ML20151Z169
| ML20151Z169 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 12/31/1984 |
| From: | Boggs R, Cheney C, Kaiser W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML18092B024 | List: |
| References | |
| WCAP-10694, NUDOCS 8602140081 | |
| Download: ML20151Z169 (91) | |
Text
.
. WESTINGHOUSE CLASS 3-CUSTOMER DESIGNATED DISTRIBUTION ANALYSIS OF CAPSULE Y FROM THE
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM R. S. Boggs C.A. Cheney W.T. Kaiser A
December 1984
. Work performed under Shop Order No. PZVJ-138 l
APPROVED: 7.D. h % rt ~
T. A. Meyer', Manager Structural Materials and Reliability Technology Prepared by Westinghouse for the Public Service Electric and Gas Company Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the*
customer's approval.
WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 g21gg{ [ g2 81808:1b-121784 P
PREFACE This report has been tech'nically' reviewed and verified.
Reviewer Sections 1 through 5 and 7' S. E. Yanichko .
t,a. / N'e Section 6 S. L. Anderson ddd d allaw Appendix A F. J. Witt N, n///U/ m I
81808:1b-101584 iii
TABLE OF CONTENTS (Cont)
Page Title Section A-1 Appendix HEATUP AND COOLDOWN LIMIT CL'RVES FOR A NORMAL OPERATION A-1 A-1. Introduction A-L A-2. Fracture Toughness Properties A-2 A-3. Criteria For Allowable Pressure-Temperature
- Relationships A-5 A-4. Heatup and Cooldown Limit Curves yi 81808:1b-101584
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a TABLE OF CONTENTS Section Title Page 1
SUMMARY
OF RESULTS 1-1 2 INTRODUCTION 2-1 3 BACKGROUND 3-1 4 DESCRIPTION OF PROGRAM 4-1 5 TESTING OF SPECIMENS FROM CAPSULE Y 5-1 5-1. Overview 5-1 5-2. Charpy V-Notch Impact Test Results 5-3 5-3. Tension Test Results 5-4 5-4. Wedge Opening Leading Tests 5-5 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1 6-1. Introduction 6 6-2. Discrete Ordinates Analysis 6-1 6-3. Nsutron Dosimetry 6-7 6-4. Transport Analysis Results 6-10 6-5. Dosimetry Results 6-21 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE Tcl 8 REFERENCES 8-1 81808:1b-121784 v k
LIST OF ILLUSTRATIONS-Title Page Figure Arrangement of Surveillance Capsules in the 4-5 4-1 Salem Unit 1 Reactor Vessel (Updated Lead Factors for Capsules Shawn in Parentheses)
Capsule Y Diagram Showing Location of Specimens, 4-6 4-2 Thermal Monitors, and Dosimeters -
Irradiated Charpy V-Notch Impact Properties for 5-17 5-1 Salem Unit 1 Reactor Vessel Intermediate Shell Plate '
B2402-3 (Longitudinal Orientation)
Irradiated Charpy V-Notch Impact Properties for 5-18 5-2 Salem Unit 1 Reactor Pressure Vessel Weld Metal Irradiated Charpy V-Notch Impact Properties for 5-19 5-3 Salem Unit 1 Reactor Pressure Vessel Weld Heat Affected Zone Metal Irradiated Charpy V-Notch Impact Properties for 5-20 5-4 Salem Unit 1 ASTM Correlation Monitor Material Charpy lapact Specimen Fracture Surfaces for 5-21 5-5 Salem Unit 1 Pressure Vessel Intermediate Shell Plate B2402-3 Charpy Impact Spectaen Fracture Surfaces for 5-22 5-6 Salem 1 Weld Metsi Charpy Impact Specimen Fracture Surfaces for 5-23 5-7 Salem Unit 1 Weld Heat Affected Zone Metal Charpy Impact Specinen Fracture Surfaces for 5-24 5-8 ~
Salem Unit 1 Reactor /essel ASTM Correlation Monitor Material Comparison of Actual versus Predicted 30 ft Ib 5-25 5-9 Transition Temperature Increases for the Salem Unit 1 Reactor Vessel Material based on the Prediction Methodr,of Regulatory Guide 1.99 Revision 1 Tensile Properties for Salem Unit 1 Reactor 5-26 5-10 Vessel Intermediate Shell Plate 82402-3 81808:1b-010885 vii
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LIST'0F ILLUSTRATIONS (Cont) l Figure Title Page Tensile Properties for Salem Unit l' Reactor 5-27 5-11 j Vessel Weld Metal Fractured Tensile Specimens of Salem Unit 1 5-28 5-12 Reactor Vessel Intermediate Shell Plate 82402-3 '
and Weld Metal Typical Stress-Strain Curve for Tension Specimens 5-29 5-13 Salem Unit 1 Reactor Geometry 6-2 f 6-1 l Plan View of a Reactor Vessel Surveillance Capsule 6-4 6-2 6-11 6-3 Calculated Azimuthal Distribution of Maximum Fast-Neutron Flux (E > 1.0 MeV) within the Pressure Vessel Surveillance Capsule Geometry 6-12 6-4 Calculated Radial Distribution of Maximum l Fast-Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel l 6-13 l 6-5 Relative Axial Variation of Tast-Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel l
6:14
~6-6 Calculated Radial Distribution of Maximum Fast-Neutrcn Flux (E > 1.0 MeV) Within the Surveillance Capsule 6-15 6-7 Calculated Variation of Fast-Neutron Flux Monitor Saturated Activity Within Capsules Located at l
l 40 Degrees 6-16 6-8 Calculated Variation of Fast-Neutron-Flux Monitor Saturated Activity Within Capsules Located at 4 Degrees Effect of Fluence and Copper and Phosphorus Contents . A-8 A-1 on ART NDT for Reactor Vessel Steels A-9 A-2 Fast Neutron Fluence (E > 1.0 MeV) as a Function of Full Power Service Life (EFPY) l Salem Unit 1 Reactor Coolant System A-10 i A-3 Heatup Limitations Applicable up to 10 EFPY Salem Unit 1 Reactor Coolant System A-11 A-4 Cooldown Limitations Applicable up to 10 EFPY yiii l 81808:1b-121084
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LIST OF TABLES Title Page Table Chemical Composition of the Salem Unit 1 Reactor 4-3 4-1 Vessel Surveillance Materials 4-4 4-2 Heat Treatment of the Salem Unit 1 Reactor Vessel Surveillance Materials Charpy.V-Notch Impact Data for the Salem Unit 1 5-6 5-1 Intermediate Shell Plate B2402-3 Irradiated at 550*F, 8
Fluence 8.91 x 10" n/cm (E > 1 MeV)
Charpy V-Notch Impact Data'for the Salem Unit 1 5-7 5-2
' Pressure Vessel Weld Metal Irradiated at 550*F, Fluence 8.91 x 102' n/cm2 (E > 1 MeV)
Charpy V-Notch Impact Data for the Salem Unit 1 5-8 5-3 Pressure Vessel Weld Heat Affected Zone Metal 2 Irradiated at 550*F, Fluence 8.91 x 10' n/cm (E > 1 MeV)
Charpy V-Notch Impact Data for the Salem Unit 1 5-9 5-4 ASTM Correlation Monitor Material Irradiated at 550*F, Fluence 8.91 x 10" n/cm 2(E > 1 MeV)
Instrumented Charpy Impact Tes't Results for 5-10 5-5 Salem Unit 1 Intermediate Shell Plate 82402-3 (Longitudinal Orientation)
Instrumented Charpy Impact Test Results for The 5-11 5-6 Salem Unit 1 Weld Metal Instrumented Charpy Impact Test Results for 5-12 5-7 Salem Unit 1 Weld Heat Affected lone Metal Instrumented Charpy Impact Test Results for 5-13 5-8 Salem Unit 1 ASTM Correlation Monitor Material The Effect of 550*F Irradiation at 8.91 x 10" 5-14 5-9 (E > 1 MeV) on the Notch Toughness Properties of The Salem Unit 1 Reactor Vessel Materials Summary of Salem Unit 1 Reactor Vessel Surveillance 5-15 5-10 Capsule Charpy Impact Test Results Tensile Properties for Salem Unit 1 Reactor Vessel 5-16 5-11 2 Material Irradiated to 8.91 x 10" n/cm 81808:1b-121084 ix
LIST OF TABLES Table Title Page 47 Group Energy Structure 6-5 6-1 Nuclear Constants for Neutron Flux Monitors 6-7 6-2 Contained in The Salem Unit 1 Surveillance Capsules Calculated Fast-Neutron Flux (E > 1.0 MeV) and 6-17 6-3 Lead Factors for Salem Unit 1 Surveillance ,
Capsules Calculated Neutron Energy Spectra at the Center 6-18 6-4 of Salem Unit 1 Surveillance Capsules -
Spectrum-Averaged Reaction Cross Sections at the 6-19 6-5 Center of Salem Unit 1 Surveillance Capsules Irradiation History of Salem Unit 1 Reactor 6-22 6-6 Vessel Surveillance Capsules Comparison of Measured and Calculated Fast-Neutron 6-25 i 6-7 Flux Monitor Saturated Activities for Capsule Y Comparison of Measured and Calculated Fast-Neutron 6-26 6-8 Flux Monitor Satufated Activities for Capsule T
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Results of Fast-Neutron Dosimetry for Capsules Y and T 6-27 6-9 6-10 Results of Thermal-Neutron Dosimetry for Capsules Y and T 6-28 Summary of Neutron Desimetry Results for Capsule Y 6-29 6-11 Summary of Neutron Dosimetry Results for Capsule T 6-29 6-12 Salem Unit 1 Reactor Vessel Toughness Data A-7 A-1 (Unirradiated) 81808:1b-112984 x
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SECTION' 1 SlM4ARY OF RESULTS The analysis of the reactor vessel material contained in Capsule Y, the second surveillance capsule to be removed from the Public Service Electric and Gas Company Salem Unit i reactor pressure vessel, led to the following conclusions:
o The capsule received an average fast neutron fluence (E > 1.0 MeV) 18 n/cm 2, of 8.91 x 10 o Irradiation of the reactor vessel intermediate shell plate B2402-3, to 8.91 x 10 n/cm}resultedin30and50ft-lbtransitiontemperature 10 increases of 110'F and 125'F, respectively, for specimens oriented parallel to-the major working direction (longitudinal orientation).
