ML20010H311
| ML20010H311 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/30/1980 |
| From: | Shaun Anderson, Mager T, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20010H308 | List: |
| References | |
| WCAP-9794, NUDOCS 8109240329 | |
| Download: ML20010H311 (52) | |
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l ANALYSIS OF CAPSULE 125 FROM THE CONSUMERS POWER COMPANY BIG ROCK POINT NUCLEAR PLANT REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM (WCAP 9794)
EPRI RESEARCH PROJECT 10213 TOPICAL REPORT September 1980 Prepared by WESTINGHOUSE ELECTRIC CORPGAATION i
Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvania 15230 T. R. Mager, Principal investigator Prepared for t
ELECTRIC POWER RESEARCH INSTITUTE i
3412 Hillview Avenue Palo Alto, California 94303 T. U. Ma'rston, Project Manager i
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ANALYSIS OF CAPSULE 125 FROM THE CONSUMERS POWER COMPANY BIG ROCK POINT NUCLEAR PLANT REACTOR VESSEL RADIAT ON SURVEILLANCE PROGRAM (WCAP 9794)
EPRI RESCARCH PROJECT 10213 TOPICAL REPORT S. E. Yanichko S. L Anderson R. P. Shogan R. G. Lott September 1980 Prepared by WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvania 15230 T. R. Mager, Principal Investigator 9
l Prepared for l
l ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue P3 o Alto, California 94303 l
T. U. Marstr '. Project Manager
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t LEGAL NOTICE This report was prepared by Westinghouse Electric Corporation (WESTINGHOUSE) as an account of work sponsored by the Electric Power Research Institute, Inc. (EPRI). Neither EPRI, members of EPRI, nor WESTINGHOUSE, nor any person acting on behalf of either:
a.
Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or b.
Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.
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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 125 5-1 5-1.
Charpy V-Notch Impact Test Results 52 5-2.
Tension Test Results 5-3 6
NEUTRON DOSIMETRY 6-1 6-1.
Introduction 6-1 6-2.
Neutron Dosimetry 6-1 6-3.
Discrete Ordinates Analysis 6-4 6-4.
Results of Dosimetry Analysis 6-5 References A-1 9
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. LIST OF ILLUSTRATIONS Figure Title Page 4-1 Big Rock Point Reactor Vessel Surveillance Capsule Assemblies 4-3 4-2 Big Rock Point Reactor Vessel Showing the Relative Locations of the Surveillance Capsule Positions 44 5-1 Typical instrumented Charr-- Curves 5-10 52 Charpy V-Notch Impact Data for Big Rock Point Reactor Vessel Base Metal 5-11 53 Charpy V-Notch Impact Data for Big Point Reactor Vessel Weld Metal 5-12 5-4 Charpy V Notch impact Data for Big Rock Point Reactor Vessel Weld Heat-Affected Zone Material 5 13 5-5 Charpy impact Specimen Fracture Surfaces for Big Rock Point Base Metal 5-14 5-6 Charpy impact Specimen Fracture Surfaces for Big Rock Point Weld Metal and Weld HAZ Metal 5-15 57 Comparison of Predicted Versus Actual 41-Joule Transition Temperature increase for Big Rock Point Base Material 5-16 5-8 Comparison of Predicted Versus Actual 41-Joule Transition Temperature increase for Big Rock Point Weld Metal 5-17 5-9 eFracture Tension Specimens from Big Rock Point Reactor Vessel Material 5-18 5-10 Typical Stress-Strain Curve for Tension Specimens (Tension Specimen No. 41M) 5-19 6-1 Surveillance Capsule Geometry 6 16 6-2 Big Rock Point Reactor Geometry 6-17 b
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1.1_
LIST OF TABLES Table Title Page 4-1 Chemical Composition and Heat Treatment of Big Rock Point Reactor Vessel Surveillance Material 4-2 5-1 Big Rock Point Charpy V-Notch Tou hness After Irradiation to 2.27 x 1019n/cm (E > 1 Mev) 5-4 5-2 Instrumented Charpy impact Test Results for Big Rock Point Capsule 125 Vessel Base Material 5-5 5-3 Instrumented Charpy impact Test Results for Big Rock Point Capsule 125 Weld Material 56 5-4 Instrumented Charpy impact Test Results for Big Rock Point Capsule 125 Weld Heat Affected Zone Material 5-7 5-5 Summary of Big Rock Point Reactor Vessel Surveillance Program Charpy impact Test Results 5-8 5-6 Tensile Test Results from Big Rock Point l
Capsule 125 5-9 6-1 Nuclear Parameters for Neutron Flux Monitors 6-6 6-2 21 Group Energy Structure 6-7 6-3 Irradiation History of Big Rock Point Surveillance Capsule 6-8 6-4 Measured and Saturated Activities of Fast Neutron Flux Monitors Removed from Capsule 125 6-13 Spectrum Averaged Reaction Cross-Sections for 6-5 r
Big Rock Point Capsule 125 6-14 6-6 Summary of Neutron Dosimetry Results for Big Rock Point Capsule 125 6-15 l
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SECTION 1
SUMMARY
OF RESULTS The analysis which compared unirradiated and irradiated material properties of the reactor vessel material contained in surveillance capsule 125 from the Consumers Power Company Big Rock Point Nuclear Power Plant reactor pressure vessel led to the following conclusions.
The capsule received an average fast fluence of 2.27 x 10 n/cm2 (E>1 Mev).
19 s
e The fast fluence of 2.27 x 1019 n/cm2 resulted in an 83*C (150 F) increase in the 68 joule (50 ft Ib) transition temperature and a 75 C (135 F) increase
,in the 41 joule (30 ft Ib) transition temperature of the weld metal. The transition temperature increases were essentially the same as those obtained of 7.1 x 10 yts performed on the same weld metal after irradiation to a fluence from earlier t 1 n/cm2 and indicate that the werd metal has reached a steady state or saturated condition.
Base metal specimens exhibited a 68 joule and 41 joule transition temperature m
increase of g7*C (2120*F) and 67*C (120 F), respectively, after irradiatica to 2.27 x 101 n/cm. These transition tempera;are increases were somewhat gher and 2.3 x 10 gsts performed after irradiation to fluence levels of 7.1 x 10 than in prior t 2
1 n/cm,
Heat affected zone specimens showed a 68 joule and 41 joule transition temperature a
increase ofy2 C (330 F) and 61*C (110 F), respectively, after irradiat on to i
2.27 x 10 n/cm. Both transition temperature increases were higher than those resulting from a prior capsule irradiated to 7.1 x 10gmewh n/cm I
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SECTION 2 INTRODUCTION This report presents the results of the examination of capsule 125, the fifth irradiated capsule from the continuing surveillance program which monitors the effects of neutron irradiation on the Consumers Power Company of Michigan Big Rock Point Nuclear Plant reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Big Rock Point reactor pressure vessel materials was designcd and recommended by the General Electric Company. Descriptions of the surveillance program have been presented in detail by the General Electric Company.Ill This report summarizes testing and the postirradiation data obtained from the fifth irradiated material surveillance capsule (capsule 125) removed from the Big Rock Point reactor vessel, and discusses the analysis of these data. The data are also compared to results of the four previously removed capsules evaluated by the U.S. Naval Research Laboratory.[2]
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2-1.
