ML20076E875
| ML20076E875 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 03/31/1983 |
| From: | James Anderson, Shogan R, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML14184A472 | List: |
| References | |
| WCAP-10304, NUDOCS 8308250076 | |
| Download: ML20076E875 (64) | |
Text
.
ANALYSIS OF CAPSULE T FROM THE '
H. B. ROBINSON UNIT 2 REACTOR VESSEL HADIATION SURVEILLANCE PROGRAM (WCAP-10304)
EPRI RESEARCH PROJECT 1021-3 TOPICAL REPORT March 1983 Prepared by WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division' P. O. Box 3g5 Pittsburgh, Pennsylvania 15230 t
T. R. Mager, Principalinvestigator Prepared for ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue Palo Alto, California 94304 T. U. Marston, Project Manager 8308250076 830819 PDR ADOCK 05000261 P PDR
ANALYSIS OF CAPSULE T FROM THE H. B. ROBINSON UNIT 2 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM (WCAP-1030*)
EPRI RESEARCH PROJECT 1021-3 TOPICAL REPORT S. E. Yanichko S. L. Anderson R. P. Shogan R. G. Lott March 1983 Prepared by WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division
.". O. Box 355 Pittsburgh, Pennsylvania 15230 T. R. Mager, PrincipalInvestigator Prepared for ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue Palo Alto, California 94304 T. U. Marston, Project Manager
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LEG AL NOTICE i i I
l 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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l TABLE OF CONTENTS l
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 T 5-1 5-1. Overview 5-1 5-2. Charpy V-Notch Impact Test Results 5-2 5-3. Tension Test Results 5-17
, 5-4. Wedge Opening Loading Tests 5-18 1
6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1. Introduction 6-1 6-2. Discrete Ordinates Analysis 6-1 6-3. Neutron Dosimetry 6-5 6-4. Transport Analysis Results 6-8 6-5. Dosimetry Results 6-15 References A-1 iii
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LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the H. B. Robinson Unit 2 Reactor Vessel 4-3 l
4-2 Arrangement of Specimens, Thermal Monitors, and Dosimeters in Capsule T 4-4 5-1 Charpy V-Notch Impact Data for H. B. Robinson Unit 2 Reactor Vessel Shell Plate W10201-6 5-5 5-2 Charpy V-Notch impact Data for H. B. RoDinson 4
Unit 2 Reactor Vessel Weld Metal 5-6 5-3 Charpy V-Notch Impact Data for H. B. Robinson Unit 2 Reactor Vessel Weld HAZ Material 5-7 5-4 Charpy V-Notch impact Data for H. B. f obinson Unit 2 ASTM A3028 Correlation Monitor Material l (Transverse Orientation) 5-8 5-5 Charpy impact Specimen Fracture Surfaces for H. B. Robinson Unit 2 Reactor Vessel Shell Plate W10201 -6 5-13 5-6 Charpy impact Specimen Fracture Surfaces for H. B. Robinson Unit 2 Reactor Vessel Weld Metal and HAZ Metal 5-14 5-7 Charpy impact Specimen Fracture Surfaces for H. B. Robinson Unit 2 ASTM A302B Correlation Monitor Material 5-15 L
5-8 Comparison of Actual Versus Predicted 41-Joule Transition Temperature increase for H. B. Robinson Unit 2 Reactor Vessel Materials Using the Prediction Methods of Regulatory Guide 1.99 Revision 1 5-17 5-9 Tensile Properties for H. B. Robinson Unit 2 Reactor Vessel Shell Plate W10201-6 5-20 5-10 Tensile Properties for H. B. Robinson Unit 2 Reactor Vessel Weld Metal 5-21 5-11 Fractured Tension Specimens from H. B. Robinson I Unit 2 Shell Plate W10201-6 and Weld Metal 5-22 v
J LIST OF ILLUST RATIONS (cont)
Figure Title Page 5-12 Typical Stress-Strain Curve for Tension Specimens 5-23 6-1 R Theta Reactor Geometry 6-2 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-3
^
6-3 Calculated AzimuthalDistribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel - Surveillance Capsule Geometry 6-9 6-4 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel 6-10 6-5 Relative AxialVariation of Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel 6-11 6-6 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Surveillance Capsules 6-12 6-7 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule T 6-13 vi
- -. . _ _ _ - - _. . . . , . . . __ _ _ = -.