18 2 o Weld metal irradiated to 8.91 x 10 n/cm resulted in a 30 and 50 f t-lb ~ transition temperature increase of 165'F and 160'F, respectively.
o The average upper shelf energy of the plate B2402-3 decreased from 130 to 113 f t-lbs and the limiting weld metal decreased from 104 to 75 ft-lbs. 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
- alem Unit i surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1, shows that the weld metal transition temperature increase was as predicted and was slightly less than predicted for the B2402-3 plate. Since the surveillance weld metal is not identical to the limiting intermediate to lower shell seam the future operating limits for the vessel, shown in Appendix A, were based on the upper limit ARTNDT fr m the regulatory guide curve.
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81808:1b-121784 1-1 L- .
r SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule Y, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Public Service Electric and Gas Company Salem Unit 1 reactor pressure vessel materials under actual cperating conditions, i
i The surveillance program for the Public Service Electric and Gas Company Salem ,
Unit i reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program '
and the preirradiation mechanical properties of the reactor vessel materials are presented by Davidson, ETAL.Ill The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based on ASTM E-185-70, " Recommended Practice for Surveillance Tests for Nuclear
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Reactors"(2) Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes testing and the postirradiation data obtained from surveillance Capsule Y removed from the Public Service Electric and Gas Company Salem Unit i reactor vessel and discusses the analysis of the data.
81808:1b-121784 2-1
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t SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an ir;portant factor in The beltline region of the reactor ensuring safety in the nuclear industry.
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 prcperties of low alloy ferritic pressure vessel steels such as SA302 Grade 8 modified (base material of the Salem Unit i reactor pressure vessel beltline) are well documented in the literature. Generally, low alley ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non-ductile Failure," Appendix G to Section Ill of the ASHE Boiler and Pressure Vessel Cede.
The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)*
RT is defined as the greater of either the drop weight nil-ductility NOT transition temperature (NDTT per ASTM E-208) or the temperature 60*F less than the 50 ft ib (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material. The RT NDT of a given material is used to index that material to a reference stress intensity factor curve (K;p curve) which appears in Appendix G of the ASME Code. The K IR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to 81808:1b-101584 3-1
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F the K IR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.
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RT and, in turn, the operating limits of nuclear power plants can be NDT 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 reacter pressure vessel steel can be monitored by a reactor surveillance program such as the Salem Unit 1 Reactor Vessel Radiation Surveillance Program,Ill in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are ;
tested. The increase in the average Charpy V-notch 30 f t Ib temperature to adjust the (aRTNDT) due to irradiation is added to the original RT NDT i r radiation embrittlement. This adjusted RTNDT (RTNDT initial +
RT NDT curve and, in turn, to aRTNDT) is used to index the material to the KIR set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
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~_3 SECTION 4 DESCRIPTION OF PROGRAN Eight surveillance capsule's for monitoring the effects of neutron exposure on the Salem Unit i reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The' capsules were posit, toned in the reactor vessel between the thermal shield and the vessel wall at locations shown in figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.
Capsule Y was removed after 3.39 effective full power years of plant
. operation. This capsule contained Charpy V-notch impact, tensile, and WOL '
specimens (figure 4-2) from-the intermediate shell plate B2402-3 and submerged are weld metal representative of the Intermediate shell vertical seams of the reactor vessel and Charpy V-notch specimens from weld heat-affected zone (HAZ) material and ASTM correlation monitor material. All heat affected zone
The chemistry and heat treatment of the surveillance material are presented in table 4-l'and table 4-2, respectively. The chemical analyses reported in I-r-
table 4-1 were obtained from unirradiated material used in the surveillance program. In addition, a chemical analysis was performed on an irradiated ,
I Charpy specimen from the weld metal.and is reported in table 4-1.
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- 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. Base metal Charpy V-notch impact specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction ~of the plate (longitudinal orientation). Charpy V notch and tensile specimens from the weld metal were orie'n ted with the 8'.i'08 :1b-101584 4-1
i The longitudinal axis of the specimens transverse to the welding direction.
WOL specimens in Capsule Y w~ere machined such that the simulated crack in the specimen would propagate normal to the major working direction for the plate -
specimen and parallel to the weld direction.
Capsule Y contained dosimeter wires of pure iron, copper, nickel, and i
aluminum-cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Neptunium (Np237) and Uranium (U238) were ,
4 contained in the capsule.
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex l
l tubes were included in the capsule and were located as shown in Figure 4-2. (
The two eutectic alloys and their melting points are:
2.5% Ag, 97.5% Pb Melting Point 579'F (304*C) ;
1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590*F (310*C) ! -
l The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule Y are shown in Figure 4-2.
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TABLE 4-1 CHEMICAL COMPOSIT10N OF THE SALEM UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS HSST 02 Plate B2402-3 ASTM Combustion Weld Metal Correlation Wastinghouse Engineering Westinghouse Monitor Analysis Analysis Analysis Material Element (a]
C 0.20 0.20 0.08/0.22 ICl 0.22 0.026 0.025 0.016/0.015(c] 0.018 S
0.22 --[b] 0.16/0.16(c] 0.14 Cu Si 0.27 0.27 0.17/0.16 ECl 0.25 Mo 0.42 0.45 0.53/0.49 ECl 0.52 0.52 0,50 1.26/1.14(c ] 0.68 Ni Mn 1.13 1.22 1.14/1.28 ICI 1.48
- IDI __(b]
Cr 0.12 0.04/0.04(c]
0.012 0.011 0.019/0.015[c] 0.012 P
--[b] 0.007
- IDI Sn 0.018
--(b] 0.01
--[b]
Al 0.048 (a] All elements not listed are less than 0.010 weight percent.
(b] Not measured.
(c) Analysis performed on irradiated weld specimen W-22 81808:1b-101584 4-3
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l TABLE 4-2 HEAT TREATMENT OF THE SALEM UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Temperature Time (br) Ccolant Material (*F) 4 Water que,ched Intermediate- 1550*/1650' 4 Furnace cooled }
f Shell Plate 82402-3 1225* ! 25" 1150' + 25' 40 Furnace cooled f l
i4-14 Furnace cooled Weldment 1150* 1 25=
1625* + 25' 4 Air Cooled Correlation Monitor 4 Water quenched Plate HSST 02 1600* + 25" 4 Furnace cooled A533 Grade 8 Class 1 1225* 1 25' 1150* + 25' 40 Furnace cooled 1
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81808:1b-101584 4-4
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Figure 41. Arrangement of Surveillance Capsules in the Salem Unit 1 Reactor Vessel (updated lead factors for the capsules shown in parentheses) 45
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Fegure 4-2 Capsuie Y Diagram Stww.ng i i
Locatum of Specimens. ?
Thermal Monsters and Dossneters l t
46
SECTION 5 TESTING OF SPECIMENS FRON CAPSULE Y 5-1. OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with~ consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H, ASTM Specification E185-82 and Westinghouse Procedure MHL 8402, Revision 0 ,
as modified by RMF Procedures 8102 and 8103.
Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8511. N No discrepancies were found.
Examination of the two low-melting 304'C (579'F) and 310*C (590'F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304'C (579'F).
The Charpy impact tests were performed per ASTM Specification E23-82 and RMF -
Procedure 8103 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of .
the Charpy machine is instrumented with an Effects Technology model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (E).
D From the load-time curve, the load of general yielding (Pgy),the time to general yielding (tGY), the maximum lead (P g ), and the time to maximum load g(t ) can be determined. Under scme test conditions, a sharp 81808:1b-101584 5-1
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P drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (P p), and the A
load at which f ast fracture terminated is identified as the arrest load (P )*
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.
Therefore, the propagation energy for the crack (E p ) is the difference between the total energy to fracture (E )D and the eiiergy at maximum load.
The The yield stress (cy) is calculated from the three point bend formula.
flow stress is calculated from the average of the yield and maximum loads, also using the three point bend formula.
Percentage shear was determined from postfracture photographs using the The ratio-of-areas methods in compliance with ASTM Specification A370-77.
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. 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 crcsshead speed of 0.05 inch perminute throughout the test. I Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were The ;
spring-loaded to the specimen and operated through specimen failure.
extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2
! per ASTM E83-67.
Elevated test temperatures were cbtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. l 81808:1b-101584 5-2
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.Secause 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 In test each end of the gage section of a dummy specimen and in each grip.
configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was The upper grip developed over the range room temperature to 550*F (288'C).
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 The yield elongation were determined directly from the load-extension curve.
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 10 2 contained in Capsule Y irradiated at 8.91 x 10 n/cm are presented in Tables 5-1 through 5-8 and Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the Capsule Y Table 5-10 summarizes the Charpy impact test material is shown in Table 5-9.
results from Capsule Y with the previous results of Capsule T.