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l SECTION 3 BACKGROUND The ability of tha large steel pressure vessel containing the reactor core and 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 the low-alloy ferritic pressure. vessel steels such as the SA 302 B (base material of the Big Rock Point reactor pressure vessel beltline) are well documented in the literature. Generally, low-alloy ferritic materials show an increase in hard-ness 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 is presented in " Protection Against Non ductile Failure," Appendix G to Section lli of the ASME Bciler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature, RTN DT-NDT s defined as the greater of the drop weight nil-ductility transition temperature (NDTT i
RT per ASTM E 208) or the temperature 60 F less than the 50 ft-lb temperature (or 35 mil lateral expansion temperature if this is gr, eater) as determined from Charpy specimens oriented normal to the rolling direction of the material. The RTNDT of a given mate.-ial is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in appendix G of the ASME Section ill 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 inde'xed to the KIR curve, allowable stress-intensity factors can be obtained for this material as a function of tempera:ure. Allowable operating limits for the plant can then be determined utilizing these allowable stress-intensity factors.
RTNDT, and in turn the operating limits of nuclear power plants, can be adjusted to account for the 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 vessel surveillance program such as the Big Rock Point Reactor Vessel Radiation Surveillance Program,N in w1ich a surveillance capsule is periodically removed from the nuclear reactor and the encapsulated specimens are tested. The increase in the Charpy V-notch transition 3-1
temperature (ARTNDT) due to irradiation is added to the original RTNDT o adjust the t
NDT nitial + ARTNDT) i NDT or radiation embrittlement. The adjusted RTNDT (RT f
RT which takea into account the effect of irradiation on the reactor vessel rnaterials, is used to index the material to the KIR curve and in turn to set operational limits for the nuclear power plant.
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1 SECTION 4
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DESCRIPTION OF PROGRAM Nine surveillance capsules for monitoring the effects of neutron exposure on the Big Rock Point reactor pressure vessel core region material were inserted in the reactor prior to initial plant startup. Test specimens including temperature and flux monitors were loaded into capsules as shown in figure 4-1, then the capsules, grouped as sets, were secured ir, baskets and the baskets placed at various locations so that all specimens within a basktt were exposed to the same nominal neutron flux. The n;ne capsules, including two thermal control capsules, were positioned in the reactor vessel at various locations between the core and the reactor vessel wall as shown in figure 4-2.
Capsule 125 was removed after approximately 16 calender years (8.6 effective full-power years) of plant operation. This capsule contained Charpy V notch impact and tensile specimens from an A3028 base metal core region plate, weld metal representative of the core region of the reactor vessel, and weld heat-affected zone (HAZ) metal. The chemical composition and heat treatment data for the curveillance material are presented in table 4-1.
Charpy V. notch specimens were machined from the one-quarter and three-quarter thickness locations of the base metal plate. All base metal Charpy V-notch specimens were oriented in the transverse rolling direction. The V notch of the base metal impact specimens was m1chined perpendicular to the plate surfaces. Tensile specimens %m. the base metal were also machined in the transverse direction.
1 Charpy V-notch specimens machined from the surveillance weldment were oriented such that the long axis of the ppecimen was perpendicular to the weld length and the V-notch was parallel to the plate surface. Tensile specimens were oriented parallel to the length of the weld. Charpy specimens from the HAZ were perpendicular to the weld length and V-notches were parallel to the plate surface. The HAZ tensile specirnens were oriented perpendicular to the weld length with the base-weld metal interface located at the center of the gage length.
Capsule 125 contained dosimeter flux wires of iron, nicksl and copper. Thermal monitors with melting points of 309 C (589*F) and 326*C (618*F) were also included in the capsule. The 3
position of test specimens and temperature and flux monitors is illustrated in figure 4 2.
4 4-1
l TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF BIG ROCK POINT REACTOR VESSEL SURVEILLANCE MATERIAL Chemical Composition (Wt. %)
C Mn Si P
S Ni Cr Mo V
Cu 8ase metal 0.30 1.42 0.25 0.016 0.018 0.18 0.13 0.51 0.02 0.10 Weld metal 0.12 1.25 0.28 0.014 0.012 0.10 0.19 0.53 0.03 0.27 Heat Treatment: Austenitized at 1600 F (871*C) i 25*F (14*C), for 4 hou7, brine quenched.
Tempered at 1225*F (663*C) 25*F (14*C), for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, furnace cooled.
Stress relieved at 1125*F(607*C) 25 F (14*C), for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, furnsco cooled.
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IDENTIFir.ATION IDENTIFICATION MARKS MARKS 6
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ALL POSITIONS ARE SURVEILLANCE CAPSULES 122 LEGEND:
E DENOTES CAPSULES RECEIVED BY NRL Figure 4-2. Big Rock Point Reactor Vessel Showing the Relative Locations of the Surveillance Capsule Positions 4-4
SECTION 5 TESTING OF. SPECIMENS FROM CAPSULE 125 The postirradiation mechanical testing of the Charpy V notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory. Upon receipt of the capsule at the laboratory, the capsule was disassembled and all specimens were inspected relative to identification number and appearance. Thi:, inspection' revealed that the Charpy specimens located at the bottom of capsules adjacent to the aluminum filler block were welded to the filler block. As a result, none of the specimens which exhibited this condition were tested.
Examination of the two types of low-melting 309 C (589*F) and 326*C (618 F) alloys contained in the capsule indicated that the lower melting point alloy did melt. This condition was also noted by the U.S. Naval Research Laboratories on thermal monitors removed from earlier capsules in this program. The melting of this alloy apparently resulted during early plant life when the reactor operating temperature had risen to approximately 316*C (600 F), but had fallen to and remained at ~306*C (582 F) for the majority of the time, it was therefore concluded that an irradiation temperature of 307'C (585 F) would most reasonably represent the neutron embrittle-ment condition of the surveillance materials.
The Charpy impact tests were performed on a Tinius-Olsen Model 74, 358 joule machine. The tup (striker) of the Charpy machine was 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 ). From the load time curve, the D
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 conditions, a sharp drop in load indicat'ive 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 fsst fracture terminated is identified a2 the arrest load (P I-A The energy at maximum load (Ey) ',as determined by comparing the energy time record and the load time record. The energy at maximum load is roughly equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (Ep) is the difference between the total energy to fracture (E ) and the energy at maximum load. Typical D
curves for instrumented Charpy tests are shown in figure 5-1.
5-1
The yield stress (a ) is calculated from the three-point bend formula. The flow stress is y
calculated from the average of the yield and maximum loads, also using the three-point-bend formula.
