; 10 Cycle 1 1. J x 10 1.60 x 10 10 1.24 x 10 10 1.75 x 10 10 l, 10 Cycle 2 1.36 x 10 2.04 x 10 10 1.56 x 10 10 2.17 x 10 10 Cycle 3 1.07 x 10 10 1,56 x 10 10 1.17 x 10 10 1.61 x 10 10 . 1.08 x 10 10
+
10 Cycle 4 1.59 x 10 1.12 x 10 10 1.50 x 10 10 Cycle 5 1.08 x 10 10 1.52 x 10 10 1.03 x 10 10 1.38 x 10 10 l t (E > 0.1 Mev) { n/cm2 .3,c) 0' 15' 30' 45' i OESIGN BASIS 3.02 x 10 10 A 66 x 10 10 4.25 x 10 10 6.11 x 10 10 Cycle 1 2.23 x 10 10 3.37 x 10 10 3.12 x 10 10 4.38 x 10 10 , Cycle 2 2.83 x 10 10 4.20 x 10 10 3.92 x 10 10 5.43 x 10 10 ,
' Cycle 3 2.29 x 10 10 3.29 x 10 10 2.94 x 10 10 4.03 x 10 10 _l Cycle 4 2.25 x 10 10 5.35 x 10 10 2.82 x 10 10 3.76 x 10 10 Cycle 5 2.25 x 10 10 3.21 x 10 10 2.59 x 10 10 3.46 x 10 10 dpa/sec 0* 15 30' 45' .
OESIGN BASIS 2.25 x 10'11 3.41 x 10 ~11 2.73 x 10 ~11 3.88 x 10 ~11 Cycle 1_ 1.66 x 10'11 2.47 x 10'11 2.00 x 10'11 2.78 x 10'11 Cycle 2 2.11 x 10 ~11 3.15 x 10 ~11 2.52 x 10'11 3.45 x 10 ~11 . Cycle 3 1.66 x 10'11 2.41 x 10'11 1.89 x 10 ~11 2.56 x 10'11 Cycle 4 1.68 x 10 ~11 2.45 x 10 ~11 1.80 x 10'11 2.39 x 10 -11 Cycle 5 1.68 x 10'11 2.35 x 10 -11 1.66 x 10'11 2.19 x 10'11 L L w ... io 6-15 .
{ TABLE 6-3 RELATIVE RA01AL OISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) } WITHIN THE PRESSURE VESSEL WALL l
.i t . Radius ]
(em) O' 15' 30' 45' j 220.27(I) 1.00 1.00 1.00 - 1.00 ! 220.64 0.979 0.979 0.980 0.979 L221.6E 0.891 0.891 0.893 0.889 222.99 0.771 0.769 0.773 0.766 l 224.31 0.655 0.652 0.658 0.648 225.63 0.552 0.549 0.555 0.543 226.95 0.463 0.459 0.467 0.452 j 228.28 0.387 0.383 0.390 0.376 ; 229.60 0.322 0.318 0.326 0.311 i
;
230.92 0.268 0.263 0.271 0.257 232.25- 0.222 0.218 0.225 0.211 233.57 0.183 0.180 0.187 0.174 234.89 0.151 0.148 0.155 0.142'
'236.22 0.125 0.121 0.128 0.116 -
0.105 0.0945 237.54 0.102 0.0992 238.86 0.0831 0.0807 0.0862 0.0762 t 240.19 0.0673 0.0650 0.0703 0.0608. 241.51 0.0539- 0.0512 0.0567 0.0472 242.17(2) 0.0508 0.0477 0.0536 0.04?t NOTES: 1) Base Metal Inner Radios
- 2) Base Metal Outer Radius 'fr 4
wn.w." " 6-13 ,
j s p ; TABLE 6-4 i y ,
- RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLVX (E > 0.1 MeV)
{: WITHIN THE' PRESSURE VESSEL WALL l i L i !^ [. Radius ; (cm)- 0* 15' 30' 45' ! 220.27 II} 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 -i
- 221.66 1.00 1.00 1.00 0.995 !
222.99 0.974 0.966 0.982 0.956 l
~224.31 0.928 0.915 0.938 0.902 l 225.63 0.875 0.859 0.886 0.843 226.95 0.819 0.802 0.832 0.782 228.28 0.762 0.743 0.777 0.722 229.60 0.705 0.686 0.721 0.663 I ; 230.92 0.649 0.629 0.665 0.605 >
232.25 0.594 0.575 0.611 0.549 !
;
23.3.57- 0.540 0.522 0.558 0.495 234.89 0.488 0.470 0.506 0.443 !
. 236.22 0.436 0.421 0.455 0.392 :
237.54 0.386 0.373 0.406- 0.343 ;
- 238.86 0.337 0.326 0.358 0.296 i 240.19 0.290 0.280 0.310 0.248
- 241.51 0.244 0.232 0.261 0.201 242.17(2) 0.233 0.219 0.249 0.188 :
t NOTES:- 1) Base Matal Inner Radius
;
- 2) Base Metal Outer Radius i
I : 1 r w.mises to 6-17 ;
;
l' - -. - _ . - - . , - - . . .
y
-tl TASLE 6 ,
RELATIVE RADIAL OISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL WALL Radius-(cm) O' 15' 30' 45' 220.27(1) 1.00 1.00 1.00 1.00 220.64 0.982 0.982 0.986 0.984 221.66 0.911 -0.910- 0.923 0.915 222.99 0.813 0.812 0.837 0.823 224.31 0.721 0.718 0.751 0.730 225.63 0.637 0.633 0.673 0.646 226.95 0.562 0.558 0.602 0.572 228.28 0.496 0.491 0.539. 0.505-
-229.60 0.438 0.433 0.481 0.447 230.92 0.387 0.381 0.430 0.394 232.25 0.341 0.335 0.38S 0.347 233.57 0.300 0.295 0.341 0.305 234.89 0.263 0.258 0.302 0.266' 236.22 0.230 0.225 0.267 0.231 237.54 0.199 0.195 0.234 0.199 238.86- 0.171 0.168 0.203 0.169 240.19 0.145 0.142 0.174 0.140-241.51 0.121 0.117 0.146 0.113 242.17(2) 0.116 0.110 0.140 0.106 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius .
l l l I-l mwmim io 6-18 1
< TABLE 6-6 NUCLEAR PARAMETERS FOR NEL' TRON FLUX MONITORS i Reaction Target Fission Monitor of Weight Response Product Yield l Material Interest fraction . Range Half-Life (%) i Copper Cu63(n a)Co60 0.6917 E> 4.7 MeV 5.272 yrs Iron Fe54(n.p)Mn54 0.0582 E> 1.0 MeV 312.2 days Nickel NiS8(n.p)CoS8 0.6830 E> 1.0 MeV 70.90 days Uranium-238* U238(n f)Cs137 1.0 E> 0.4 MeV 30.12 yrs 5.99 Neptunium-237* Np237(n,f)Cs137 1.0 E> 0.08 MeV 30.12 yrs 6.50 . Cobalt-Aluminum
- CoS9(n,1)Co60 0.0015 0.4ev<E< 0.015 MeV 5.272 yrs Cobalt-Aluminum CoS9(n,r)Co60 0.0015 E< 0.015 MeV 5.272 vrs ,
- Denotes that monitor is cadmium shielded.
sm.ause i. 6-19
p: L TABLE 6-7 L IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINE0 IN CAPSULE X Irradiation P pf Irradiation Decay- r li- 3 Period (MW) t Pp ,f, Time (days) Time (days)' ; 10/81 299 0.088 31 2671 11/81 1100 .322 30 2641 12/81 114 .034 31 2610 1/82 1712 .502 31 2579 2/82 1569 .460 28 2551 3/82 713 .209 31 2520 4/82 1656 .486 30 2490 5/82 1995 .585 31 2459 6/82 1758 .516 30 2429 7/82 755 .221 31 2398 i 8/82 1973 .578 31 2367 9/82 2020 .592 30 2337 10/82 2758 .809 31 2306 11/82 733 215 30 2276 < 12/82 1996 .565 31 2245 : 1/83 1152 .338 31 2214 2/83 0 .000 28 2186 3/83 0 .000 31 2155 4/83 0 .000 30 2125 5/83 252 .074 31 2094 6/83 2714 .796 30 2064 7/83 26C .778 31 2033 8/83 1518 .445 31 2002 9/83 3140 .921 30 1972 10/83 2601 .763 31 1941 11/83~ 2255 .661 30 1911 12/83 2704' .793 31 1880 1/84 2988 .876 31 1849 2/84 2715 .796 29 1820 l l no.mine in 6-20 l.
-~ _
w-h [
-TABl.E 6-7 (Cont'd)
IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED'IN CAPSULE-X Irradiation P pj Irradiation Decay 3 Period (MW) P pg, Time (days) Time (days) t 3/84 0 .000 31 1789 4/84 0 .000' 30 1759 5/84 2681 .786 31 1728
'6/84 3148 .923 30 1698
7/84 3118 .914 31 1667 l 8/84 3186 .934 31 1636 9/84 3445 1.00 30 1606
.10/84 2745 . 805 31 1575 11/84 2157 .632 30 1545 12/84 305 .089 31 1514 1/85 3168 .929 31 1483 l 2/85 3108 .911 28 1455 L
-3/85 2277 .668 31 1424 3 4/85 829 .243 30 1394 5/85 0 .000- 31 1363 i 6/85 164 .048 30 1333 !
7/85 3151 .924 31 1302 , 8/85 3405 .998 31 .1271 .
'9/85 3236 .949 30 1241 '
10/85 3406 .999 31 1210 11/85 2139 .627 30 1180 - 12/85 3260 .956 31 1149 1 1/86 3322 .974 31 1118 2/86 3010 .882 28 1090 3/86 2808 .825 31 1059-4/86 3062 .898 30 1029 5/86 1346 .395 31 998 6/86 0 .000 30 968 7/86 0 .000 31 937 6
= >.40. a 6-21
'k y p
TABLE-6-7_(Cont'd)
- IRRADIATION HISTORY OF NEUTRON SENSORS l
CONTAINED IN CAPSULE X ,a Irradiation P pj Irradiation Decay 3 Period Time (days) (MW) t Pp ,f, Time (days)- 8/86 0 .000 31 906 -
'9/86 1311 .385 30 876 -
-10/86 3082 .904 31 845 i 11/86 62 .018 30 815 1 12/86 3383 .992 31 784 ;
1/87 3407 .999 31 753 2/87 2769 .812 28 725 3/87 3411 1.00 31 694 4/87 3177 .931 30- 664 5/87 3354 .983 31 633 , 6/87 3391 .994 30 603 ! 7/87 3407 .999 31 572 8/87 2484 .728 31 541 9/87 308 .090 30 511 10/87 0 .000 31 480 i 11/87 1530 .449 30- 450 12/87 3026 .887 31 419 . 1/88 3199 .938 ' 31 388 2/88 3326 .975 29 359 ; 3/88 3200 .938 31 328 i 4/88 3215 .943 30 298 : 5/88 3399 .997 31 267 6/88 3168 .929 30 237
.7/88 3341 .979 31 206 I 8/88 3390 '
.994 31 175 9/88 3386 .993 30 145 20/88 3119 .914 12 133 l
NOTE: Reference Power = 3411 MW t f i 1 w.wm io 6-22
{l
.; ?: ,
TABLE 6-8' ; MEASURED SENSOR ACTIVITIES AND REACTION RATES :
- i. !
. . ;T
;
F . i* F , . . Measured Saturated . Reaction P' Monitor'and Activit Activity Rate Axial location. (dis /see amy ). (dis /sec-om) (RPS/ NUCLEUS) I [' Cu-63 (n,a) C0-60 j I. Top l'.26 x 10 5 3.44 x 10 5 i Y -Middle 1.25 x 10 5 3.41 x 10 5 l r Bottom 1.29 x 10 5 3.52 x 10 5 ; l Average 1.27 x 10 5 3.46 x 10 5 5.28 x 10'I7 ]
-l Fe-54(n.p) Mn 'f
?
r 6 6 Top 1.65 x 10 3.22 x 10 [ Middle 1.64 x 10 6 3.20 x 10 6 , Bottom '1.73 x 10 6 3.37 x 10 6 l Average 1.67 x 10 6 3.26 F 106 5.20 x 10 -15
.Ni-58(n.p)Co-58 .l
;
7 7 Top 1.10 x 10 5.15 x 10 "l Middle 1.07 x 10 7 5.01 x 10 7 ! Bottom 1.12 x 10 7 5.24 x 10 7 7.33 x 10 -15 7 Average 1.10 x 10 5.13 x 10 7 U-238 (n f) Cs-137 (Cd) . I: Middle 5.33 x 10 5 5.81 x 10 6 3.83 x 10 -14 ,
;.-
+
w,wm io 6-23 I
p - , k ,, i TABLE 6-8
;
i MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd' . t
. Measured Saturated ' Reaction Monitor and Activity Activity Rate :
i . Axial location (dis /see am) (dis /see am) -(RPS/ NUCLEUS) ( Np-237(n.f) cs-137 (Cd). ; Middle 3.91 x 10 0 4.26 x 10 7 2.58 x 10'13 . 00-59(nr)Co-60 o !
;
Top 3.14 x 10 7 8.57 x 10 7 !