Irradiation of vessel intermediate shell plate 82402-3 material spec mens to 10 8.91 x 10 n/cm2 (Figure 5-1) resulted in both 30 and 50 ft-lb transition temperature increases of 110*F and 125'F, respectively, and an upper shelf energy decrease of 17 ft-lb.
81808:1b-101584 5-3 A
T '
4 10 n/cm2 (Figure 5-2) resulted in both Weld metal irradiated to 8.91 x 10 30 and 50 f t-lb transition temperature increases of 165'F and 160*F, respectively, and an upper shelf energy decrease of. 29 f t-lb.
10 n/cm2 (Figure 5-3) resulted in Weld HAZ metal irradiated to 8.91 x 10 both 30 and 50 ft-lb transition temperature increases of 160*F and 125'F, respectively, and an upper shelf energy decrease of 19 f t-lb.
ASTM correlation monitor material irradiated to 8.91 x 1010 (Figure 5-4) resulted in both 30 and 50 ft-lb transition temperature increases of 125'F and 135'F, respectively, and an upper shelf energy decrease of 20 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 appearar:e with increasing test temperature.
Figure 5-9 shows a comparison of the 30 f t-lb transition temperature increases for the various Salem Unit 1 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision This 1.[3]
l comparison shows that the transition temperature increase resulting from 2
irradiation to 8.91 x 10 n/cm is less than predicted by the Guide for 10
- plate B2402-3 (longitudinal orientation). The weld metal transition 18 2 temperature increase resulting from 8.91 x 10 n/cm is equal to the Guide prediction.
l 5-3. TENSION 'EST RESULTS I
The results of tension tests performed on plate B2402-3 (longitudinal 18 2 orientation) and we'id metal irradiated to 8.91 x 10 n/cm are shown in f Table 5-10 and Figures 5-10 and 5-11, respectively. These results shown that irradiation produced an increase in 0.2 percent yield strength of 12 to 15 ksi for plate B2402-3 and approximately 20 ksi for the weld metal. Fractured tension specimens for each of the materials are shown in Figure 5-12. A typical stress-strain curve for the tension specimens is shown in Figure 5-13.
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81808:1b-101584 5-4
1 5-4. WELD OPENING LOADING TESTS Test results for aeld opening loading (WOL) fracture mechanics specimens centained in Capsule Y will be reported at a later time.
81808:Ib-101584 5-5 e.
l TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE SALEM UNIT 1 INTERMEDIATE SHELL PLATE B2402-3 10 IRRADIATE 0 AT 550'F, FLUENCE 8.91 x 10 n/cm2 (E > 1 WeV)
Temperature Impact Energy Lateral Expansion l
Sample No. 'C('F) Joules (ft-lbs) m (mils) % Shear E41 -18 ( 0) 8.0 ( 6.0) 0.18 ( 7.0) 2 0.60 (23.5) 17 E44 10 ( 50)~ 38.0 ( 28.0) 0.97 (38.0) 30
. E42 38 (100) 68.0 ( 50.0) 1.24 (49.0) 36 i E45 66(150) 77.5 ( 57.0)
I 1.51 (59.5) 57 I E47 93(200) 107.0 ( 79.0) 2.11 (83.0) 94 E43 121(250) 164.0 (121.0) 149.0 (110.0) 2.03 (80.0) 98 E48 149 (300) 1.94 (76.5) 100 E46 177(350) 149.0(110.0) i i
l l
I I
i iI I i i
81808:1b-101584 5-6
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE SALEW UNIT 1 PRESSURE VESSEL WELD WETAL 1RRADIATED AT 550'F, 18 FLUENCE 8.91 x 10 n/cm2 (E > 1 WeV)
Temperature Impact Energy Lateral Expansion Sample No. *C (*F) Joules (ft-lbs) m (mils) % Shear W23 -18 ( 0) 28.5 ( 21.0) 0.52 (20.5) 7 W17 -4(25) 42,0 ( 31.0) 0.67 (26.5) 19 46.0 ( 34.0) 0.57 (22.5) 25 W22 10 ( 50) 0.95 (37.5) 50 W24 38 (100) 57.0 ( 42.0) 70.5(52.0) 1.10 (43.5) 57 W21 52 (125) 100.5(74.0) 1.63 (64.0) 100 W18 66(150) 96.5(71.0) 1.37 (54.0) 100 W20 93(200) 107.0 ( 79.0) 1.47 (58.0) 100 W19 149(300) 81808:1b-101584 5-7
C i l l l
t
! l TABLE 5-3 I f CHARPY V-NOTCH IMPACT DATA FOR THE SALEM UNIT 1 f
l PRESSURE VESSEL WELD HEAT AFFECTED ZONE METAL 18 IRRADIATED AT 550*F, FLUENCE 8.91 x 10 n/cm2 (E > 1 MeV) f I
Temperature Imoact Energy Lateral Expansion mm (mils) % Shear Sample No. *C (*F) Joules (ft-lbs) l 8 H20 -46 (-50) 4.0 ( 3.0) 0.24 ( 9.5) L l 0.51 (20.0) 19 ;
H21 -18 ( 0) 36.5 ( 27.0) 0.88(34.5) 35 ,
i H24 10 ( 50) 72.0 ( 53.0) 0.58(23.0) 24 H23 10 ( 50) 36.5 ( 27.0) 1.18 (46.5) 55 H18 24 ( 75) 101.5 ( 75.0) 124.5 ( 92.0) 1.45 (57.0) 61 l H22 38 (100) 'l 1.71(67.5) 100
! H17 93 (200) 137.0 (101.0) 1.77 (69.5) 100 l H19 149 (300) 119.5 ( 88.0) f f
I I
i i
f I
t I
r 81808:1b-101584 5-8
TABLE 5-4 CHARPY V-NOTCH !MPACT DATA FOR THE SALEM UNIT 1 ASTM CORRELATION MONITOR MATERIAL 18 IRRADIATED AT 550*F, FLUENCE 8.91 x 10 n/cm2 (E > 1 MeV)
Temperature Impact Energy Lateral Expansion Sample No. *C (*F) Joules (f t-lbs) mm (mils) % Shear R62 38 (100) 13.5 ( 10.0) 0.22 ( 8.5) 16 R59 66 (150) 31.0 ( 23.0) 0.56 (22.0) 24 R57 79 (175) 38.0(28.0) 0.79 (31.0) 27 R64 93 (200) 51.5 ( 38.0) 0.76 (30.0) 33 R58 107 (225) 73.0 ( 54.0) 0.94 (37.0) 48 R60 121 (250) 115.0 ( 85.0) 1.55 (61.0) 64 R63 177 (350) 150.5 (111.0) 2.08 (82.0) 100 R61 204 (400) 127.5 ( 94.0) 1.83 (72.0) 100 81800:1b-101584 5-9 y _
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TABLE 5-10
SUMMARY
OF SALEM UNIT NO.1 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS 68 Joule 41 Joule 50 f t-lb 30 ft-lb Decrease in Trans. Temp. Trans. Temp. Upper Shelf Increase Increase Energy Fluence 10 10 n/cm 2
(.7) (,E) (ft-lb)
Material .-
120 100 19 Plate B2402-1 2.40 110 100 13 Plate B2402-2 2.40 85 75 Ncne Plate 92402-3 2.40 125 110 17 Plate B2402-3 8.91 160 165 29 Weld Metal 8.91 125 150 19 HAZ Metal 8.91 80 60 8 Correlation Monitor 2.40 135 125 20 Correlation Monitor 8.91 4
I l
5 81808:1b-121784 5-15
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W n
W W
n l
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l
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$ 5 5 ? E
. y g .n W h .o
- s.i6 l ,
_ ____________ ___Eum_.
10882-7 TEMPER ATURE t*Cl 100 -50 0 50 100 150 200 I I
,f. ,-e 1 I I iOO -
3 os3 60 -
- -d 60 -
2 O G
w 40 -
O 3 LEGENO:
go _
/ O UNIRRADIATED e IRR ADIATED 2s /
2.5
, 100
. _ 2s .-. .
- 2.
5 u_
i a s0 - e -
i i e i
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E* 3258r - -*/ - 1"
$ el - 05 20 -
O f 0 0
140 0 - 180 qg _
UNIR R ADI ATE D [ g _
e e - 140 we 100 -
BR R ADI AT E D -
120
$ O 15500FI A ~ 2 5 8 91 a 1018 n/cm -
100 3
> - 80 E 60 -
g E
125cp . ;;
w 60 g _
110*F - dr)
E0 ~ O - go 0
0 0 200 300 400 100 , 100 TEMPER ATURE (8F)
Figure 51. Irradiated Charpy V. Notch Impact Properties for Salem Unit 1 Reactor Vessel Intermediate Shell Plate 02402 3 (Longitudinal Orientation) 5 17 7 + - - . - . . . - _
6
' s 10882 6 TEMPERATURE (*Cl 0 50 100 150 150 100 -50 300 _ l l l l 3-I. .I -,
80 -
2N bM
- g ,
O LEGEND:
N 2
e O UNIRRADIATED e IRRADIATED O s2 20 -
g
~
0 25
_ 100 0_ - 2.0 E 80 - .