Percent shear was determined from postfracture photographs using the ratio-of areas method in compliance with ASTM Specification A37D 74. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
Tensile tests were performed on a 20,000 lb Instron, split-console test machine (Model 1115) per ASTM Specifications E8 and E21. Special adapters used in the testing were made from hardened Inconel 750 material. The tests were conducted at a constant cross head speed of 0.05 inch / minute throughout the test. Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specinien and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83.
Elevated test temperatures were ol ained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. 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, wius a slight load on the specimen, a plot of specimen temperatu're vw;us upper and lower grip and controller temperatures was developed over a range from room tem-perature to 288 C (550 F). The upper grip was used to control the fumace temperature.
During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments indicated that this method is accurate to 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 fractury stress (trus stress at fracture) and percentage of reduction in area was computed using the final diameter measurement.
5-1.
CHARPY V NOTCH IMPACT TEST RESULTS The toughness results of impact tests performed on the various materials in capsule 125 after Irradiation to 2.27 x 1019 n/cm2 are presented in table 5-1 and figures 5-2 through 5-4.
Instrumented Charpy impact' test results for the various materials are presented in tables 5-2 through 5-4. Ti.e fractured surfaces of the Charpy impact specimens are'shown in figures 5-5 and 5-6.
5-2
Irradiation of the base material to 2.27 x 1019 n/cm2 as shown in figure 5 2 resulted in a 68 Joule (50 ft Ib) and 41 Joule (30 ft Ib) transition tamperature increase of 67 C (120 F) and an upper shelf energy decrease of 19 Joule (14 ft Ib).
The irradiation of the weld metal to 2.27 x 1019 n/cm2 as shown in figure 5-3 resulted in a 68 Joule (50 ft Ib) and 41 Joule (30 ft lb) transition temperature increase of 83*C (150*F) and 75*C (135*F), respectively, and an upper shelf energy decrease of 28.5 Joule (21 ft Ib).
The irradiation of HAZ metal as shown in figure 5-4 resulted in a 68 Joule (50 ft Ib) and 41 Jcule (30 ft Ib) transition temperature increase of 72 C ('30*F) and 61*C (110 F), respectively, and an upper shelf energy decrease of 27.1 Joule (20 ft ab).
A summary of the surveillance capsule impact tests results performed to date on the Big Rock Point surveillance materials is presented in table 5-5. A comparison of the 41 Joule (30 ft Ib) transition temperature increase with predicted increases using the methods of NRC Regulatory Guide 1.99 and Westinghouse predictico curves is shown for the base metal and weld metal in figures 5-7 and 5 8, respectively. The comparison of the base metal (0.10 Cu) test results 19 2
shown in figure 5 7 indicates fairly good agreement up to a fluence of 2.3 x 10 n/cm,
but then tends to indicate that the Regulatory Guide overpredicts at higher fluence levels. For the weld metal (0.27 Cu), figure 5-8 shows that actual shifts are significantly lower.than Regulatory Guide and Westinghouse predictions, and the diffe ance tends to get progressively greater at 19 2
fluences exceeding approximately 1 x 10 n/cm. These results therefore indicate that for the Big Rock Point surveillance materials, Westinghouse and especially the Regulatory Guide methods, overpredict shifts in transition temperature. These overpredictions appear to be the result of the material approaching a saturated or steady state condition.
52.
TENSION TEST RESULTS The results of tension tests performed at various temperatures for the surveillance materials after irradiation to 2.27 x 1019 n/cm2 are shown in table 5-6. These results indicate that the yield and ultimate tensile, strength dauease with increasing test temperature. The room temperature yield strength of 573 MPa (83.1 Ksi) to 587 MPa (85 Ksi) indicate that the materials are not highly sensitive (c irradiation as indicated by the Charpy test results. Fractumd tension specimens, shown in figuni 5-9, show that the materials are relatively ductile after irradiation. A typical stress strain curve for the tension specimens is shown in figure 5-10.
5-3
TABLE 5-1 BIG ROCK POINT CHARPY V. NOTCH TOUGHNESS AFTER 1RRADIATION TO 2.27 x 10 N/CM2 (E > 1 MEV) 19 Base Metal Lateral Specimen Test Temp.
Energy Expansion Shear Number
(*C)
(* F)
(J)
(ft.lb)
(mm)
(mils)
(%)
41A
-18 0
13.6 10
.14 5.5 0
414 10 50 31.2 23
.51 20 5
41B 24 75 28.5 21
.39 15.5 10 41C 38 100 24.4 18
.52 20.5 12 41D 52 125 47.5 35
.69 27 19 415 66 150 59.0 43.5
.99 39 36 413 66 150 51,5 38
.93 36.5 30 412 93 200 -
86.8 64 1.37 54 100 411 121 250 94.9 70 1.61 63.5 100 416 149 300 90.8 67 1.52 60 100 Weld Metal 4K6
-46
-50 23 17
.39 15.5 0
4K3
-18 0
23 17
.24 9.5 5
4K2 10 50 25.8 19
.44 17.5 10 4JY 24 75 58.3 43
.95 37.5 52 4JB 38 100 66.4 49 1.09 43 49 I
4JC 52 125 61.7 45.5 1.07 42 62 4K4 66 150 73.2 54 1.31 51.5 72 4JT 93 200 74.6 55 1.35 53 85 4K1 107 224 94.9 70 1.63 64 100 4JU 121 250 104.4 77 1.83 72 100 4K5 149 300 99 73 1.79 70.5 100 HAZ Metal 552
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.76 29.8 50 557 24 75 56.9 42
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551 56 150 115.2 85 1.77 69.5 100 518 93 200 86.8 64 1.33 52.5 92 s
L 553 121 250 94.9 70 1.50 59 100
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INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR i
BIG ROCK POINT CAPSULE 125 VESSEL BASE MATERIAL i
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i Normalized Energies