-Middle 3.35 x 10 7 9.14 x 10 7 !
7 Bottom' 3.00 x 10 7 8.19 x 10 ! 7 Average 3.16 x 10 7 8.63 x 10 5.63 x 10-12 l
- . I
~ Co-59(nr)Co-60(Cd) o Top 1.68 x 10 7 4.59 x 10 7 [
' Middle 1.68 x 10 7 4.59 x 10 7 I Bottom 1.55 x 10 7 4.23 x 10 7 !
Average 1.64 x 10 7 4.47 x 10 7 2.91 x 10
?
.P (A
E
'h r
I
+
1 w w w i.= io 6-24 l I
( t-
~~.
, ,g .
TABLE 6 l
SUMMARY
OF NEUTRON DDSIMETRY RESULTS f i i U " i TIME AVERAGED EXPOSURE RATES f 0 2 1.01 x 10 11 9.'(E> 1.0 MeV) (n/cm -see) 1_ 8% 2 fo + (E> 0.1 MeV)'(n/cm -sec) 4.26 x 10 11 ' i 15% t dpa/sec 1.89 x 10 -10 11% p(E< 0.414 eV) (n/cm2 sec) 4.22 x 10 10 29%
.?
;
INTEGRATED CAPSULE EXPOSURE
-j 2
6 (E> 1.0 MeV) (n/cm ) 1.38 x 10 19 1 8% 2 f'(E> 0.1 MeV) (n/cm ) 5.82 x 10 19 1 15% l 2.58 x 10 -2 33g . 1dpa_ 6 (E< 0.414 eV) (n/cm ) 2 5.77 x 10 18 29% NOTE: Total Irradiation Time = 4.33 EFPY o , h 1
.i r
i ~
. mr.w a 6-25
h '{{f-
- . i .
TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER i: Adjusted p Reaction Measured Calculation- C/M i Cu-63in,a)Co-60 -17 ' 5.28x10 5.26x10'17 1.00 Fe-54(n.p)Mn-54 5.20x10
-15 5.34x10
-15 1.03 Ni-58 (n p) Co-58 7.33x10
-15 -15 o 7.35x10 1.00 U-238-{nf)Cs-137(Cd) 3.83x10'I4 3.20x10'14 0.83 Np-237 (n,f) Cs-137 (Cd) -13 2.58x10 2.88x10'13 1.12-
. Co-59. (n,r) .Co-60 (Cd) -12 -12 "
2.91x10 2.91x10 1.00 Co-59(n,r)_Co-60 -12 -12 5.63x10 5.62x10 1.00 l mn.=i.= io 6-26
TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER Energy AdjusjedFlux Energy Adjysted Flux Group ( _Mev) (n/cm -sec) Group (Mev) (n/cm -sec) 1 1.73x10f 4.23x10f 28 9.12x10j 1.98x10$ 2 1.49x10 1.04x10 29 5.53x10 2.60x10 1.35x10 1 4.83x10 7 -3 9 3 1 8 30 3.36x10 3 8.24x10 4 1.16x10 1.24x10 8 31 2.84x10-3 7.98x10 9 3 9 5 1.00x10 3.02x10 32 2.40x10 7.77x10 0 8 -3 6 8.61x10 0 5.50x10 9 33 2.04x10 3 2.20x10 10 10 7 7.41x10 1.32x10 34 2.04x10 10 6.07x10 0 9 1.23x10 4 8 0 1.92x10 35 7.49x10 1.90x10 10 9 9 4.97x10 4.07x10 36 4.54x10.4 1.82x10 0 9 10 10 3.68x10 5.43x10 37 2.75x10.4 1.96x10 0 10 11 2.87x10 0 1.14x10 10 38 1.67x10.4 2.24x10 10 10 12 2.23x10 1.57x10 39 1.01x10 4 2.12x10 0 10 5 10 13 1.74x10 2.19x10 40 6.14x10 2.0Sx10 1.35x10 0 2.37x10 10 -5 10 14 41 3.73x10 2.00x10 10 1.11x10 0 4.28x10 10 2.26x10
-5 15 10 42
-5 1.91x10 10 16 8.21x10.3 4.84x10 43 1.37x10 1.83x10 10 5.00x10 10 8.32x10
-6 17 6.39x10.3 10 44
-6 1.72x10 10 18 4.98x10.3 3.61x10 45 5.04x10 1.55x10 10 3.88x10 1
5.15x10 10 -6 19
-1 10 46 3.06x10
-6 1.42x10 10 20 3.02x10 5.11x10 47 1.86x10 1 ?8x10 10 -
9 21 1.83x10.1 5.12x10 48 1.13x10.6 9.56x10 10 9.97x10 910 7 22 1.11x10 1 4.11x10 39 49 6.74x10 22 2.85x10 6.83x10 7 23 10 50 4.14x10 7 1.11x10 24 4.09x10
-2 1.62x10 10 51 2.51x10 7 9.09x10 9 25 2.55x10 -2 2.23x10 10 52 1.52x10
-8 7.28x10 9 !
10 26 1.99x10 1.07x10 53 9.24x10 1.47x10
-2 10 27 1.50x10 1.35x10 NOTE: Tabulated energy levels represent the upper energy of each group.
;
i i m u. w . io 6-27
^'
- g. .
6
- TABLE 6-12
-l COMPARISON OF CALCULATED AND MEASURED 4 EXPOSURE LEVELS FOR CAPSULE X l I
f A- Calculated! Measured C/M s 1.22 x 10 19 2
- h. . 6(E>1.0MeV)(n/cm) 1.38 x 10 19 0.88' j w .
h
~
- +(E> 0.1 MeV) (n/cm )
2 5.73'x 10 19 5.82 x 10 19 0.98 i t 2.43 x 10 -2 2 F - dpa 2.58 x 10 0.94 ! 2 18 10 L 6(E< 0.414 eV)-(n/cm ) 4.81 x 10 5.77 x 10 0.83
.l
;
v t e
?
t t t t P S
;
t E . i i:
)
f- l l L. -w.mi.m io 6-28 1 l J W ^ c--,w-g e- - g.-e- V 4- g- g- 3 yr-g- e.- -ev,wr m ww pa- y w--m s.
TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE' METAL INTERFACE , AZIMUTHAL, ANGLE O' 15' 30' 45' 4.33 EFPY 6(E>-1.0 MeV) 1.74 x 10 18 2.57 x 10 18 1.89 x 10 18 2.60 x 10 18 2 (n/cm) f(E> 0.1 MeV) 3.25 x 10 18 4.87 x 10 18 4.27 x 10 18 5.85 x 10 18 2 (n/cm) 1 3.71 x 10 -3 2.86 x 10 ~3 3.87 x 10 -3 i dpa 2.53 x 10'3 16.0 EFPY f(E>1.0MeV) 6.43 x 10 18 9.50 x 10 18 6.98 x 10 18 9.61 x 10 18 2 (n/cm) l t(E> 0.1 MeV) 1.20 x 10 19 1.80 x 10 19 1.58 x 10 19 2.16 x 10 19 z 2 (n/cm) dpa 9.35 x 10 ~3 1.37 x 10
-2 1.06 x 10 -2 1,43 x 10 -2 32.0 EFPY f(E> 1.0 MeV) 1.29 x 10 19 1.90 x 10 19 1.40 x 10 19 1.92 x 10 19 2
(n/cm) l f(E> 0.1 MeV) 2.40 x 10 19 3.60 x 10 19 3.16 x 10 19 4.32 x 10 19 2 (n/cm) dpa 1.87 x 10 -2 2.74 x 10 -2 2.11 x 10 -2 2.86 x 10 -2 un.raunao 6-29
TABLE 6-14 :- NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES 16 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE dpa SLOPE
- 2 2 (n/cm ) (equivelent n/cm )
Surface 1/4 T 3/4 ' _ Surface 1/4 T 3/4 T O' 6.43 x 10 18 3.50 x 10 18 7.50 x 10 17 6.43 x 10 18 4.05 x 10 18 1.41 x 10 18 15' 9.50 x 10 18 5.14 x 10 18 1.07 x 10 18 9.50 x 10 18 5.95 x 10 18 2.03 x 10 18 30' 6.98 x 10 18 3.82 x 10 18 8.35 x 10 I7 6.98 x 01 18 4.65 x 10 18 1.78 x IC 18 45' 9.61 x 10 18 5.14 x 10 I0 1.04 x 10 18 9.61 x 10 18 6.14 x 10 18 2.11 x 10 I0 b 32 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE dpa SLOPE 2 (n/cm ) (equivalent n/cm2 ) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 0* 1.29 x 10 19 7.02 x 10 18 1.50 x 10 IO 1.29 x 10 19 8.10 x 10 18 2.82 x 10 18 15* 1.90 x 10 19 1.03 x 10 19 2.14 x 10 18 1.90 x 10 I9 - 1.19 x 10 19 4.06 x 10 18 30* 1.40 x 10 19 7.66 x 10 18 1.67 x 10 I0 1.40 x 01 19 9.30 x 10 I0 3.56 x 10 18 45* 1.92 x 10 19 1.03 x 10 19 2.08 x 10 18 1.92 x 10 19 1.23 x 10 I9 4.22 x 10 IO 3897s/081489 10 _ _ _ . _ __.____-.m___.m._-__ a t g w
- g 3- w g3 - y-n 4
. . ~ ,- . m -~ . - - -
,),'
, )( ;), u y ,-
. TABLE'6-15
- 1 UPDATED LEAD FACTORS.FOR McGUIRE-- ;
n, ,-
.- _ d- ,-. UNIT 1 SURVEII. LANCE CAPSULES-e.
I. w
~
3 1 i9 ., Capsule - Lead Factor: .' h: U- 5.33(a) ,w f' r-X. - 5.31(*)
.W- 5.31 =!
2 5.31 l V 4.76 t
'Y 4.76 w:
b ]
- (a).Plantspecificevaluation j ai -;
'f
'I
-;
,t
.t ,
5 (i A
.I ge 'i
;
s
, a 4
t 1 3 t t'. [ e9-
. s d
- = >. w e io 6-31 .
..s
+ - e , 6 l. . r.i-I... s_,, , , , , . + , - , . -n......,- -e an,,,- ,,-,,,,,,--~--e,-w,er,+,.,m,-n.,w,-4-.-,-,,, .a.- , + + s. .-g-.-<-,w~.
SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHE 0ULE j The following removai :rhedule meets ASTM E185-82 and is recommended for futura capsules to be removed from the McGuire Unit I reactor vessel: I Estimated Capsule L n Lead Removal Fluence Capsule 2 (deo) Factor Time (a) (n/cm ) U 56 5.33 1.06 4.14 x 1018(b) X 236 5.31 4.33 1.38 x 10 19(b) V 58.5 4.76 7 2.00 x 10 19(C) Y 238.5 4.76 10 2.86 x 10 19 W 124 5.31 Standby -- Z 304 5.31 Standby -- a) Effective full power years from plant startup b) Actual fluence c) Approximate fluence at vessel inner wall at and of life (32 EFPY)
=wo.im io 71
t SECTION 8 REFERENCES
- 1. Davidson, J.A., and Yanichko, S.E., 'Oune Power Company William B.
McGuire Unit No. 1 Reactor Vessel Radiation Surveillance Program " WCAP-9195, November 1977. 1
- 2. Yanichko, S.E., et al., " Analysis of Capsule U from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiction Surveillance Program, WCAP-10786, February 1985.
- 3. Code of Federal Regulations,10CFR50, Appendix G,
- Fracture Toughness Requirements" and Appendix H, " Reactor Vessel Material Surveillance Program Requirements", U. S. Nuclear Regulatory Commission, Washington, D.C.
- 4. Regulatory Guide 1.99, Revision 2. " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May, 1988.
- 5. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)-034, Vol. 5. August 1970.
- 6. "0RNL RSCI Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".
1
- 7. A. Saeed, et al., "The Nuclear Design and Core Physics Characteristics of the W. 8. McGuire Unit 1 Nuclear Power Plant - Cycle 1, "WCAP-9323, May 1978. (Proprietary)
- 8. J. R. Lesko, et al., "The Nuclear Design and Core Physics Characteristics of the McGuire Unit 1 Nuclear Power Plant - Cycle 2," WCAP-10463, January 1984. (Proprietary) un,mssee so g.1
- 9. J. R. Lesko, et al., "The Nuclear Design and Core Physics Characteristics of the McGuire Unit 1 Nuclear Power Plant - Cycle 3," WCAP-10782, February 1985. (Proprietary) t
- 10. J. R. Lesko, et al., "The Neelear Design and Core Physics Characteristics of the McGuire Unit 1 Nuclear Power Plant - Cycle 4." WCAP-11141, May 1986. (Proprietary)
;
- 11. J. R. Lesko, et al., "The Nuclear Design and Core Physics Characteristics of the McGuire Unit 1 Nuclear Power Plant - Cycle 5," WCAP-11589, October 1987. (Preprietary)
- 12. ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12. American Society for Testing and katerials, Philadelphia, PA, 1984, t
- 13. ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Desimetry Results", in ASTM Standards, Section 12. American Society for Testing and Materials, Philadelphia, PA, 1984.
- 14. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)",
in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, PA, 1984.
- 15. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 16. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in AS1M Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
no,m un io - 8-2
- 17. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Stendards, )
Section 12, American Society for Testing and Materials, Philadelphia, PA, i 1984.