2 N e e- 1.s _
g .0 - O ,
f "2 -
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- j D 40 -
20 0 -
0.5 5 , 2-[0 a 140 _ ts0
- 180 120 - UNIRRADI AT ED O lRR ADI ATE D _ g4o
- 15500Fl
_ 100 -
891a1018 n/cm2 - 120 E
ia -
.s , _
,00 _
=
- 80 a .0 -
$ 160*F :
40 -
,_e > ,.5., - .0 20 - 20 i l i i 0 0
0 100 200 M0 300 200 100 TEMPERATURE (*FI Figure 6 2. Irradiated Charpy V. Notch Impact Properties for Salem Unit 1 Reactor Pressure Vessel Weld Metal i
5 18 r
I f
k"
108825 TEMPERATURE (CC) 40 0 50 100 150 150 100
,00 _ l l I I al .I 4 2 3 so -
O e .
e .0 -
,2 m
b 40 - O
" 0
}s 0
25 e 100 3 '
i so -
o j _ 2, i . _
o .- -
1 e*
5 gO 2
- in b: __ noo, u i
2O o -
g 28 O g 0 a 0 140 ~
0 O _ is0 120 -
UNIRRADIATED 0 g _ m C g - 120 t*
~
IRR ADI ATE D 100 ,
3 O s si 10 s of,,2 1
,o o a0 - [ _
- O ?- ,2s.F - a 5 o ,0 1600F - 40 g
20 2 0 - 20 i 8 l l i l 0 0 i 100 0 100 200 300 300 200 TEMPER ATURE ('Fl Figure 5 3. Irradiated Charov V. Notch Impact Properties for Salem Umt 1 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5 19 8
m
)
10882 4 TEMPERATUREl*Cl 100 150 200 100 50 0 50 200 _
i I I J ._J.
\2 3 LEGEND:
to - O UNIRRADI ATED p G IRRADIATED O e e0 t 40 -
3g IRR ADI ATED (5500F) 8.91 a 1018 n/cm2 20 - Is 2 p j '#
0-2.5 100 U b e __
mo im -
o 6 5 * -
i.5 -
g e0 -
1
~
2 g
- 1.0 5 40 -
3$ ,p 0'e
' - 0.5 20 - \
. , . 0
' 0 140 - 150
- 1e0 120 - 2 s 0 9
UNIR R ADI ATED -
140 100 - e - 120 f so -
O - 100 3 5
- 80 ~
0 60 -
N 1350F y - g z
"' 40 -
g &
- ~g 1250F - 40 O - 20 2g2 g gp 0
0 400 100 200 300 100 0 TEMPERATURE (*F)
Figure 54. Irradiated Charpy V. Notch Impact Properties for Salem Unit 1 .
ASTM Correlation Monitor Material 5 20 h k
1 i
) f i
I 10882 12 l
} l l
)
l l
i s.
- 1. . .
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- r. , p ,
l V
}
1 i
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i j
i .
i 6 I i i
l l -
e r,
l
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e j
I h r
(
- v. -
E47 E43 E48 E46 Figure 5 5. Charpy impact Specimen Fracture Surfaces for Salem Unit 1 Pressure Vessel Intermediate Shell Plate B2402 3
, 5 21 x
l n
I 10887ti l l l
t
- l a
/
- i, i
i 7
4 .
1 i
f i
l W23 W17 W22 W24 f i
?
i atte*
t l
l 1
l l
l l
t W21 W18 W20 W19 i
l I
F Figure 5 6. Charpy impact Specimen Fracture Surfaces for Salem Unit 1 Weld Metal L i
l 5 22 i
d '
h
3f l
1 10882 10 l 11 l
i
-i '
I I
I
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l l
r l ..
4
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i l
i
(
k h
a, t
i p : .
H18 H22 H17 H19 i
l Figure 5-7. Charpy impact Specimen Fracture Surfaces for Salem Unit 1 ,
Weld Heat Affected Zone Metal 5-23 i '
h i
.w
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l l i-I c
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SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6 -1. INTRODUCTION Knowledge of the neutron environment within the pressure vessel / surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation-induced properties, changes observed in materials test specimens and the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second, in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship must be established between the 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.
ransport analysis performed j This section describes a discrete ordinates S n for the Salem Unit i r ractor to determine the fast neutron (E > 1.0 MeV) flux and fluence as w?ll c, the neutron energy spectra within the reactor vessel and surveillance ci ssules. The analytical data were then used to develop lead factors for c.. in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averaged reaction cross sections derived from this calculation, the analysis of the neutron dosimetry contained in Capsule Y is discussed and a updated eva'.uation of dosimetry from Capsule T is presented.
6-2. DISCRETE ORDINATES ANALYSIS A plan view of the Salem Unit 1 reactor geometry at the core midplane is shown in figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a zero- to 45-degree sector is depicted. Eight irradiation capsules attached to 1921e:ld/101084 6-1 Jk . - .
~
f
.-T 16.3584 00 [CAPSULESl
( U,V,W,X J I ' ' / / / /./ /
h ff'I//rity Rg4'To l
1 n M
i ESSgg 40'
/ (S,T,Y,2 J
[/ j l l
/
/45' j
'J///// t
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. I i r l , '
I {
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/ / *
/
l l l
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, i
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i l / ,I I
/
',I /
1
/ /
/
I' /'
! /
- I
//
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l i
[
i.
l Figure 6-1. Salem Unit 1 Reactor Geometry I
b
' 6-2
~ m
)I the thermal shield are included in the derign to constitute the reactor vessel surveillance program. Four capsules are located symmetrically at 4 and 40 degrees from the cardinal axes as shown in figure 6-1.
A plan view of a single surveillance capsule attached to the therinal shield is shown in figure 6-2. The stainless steel specimen container is 1-inch square and approximately 38 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot-high reactor core. .
From a neutronic standpoint, the surveillance capsule structures are significant. In fact, as is sh'own later, they have a marked effect on the distributions of neutron flux and energy spectra in the water annulus between the thermal shield and the reactor vessel. Thus, in order to properly ascertain the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model. Use of at least a two-dimensional computation is therefore mandatory.
In the analysis of the neutron environment within the Salem Unit I reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the 00T E43 two-dimensional discrete ordinates code. The radial and azimuthal distributions were obtained from an R,0 computation wherein the geometry shown in figures 6-1 and 6-2 was described in the analytical model.
In addition to the R,0 computation, a second calculation in R,Z geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest. In the R,Z analysis, the reactor core was treated as an equivalent volume cylinder and, of course, the surveillance capsules were not ir.cluded in the model.
Both the R,0 and R,Z analyses employed 47 neutron.' energy groups and a P 3 expansion of the scattering cross sections. The cross sections used in the analyses were obtained from the SAILOR cross section library El which was developed specifically for light water reactor applications. The neutron energy group structure used in the analysis is listed in table 6-1.
1921e:ld/101084 6-3 2 t
5396A-17 0
(40 OR 40 )
0 (3 OR 39 )
r CHARPY SPECIMEN
1.0 MeV) to measured material property changes. To properly account for burnout of the product isotope generated by fast neutron reactions, it is necessary to also determine the magnitude of the thermal neutron flux at the monitor location. Therefore -
bare and cadmium-covered cobalt-aluminum monitors were also included.
TABLE 6-2 NUCLEAR CONSTANTS FOR NEUTRON FLUX MONITORS CONTAINED IN THE SALEM UNIT 1 SURVEILLANCE CAPSULES Target Fission Weight Product Yield Fraction Half-life (%)
Monitor Material Reaction of Interest 60 5.27 years Copper Cu (n a) Co 0.6917 6 314 days Iron Fe (n.p) Mn 0.0585 Ni (n,p) Co 0.6777 71.4 days Nickel (n,f) Cs 1.0 30.2 years 6.3 Uranium-238(*} U I 1.0 30.2 years 6.5 Neptunium-237.*I Np (n,f) Cs 60 5.27 years Cobalt-aluminum (*) Co 9 (n,y) Co 0.0015 CoS9 (n,Y) Co60 0.0015 5.27 years Cobalt-aluminum
- a. Denotes that monitor is cadmium-shielded 1921e:1d/113084 6-7 J Lw - - , ~ . . - - _ . _ . - . __
s Y '
The relative locations of the various monitors within the surveillance capsule are shown in figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptuntum and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in table 6-2 does not yield a direct measure of the energy-dependent flux level at the point of interest.
Rather, the activation or fission process is a measure of the integrated ef feet 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 from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest.
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 analysis of the passive monitors and subsequent derivation of the average First, the disintegration neutron flux requires completion of two operations.
Second, rate of product isotope per unit mass of monitor must be determined.
in order to define a suitable spectrum-averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated.
The specific activity of each of the monitors is determined using established ASTM procedures.D,8,9,M,ll] Following sample preparation, the activity of each monitor is determined by means of a lithium-drif ted germanium, Ge(Li),
garme. spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and For the samples removed from the acceptable error in detector calibration.
Salem Unit 1, the overall 2a deviation in the measured data is determined to The neutron energy spectra are determined be plus or minus 10 percent.
analytically using the method described in paragraph 6-1.