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Test Charpy Lateral Charpy Maximum Prop Yield Time to Maximum Time to Fracture Arrest Yield Flow Sample Tenip Energy Espansson Shear E IA Eg/A E /A Load Yield Load a Mammeum Load Load Stress Stress D
p Number (C)
(J)
(nun) 1%)
(kJ/m g ggjf,23 ggjf,2)
(N)
(ps)
(N)
(us)
(N)
(N)
(MPs)
(MPs) 2 l
41A
-18 13.6 0.14 0
169 151 18 15600 100 16700 170 16700 0
801 830 414 10 31.2 0.51 5
380 336 53 15300 110 18000 350 18000 400 790 858 418 24 28.5 a39 10 355 200 146 12200 130 15800 280 15000 629 721 41C 38 24.4 0.52 12 306 175 129 14500 100 15600 200 15600 3300 744 772 41D 52 47.5 0.60 19 593 427 165 14000 120 17100 460 17100 4000 721 801 9
ui 415 66 50.0 0.99 36 737 491 246 13800 120 17300 500 17300 69tC 700 801 413 66 51.5 0.93 30 644 331 312 14500 100 16900 350 16900 8900 744 807 412 93 86.8 1.37 100 1084 459 625 14000 100 16000 480 721 795 411 121 94.9 1.61 100 1186 475 710 13600 130 17800 490 698 807 416 149 90.8 1.52 100 1135 459 675 12500 110 16900 520 641 755 i
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TABLE 5-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR BIG ROCK POINT CAPSULE 125 WELD MATERIAL Nermalized Energies Test Charpy Laseral Clw py Menimum Prop Ywid Time to Maximum Time to Fracture Arrest Yieki Flow E /A E /A E /A Load Yield Load Maximum Load Load Strees Stress Sample Tony Energy Emparwon Sheer o
M p
2 2
2 Number (*C)
(J)
(mm)
(%)
(kJ/m ) (kJ/m )
(kJ/m )
(N)
(us)
(N)
(ps)
(N)
(N)
(MPa)
(MPa) 4KS
-46 23.0 a39 0
288 47 240 12700 110 16700 500 652 755 4K3
-18 23.0 a24 5
288 225 62 12900 120 16500 270 16500 664 755 4K2 10 25.8 Q44 10 322 258 63 14700 100 16200 260 16200 1800 755 795 4JY 24 58.3 E95 52 728 538 190 14500 110 17100 560 17100 7100 744 812 I
(P 4J8 38 66.4 1.09 49 830 491 339 14500 110 17100 510 15300 8700 744 812 l
4JC 52 61.1 1.07 62 771 411 359 14200 110
'6700 430 15300 10900 732 795 4K4 66 73.2 1.31 72 915 507 408 11800 140 16700 560 13100 606 732 4JT 93 74.6 1.35 A5 932 411 520 12900 100 16000 480 12700 8700 664 744 4K1 107 94.9 1.63 100 1186 491 694 12500 110 16700 540 641 750 4JU 121 104.4 1.83 100 1304
- 569, 735 13100 130 17100 630 675 778 4 K!,
149 99.0 1.79
-100 1237 459
- 177
'12000 110* 15600 '
- 560 618 709 i
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TABLE 54 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR BIG ROCK POINT CAPSULE 125 WELD HEAT-AFFECTED ZONE MATERIAL Normalised Energies Test Charpy Lateral Charpy Maximum Prop Yield Time to Maximum Time to Fracture Arrest Yield Flow Sample Ten, Faergy Exos wien Sheer E /A E /A Ep/A Load Yield Load Maximum Load Load Stress Stress g
M 2
2 2
Number (*C)
(J)
(m n)
(%)
(kJ/m ) (kJ/m )
(kJ/m )
(N)
(ps)
(N)
(ps)
(N)
(N)
(MPa)
(MPa) 552
-46 28.5 OL33 0
355 335 20 17000 100 19600 320 19600 0
916 961 51C
-18 31.9 0.30 5
398 375 22 16200 120 19600 37')
19600 0
835 921 555 10 36.6 0.63 26 457 359 98 15600 100 18500 340 18500 2000 801 875 554 24 69.1 a76 50 864 522 341 12500 130 18700 560 17600 641 801 Y
557 24 56.9 0.97 48 711 538 173 15800 110 18500 530 18500 5100 812 881 556 38 74.6 1.18 41 932 569 362 15300 100 18900 540 16700 3600 790
' 881 54U 52 70.5 1.02 45 881 581 299 15300 100 17800 590 17P00 9000 790 853 551 66 115.2 1.77 100 1440 600 839 13800 100 17800 QO 709 812 518 93 86.8 1.33 92 1084 491 593 14500 130 17600 530 744 824 553 121 94.9 1.50 100 1186 443 742 13600 130 17300 490 698 795 54T 149 134.2 1.94 100
- 1677, 600 1076 13300 90 17100 650 687 784
o TABLE 5-5
SUMMARY
OF BIG ROCK POINT REACTOR VESSEL SURVEILLANCE PROGRAM CHARPY IMPACT TEST RESULTS 41 Joule 68 Joule 30 ft-lb 50 ft-lb Upper Shelf Trans. Temp Trans. Temp Energy Fluence Ir. crease increase Decrease Material (1019 n/cm2)
(* C's
(*F)
(*C)
(*F)
Joule (ft lb)
Base metal
.15 0
0 0
0 0
0
.71 33 60 44 80 12.2 9
2.27 67 120 67 120 19 14 2.3 39 70 50 90 27.1 20 10.7 83 150 75 135 16.3 12 31 55 56 100 20.3 15 Weld metal
.15
.71 75 135 78 140 33.9 25 2.27 75 135 83 150 28.5 21 2.3 72 130 111 210 51.5 38 10.7 I*I 94 170 128 230 40.7 30 HAZ
.15 0
0 0
0 0
0
.71 14 25 33 60 29.8 22 2.27 61 110 72 130 27.1 20 2.3 44 80 72 130 28.5 21 10.7 89 160 106 190 33.9 25 Standard
.15 8
15 11 20 35.3 26 reference
.71 22 40 22 40 19 14
- e. Trarmiten temperature increases highly questionable due to large scatter in data.
f l
l 5-8 c-
TABLE 5-6 TENSILE TEST RESULTS FROM BIG ROCK POINT CAPSULE 125 Test
- Yield Ultimate Fractuee Fracture Uniform Totas Reduction Fracture Sample Temp Strength Strength Load Strength Elongation Elongation in Area Stress Number Material
(*C)
(MPa) (psi)
(MPa) (psil (N)
(MPa)
(%)
(%)
(%)
(MPa) 41K Base RT 587 85.0 745 107.9 16000 534 12.0 23.2 62.5 1425 41M 8ase 93.3 555 80.4 675 97.7 15400 484 9.7 19.5 59.5 1196 41L Base 287.8 507 73.5 667 96.7 16700 526 10.1 18.7 59.1 1287 51K HAZ RT 573 83.1 682 98.8 15600 492 9.4 20.9 67.2 1500 51J HAZ 93.3 527 76.4 642 93.1 13500 427 9.2 21.3 67.2 1303 4DE Weld RT 576 83.5 686 99.4 15600 493 10.7 21.5 63.9 1365 y
4DK Weld 93.3 542 78.6 639 -
92.7 14300 452 9.6 20.6 62.5 1204 4DJ Weld 287.8[a]
e.
Sample last because of equipment failuse
17468 3 Pm Pp PGy i-I I
o<
b l
P^
I l
l l
I I
I l
l I
I l
i i
x t
tGy m
ED a
e p
E m
g w
5 I
I I
I I
l i
I i
l
'm TIME Figure 5-1.
Typical Instrumented Charpy Curves li 5-10
17468-4 TEMPERATURE ( C)
-50 0
50 100 150 200 120 I
I I
I I
I 6
100 80 7
k 60 6
40 6
20 2
0 80 2.0
- d E. 60 1.5 x
l 1.0 40
~
20 O
0.5 E'
o 0.0 120 160 100 UNIRRADIATED 120
$ 80 6
6
(
D 0
3 80 $
O 5 60, 670C(120 F) b 0
g 40 0
W 67 C (120 F) 40 O
A O IRRADIATED 2.27 x 1019 2
20 n/cm 9
I I
I O
O 100 0
100 200 300 400 TEMPERATURE (OF)
Figure 5-2. Charpy V Notch impact Data for Big Rock Point Reactor Vessel Base Metal 5-11
17468 5 TEMPERATURE (OC) 100 50 0
50 100 150 200 120 100 80
--6 k
60 M
O 40 20 O
O C-80 2.0
_d F
60 1.5 a.