i LIST OF TABLES Table Title Page s
4-1 Chemical Composition and Heat Treatment of Material for the H. B. Robinson Unit 2 Reactor Vessel Surveillance Program 4-5 i
5-1 Charpy impact Data for H. 8 Robinson Unit 2 Reactor Vessel Shell Plate W10201-6 tirradiated at 4.11 x 1019 n/cm2) --
5-3 5-2 Charpy impact Data for H. B. Ro'binson Unit 2 Reactor Vessel Weld Metal and H4Z-Material (Irradiated at 4.11 x 1019 n/cm2) 5-3 5-3 Charpy impact Data for the, ASTM, A3028 Correlation Monitor Material (irradiat ad at 4.11 x '
10 39 n/cm2) 1 s
- - 5-4
~
54 Instrubented Charpy impact Test Resuits for H. B. Robinson Unit 2 Reactor \iccsel ~
Shell Plate W10201-6 < 5-0 l( ,
5-5 Instrumented Charpy Impact Test Results for H. B. Robinson Unit 2 Weld Metal and HAZ
" '~
Materini , x 5-10 5-6 Instrumented Charpy impact Test Results ;o
< for ASTM '
A3028 Correlation Monitor Materia! 5-11
. s ,
The Effect of'288 #C frradiation at 4.i 1 x 1019 n/cm2
. 5-7 (E > 1.0 MeV) on the Notch Toughness Propeities of
- H. B. Robinson Unit.2 ReactorVessel Materials 5-12
'5-8 Summary of H.O. Robinson Unit 2 Reactor Oe'ssel Surveillanco Capsule Charpy impact Tet! Results 5'-16 5-9 ' Tensile Properties'for H. B. Robinson Unit 2 Reactor Vessel Materials !rradiated to
~~
4.11 x 1019 n/cm2.
~
5'-19 6-1 21 Group Energy Structure .- 6-4 6-2 Nuclear Parameters for Neutron Flux Monitors 6-6 s
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4
, - , . - s-#_
LIST OF TABLES (cont)
Table Title Page 6-3 Calculated Fast Neutron Flux (E > 10 MeV) and Lead Fictors fcr H. B. Robinson Unit 2 Surveillance Capsules 6 6-4 Calculated Neutron Energy Spect:a at the Center of the H. B. Robinson Unit 2 Serveillance Capsules 6-14 6-5 Spectrum Averaged Reaction Cross Sections at the Dosimeter Block Location for H. B. Robinson Unit 2 Suveillance Capsules 6-15 6-6 Irradiation History of Capsule T 6-16 6-7 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule T 6-20 6-8 Results of Fast Neutron Dosimetry for Capsule T 6-21 6-9 Results of Thermal Nautron Dosimetry for Capsule T 6-21 6-10 Summary of Neutrco Dosimetry Results for Capsule T 6-22 0-Vii
1 SECTION 1
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in Capsule T, the third reactor vessel material surveillance capsule which was removed from the Carolina Power and Light Company H. B. Robin-son Unit 2 reactor pressure vessel after approximately 7.2 effective full power years of operation, led to the following conclusions:
m The surveillance capsule received an average fast neutron fluence (E >2 1 MeV) of 4.11 x 1019 n/cm compared to a calculated fluence of 3.81 x 1019 n/cm .
m Irradiation of base metal plate material to 4.11 x 1019 n/cm2 resulted in a 42 C in-crease in the 41-joule transition temperature when compared with unirradiated values. The transition temperature increase, when compared with prior results after irradiation to 3.69 x 1018 n/cm2, indicates that the increased neutron fluence produced an additional 34 C increase in transition temperature.
m Submerged arc weld metalirradiated to 4.11 x 10 19 n/cm2showed a 41-joule transition temperature increase of 158 C compared to a 97 C increase for prior tests irradiated at 5.84 x 1019 n/cm2, a Comparisons of the 41-joule transition temperature increases for the H. B. Robinson Unit 2 surveillance materials with Regulatory Guide 1.99 Revision 1 predictions show that the increases after irradiation to 4.11 x 1019 n/cm2 are significantly less than predicted.