6-8 1921e:Id/101084
b Having the measured activity of the monitors and the neutror, energy spectra at the locations of interest, the. calculation of the neutron flux proceeds as follows. The reaction product activity in the monitor is expressed as R= a(E)$(E)dE I (1-e-At)) , -At d (6-2) f lY/ E j=1 max where R = induced product activity N = Avogadro's number A = atomic weight of the target isotope fg = weight fraction of the target isotope in the target material Y
= number of product atoms produced per reaction a(E) = energy dependent reaction cross section
$(E) = energy dependent neutron flux at the monitor location with ,
the reactor at full power
= average core power level during irradiation period j P) = maximum or reference core power level P
A = decay constant of the product isotope
= length of 1rradiation period j
~
t) = decay time following irradiation period j t
d Because neutron flux distributions are calculated using multigroup transport methods and, further, because the prime interest is in the fast neutron flux above 1.0 Mev, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-2) is replaced by the following relation, f
sE a(E)+(E)dE = i $ (E > 1.0 Mev) where N
I a$ 99
/o*a(E)$(E)dE ,
q=1
["
/ 1.0 MeV
$(E)dE '9 9"9 1.0 MeV 1921e:1d/101084 6-9
I Thus, equation (6-2) is rewritten R= f, Y a $ (E > l.0 Mev) I (1-e-Xt)) ,-Xtd j=1 max or, solving for the neutron flux, R
$(E > 1.0 MeV) =
(1-e- t)) ,-Xt d (6-3) fg Ya I j=1 max The total fluence above 1.0 MeV is then given by i
n P l 4(E) 1.0 MeV) = $(E> 1.0 MeV) I p 1 t j (6-4) i j=1 max }
where i t
i b t total effective full power seconds of reactor j = operation up to the time of capsule removal .
),) P l'
An assessment of the thermal neutron flux levels within the surveillance 60 capsules is obtained from the bare and cadmium-covered Co 9 (n,y) Co data by means of cadmium ratios and the use of a 37-barn, 2,200 m/sec cross section. Thus,;
0-1 R
bare
" (6-5) fg Yo g ),,-Atg ,-Xtd j=1 max where D is defined as R /R Cd covered *
. 6-4 TlANSPORT ANALYSIS RESULTS s
Results of the S transport calculations for the Salem Unit 1 reactor are n
sumarized in figures 6-3 through 6-8 and in tables 6-3 through 6-5. In i R
figure 6-3, the calculated maximum neutron flux levels at the surveillance ;
capsule centerline, pressure vessel inner radius, 1/4 thickness location, and f 3/4 thickness location are presented as a function of azimuthal angle. The 4 N
i i
l 1921e:1d/101084 6-10 ,
. 1
f 5396 A.18 11 10 SURVEILLANCE 8 -
CAPSULES R = 211.41 cm 6 -
4 -
PRESSUR E ,
VESSEL IR f 2 -
1/4T LOCATION "g
X 10 3 10
$ 8 -
a g 6 -
Thus, E 4 -
3/4T LOCATION 2 -'
108 O 10 20 30 40 50 60 A2IMUTHAL ANGLE (deg)
Figure 6-3. Calculated Azimuthal Distribution of Maxmum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure vessel - Surveillance Capsule Geometry 1921e:1d/101084 6-11
l I
5396 A.19 l
l
(
l 10 11 _
8 -
6 -
4 -
219.71 I
I g IR 225.19 ,
n 2 l
?
! I .
- 1/4T X
10 _.
3 10 1 8 -
5 6 -
m 5
w 236.14 .
4 l z -
1 i+
3/4T 241.62 {
n 2 -
l l- 1 I l l l l OR_ {
l ! l l l g g 214 216 218 220 222 224 226 228 230 232 234 236 238 240 242 F
-t R ADIUS (cm)
L i
k r
Figure 6-4. Calculated Radial Distribution of Maximum Fast Neutron Flux L' (E > 1.0 MeV) Within the Pressure Vessel L I
t 6-12 1921e:1d/101084
- A-. .
f 5396 A 20 100 _
8_ -
6 .
4 -
2 y 10-1 --
d 82 2 6 -
e 4 -
g E
l5 -
y 10-2 --
8 _
6 4 -
1 CORE MIOPLANE 2 - TO VESSE L
~ CLOSURE HEAD
! I l !
I !
10-3 ~
200 300 400
-100 0 100
-300 -200 DISTANCE FORM CORE MIDPLANE (cml Figure 6-5. Relative Axial Variation of Fast Neutron Flux (E > 1.0 MeV) Within the Pressure vessel 1921e:1d/101084 6-13
j 5396A.21 I
i 10 12 -
8 -
6 -
4 -
~
2- -
211.41 400CAPSULES 10 11 -
~
8 -
V -
e* 6 -
E 4 -
A 4 - 40CAPSULES X
3 m
2 - -
CAPSULE g CENTER i
10 10 3 J-6-
t-4 -
l
- 2 - ,
THERMALj SHIELD :
j c h
- 5
- M TEST SPECIMEN [
9 3:
l3 l !C UN I I MU' '
1' 21o 211 212 213 214 207 20s 209 RADIUS (cm)
I l Figure 6-6. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Surveillance Capsule 1921e:1d/101084 6-14 ;
i 1- -
L
1
~7F 53t6A 22 s ,
108 _
8_
6-4 -
2 -
1E ~
nim (n p) Co 6 4 ,
Np (n, f) Cs gs* 4 -
.s -
D C 2 -
211.41 5
E N 107 -
6 F (n. )Mn ,
3 4 -
5 -
2 CAPSULE CENTER 106 __
8 _
6 -
Cu63 (n, a) Co#
4 -
- w' 2
THERMALl SHIELD i,
! ) TEST SPECIMENS I,f, jgi,!, j ;', j j M #'
1E 212 M M 210 211 207 208 R AOfUS (cm)
Figure 6-7. Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules 1.ocated at 40 Degrees 1921e:1d/101084 6-15 2 e '
5396A 23 108,_
8 6_
NiS3 (n p) Co S8 4 -
~
Np237 (n, f) Cs l37 7
10 ---
l 8_
- 6 -
211.41 fem (n,p) Mn N e> 2 - U238 (n, f) Cs 137
~
2 10 6 h 8 -
c - CAPSULE CENTER
- 6 -
4 e 4 - -
- CuS3 (n, d) Co60 2 -
105 8 --
6 4 -
n u m: .
2 THERMAL! !$ !!k! TEST SPECIMENS !yU 9.
SHIELD ', !5j!u ,
o : :1 I -
l 1:
212 M N 210 211 208 209 207 R ADIUS (cml Figure 6-8, Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules Located at 4 Degrees 1921e:1d/101084 6-16
7 F' TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MeV) AND LEAD FACTORS FOR' SALEM UNIT 1 SURVEILLANCE CAPSULES Capsule Azimuthal $(E > 1.0 MeV) Lead Identification location (deo) 2 (n/cm - see) Factor u 4 2.97 x"10 10 1.02 V 4 2.97 x 10 10 1.02 W 4 2.97 x 10 1.02 x 4 2.97 x 10 10 1.02 Y 40 10 9.23 x 10 3,37 2 40 9.23 x 10 10 3,37 5 40 9.23 x 10 3.17 T 40 9.23 x 10 10 3.17 1921e:Id/113084 6-17 N
I s
TABLE 6-4 I CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF THE SALEM UNIT 1 SURVEILLANCE CAPSULES Group 2
4 (n/cm -sec) GROUP e (n/cm -sec)
No. 4* Caosules 40' Capsules No. 4* CaDsules 40* CaDsules 1 1.32x10 2.03x10 25 8.39x10' 3.34x10 9 10 4.74x10 7.48x10 26 7.92x10 3.22x10 l 2 8 9 10 9 27 6.35x10 2.61x10 3 1.52x10 2.61x10 8 9 10 8 28 4.69x10 1.94x10 4 2.68x10 4.78x10 8 6.76x10 5 4.22x10 8.01x10 29 1. 64 x10' 9 9 9 8
1.81x10 30 1.02x10 4.17x10 6 9.12x10 9 1.15x10 2.51x10 31 1.67x10 6.99x10 7
9 9 9 9 1.03x10 4.31x10 -
8 2.02x10 5.05x10 32 9 9 10 1.58x10 9
4.44x10 33 2.46x10 1.03x10 9
9 9 10 9 34 4.11x10 1.71x10 I 10 1.24x10 3.63x10 9 10 11 1.43x10 4.20x10' 35 5.53x10 2.30x10 9 9 10 8 36 5.04x10 2.12x10 12 7.03x10 2.14x10 8 9 10 8 7.61x10 3.23x10 13 2.07x10 6.37x10 37 9 9 10 1.02x10 9
3.15x10 38 4.32x10 1.84x10 14 9 9 10 f 9 39 4.57x10 1.96x10 15 2.61x10 8.18x10 10 9 10 3.14x10 9
1.03x10 40 6.13x10 2.65x10 16 10 9 10 9 7.42x10 3.23x10 17 4.56x10 1.53x10 41 ,
10 9 10 .
8.26x10 9
2.91x10 42 4.23x10 1.86x10 18 10 9 10 5.60x10' 2.04x10 43 5.12x10 2.26x10 19 10 10 9 2.77x10 9
1.02x10 44 3.38x10 1.50x10 20 10 9 10 9 45 2.86x10 1.26x10 21 7.96x10 3.21x10 10 9 10 9 46 5.46x10 2.36x10 22 6.07x10 2.39x10 10 10 10 9 47 1.38x10 5.53x10 23 7.29x10 2.86x10 9 10 24 6.36x10 2.53x10 l
l 1921e:1d/101084 6-18
_ s ht
3 TABLE 6-5 SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF SALEM UNIT 1 SURVEILLANCE CAPSULES I
l a
_a (barns)
Reaction Capsules at 4' Capsules at 40' Fe54 (n.p) Mn 54 0.0980 0.0735 60
- 0.000659 Cu63 (n a) Co 0.00112 S
Ni 0 (n.p) Co 0.127 0.0993 Np (n,f) Cs 2.62 2.83 U (n,f) Cs 0.385 0.385
_ a(E)+(E)dE
- a. a=>
'" $(E)dE s1 MeV 1921e:Id/101084 6-19 a ' .
influence of the surveillance capsules on the fast neutron flux distribution is clearly evident. In figure 6-4, the radial distribution of maximum fast neutron flux (E > 1.0 Mev) through the thickness of the reactor pressure vessel is shown. The relative axial variation of neutron flux within the vessel is given in figure 6-5. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figure 6-3 or 6-4 by the appropriate values from figure 6-5.