O X
E O
40 O
i,o e
20 O
U'8 0.0 0
120 160 100 UNIRRADIATED 120
(
3 80 O
n O
l W
U
>. 60 i-O 80 $o o
o s
y 40 g
83 C (150 F) 3 0
0 m
a O
0 0
75 C (135 F) 40 w
1RRADIATED 20 O
2.27 x 1019 n/cm2 0
o
-200 100 0
100 200 300 400 TEMPER ATURt: (OFi Figure 5-3.
Charpy V-Notch impact Data for Big Rock Point Reactor Vessel Weld Metal 5-12 i
17468 6 TEMPERATURE (OC) 100 50 0
50 100 150 200 120 I
I I
I I
I I
100 O
80
-d o O 60 9
O 40 6
2 O
6 20 2
0 v
80 O
2.0
- d O
3.
60 O
1.5 a.
O x
o 40 O
O 1.0 l 20 0.5 E"
0 0.0 120 160 100 0
UNIRRADIATED T g
120 J
80 6
t a
3 O
O 5
60 O
go g 0
0 C
72 C (130 F)
Q f, 61 C (110 F) 6 40 d
0 0
40 19 n/cm2 20 IRRADIATED 2.27 x 10 0
0 200 100 0
100 200 300 400 TEMPERATURE (OF)
Figum 5-4.
Charpy V Notch impact Data for Big Rock Point Reactor Vessel, Weld Heat Affected Zone Material 5-13
17468 7 I
i l
i l
- ' 4${
'}
m, % m,
w
- z.,c m
.... N, as,
.. u 41A 414 418 41C 41D
)
Y a'
p 1
i i
415 413 412 411 416 t
Figure 5 6 Charpy impact Specimen Fracture Surfaces for Big Rock Point Base Metal 5-14
- l
17468-8 3- _, a s
y
~..s.,,,
.i 4
. s
- - ~
..a.
- .-4 L
s
(
d.
'T l
.T s
- .;s;;9
,-,+ 6 i
' n :=
yn :.s
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'h.,
- ' *.-, _ 3.}
I' l3 ?
^.x')
My* _
l y
- ' fy 1.z.' ",
n,e -
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- s; a;
- -,.g..
ist Tg 4
', '.a f
.A
- g er,
%.c 48
, -. - ]
2 g
c.
mm..
8 I
4KS 4K3 4K2 4JY 4J8 4JC 1
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=
3 9
t.
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4K4 4JT 4K1 4JU 4K5 WELD METAL
~
l.'.
'4W w
im., -
x e-
~. *
- w..A,. 4,
.,/
3.
n.
A> v_t v
4-
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SG2' 51C 565 564 557 556 r
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3 E. 4I n-r.
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--.g.
-2
,o.
u w.
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.__.x t
_. _ y 54U 561 518 563 54T l
WELD HAZ METAL Figure 5-6 Charpy Impact Speciruen Fracture Surfaces for Big Rock Point Wald Metal and Wold HAZ Metal 5-15
500 400 -
WESTINGHOUSE 200
/
300
/
u p
~
REG. GUIDE 1.99 n
4 200 y
100 o
E "A
m K
3F 100
/
50 CAPSULE 125 I
90 y
80
/
g i
2 70
/
O 40 m
W 6G 2
30 ep O
50
/
m E
40 1
2
/
20 N
30
[
m s
/
k 20
[
/
10 q
/
I I
I I I I I I
I I I I 10 2
3 4
5 6 7891020 1018 2
3 4
5 67891019 2
FLUENCE (n/cm )
Figure 5-7.
Comparison of Predicted Versus Actual 41 Joule Transition Temperatt.rc h
increase for Big Rock Point Base Material 6
l
T
- l 500 400 REG. GUIDE 1.99 j
200 3M
_u.
o
.~
m Q
200 m
WESTINGhvuSE 100
'g m
CAPSULE 125 C
g-
/
2 m
S 100
/
l 50 e
70
/
40 g
P 30 g
o 50 H
i G
40 20 30 uJ J
3 20 G
10 I
10 18 10 2
4 5 6 7891019 2
3 4
5 6 7891020 2
FLUENCE (n/cm )
3+
Figure 5-8.
Comparison of Predicted Versus Actual 41 Joule Transition Temperature E
increase for Big Rock Point Weld Metal 5
i l
I 4
i i
i 4
i l
I 4
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4DE 4DK a
M Figure 5-9 Fractured Tension Specimens from Big Rock Point Reactor Vessel Material h
a a
4 J
l
17468-12 800 100 600 -
75 E2 m
mM 400 E
?
50 $w m
cc e
200 25 0
0 0
2 4
6 8
10 12 14 16 18 20 STR AIN (%)
)
Figure 5-10. Typical Stress-Strain Curve for Tension Specimens (Tension Specimen No. 41M) 5-19 e-
+w w
y g- -,
e--w-
SECTION 6 NEUTRON DOSIMETRY 61.
INTRODUCTION Knowledge of the neutron environment within the pressure vessel surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons.
First, in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second, in relating the changes observed in the test specimens to the present and future candition of the reactor pressure vessel, a relationship between the environment at various positions within the reactor vessel and that experienced by the test specimens must be established. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained wi h passive neutron flux monitors t
contained in ea:h of the surveillance capsules. The latter information, on the other hand, is derived solely from analysis.
This section describes the analysis of neutron dosimeters removed from Big Rock Point capw!e 125. A% discussed are the results of a discrete ordinate Sn transport analysis per-forrned fc,r the Big Rock Point reactor to derive spectrum averaged reaction cross sections which are used in the dosimeter evaluations.
6 2.
NEUTRON DOSIMETRY The passive neutron flux monitors included in Big Rock Point capsule 125 are listed in table 6-1. Each of the reactions listed in table 6-1 is used as a fast neutron monitor to relate neutron fluences (E > 1.0 Mev) to measured material property changes. The relative locations of the various monitors within a surveillance capsule are shown in figure 6-1. The flux monitors were located along the ends of the Charpy impact specimens which were irradiated in subcapsules designated L6, M6, N6, and N5.
The use of passive monitors suc5 as those listed in table 61 does not yield a direct measure of the energy-dependent flux levd at the point of interest. Rather, the activation 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 6-1
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:
e The operating history of the reactor a
The energy response of the monitor The neutron energy spectrum at the monitor location a
The physical characteristics of the monitor a
The analysis of the passive monitors and the subsequent derivation of the average neutron flux requires completion of two procedures. First, the disintegration rate of product isotope per unit mass of monitor must be determined. Second, in order to define a suitable spectrum averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated.