- 18. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron .l Flux by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 19. ASTM Designation E263-82, " Standard Method for Determining fast-Neutron ,
flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, l American Society for Testing and Materials, Philadelphia, PA, 1984.
- 20. ASTM Designation E264-82, " Standard Method for Determining fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
I
- 21. ASTM Designation E481-78, " Standard Method for Measuring Neutron-Flux :
Density by Radioactivation of Cobalt and Silver", in ASTM Standards, - Section 12, American Society for Testing and Materials, Philadelphia, PA, ! 1984.
- 22. ASTM Designation E523-82, " Standard Method for Determining fast-Neutron '
Flux Density by Radioactivation of Coppar", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 23. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, PA, 1984.
- l. >
l 24. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237". in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984. i w.m.n io 8-3
i
- 25. ASTM Designation E1005-84, " Standard Method for Application and Analysis ,
of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984. i
- 26. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory Richland, WA, September 1979.- .i
- 27. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation, AFWL-TR-67-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, ,
July 1967.
- 28. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Desimetry. Applications", R. E. Maerker, et al.,1981.
t l t i i 1 4 I w,w ." 8-4
ci - APPENDIX A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERA 110N A-1. INTRODUCTION
- Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel. The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture tough-ness properties and estimating the radiation-induced ARTNDT. RT NDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.
RT increases as the material is exposed to fast-neutron radiation. NDT Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ART due to the radiation exposure associated with that NDT time period must be added to the original unirradiated RTNDT. The extent of the shift in RT NDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory 3 Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)(A-1) The value, f", given in figure A-1 is the calculated value of the neutron fluence at the location of interest (inner surface,1/4T, or 3/4T) in the vessel at the lecation of the postulated defect, n/cm2 (E
> 1 MeV) divided by 10 19 The fluence factor is determined from figure A-1. ,
A-2. FRACTURE TOUGHNESS PROPERTIES l The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan [A-2) The pre-irradiation fracture-toughness , properties of McGuire Unit 1 of the reactor vessels are presented in table A-1.
._ no.=n.. io A-1
i 1 I A-3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, f r the metal temperature at that time. K IR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code [A-3) , The KIR curve is given by the following equation: KIR = 26.78 + 1.223 exp (0.0145 (T-RTNDT + 160)) (1) where KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT l Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code [A-3) as followsi l CKIM + EIT
- KIR (2) l where l
KIM = stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR = function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits C = 1'.5 for hydrostatic and leak test conditions during which the reactor core is not critical a m . a r m ,o A-2
l At any time during the heatup or cooldown transient, KIR is determined by ) the inetal temperature at the tip of the postulated flaw, the appropriate value , for RTNDT, and the reference fracture toughness curve. The thermal stresses y resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, i K37, for the reference flaw are computed. From equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. 1 r i for the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal l gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cocidown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor. coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant l_ temperature, the AT developed during cooldown results in a higher value of K at the 1/4 T location for finite cooldown rates than for steady-state IR l operation. Furthermore, if conditions exist so that the increase in K IR exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling'is decreased at various n o ,= m ein A-3 l
- - - - , n
'1 intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as , finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K IR for the 1/4 T crack during heatup is lower than the K IR for the 1/4 T crack during steady-state conditions at the same tyce coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIR s do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite hrsatup ratea when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any , pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatep rate must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by point comparison of the steady-state and finite heatup rate data. At any given temperature, the w,an" " A-4
allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the compo:ite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the cours,e of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must , at all times be based on analysis of the most critical criterion. Finally, the 1983 Amendment to 10CFR50(A-4) has a rule which addresses the - metal temperature of the closure head flange and vessel f.lange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RT NDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure. , Table A-1 indicates that the limiting RTNDT f 40*F occurs in the closure ' head flange of McGuire Unit 1, so the minimum allowable temperature of this region is 160'F. These limits are less restrictive than the curves shown on ' figures A-2 and A-3. A-4. HEATUP AND C00LDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in section 3.0, and the procedure is presented in reference A-5. Figure A-2 is the heatup curve for 60'F/hr and applicable for the first 32 EFPY with margins for possible instrumentation errors. Figure A-3 is the cooldown curve up to 100*F/hr and applicable for the first 32 EFPY with margins for possible instrumentation errors. i Allowable combinations of temperature and pressure for specific temperature
- change rates are below and to the right of the limit lines shown in figures A-2 and A-3. This is in addition to other criteria which must be met before the reactor is made critical.
I The leak limit curve shown in figure A-2 represents minimum temperature requirements at the leak test pressure specified by applicable codes [A-2,A-3) , l The leak test limit curve was determined by methods of references A-2 and A-4, 1
** = 22" A-5 i
s E 1 a' h _ Figures A-2 and A-3 define-limits for ensuring prevention of nonductile
)Es 1 g failure for the McGuire Unit 1 Primary Reactor Coolant System.
i A-5.. ADJUSTED REFERENCE TEMPERATURE
.l c From Regulatory Guide 1.99 Rev. 2 [A-1] the adjusted reference temperature (ART) for each material in the beltline is given by the' following expression:
1 ART = Initial RTNDT + ARTNDTfMargin (3) Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of_Section III of the ASME Boiler and Pressure Vessel Code. If n.easured values of initial RT NDT f r the material in question are not available, generic mean values for that class of material may j be used if there are sufficient test results to establish a mean and standard deviation for the class. ART is the mean value of the adjustment in reference temperature caused NDT , by irradiation and should be calculated as follows: (4) ART NDT = (CF]t(0.28-0.10 log f). , To calculate ART NDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth, f(depthX)"isurface (5) where x (in inches) is the depth iato the vessel wall measured from the vessel inner (wetted) surface. The resultant fluence is then put.into equation (4) to calculate ART NDT at the specific depth. CF ('F) is the chemistry factor, obtained from reference A-1. Beltline region materials of McGuiro Unit 1 are considered for the limiting material. Limiting material is found to be the lower shell longitudinal weld located at the 30' azimuthal angle. The calculation of ART for the limiting material is shown in table A-2. This calculation was used to develop heatup and cooldown curves for McGuire Unit 1. no.=2 uao A-6
[-j=.lt : - - Il ! t
, I l
I q . TABLE A-1 MCGUIRE UNIT 1 REACTOR VESSEL TOUGHNESS TABLE (Unirradiated) Material Code Cu Ni T RT NDT NDT Component Number (%)- (%) (*F) (*F) c Closure head dome B5086-1 0.11 0.48 20 Closure head segments B5087 0.11 0.62 10 37(c) 10( ) Closure head flange B5002 -- 0.75 40(c) 40(c) Vessel Flange B4701 -- 0.73 29(c)
- Inlet nozzle B5003 0.12 0.68- 29((c) 60 c) ICI
- nlet nozzle B5003-2 0.10- 0.71 60(c) 60(c) 60 Inlet nozzle B5003-3 0.10 0.69 60(c) 60(c)
Inlet nozzle B5003-4 0.10 0.69 c) Outlet nozzle B5004-1 -- 0.74 60(C3 60 I 60((c) 60 c) Outlet nzzle B5004-2 -- 0.74 60(c) 60(c) . Outlet. le B5004-3 -- 0.71 ICI Outlet t. le B5004-4 -- 0.79 60((c) 60 c) 60(c) Upper she. B5453-2 0.14 0.58 10 60(c) 15 Upper shell B5011-2 0.10 0.54. 10 27(c) Upper shell B5011-3 0.13 0.56 0 0(c) , Intermediate shell B5012-1 0.13 0.60 -30 34 Intermediate shell B5012-2 0.13 0.62 0 0 Intermediate shell B5012 0.10 0.66- -20 -13 Lower shell .85013-1 0.14 0.56 -10 0 Lower shell B5013-2 0.10 0.52- '-10 30 Lower shell B5013-3 0.10 0.55 0 15
' Bottom head segment B5458-1 0.14 0.60 -70 -26(c)
Bottom head segment c B5458-2 0.15 0.54 -30 Bottom head segment B5458-3 0.13 0.56- -20 -15[(c]) 2 Bottom head' dome B5085 0.13 0.53 0 10(c) Intermediate shell longitudinal M1.22g) 0.21 0.88 -60 -50
, weld seams l Intermediate shell to lower shell G1.39 0.05 <0.20 -70 -70 .