In figure 6-6, the radial variations of fast neutron flux within surveillance These data, in conjunction with capsules at 4 and 40 degrees are presented.
the maximum vessel flux, are used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E > 1.0 Mev) at the dosimeter block location (capsule center) to the Updated lead maximum fast neutron flux at the pressure vessel inner radius.
factors for all of the Salem Unit 1 surveillance capsules are listed in table .
6-3.
5 Since the neutron flux monitors contained within the surveillance capsules are not all located at the same radial location, the measured disintegration rates are analytically adjusted for the gradients that exist within the capsules so that flux and fluence levels may be derived on a comon basis at a comon j
location. This point of comparison was chosen to be the capsule center.
Analytically determined reaction rate gradients for use in the adjustment procedures are shown in figures 6-7 and 6-8 for capsules at 4 and 40 degrees.
All of the applicable fast neutron reactions are included, i
In order to derive neutron flux and fluence levels f rom the measu i disintegration rates, suitable spectum-averaged reaction cross sections are required. The neutron energy spectrum calculated to exist The at the center of each of the Sequoyah surveillance capsules is listed in table 6-4.
associated spectrum-averaged cross sections for each of the fast neutron. '
reactions are given in table 6-5.
t i
i 6-20 1921e:1d/101084
'~
'~~ .- _- .
6-5. 00SIMETRY RESULTS I
The irradiation history of the Salem Unit I reactor up to the time of removal j
of Capsule Y is listed in table 6-6. Comparisons of measured and calculated l
saturated activity of the flux monitors contained in Capsule Y and T based on the irradiation history shown in table 6-6 are listed in table 6-7, and 6-8.
The data are presented as measured at the actual monitor locations as well as adjusted to the capsule center. All gradient adjustments to the capsule center were based on the data presented in figure 6-7.
The fast neutron (E > 1.0 Mev) flux and fluence levels derived for Capsule Y and T are presented in table 6-9. The thermal neutron flux obtained from the Due to the relatively cobalt-aluminum monitors is sunnarized in table 6-10.
low thermal neutron flux at the capsule location, no burnup correction was made to any of the measured activities. The maximum error introduced by this assumption is estimated to be less than 1 percent for the N1 0 (n.p)Co$
reaction and even less significant for all of the other f ast neutron reactions.
An examination of table 6-9 shows that the fast neutron flux (E >101.0 Mev) derived f rom the five threshold reactions ranges fers. 7.66 x 10 to 0 2 9.44 x 10 n/cm -sec, a total span of less than 23 percent. It may also 10 2 be noted that the calculated flux value.of 9.23 x 10 n/cm -sec exceeds all but one of the measured values, with calculation to experimental ratios ranging from 0.98 to 1.21.
Comparisons of measured and calculated current fast neutron exposures for Capsules Y and T as well as for the inner radius of the pressure vessel are presented in table 6-11 and 6-12. The measured value is given based on the average of all five threshold reactions. Based on the data given in table 6-11, the best estimate exposure of Capsule Y is 18 4T = 8.91 x 10 n/cm2 (E > 1 Mev) 1921e:Id/101084 6-21
~. _
TA8tE 6-6 IRRADIATION HISTORY OF SALEM UNIT 1
. REACTOR VESSEL SURVEILLANCE CAPSULES ;
Irradiation Time Decsy Time P) P,, P /P,,
(Days) (Day)
Month Year (MW) (MW) 0.006 21 2790 12 1976 20 3483 0.063 31 2759 1 1977 220 3483 0.244 28 2731 2 1977 849 3483 3483 0.369 31 2700 3 1977 1286 0.534 30 2670 4 1977 1859 3483 0.560 31 2639 5 1977 2021 3483 0.026 30 2609 6 1977 90 3483 0.579 31 2578 7 1977 2018 3483 0.783 31 2547 8 1977 2727 3483 0.474 30 2517 9 1977 1651 3483 0.000 31 2486 10 1977 0 3483 0.030 30 2456 11 1977 104 3483 -
0.714 31 2425 12 1977 2487 3483 0.520 31 2394 1 1978 1811 3483 0.964 28 2366 2 1978 3358 3483 0.443 31 2335 3 1978 1543 3483 0.000 30 2305 4 1978 0 3483 0.000 31 2274 5 1978 0 3483 0.288 30 2244 6 1978. 1002 3483 '
0.761 31 2213 7 1978 2652 3483 0.779 31 2182 8 1978 2713 3483 0.808 30 2152 9 1978 2814 3483 0.207 31 2121 10 1978 721 3483 0.406 30 2091 11 1978 1414 3483 0.634 31 2060
- 12. 1978 2208 3483 0.891 31 2029 1 1979 3105 3483 0.866 28 2001 2 1979 3015 3483 0.831 31 1970 t 3 1979 2895 3483 t 0.082 30 1940 4 1979 285 3483 CAPSULE T REMOVED 0.000 31 1909 5 1979 0 3483 0.000 30 1879 6 1979 0 3483 0.000 31 1848 7 1979 0 3483 0.000 31 1817 8 1979 0 3483 '
0 0.000 30 1787 9 1979 3483 0.000 31 1756 10 1979 0 3483 3483 0.000 30 1726 11 1979 0 0.015 31 1695 12 1979 53 3483 0.694 31 1664 1 1980 2417 3483 0.813 29 1635 2 1980 2833 3483 4
1921e:1d/101084 6-22 6
TABLE 6-6 (Cont'd)
IRRADIATION HISTORY OF SALEM UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULES P P /P Irradiation Time Decay Time P) (Day)
Month- Year (MW) (MW) (Days) 3 1980 3398 3483 0.976 31 1604 4 1980 3303 3483 0.948 30 1574 5 1980 3281 3483 0.942 31. 1543
- 6 1980 2861 3483 0.821 30 1513 7 1980 0 3483 0.000 31 1482 8 1980 2570 3483 0.738 31 1451 1980 1007 3483 0.289 30 1421 9
10 1980 0 3483 0.000 31 1390 11 1980 0 3483 0.000 30 1360
' 12 1980 178 3483 0.051 31 1329 1 1981 2659 3463 0.763 31 1298 2 1981 2964 3483 0.851 28 1270
. 3 1981 2391 3483 0.687 31 1239 4 1981 124 3483 0.035 30 1209 l
I 5 1981 2421 3483 0.695 31 1178 6 1981 737 3483 0.212 30 1148 7 1981 2882 3483 0.828 31 1117 8 1981 2998 3483 0.861 31 1086 l
9 1981 3010 3483 0.864 30 1056 10 1981 2650 3483 0.761 31 1025 11 1981 2172 3483 0.624 30 995 12 1981 2318 3483 0.666 31 964 1 1982 45 3483 0.013 31 933 2 1982 0 3483 0.000 28 905 3 1982 0 3483 0.000 31 874 4 1982 625 3483 0.180 30 844 5 1982 3089 3483 0.887 31 813 6 1982 3201 3483 0.919 30 783 7 1982 3135 3483 0.900 31 752 3060 3483 0.879 31 721 8 1982 9 1982 3211 3483 0.922 30 691 10 1982 1639 3483 0.471 31 660 11 1982 0 3483 . 0.000 30 630 12 1982 0 3483 . 0.000 31 599 1 1983 0 3483 '0.000 31 568 2 1983 13 3483 0.004 28 540 3 1983 0 3483 0.000 31 509 4 1983 0 3483 0.000 30 479 5 1983 916 3483 0.263 31 448 6 1983 3470 3483 0.996 30 418 7 1983 3421 3483 0.982 31 387 8 1983 2556 3483 0.734 31 356 1921e:1d/101084 6-23
7 ~m7 3
s 1ABLE 6-6 (Cont'd)
IRRADIATION HISTORY OF SALEM UNIT 1
~
REACTOR VESSEL SURVEILLANCE CAPSULES Irradiation Time Decay Time P
P) P)/P (Days) (Day)
0.920 30 326 9 1983 3204 3483 1.008 31 295
- 10 1983 3512 3483 0.948 30 265 11 1983 3302 3483 0.928 31 234 12 1983 3234 3483 0.753 31 203 1 1984 2623 3483 0.817 29 174 2 1984 2846 3483 EFPS = 1.07E+08 SEC or EFPY = 3.39 years Decay time is referenced to 8/15/84.