The specific activity of each of the monitors is determined using established ASTM pro-cedures.[3AS) Following sample preparation, the activity of each monitor is determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncer-tainty in counting, and the acceptable error in detector calibration. For the samples removed from Big Rock Point capsule 125, the overall 2o deviation in the measured data is determined to be 10 percent. The appropriate neutron energy spectra are determined analytically using the method described in section 6-3.
Having the measured activity of the monitors and the neutron energy spectra at the focations of interest, the calculation of the neutron flux proceeds as follows.
l The reaction product activity in the monitor is expressed as N
N II -,-At-),-At j
d (6-1) o R = - f; y (E) $(E) dE Pj/Pmax E
j= 1 where:
R
= induced product activity l
N
= Avagadro's number o
A
= atomic weight of the target !sotope l
6-2 N
""" +--"-
-r
f;
= weight fraction of the target isotope in the target material y
= number of product atoms prodeced per reaction ciE)
= energy-dependent reaction cross section 1
c(E)
= energy-dependent neutron flux at the monitor location with the reactor at full power P;
= average core power level during irradiation period j P
= maximum or reference core power level max A
= decay constant of the product isotope tj
= length of irradiation period j
= decay time following irradiation period j td Since neutron flux distributions are calculated using multigroup transport methods and, further, since the prime interest is in the fast neutron flux above 1.0 Mev, spectrum averaged reaction cross sections are defined such that the integral term ir. equation 6-1 is replaced by the following relation.
E where:
N oo
[ a(E) c(E) dE'
- G 4G g,o
,G=1 co N
(E) dE b
4G r
.0 Mev G=G.0 Mev 1
Thus, equation 6-1 is rewritten N
N
-At-
-At R = j f; y a c(E > 1.0 Mev) [ Pj/Pmax (1 - e be d
j=1 6-3
~
or, solving for the neutron flux, 0(E > 1.0 Mev) =
(6-2)
N N
-At-) e-At 7 y y a [ Pj/Pmax (1 - e o
3 d
f j=1 The total fluence above 1.0 Mev is then given by N
4(E > 1.0 Mev) = $(E > 1.0 Mev) [ Pj/Pmax j (6-3) t j=1 where:
N
[ P /Pmax ; = total effective full power seconds of reactor operation t
j j=1 up to the time of capsule removal 63.
DISCRETE ORDINATES ANALYSIS The irradiation basket containing Big Rock Point capsule 125 was positioned in the water annulus between the thermal shield and the reactor vessel itself. A schematic plan view of a 45* sector of the reactor geometry showing the location of capsule 125 is depicted in figure 6 2.
l From a neutronic standpoint, the surveillance capsule structures as well as the remainder of the reactor internals components are significant in that they have a marked impact on the neutron energy spectrum at the test specimen location. Thus, in order to properly ascertain the spectrum averaged reaction cross sections at the dosimeter positions, the capsule itself must be included in the analytical model. Use of at least a two-dimensional computation is, there-fore, mandatory.
e in the analysis of the neutron environment within capsule 125, predictions of the neutron energy spectrum and, hence, the spectrum averaged reaction cross sections were made with the DOTl61 two dimensional discrete ordinates code. Radial and azimuthal distributions were l
obtained from an R, o computation wherein the geometry shown in figure 6-2 was described in the analytical model. The analysis employed 21 neutron energy groups, an Sg angular quadrature, and' a Pj cross-section expansion. The cross sections were generated via the Westinghouse GAMBIT I71 code system with broad group processing by the APPROPOS I8I and ANISNI8I codes. The energy group structure used in the analysis is listed in table 6-2.
l 6-4
64.
RESULTS OF DOSIMETRY ANALYSIS The irradiation history of the Big Rock Point reactor up to the time of removal of capsule 125 is given in table 6-3. The measured actility of the iron, nickel, and copper neutron dosimeters removed from capsule 125 are given in table 6-4. Also listed in table 6-4 are the dosimeter ativities corrected to saturation. Each of the dosimeter wires has been segmented to provide two measurement points per flux wire. The spectrum average cross sections derived from the discrete ordinates calculations are summarized in table 6-5 for each of the dosimeter reactions of interest.
The fast neutron (E > 1.0 Mev) flux and fluence levels derived for capsule 125 are summarized in table 6-6. The data listed in table 6-6 show very slight variation from location to location within tre capsule as well as among the three flux monito, materials. In fact, all of the fluence data listed in table 6-6 fall within 5 percent of the average value.
Using the Fe54 (n,p) Mn54 data from table 6-6, the fast neutron exposure for capsule 125 is determined to be 2.27 x 1019 n/cm2 following 2.72 x 108 seconds of full-power operation.
f 6-5
TABLE 6-1 l
NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS
~
Monitor Reaction of Target Weight
Response
Product Material Interest Fraction Range Half-Life Copper Cu63 (n,a) Co60 0.6917 E > 4.7 Mev 5.27 years Iron Fe54 (n,p) Mn54 0.0585 E > 1.0 Mev 314 days Nickel NiS8 (n,p) CoS8 0.6777 E > 1.0 Mev 71.4 days
.f t
6-6
-v
.,..,._,n,.;,._n.
.n-
,w---
e.-., -,
.-..~__,.-,n,.-,,
.-..nn n
c
TABLE 6-2 21 GROUP ENERGY STRUCTURE Group Lower Energy (Mev) 1 7.79 'I I
2 6.07 3
4.72 4
3.68 5
2.87 6
2.23 7
1.74 8
1.35 9
1.05 10 0.821 11 0.388 12 0.111 13 4.09 x 10 2 14 1.50 x 10-2 15 5.53 x 10-3 16 5.83 x 10 4 17 7.89 x 10-5 18 1.07 x 10 5 19 1.86 x 104 20 3.00 x 10~7 21 0.0 s.
Upper energy of group 1 is 10.0 Mew m
6-7
=
4 I
TABLE 6-3
~
IRRADIATION HISTORY OF BIG ROCK POINT SURVEILLANCE CAPSULE P
P, Irradiation Time Decay Time *I I
j Month (Mw)
(Mw)
Pj/P, (days)
(days)
)
12/62 8/66 79 240
.327 1370 4884 9/66 78 240
.323 30 4854 10/66 0
240
.000 31 4823 11/66 103 240
.426 30 4793 12/66 175 240
.731 31 4762 1/67 144 240
.599 31 4731 2/67 138 240
.574 28 4703 3/67 224 240
.934 31 4672 4//7 225 240
.937 30 4642 5/67 137 240
.572 31 4611 6/67 120 240
.498 30 4581 7/67 ~
232 240
.967 31 4550 8/G7 231 240
.962 31 4519 9/67 216 240
.902 30 4489 10/67 224 240
.934 31 4458 11/67 233 240
.969 30 4428 12/67 195 240
.813 31 4397 l
1/68 189 240
~.787 31 4366
~
2/68 63 240
.264 29 4337 i
3/68 72 240'
.300 31 4306 4/68 153 240
.639 30 4276 5/I68 195 240
.813 31 4245 6/68 125 240
.521 30 4215 l
7/6a 102 240
.424 31 4184 l
8/68 235 240
.981 31 4153 9/68 215 240
.895 30 4123 10/68 194 240
.810 31 4092 l,
11/68 199 240
.829-30 4062 12/68 215 240
.894 31 4031 l
6-8 1
~
TABLE 6-3 (cont) lRRADIATION HISTORY OF BIG ROCK POINT SURVEILLANCni CAPSULE l
j P.