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SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule T, the third capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irra-diation on the H. B. Robinson Unit 2 reactor pressure vessel materials under operating conditions.
The surveillance program for the reactor pressure vessel materials was designed and recommend-ed by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented in WCAP-7373 Ill. The surveillance program was planned to cover the 40-year life of the rea'ctor pressure vessel and was based on ASTM E185-66. Westinghouse Nuclear Technology Division 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 the third surveillance capsule (Capsule T) removed from the H. B. Robinson Unit 2 reactor vessel and discusses the analy-sis of these data. The data are also compared to results obtained from the two previous surveil-lance capsules (Capsules S and V)[2,3,4] which were tested as part of the program.
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SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important f actor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vesselis the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the me-chanical properties of low alloy ferritic pressure vessel steels such as SA302 Grade A (base mate-rial of the H. B. Robinson Unit 2 reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.
A method of performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non-ductile Failure," Appendix G to Section 111 of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature RTNDT-RT NDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the temperature 60 F less than the 50 f t Ib (or 35-millateral expansion) temperature as determined from Charpy specimens oriented normal to the major work-ing direction of the material. The RTNDT of a given materialis used to index that material to a refer-ence stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given materialis indexed to the KIR curve, al-lowable stress intensity factors can be obtained for this material as a function of temperature. Al-lowable operating limits 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 surveillance program such as the H. B. Robinson Unit 2 Reactor Vessel Radiation Surveil-lance Program,IU in which a surveillance capsule is periodically removed from the operating nucle-ar reactor and the encapsulated specimens are tested. The increase in the Charpy V-notch 30 f t Ib temperature (iRTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDTf 0f 3-1
, i W' radiation embrittlement. This adjusted RTNDT (RT NDT initial + .1RTNDT) is used to index the mate-rial to the KlR curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
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SECTION 4 DESCRIPTION OF PROGRAM Eight surveillance capsules for monitoring the effects of neutron exposure on the H. B. Robinson Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The eight capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.
Capsule T was removed after approximately 7.2 effective full power years of plant operation. This capsule, shown in figure 4-2, contained Charpy V-notch impact, tension, and IX-wedge opening loading (WOL) fracture mechanics specimens from the vesselintermediate shell plate W10201-6 and submerged arc weld metal representative of the core region of the reactor vessel and Charpy V-notch specimens from weld heat-affected zone (HAZ) material. The capsule also contained Charpy V-notch specimens from the 6-inch-thick ASTM correlation monitor material (A302 Grade B). The chemistry and heat treatment of the surveillance materials are presented in table 4-1. The chemica! analyses reported in table 4-1 were obtained from unirradiated material.used in the sur-veillance program. In addition, a chemical analysis was performed on two irradiated Charpy speci-mens from the weld metal and is also reported in table 4-1. The analysis of the irradiated weld metalindicates that the copper, nickel, and phosphorus contents are in good agreement with the unirradiated analysis.
Test specimens from the plate material were machined from the 1/4 thickness location of the plate. Test specimens represent material taken at least one plate thickness from the quenched end of the plate. All base metal Charpy V-notch specimens were oriented with the longitudinal axis of the specimen parallel to the principal rolling direction of the plate. Base metal tension specimens were oriented with thelongitudinal axis of the specimen transverse to the rolling direction of the plate. Charpy V-notch and tension specimens from the weld metal were oriented with the longi-tudinal axis of the specimens transverse to the welding direction.