I 1921e:1d/101084 6-24 1l
- ' -(
~
s TABLE 6-7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX l
HONITOR SATURATED ACTIVITIES FOR CAPSULE Y Saturated Activity Adjusted Saturated Activity Reaction Radial ( '"' ') (dis" /5) and Location Axial Position (cm) CaDsule Y Calculated CaDsule Y [alculated Fe (n.D) Mn$
6 6 6 E 41 211.18 3.83x10 4.59x10 3.66x10 6 6 6 E 45 211.18 3.49x10 4.59x10 3.34x10 6 6 6 E 48 211.18 4.00x10 4.59x10 3.83x10 6 6 6 H 17 212.18 3.15x10 3.81 x10 3.62x10 6 6 6 H 21 212.18 3.22x10 3.81x10 3.70x10 6 6 6 H 24 212.18 3.39x10 3.81x10 3.90x10 6 6 6 6 Average 3.51x10 4.20x10 3.68x10 4.43x10 gu63 (n.a) Co 60 5
Top-Middle 211.18 4.00x10 3.80x10 Middle 211.18 5 5 Bottom-Middle 211.18 3.72x10 3.54x10 5 5 0 5 Average 3.86x10 4.22x10 3.67x10 4.01x10 N 1, (n.D) Co I 7 7 Middle 211.18 5.81x10 6.84x10 5.48x10 6.45 x10 NP (n.f) Cs 137 7 7 7 Middle 211.41 4.41x10 4.31x10 4.41x10 4.31x10 U238 (n.f) Csl 6 6 6 6 Middle 211.41 5.88x10 5.19x10 5.88x10 5.19x10 1921e:1d/121884 6-25 a c
~
F .lI_
.i 1
TABLE 6-8 I
COMPARISON OF MEASURED AND CALCULATD FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE T 1
i Saturated Activity Adjusted Saturated I.ctivity
$ )
Radial ( $) (
Reaction " --
and Location Calculated cad 5ule T Calculated ,
Axial Position _ (cm) cad 5ule T 54 Fe54 (n.D) Mn 6 6 6 3.35x10 4.59x10 3.21 x10 P8 211.18 6 6 6 3.51x10 4.59x10 3.36x10 S1 211.18 6 6 6 3.07x10 4.59x10 2.94x10 P5 211.18 6 6 6 2.58xiO 3.81x10 2.96x10
- E8 212.18 6 6 6 2.84x10 3.81 x10 3.27x10 i E5 212.18 6 6 6 ,
3.05x10 3.81x10 3.51x10 R1 212.18 6 6 6 6 4.43x10 3.07x10 4.20x10 3.21x10 Average 60 Cu63 (n.a1 Co 5 5 3.63x10 3.45x10 Top 211.18 3.57x10' 3.39x10' Bottom 211.18 5 5 5
4.22x10 5
3.42x10 4.01x10 Average 3.60x10 Ni$0 (n.0) Co$
6.84x10 4.75x10 6.45r.10 ddle 211.18 5.04x10 i.
tj 6-26 j 1921e:1d/101084
,dA ._ _
a6
i p _ _ _ _
TA8LE 6-9 RESULTS OF FAST NEUTRON 00SIMETRY FOR CAPSULES Y AND T e
I h 4 (E > 1.0 Mev)
Adjusted Saturated Activity 4 (E > 1.0 Mev)
[-. (n/cm -sec (n/cm )
$ (DPS/cm)
Capsule Reaction Measured Calculated Measured Calculated Measured Calculated Fe54(a,p)Mn54 3.68x106 4,43xio6 7.66x1010 9.23x1010 8.17x1018 9.87x1018 Y
4.02x105 8.42x1010 a,97x1018 Cu63(n.a)Co60 3.67x105 NiS8(n.p)CoS8 5.48x107 6.45x107 7.84x1010 8.35x1018 Np237(n,f)Cs137 4.41x107 4.31x107 9.44x1010 1.01x1019 U238(n,f)Cs131(a) 5.17x106 5.19x106 8.44x1010 8.98x1018 Average 8.91x1018 Fe54(n p)Mn54 3.21x106 4.43x106 6.69x10ll 9.23x1010 2.26x1018 3.11x1018 T
Cu6 3(n,a)Co60 3.42x10 5 4.02x105 7.85x10ll 2.65x1018 N158(n.p)CoS8 4.75x107 6.45x107 6.80x10ll 2.29x1018 Average 2.40x1018
- a. U 238 adjusted saturated activity has been multiplied by 0.88 to correct for 350 ppm U 235 impurity.
- w
Nf
~
s TABLE 6-10 RESULTS OF THERMAL NEUTRON 00SIMETRY FOR CAPSULES Y and T Saturated Activity (dos /am)
$Th Cd-covered (n/cm -sec)
Caosule Axial location 8m 10 6.55x10 3.21/10 5.93x10 Y Top 10 7 I 6 44x10 2.73x10 6.59x10 Middle I I 7.08x10 10 Bottom s'.66x10 2.62x10 I 10 5.32x10 I
2.24x10 5.44x10 T Top 10 I
5.70x10 7
2.28x10 6.07x10 Middle 10 5.94x10 I
2.38x10 6.29x10 Bottom 1921e:1d/101084 6-28 d k
s I
TABLE 6-11
SUMMARY
OF NEUTRON 00SIMETRY RESULTS FOR CAPSULE Y Current 4 (E > 1.0 mev) EOL $ (E > 1.0 mev)
(n/cm ) (n/cm )
location Measured Calculated Measured Calculated 18 I8 Capsule Y 8.91x10 9.87x10 .
I8 18 I9 I9 Vessel IR 2.81x10 3.12x10 2.65x10 2.94x10 18 18 I I Vessel 1/4T 1.56x10 1.73x10 1.47x10 1.63x10 '
ll II I8 3.36x10 I8 Vessel 3/4T 3.21x10 3.56x10 3.03x10 TABLE 6-12
SUMMARY
OF NEUTRON 00SIMETRY RESULTS FOR CAPSULE T Current $ (E > 1.0 mev) EOL $ (E > 1.0 mev) 2 2 f (n/cm ) (n/cm )
- Location Measured Calculated Measured Calculated l0 18 Capsule T 2.40x10 3.10x10 l
l ll lI I9 I9 Vessel IR 7.57x10 9.77x10 2.27x10 2.94x10 ll II I9 I9 Vessel 1/4T 4.20x10 6.01x10 1.26x10 1.63x10 16 II 18 18 Vessel 3/4T 8.65x10 1.14x10 2.59x10 3.36x10 Note: EOL fluences are based on operation at 3483 MWt for 32 effective full-power years.
1921eild/101084 6-29 Ak '
SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule per ASTM E185-82 is recommended for future capsules to be removed from the Salem Unit I reactor vessel:
Vessel Estimated Location Lead Removal Fluence 2
Capsule (deg) Factor ' Time (a) (n/cm )
18 T 140* 3.17 1.08(removed) 2.40 x 10 18 Y 40' 3.17 3.39 (removed 8.91 x 10 Z 220' 3.17 6 1.57 x 10 19[b]
S 320' 3.17 10 2.62 x 10 19ICI 19 V 184* 1.02 32 2.70 1 10 0 356* 1.02 standby --
X 176* 1.02 standby --
W 4' 1.02 standby --
- a. Effective full power years from plant startup
- b. Approximate fluence at 1/4 thickness vessel wall at end of life
- c. Approximate fluence at vessel inner wall at end of life 81808:1b-112984 7-1 2
~ ~ ~ ~ - e...~ .. . _
S
SECTION 8 ,
REFERENCES
- 1. Davidson, J. A., Phillips, J. H., and Yanichko, S. E., "Public Service Electric and Gas Co. Salem Unit No. 1 Reactor Vessel Radiation
- Surveillance Program," WCAP-8511, November, 1975.
- 3. Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, April 1977.
- 4. Soltesz, R. G., Disney, R. K., Jedruch, J., and Ziegler, S. L., " 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.
- 5. SAILOR RSIC Data Library Collection OLC-76, Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library For Light Water Reactors.
- 6. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology (To be published).
I
81808:1b-121784 8-1
.r fl 6
- Philadelpnia, Pa., 1982.
,f Nuclear Standards, pp 957-961, America Society for Testing and Materials, Philadelphia, Pa., 1982.
i i
l
.p t, i il
':j 8180B:1b-112984 8-2
.0
APPENDIX A ,
HEATUP AND C00LOOWN LIMIT CURVES FOR NORMAL OPERATION A-1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature). The most limiting RTNOT
~
of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties ~and estimating the radiation-induced ART g. RT e gnated as ne MgW of eMer NDT the drop weight nil-ductility transition temperature (NOTT) or the temperature at which the material exhibits at least 50 ft Ib of impact energy and 35-mil lateral expansion (normal to ths major working direction) minus 60*F. ,
RT increases as the material is exposed to fast-neutron radiation. Thus, NOT ~
to find the most limiting RTNOT * ""Y ** E " "" *"
ART due to the radiation exposure associated with that time period must NOT be added to the original unirradiated RT ** " ' "
NOT*
RT is enhanced by certain chemical elements (such as copper and NOT phosphorus) present in reactor vessel steels. Design curves which show the effect of fluence and copper and phosphorus contents on ART NOT or nac w vessel steels are shown in Figure A-1.
Given the copper and phosphorus contents of the most limiting material, the radiation-induced ART NOT can be estimated from Figure A-1. Fast neutron fluence (E > 1 MeV) at the vessel inner surface, the 1/4 T (wall thickness),
and 3'/4 T (wall thickness) vessel locations are given as a function of full-power service lif e in Figure A-2. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RT .
1 l
l 79390:10/101884 A-1 x
lI._
A-2. FRACTURE TOUGHNESS PROPERTIES The preirradiation f racture-toughness properties of the Salem Unit I reactor vessel materials are presented in Table A-1. 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. Ell The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the Salem Unit i Vessel Material Surveillance Program.