Irradiation Time Decay Time 'I P
Month (Mw)
(Mw)
P /P, (days)
(days) j 1/69 186 240
.775 31 4000 2/69 183 240
.762 28 3972 3/69 145 240
.603 31 3941 4/69 100 240
.418 30 3911 5/69 124 240
.516 31 3880 6/69 161 240
.669 30 3850 7/69 155 240
.647 31 3819 8/69 158 240
.659 31 3788 9/69 169 240
.706 30 3758 10/69 138 240
.576 31 3727 11/69 162 240
.676 30 3697 12/69 169 240
.703 31 3666 1/70 - 7R0 125 240
.520 212 3454 8/70 170 240
.708 31 3423 9/70 145 240
.606 30 3393 10/70 168 240
.702 31 3362 1190 144 240
.599 30 3332 12,'70 167 240
.694 31 3301 191 158 240
.660 31 3270 2R1 57 240
.237 28 3242 391
' 98 240
.410 31 3211 4/71 161 240
.672 30 3181 591 162 240
.676 31 3150 6/71 154 240
.642 30 3120 7R1 160 240
.665 31 3089 8/71 160 240
.668 31 3058 9/71 144 240
.598 30 3028 10/71 146 240
.609 31 2997 11/71 134 240
.560 30 2967 1291 137 240
.D0 31 2936 I
6-9 l
TABLE 6-3 (cont) lRRADIATION HISTORY OF BIG ROCK POINT SURVEILLANCE CAPSULE j
P, Irradiation Time Decay Time 'I l
P Month (Mw)
(Mw)
P /P (days)
(days) j max 1/72 119 240
.494 31 2905 2n2 119 240
.496 29 2876 392 61 240
.255 31 2845 4/72 0
240
.000 30 2815 Sn2 78 240
.327 31 2784 6/72 165 240
.687 30 2754 7/72 188 240
.785 31 2723 8/72 201 240,
.837 31 2692 9/72 198 240
.824 30 2662 10n2 192 240
.801
'll 2631 11/72 156 240
.650 3t!
2601 12n2 188 240
.784 31 2570 1/73 158 240
.658 31 2539 2/73 157 240
.655 28 2511 3/73 9
240
.039 31 2480 4R3 95 240
.395 30 2450 5/73 214 240
.890 31 2419 6/73 210 240
.914 30 2389 l
7/73 218 240
.907 31 2358 8n3 217 240
.905 31 2327 993 217 240
.903 30 2297 1003 219 240
.914 31 2266 11/73 186 240
.775 30 2236 12/73 45 240
.187 31 2105 1/74 108 240
.451 31 2174 2n4 167 240
.697 28 2146 3n4 120 240
.498 31 2115 4/74 0
240
.000 30 2085 5/74 160 240
.065 31 2054 6n4 11 240
.044 30 2024 l
1 6-10
TABLE 6-3 (cont)
IRRADIATION HISTORY OF BIG ROCK POINT SURVEILLANCE CAPSULE Pj P
irradition Time Decay Time *I I
max Month (M,)
(M,)
P;/P (days)
(days) max 7/74 19 240
.081 31 1993 8/74 1
240
.748 31 1962 9/74 200 240
.833 30 1932 10/74 201 240
.839 31 1901 11/74 196 240
.817 30 1871 12/74 200 240
.835 31 1840 105 99 240
.413 31 1809 2/75 595 0
240
.000 120 16o9 6R5 134 240
.557 30 1659 7/75 204-240
.850
?.1 1628 8/75 197 240
.822 31 1597 995 197 240
.820 30 1567 10/75 187 240
.778 31 1536 1195 168 240
.700 30 1506 12n5 153 240
.639 31 1475 1/76 146 240
.603 31 1444 2/76 - 6/76 0
240
.000 151 1293 7/76 12 240
.049 31 1262 8/76 155 240
.644 31 1231 996 202 240
.840 30 1201 1096
@7 240
.862 31 1170 11/76 203 240
.844 30 1140 12/76 204 240
.850 31 1109 1/77 206 240
.858 31 1078 2R7 204 240
.849 28 1050 3/77 203 240
.844 31 1019 4/77.
199 240
.828 30 989 5/77 190 240
.791 31 958 6/77 170 240
.709 30 928 6-11
-.....-.-_,.__,._.,,_.-..m,
l TABLE 6-3 (cont)
IRRADIATION HISTORY OF BIG ROCK POINT SURVEILLANCE CAPSULE P
P frradiation Time Decay Time *I I
j max Month (Mw)
(Mw) '
P /P (days)
(days) j max 7/77 109 240
.454 31 897 8/77 9/77 0
240
.000 61 836 10/77 20 240
.084 31 805 11/77 180 240
.752 30 775 12/77 201 240
.838 31 744 1/78 83
.240
.346 31 713 2/78 187 240
.778 28 685 3/78 209 240
.872 31 654 4/78 187 240
.779 30 624 5/78 207 240
.864 31 593 6/78 193 240
.803 30 563 7/78
.200 240
.834 31 532
~
8/78 199 240
.830 31 501 9/78 26 240
.108 30 471 10/78 0
240
.000 31 440 11/78 180 240
.750 30 410 12/78 192 240
.801 31 379 1I79 - 2/79 175 240
.729 32 347 Tota'l EFPS = 2.72 x 108 a.
Decay time is referenced to January 15. 1980, 08:00 1
6-12 l
l
TABLE 6-4 MEASURED AND SATURATED ACTIVITIES OF FAST NEUTRON FLUX MONITORS REMOVED FROM CAPSULE 125 Reaction and Measured Activity Saturated Activity Dosimeter ID (dPS/gm)
(dPS/gm)
Fe54 (n,p) Mn54 L6 - A 2.28 x 106 8.18 x 106 6
L6 - B 2.31 x 106 8.29 x 10 6
M6 A
'2.30 x 106 8.26 x 10 6
8.22 x 106 M6 - B 2.29 x 10 N6 A 2.43 x 106 8.72 x 106 N6 - B 2.41 x 106 8.65 x 106 6
NS - A 2.36 x 106 8.47 x 10 N5 B 2.41 x 106 8.65 x 106 i
NiS8 (n,p) CoS8 8
L6 - A 2.46 x 100 1.12 x 10 L6 - B 2.37 x 106 1.08 x 108 M6 - A 2.45 x 106 1.12 x 108 M6 - B 2.5'5 x 106 1.17 x 108 N6 - A 2.58 x 106 1.18 x 108 8
N6 B 2.58 x 106 1.18 x 10 8
NS - A 2.52 x 106 1.15 x 10 NS B 2.48 x 106 1.13 x 108 cub 3 (n,a) Co60 5
L6 - A 3.95 x 105 9.13 x 10 L6 - B 4.08 x 105 9.43 x 105 5
5 9.50 x 10 M6 - A 4.11 : 10 5
M6 - B 4.03 x 195 9.31 x 10 5
N6 A 4.32 x 105 9.98 x 10 5
9.80 x 105 N6 B 4.24 x 10 NS.- A 4.09 x 105 9.45 x 105 5
NS - B 4.16 x 105 9.61 x 10 i
6-13
. rn :.