H weld. ' ' ower shell longitudinal weld seam M1.32
. 0.20 0.87 -- (d) b] 0.21 0.68 -56(d)
' Lower shell longitudinal weld seam M1.33([b) -- -
56 Lower shell longitudinal weld seam M1.34 0.30 0.64 --
-56(d)
- a. Used in reactor vessel surveillance weldment
- b. Used in weld root region only
- c. Estimated per U.S. NRC Standard Review Plan (A-2]
- d. Generic mean values per Ref. A-1 un.mme,o A-7 f
*- - - ~ -
L TABLE A-2 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING MCGUIRE UNIT 1 REACTOR VESSEL MATERIAL - LONGITUDINALWELD(LOWERSHELL) Regulatory Guide 1.99'- Revision 2 32 EFPY Parameter 1/4 T- 3/4 I c-s U Chemistry Factor, CF (*F) 204.15 204.15 19 2 Fluence,f(10 n/cm)(a) .842 .305 Fluence Factor, ff .952 .675
-ART 194.3 '137.8 NDT Initial RT = CFI (xF)ff NDT, (*F) (b) -56 -56 Margin, M (*F) (c) 65.5 65.5
******************************************************************************* ;
j Re' vision 2-to Regulatory Guide 1.99 , Adjusted Reference Temperature, 203.8 147.3 ART = Initial RTNDT + ARTNDT + Margin 19 2 (a) Fluence, f, is based upon fsurf (10 n/cm , E>l Mev) = 1.4 at 32'EFPY. .The McGuire Unit I reactor vessel wall thickness is 8.465 inches at the beltline region. l l (b) The initial RTNDT (I) value f r e weld is a generic value. l (c) Margin is calculated as, M = 2 ( 1 + a)*. The standard deviation for the initial RTNDT margin term (cy) is assumed to be 17'F since the initial RT NDT is a generic mean value. The standard deviation for ARTNDT, ( a) is 28'F for the weld. no.mme:ia A-8
g t ' C
=
p L
- a y
! l ,1 1 111llllllll E " E lillllllIll lit gt ; l . ~lllllllllWOBE E : lillllli WM 1.lilililililililllllin illllE - 11llllll illilill!" . iiEL iIR illllillllllip Ill!!!jIllWWF
~ lillilllilililllLIE = lile i ; '
i filililllililililllllE llE l11lllll lilid!!! Vile 11lllllllllllllil1 ll1C ) i IIIIlllllllllllllllF ll F Illlll!!..iilillilW lillW BI11111ll111lll1 m W :..i j . lBillilillilllllllllW ID .Hiililllllliilllir IR lllllllllllllllllF ljir ' ilt llllllllll1lllllll[ llis - lillilllllillillir lill .1
.I 1 lllllllll11111llllll!0 I 1 l lll ll11111111 E : lT 111Illllll lllllllE ' 14" IlIllilllilllllW Hi? o 1
Illillllllllllll & lil11lE 'IlllllllllilllllW E i
.,lllllll1141iiWIF W l111lll1111lllllllF l
lllllllllllllllll10 lE ,~ l lllllllllllll m illF .!K 'i 3 IlililllO,WWK , . . ktillllllWWillit IiIlllllWWWE
.,.3 , ,
. . .,..; , ,
L
. .....;
menee, we=* It > t anon i
;
Figure A-1. Fluence Factor for Use in the Expression for ARI NDT m u m m eio
e- q s-MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: LONGITUDINAL WELD INITIAL RTNDT:
-56'F ]
RT AFTER 32 EFPY: 1/4T, 203.8'F NDT 3/4T, 147.2*F 1 CURVES APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. CONTAINS MARGIN OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. 2500 a,,; i n p iiiiiiiiiii ; I I eIi II i 1 I I f Leak Test [ i J 2250 I l ! I Il I' I I 2000 ? ? , I I 1750 [ J Unacceptable i I Operation I
/ ,
G 1500 r [ .
? l l w 1250 <
) )
$ I g a t.p n t.. j w Up to E 1000 60*F/Br /,
O
/
5 j u 750 i S
/
500 - j criticality Limit-
# Acceptable _ _ 3ased on Inservic[
Operation nydrostatic Test Temperature (349'N 250 for the Service - period Up to 32 ~_ arrr _ 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURC (DEC.F) Figure A-2. McGuire Unit 1 Reactor Coolant System Heatup Limitations Applicable for the First 32 EFPY (With Margins For Instrumentation Errors) ne woen..ao A-10
;
- MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: LONGITUDINAL WELD .
INITIAL RTNDT: -56*F
- RT NDT AFTER 32 EFPY: 1/4T, 203.8'F , ;
3/4T, 147.2'F
-CURVES APPLICABLE FOR C00LDOWN RATES UP TO 100*F/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. CONTAINS MARGIN OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.
2500 m I ne i i ii ,ii J
'2250 l I ! ,,'
i , ,, , 2000 / ,'
,l' r ,; i>ii J * !
1750 [ ,
]
E-1500 [ G I 6 _ Unacceptable [ " Operation r w 1250
?
O I E 1000 / a r M ^ siI Acceptable r Operation
$ 750 ,
; m.