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, K , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity f actor, K is obtained from the K IR, f r the metal temperature at that time. IR reference fracture toughness curve, defined in Appendix G of the ASME Code.E I The K curve is given by the equation:
IR K
gg
= 26.78 + 1.223 exp [0.0145 (T-RTNDT
- l I where K is the reference stress intensity factor as a function of the IR metal temperature T and the metal ref erence nil-ductility temperature RT us, e g verning equadon for De Mahp-cooMown anahsis is NOT.
defined in Appendix G of the ASME Code (2] 33 g,)),,,,
IA'2I CKgg
- Kit <K IR where IM is the stress intensity factor caused by membrane (pressure) stress K is the stress intensity f actor caused by the thermal gradients it
, 79390:10/101984 A-2 k . .
K gg is a function of temperature relative to the RTNOT of the material C = 2.0 for Level A and Level 8 service limits C = 1.5 f'or hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value
" " "C' "" "" * "9 "'55 C " ' S"**
for RTNOT'
- resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, kit, f r the reference flaw are computed. From Equation (A-2), the pressure stress intensity f actors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure-versus-coolant temperature during cooldown, the Code reference flaw 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 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 cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on 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 10. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of Kg , at the 1/4 T location for finite cooldown rates than for steady-state i
19390:10/101884 A-3 Aw -.-.----
7 operation. Furthermore, if conditions exist such that the increase in KIR exceeds K g , 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 I temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown rano. The use of the composite curve eliminates this problem and insures 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 vessel wall. The thermal gradients during heatup produce compressive stresses 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 KIR f r the 1/4 T crack during heatup is lower than the KIR f r the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower XIR '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 heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates u obtained.
The second portion of the heatup analysis concerns the calculation of 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 thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of 79390:10/101884 A-4
heatup and the time (or cColant temperature) along the heatus ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup 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 heatJp rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course 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 new 10CFR50 I3l rule which addresses the metal. temperature of the closure head flange and vessel flange regions is considered. This 10CF150 rule states that the metal temperature of the closure flange regions must exceed the material RT by at least 120*F for normal operation when the NOT pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Salem Unit 1). Table A-1 indicates that the limiting RT NOT f 50*F occurs in the vessel flange of Salem Unit 1, and the minimum allowable temperature of this region is 170*F at pressures greater than 621 psig.
A-4. HEATUP AND C00LOOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in Section A-3. The derivation of the limit curves is presented in the NRC Regulatory Standard Review Plan.E*I l
7939Q:10/101984 A-5
1 y-Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly f rom the reactor pressure j vessel surveillance program. Charpy test specimens from Capsule Y indicate
- f that the surveillance weld metal and core region intermediate shell plate code j no. 82402-3 exhibited shifts in RTNOT f 165'F and 110*F, respectively. The 18 2 j surveillance weld metal shif t at a fluence of 8.91 x 10 n/cm is on the appropriate design curve (Figure A-1) prediction, and the intermediate shell
( exhibited a shift which is within the Figure.A-1 prediction. Unfortunately, lf I the surveillance weld metal is not identical (same heat of weld wire and lot flux) to any of the beltline weld metal. As a result, the heatup and cooldown il curves are based on the upper limit ART NDT 9 " " 9# ^~I
- resultant heatup and cooldown limit curves for normal operation of the reactor vessel are presented in Figures A-3 and A-4 and represent an operational time l
period of 10 ef fective full power years. These limit curves are not impacted b by the new 10CRF50 rule.
Allowable combinations of temperature and pressure for specific temperature l change rates are below and to the right of the limit lines shown on the heatup j and cooldown curves. The reactor must not be made critical until i.
pressure-temperature combinations are to the right of the criticality limit line shown in Figure A-3. This is in addition to other criteria which must be met before the reactor is made critical.
L .
The leak test limit curve shown in Figure A-3. represents minimum temperature requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of References 2 and 4.
J i Figures A-3 and A-4 define limits for insuring prevention of nonductile r
i failure.
p :
o .
l i l a
d 9
f Lt .
m.
79390: 10/101984 A-6
]
i TABLE A-1 SALEM UNIT 1 REACTOR VESSEL TOUGHNESS DATA (UN!RRADIATED) 50 f t Ib 35-Mil Average Shelf Energy T RT Material Cu P NDT T NDT NWPU WD Component Heat No. Code No. Type (%) (1) (*F) (*F (*T) (ft Ib) (ftIb)
B2407-1 A533B, C1.1 0.20 0.011 -30 99* 39 -
110 C1 Hd Dome A0610 B2406-1 A5338. C1.1 0.13 0.010 -20 89* 29 - 125 C1 Hd Segment C1544 122 C1 Hd Segment C1544 B2406-2 A533B, C1.1 0.16 0.012 -30 85* 25 -
A5338 C1.1 0.10 0.009 -50 .66* 6 - 132 C1Hd Segment B5852 B2406-3 0.010 28* 22* 28 - 199 C1 Hd Flange 123P409 B2811 A508. C1.2 -
A508, C1.2 0.009 00* 0* 50 - 145 Vessel Flange SP1191 B2410 -
TPTBTF 0.010 50* 43* 50 - 144 Inlet Nozzle 123P403 B2408-1 A508. C1.2 -
A508, C1.2 0.011 46* 26* 46 - 157 Inlet Nozzle 125P544 B2408-2 -
0.010 47* 37* 47 - 161 Inlet Nozzle 123P403 82408-3 A508. C1.2 -
A508, C1.2 0.010 9* 17* 9 - 167 Inlet Nozzle 125P544 B2408-4 -
0.010 60* 95* 60 - 75 Outlet Nozzle ZT2550 B2409-1 A508. C1.2 -
A508, C1.2 0.011 60* 95* 60 - 78
, Outlet Nozzle ZT2550 B2409-2 -
0.013 60* 10* 60 -
121 L Outlet Nozzle ZT2585 B2409-3 A508. C1.2 -
A508 C1.2 0.012 60* 13* CD - 126 Outlet Nozzle ZT2585 B2409-4 -
A533B, C1.1 0.22 0.012 -30 87* 27 - 114 Upper Shell A0497 B2401-1 A533B, C1.1 0.19 0.011 0 80* 20 - 122 Upper Shell A0495 B2401-2 A533B, C1.1 0.24 0.011 -10 114* 34 - 96 upper Shell A0512 B2401-3 A5338, C1.1 0.24 0.010 -30 105 45 73.0 97 Inter Shell C1354 B2402-1 0.010 -30 55 -5 91.5 112 Inter Shell C1354 B2402-2 A5338. C1.1 0.24 82402-3 A5338 C1.1 0.22 0.011 -40 57 - -3 104.0 127 Inter Shell C1397 0.011 -40 70 10 99.0 143 Lower Shell C1356 B2403-1 A5338 C1.1 0.19 C1356 82403-2 A5338, C1.1 0.19 0.012 -70 86 26 94.0 128 Lower Shell 30 102.0 Lower Shell C1356 B2403-3 A533B, C1.1 0.19 0.010 -40 90 1 31 A533B, C1.1 0.10 0.009 10 48* 10 - 120 Bot Hd Segment A07J5 B2404-1 132 Bot Hd Segment A0705 B2404-2 A5338, C1.1 0.11 0.010 -50 60* O -
A5338 C1.1 0.12 0.008 10 47* 10 -
126 Bot Hd Segment A0705 52404-3 A0705 B2405-1 A5338, C1.1 0.15 0.010 -20 57* -3 - 106 Bot Hd Dome Surveillarice 0.16 0.01 9 0* - 38 *
- 0 104**
Wald NMWD - Normal to Major Working Direction WD - Major Working Direction
- - Estimated per NRC Standard Review Plan Branch Technical Position MTEB 5-2 ~
- - Actual transverse data obtained from surveillance program (from minimum data points).
C_.F=~CnhdKC l dC%~ # -
. -s u i n i n u o . u u t.u.u.. o u . u i n n gn - ._
i j ;
ng'..u..u.a.uiuuu.moc..u....._.
a 40,*(40 + 10C0 (% Cu - 0.08) + 5000 (% P - 0.008)l [f/1019] 1/2 l l ji _
E- g g ,
1l m 400 ;, g , ,
,; g i , ; ; ., ; q , c- -
1 j, j
.j ,j,, 7.;
ggg y ..
.p .l .
j -
s..p , ,,,
S M W'-Q
+-- 4 ' L- *
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MATERIAL PROPERTY BASIS UPPER LIMIT OF REG. GUIDE TREND CURVES (FIGURE A-1)
COPPER CONTENT : 0.35 WT%
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RT INITIAL : 0*F !
OT : 1/4T,236*F '
NOT : 3/4T,107'F CURVES APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY 3000 , I l.i. -,
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l x
~9 MATERIAL PROPERTY BASIS UPPER LittIT OF REG. GUIDE TREND CURVES (FIGURE A-1)
COPPER CONTENT : 0.35 WT:
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I Y
APPENDIX A REFERENCES
- 1. " Fracture Toughness Requirements," Branch Technical Position MTE8 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.
l I
- 2. ASME Boiler and Pressure Vessel Code, Section Ill, Division 1 -
Appendices, " Rules for Construction of Nuclear Vessel,s," Appendix G,
" Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, American Society of Mechanical Engineers, New York, 1983.
- 3. Code of Federal Regulations, 10CFR50, A)pendix G, " Fracture Toughness l
. Requirements," U.S. Nuclear Regulatory Comission, Washington, D.C.,
' Amended May 17, 1983 (48 Federal Register 24010),
i l
- 4. " Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants. LWR Edition, NUREG-0800, 1981.
l I
f 79390: 10/101884 A-12 s ,