I i
i TABLE 64 MEASURED AND SATURATED ACTIVITIES OF FAST NEUTRON
(
FLUX MONITORS REMOVED FROM CAPSULE 125 Reaction and Measured Activity Saturated Activity Dosimeter ID (dPS/gm)
(dPS/gm)
Fe54 (n,p) Mn54 i
L6 - A 2.28 x 10 8.18 x 106 6
L6 - 8 2.31 x 106 8.29 x 106 6
M6 A
'2.30 x 106 8.26 x 10 6
l M6 - B 2.29 x 106 8.22 x 10 2.43 ' x 106 8.72 x 106 l
N6 A 6
N6 - B 2.41 x 106 8.65 x 10 6
N5 A 2.36 x 106 8.47 x 10 N5 - B 2.41 x.106 8.65 x 106 Ni58 (n.p) CoS8 L6 - A 2.46 x 106 1.12 x 108 8
L6 - B 2.37 x 106 1.08 x 10 M6 - A 2.45 x 106 1.12 x 108 M6 - B 2.55 x 106 1.17 x 108 8
N6 - A 2.58 x 106 1.18 x 10 8
N6 B 2.58 x 106 1.18 x 10 8'
N5 - A 2.52 x 106 1.15 x 10 NS - B 2.48 x 106 1.13 x 108 Cu63 (n,a) Co60 LS - A 3.95 x 105 9.13 x 105 5
L6 - B 4.08 x 105 9.43 x 10 5
M6 - A 4.11 x 105 9.50 x 10 M6 B 4.03 x 105 9.31 x 105 N6 - A 4.32 x 105 9.98 x 105 N6 B 4.24 x 105 9.80 x 105 NS - A 4.09 x 105 9.45 x 105 N5 - B 4.16 x 105 9.61 x 105 6-13
e TABLE 6-5 SPECTRUM AVERAGED REACTION CROSS SECTIONS FOR BIG ROCK POINT CAPSULE 125 o
Reaction (barns) fen (n,p) Mn54 0.155 N 58 (n p) CoS8 0.194 CuS3 (n,a) Co60 0.00173 o.
[o a(E) O(E) dE 1
[1.0 Mcv d(E) dE O
t a
I 6-14
TABLE 6-6
~
SUMMARY
OF NEUTRON DOSIMETRY RESULTS FOR BIG ROCK POINT CAPSULE 125 Reaction and
&(E > 1.0 Mev) 4(E > 1.0 Mov) 2 2
Dosimeter ID (n/cm -sec)
(n/cm )
Fe54 (n,p) Mn54 10 19 L6 - A 8.09 x 10 2.20 x 10 l
L6 - B 8.19 x 1010 10 2.23 x 10 M6 - A 8.16 x 1010 10 2.22 x 10 M6-B-8.12 x 1010 19 2.21 x 10 N6 - A 8.62 x 1013 2.34 x 1019 10 N6 - B 8.55 x 10 2.33 x 1019 N5 - A 8.38 x 1019 19 2.28 x 10 N5 - B 8.56 x 1010 2.33 x 1019 Average 8.33 x 1010 19 2.27 x 10 NiS8 (n,p) CoS8 10 L6 - A 8.24 x 10 2.24 x 1019 L6 - B 7.94 x 1010 2.16 x 1019 l
M6 - A 8.20 x 1010 2.23 x 1019 M6 B 8.54 x 1010 2.32 x 1019 N6 - A 8.64 x 1010 2.35.c 1019 l
N6 B 8.64 x 1010 2.35 x 1019 N5 A 8.44 x 1010 2.30 x 1019 N5-B 8.31 x 1010 2.26 x 1019 Average r 8.38 x 1010 2.28 x 1019 Cu6 3 (n,a) Co60 10 L6 - A 7.98 x 10 2.17 x 1019 L6 - B 8.24 x 1010 2.24 x 1019 i
M6 - A 8.30 x 1010 2.26 x 1019 10 M6 B 8.14 x 10 2.21 x 1019 N6 - A 8.73 x 1010 2.37 x 1019
' N6 - 8 8.57 x 1010 2.33 x 1019 N5 - A 8.26 x 1010 2.25 x 1019 N5 B 8.40 x 1010 19 2.28 x 10 Average 8.33 x 10ll 2.27 x 1019 6-15
17468-13
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SPECIMEN ARRAY l
1 1 : 1 1
1 1 1 1
1 1
1 1
1 1
1 1
1 1
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. Figure 6-1.
Surveillance Capsule Geometry 6-16
17468-14 0
0 0
45 l
/
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79 PRESSURE VESSEL CAPSULE 125
////////////,
/
THERMAL SHIELD
/'
REACTOR s / / / /f
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CORE
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Figure 6-2.
Big Rock Point Reactor Geometry e
'6-17
4 --
.g.,.
REFERENCES 1.
Brandt, F.A., " Reactor Pressure Vessel Material Surveillance Program at the Consumers Power Company Big Rock Point Nuclear Plant," GECR 4442, December 1963.
l l
2.
Serpan, C. Z., and Watson, H. E., " Mechanical Property and Neutron Spectral Analyses of the Big Rock Point Reactor Pressure Vessal," Nuclear Engineering and Design, Volume ll, No. 3, April 1970.
3.
ASTM Designation E261-70, Standard Method for Measuring Neutron Flux by Radioactivation Techniques," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 745-755, Am. Society for Testing and Materials, Philadelphia, Pa.,1975.
4.
ASTM Designation E-263 70, " Standard Method for Measuring Fast-Neutron Flux by Radioactivation of iron," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 764-769, Am. Society for Testing and Materials, Philadelphia, Pa.,1975.
5.
ASTM Designation E264 70, " Star;dard Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 770-774, Am. Society for Testing and Materials, Philadelphia, Pa.,1975.
6.
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-Dimension Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.
7.
Collier, G., et al, "Second Version of the GAMBIT Code," WANL TME-1969, November,1969.
I 8.
Soltesz, R. G., et al, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation - Volume 3, Cross-Section Generation and Data Processing Techniques," WANL-PR-(LL)-034, August,1970.
1 9.
Soltesz, R. G., et al, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation - Vo!ume 4 - One-Dimensional Discrete Ordinates Transport Technique," WANL-PR-(LL) 034, August,1970.
l t A-1
-