-. n 7
- Cooldown -
500 - Rates Hr i s- s ( N 2 7 ; ;{ f
- 4a - -- e j~
250 - _-. ? OO ' ' ' '50 100 150 200 250 300 350 400 450 500 INolCATED TEMPERATURE (DEC.F) Figure A-3. Mcguire Unit 1 Reactor Coolant System Cooldown Limitations Applicable for the First 32 EFPY (With Margins For Instrumentation Errors) l l me.an.e:io A-11
n. L i y A-7. REFERENCES , A-1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May,1988. A-2 " Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981. A-3 ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Appendixes, " Rules for Construction of Nuclear Power Plant Components, , Appendix G, Protection Against Nonductile Failure," pp. 558-563, 1986
. Edition, American Society of Mechanical Engineers, New York, 1986.
A-4 Code of Federal Regulations, 10CFR50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Vol. 48 No. 104, May 27, 1983. A-5 " Procedure for Developing Heatup and Cooldown Curves," Westinghouso Electric Corporation, Generation Technology Systems Division Procedure GTSD-A-1.12 (Rev. 0), July 13,1988. l l un,mme in A-12
0 .- 4 4 F t t ATTACHMENT A DATA POINTS FOR HEATUP AND C00LDOWN CURVES .
.(With Margins-for Instrumentation Errors) a 9
?
f L r h i nei.ause io' A-13
;, _ - _ . , - . - .-
.a
-n OAP COOLDOWN CURVES REG. GUIDE 1.99,REV.2 08/18/89 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOt#4 PROFILE t ( STEADY-STATE COOLDOWN )-
IRRADIATION PERIOD = 32.000 EFP YEARS FLAW OERTH = AOWIN T INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERAf tere PRESSURE TEMPERATURE PRESSURE (DEG.F ) (PSI) (DEG.F) (PSI) (DEG. F ) . (PSI) 1 85.000 426.26 21 185.000 532.03 40 280.000 940.10 2 90.000 428.71 22 190.000 542.44 41 '285.000 980.91 3 95.000 431.25 23 195.000' 553.64 42 290.000 1024.78 4 100.000 434.08 24 200.000 565.66 43 295.000 1071.82 5 105.000 437.12 25 205.000 578.47 44 300.000 1122.37 6 ff0.OOO 440.39 26 210.000 592.39 45 305.000 1976.68-7 115.000 443.90 27 215.000 607.35 46 390.000 1234.73 8 120.000 447.68 28 220.000 623.26 47 315.000 1297.39 ' 9 125.000 451.74 29 225.000 640.55 48 320.000 '1364.53 10 130.000 456.tt 30 230.000 659.12 49 325.000 1436.39 11 135.000 460.80 31 235.000 678.92 50 330.000 1513.31 12 140.000 465.85 32 240.000 700.38 51 335.000 1596.25 y 13 145.000 471.28 33 245.000 723.28 52 340.000 1684.86 ' e 14 150.000 477.01 34 250.000 748.08 53 345.000 1779.67
- 15 155.000 483.28 35 255.000 774.54 54 A 350.000 1881.21 16 160.000 490.02 36 260.000 '803.14 55 355.000 1989.98 17 165.000 497.28 37 265.000 833.78 56 360.000 2106.07 18 170.000 505.07 38 270.000 866.62 57 365.000 2230.10 19 175.000 513.45 39 275.000 902.08 58 370.000 2362.43 20 180.000 522.45 n
i
DAP COOLDOWN CURVES REG. GUIDE 1.99 REv.2 08/18/89 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 2 (20 DEG-F f HR COOLDOWN' ) 1RRAOIATION PERIOO = 32.000 EFP YEARS FLAW OEPTH = AOWIN T INDICATED Ifd31C AT ED INDICATED INDICATED IpOICATED TEMPERATURE PRESSURE INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) (PSI) (DEG.F) (PSI) 1 85.000 382.81 16 160.000 448.35 30 2 90.000 385 22 230.000 627.60 3 17 -165.000 455.95 31 235.000 648.97 95.000 387.85 18 170.000 464.12 32 4 100.000 390.67 19 240.000 671.74 175.000 472.95 33 245.000 696.48 5 105.000 393.74 20 180.000 482.32 6 110.000 397.03 34 250.000 722.85 21 185.000 -492.56 35 255.000 7 115.000 400.61 22 751.43 190.000 503.55 36 260.000 782.01 8 120.000 404.45 23 195.000 515.42 ' 9 125.000 37 265.000 814.86 408.62 24 200.000 528.04 38 270.000 to 130.000 413.10 25 850.38 205.000 541.79 39 275.000 888.42 11 135.000 417.95 26 210.000 556.57 12 40 280.000 829.31 140.000 423.16 27 215.000 572.36 13 145.000 428.73 28 45 285.000 973.30 220.000 589.48 42 7 w 14 15 150.000 155.000 434.76 441.31 29 225.000 607.93 43 290.000 295.000 1020.50 1071.26 on - I w _. _- _ . - . - - -- - - - _ - - . ._
-cu 4
DAP COOLDOWN CURVES REG. GUIDE f.99.REV.2 ~ 08/18/89 IHE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 3 (40 DEG-F / HR COOLDOWN ) IRRADIATION PERIOD = 32.000 EFP YEARS FLAW DEPTH = AOWIN T INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE INDICATED IWICATED
'(DEG.F) (PSI) TEMPERATURE PRESSURE (DEG.F) TEMPERATURE PRESSURE (PSI) (DEG. F ) (PSil 1 85.000 338.45 15 2 90.000 340.84 '155.000 398.59 29 225.000 3 16 160.000 575.53 95.000 343.47 405.97 30 230.000 4 17 165.000 413.96 596.75 100.000 346.25 18 39 235.000 619.49 5 105.000 349.35 170.000 422 56 32 240.000 6 19 175.000 4.3.79 644.11 910.000 352.70 20 33 245.000 7 115.000 356.35 180.000. 44I.81 34 250.000 670.48 21 105.000 452.65 699.03 8 120.000 360.28 22 190.000 35 255.000 729.60 9 125.000 364.57 23 195.000 464.3t 36 260.000 762.45 to 130.000 369.18 476.80 37 265.000 798.07 tt 335.000 374.21 24 200.000 490.36 38 270.000 12 25 205.000 505.01 836.20 140.000 379.54 26 210.000 520.75 39 275.000 877.22 93 145.000 385.41 40 280.OCO 14 27 215.000 537.64 41 921.30 150.000 39t.73 28 220.000 555.93 285.000- 968.77 3= 42 290.000 0 1019.80
>=*
l
4 I i c-~ ~ ~
*1w a
~-_ a ______=_.a e .~~s.a .n .' ~ - . ' ~ --. u s * =
< ;-y-t DAP COOLDOWN CURVES REG. GUIDE t.99.REv.2 08/18/89 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 4 (60 DEG-F'/ HR COOLC M ' )
1RRADIATION PERIOD = 32.000 LFP YEARS FLAW DEPTH = ADWIN T i INDICATED INDICATED INDICATED ' INDICATED l T ElePERA TURE PRESSURE INDICATED INOICATED TElePERATURE PRESSURE T E 94PER A TURE PRESSURE (DEG.F) (PSI) (DEG.F) (PSI)
~
(DEG. F ) (PSI) 1 85.000 292.80 15 155.000 355.07 28' 2 90.000 295.28 16 225.000 543.73 3 160.000 362.84 30 230.000 5E6.59 95.000 297.93 17 165.000 371.28 31 4 100.000 300.79 235.000 590.98 5 18 170.000 380.30 32 -240.000 617.43 105.000 303.89 19 175.000 390.16 33 ! 6 110.000 307.30 245.000 645.87 - 7
- 20. 180.000 400.79 34 250.000 676.43 3 115.000 311.05 21 185.000 412.30 35 8 120.000 315.09 255.000 .709.54-9 22 190.000 424.79 36 260.000 745 09 125.000 319.52 23 195.000 438.05 37 to 130.000 265.000 783.33 '
i 11 324.31 24 200.000 452.51 38 270.000 824.47 135.000 329.53 25 205.000 12 140.000 335 16 26 468.15 39 275.000 868.80 210.000 484,89 40 280.000 916.48 13 145.000 341.30 27 215.000 503.09 14 41 285.000 967.85 150.000 347.87 28 220.000 522.68 42 290.000 1023.06 M
- l l
p l t i I L m I
.* w _. . . . . . we s - .., , , , , g- ,,p.- e ,ew y- y, ,g y- m u. ., y
t" -
; Q.
.m '1 i
DAP COOLDOWN CURVES REG. GUIDE 1.99,REV.2 08/18/89 . THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 5 '( 100 DEG-F/HR COOLDOWN .h - IRRADIATION PERIOD = 32.000 EFP YEARS FLAW DEPTH = ADWIN T 4 INDICATED IWICATED INDICATED INDICATED IWICATED IMICATEl' TEMPERATURE l'RESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG. F ) (PSI) (DEG.F) (PSI) (DEG. F ) -(PSI)' t 85.000 198.10 '15 155.000 265.60 29 225.000; 2 90.000 .482.24 1-l 200.51 16 160.000 274,32 30 230.000 508.73 3 95.000 203.23 17 165.000 31 235.000 537.2C 4 283.83 ~ 100.000 206.19 18 170.000 294-12 . 32 240.000 568.11 4 5 105.000 209.50 19 175.000 305.27 33 245.000 601.34 6 110.000 213.10 20 180.000- 317.37' 34 250.000 637.11 7 115.000 217.10 21 105.000 330.51 35 255.000 675.72 8 120.000 221.45 22 190.000 344.65 36 260.000 717.30 9 125.000 226.22 23 195.000 360.06 37 265.000 762.16 10 130.000 231.43 24 200.000 376.69- 38 270.000 810.46 11 135.000 237.17 25 205.000 394.64 ' 39 275.000 862.55 i 12 140.000 243.38 26 210.000 414.09 40 280.000 918.61 .s 13 145.000 250.19 ^27 2t5.000 -435.06 '41 285.000 " ., p 14 150.000 257.57 28 220.000 457.77 978.99 03
- i 1
i ! 'N. l I e e i i __ _ .,. .. . _ . . ,,-.~, ^ s -- - - 4 ~
~ . _ _ - .- , ray ..:s,,,..~.. ...-
DAP 6OF/HR HEATUP CURVE REG. GUIDE 1.99,REV.2 08/ %#89 COMPOSITE CURVE PLOTTED 50R HEATUP PROFILE 2-HEATUP RATE (S) (DEG.F/HR) =
'60.0 IRRADIATION PERIOD = 32.000 EFP YEARS FL AW OEPTH = (1-AOWIN)T INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE INDICATED (DEO.F ) (PSI) TE tr>E RA TURE PRESSURE (DEG.F) TEMPERATURE PRESSURE
.(PSI) (DEG.F ) (PSI) 1 85.000 426.26 21 185.000 493.44 2 90.000 428.79 41 285.000 980.91, 3 22 190.000 507.72 42 95.000 423.22 23 195.000 523.08 290.000 '1024.78 4 100.000 413.88 43 295.000 1071.82 5 24 200.000 539.79 105.000 407.42 25 205.000 557.85 44 300.000 1122.37 6 110.000 403.11 45 305.000 t176.68 7 915.000 26 210.000 577.16 ' 46 310.000 400.87 27 '215.000 598.15 1223.75 8 120.000 400.18 28 47 315.000 1273.98 9 125.000 220.000 620.57 48 320.000 409.05 29 225.000 640.55 1327.70 10 130.000 403.12 30 230.000 659.12 49 325.000 1385.57 11 135.000 406.45 50 330.000 1447.47 12 140.000 31 235.000 678.92 51 335.000 1513.44 410.77 32 240.000 700.38 y 13 145.000 415.18 52 340.000 1584.61' 33 245.000 723.28 e 14 150.000 422.49 34 53 340.000 1660.47
- 15 155.000 429.79 250.000 748.08 54 350.000 1741 83 35 255.000 774.54 55 16 160.000 437.90 36 260.000 803.14 356.000 1828.74 17 165.000 447.07 56 360.000 1921.76 37 265.000 833.78 57 18 170.000 457.16 38 270.000 866.62 365.00J 2021.21 19 175.000 468.27 39 275.000 902.08 58 370.000 2127.41 20 180.000 480.26 40 280.000 58 375.000 2240.85 940.10 60 380.000 2361.62 8
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= _ - - = - . :_ - , - . = - . . _ __ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ = . .. .-
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DAP 60F/HR HEATUP CURVE REG. GUIDE 1.99,REV.2 og/ g s/ag ' . THE FOLLOWING DATA WERE CALCULATEDFOR THE INSERVICE HYDROSTATIC LEAK TEST. MINIMUN INSERVICE LEAK TEST TEMPERATURE ( 32.000 EFPY) PRESSURE (PSI) .TFMPERATURE (DEG. F ) 2000 329
.2485 349 3a ' PRESSURE PRESSURE STRESS 1.5 K1M 8
3 (PSI) (PSI) (PSI SQ.RT.IN.) O 2000 22165 92673 2485 27384- 115553 I _ . _ _ _ _ _ _ . _ .,. . . .. ,a -. ~ . . . .,, . . :~- , .. .. ,_ _ ~ . - ,.
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