4-1
Capsule T contained dosimeter wires of pure copper, nickel, and aluminum-0.15 Wt% cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Np237 and
.U230 were contained in the capsule. The test specimens served as iron dosimeters. Thermal moni-tors made from two low melting eutectic alloys sealed in quartz tubes were included in the capsule and were located as shown in figure 4-2. The two eutectic alloys and their melting points are as ;
follows:
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 1
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TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF MATERIAL FOR THE H. B. ROBINSON UNIT 2 REACTOR VESSEL SURVEILLANCE PROGRAM Chemical Composition (Wt%)
Plate Plate Plate Weld Weld Metal la] Correlation Element W10201-4 W10201-5 W10201-6 Metal W21 W20 Monitor C 0.19 0.20 0.19 0.16 0.075 -
0.24 Mn 1.35 1.29 1.32 0.98 1.36 -
1.34 P 0.007 0.010 0.010 0.021 0.026 -
0.011 S 0.019 0.021 0.015 0.014 0.033 -
0.023 Si 0.23 0.22 0.19 0.34 0.21 -
0.23 Mo 0.48 0.46 0.49 0.46 0.24 -
0.51 Cu 0.12 0.10 0.09 0.34 0.33 0.35 0.20 V - - -
0.001 - - -
Ni - - -
0.66 0.63 0.69 0.18 Cr - - -
0.024 0.02 -
0.11 Co - - - -
0.007 - -
Heat Treatment Plate W10201-4 1550 -1600 F,4 hr, water quench I
Plate W10201-5 > 1200 -1250 F,4 hr, air cooled l
1125 -1175 F,151/2 hr, furnace cooled to 600 F Plate W10201-6[
Weld Metal 1125 -1175 F,30 hr, furnace cooled to 600 F Correlation Monitor 1650*F - 4 hr, water quenched 1200 F - 6 hr, air cooled
- a. Analysis performed on irradiated Charpy specimens from Capsule T.
4-5
SECTION 5 TESTING OF SPECIMENS FROM CAPSULE T 5-1. OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tension specimens was per-formed at the Westinghouse Research and Development Laboratory with consultation by Westing-house Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendixes G and H.
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-7373 Ill. 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 on a Tinius-Olsen Model 74,358-joule machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology Model 500 instru-mentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve, the load of general yielding (PGY), the time to general yielding (tGY), the maximum load (PM), and the time to maxi-mum load (tM) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pp), and the load at which fast fracture terminated is identified as the arrest load (PAI-The energy at maximum load (EM) was 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 initi-ate a crack in the specimen. Therefore, the propagation energy for the crack (Ep) is the difference between the total energy to fracture (ED) and the energy at maximum load.
The yield stress kry) is calculated from the three-point bend formula. The flow stress is calculated from the average of the yield and maximum loads, also using the three-point bend formula.
5-1
Percent shear was determined from postfracture photographs using the ratio-of-areas method in compliance with ASTM Specification A370-74. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
Tension tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8 and E21, and MHL Procedure 7604 Revision 2. Al! pull rods, grips, and pins were made of Inconel 718 hardened to Rc 45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test.
Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83.
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air.
Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the follow-ing procedure was used to monitor specimen temperature. Chromel-alumel thermocouples were in-serted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. in test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range room tem-perature to 550 F. The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures. Experiments in-dicated that this method is accurate to plus or minus 2*F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were deter-mined directly frorn 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 i 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 l
diameter measurement.
5-2. CH ARPY V-NOTCH IMPACT TEST RESULTS The toughness results from Charpy V-notch impact tests performed on the various surveillance materials in Capsule T af ter irradiation to 4.11 x 1019 n/cm2 are presented in tables 5-1 through 5-3 and figures 5-1 through 5-4. Instrumented CMrpy impact test results for the various materials are shown in tables 5-4 through 5-6. A summary of the surveillance test results is presented in table 5-7. The fractured surfaces of the impact specimens are shown in figures 5-5 through 5-7.
5-2
TABLE 5-1 CHARPY IMPACT DATA FOR H. B. ROBINSON UNIT 2 REACTOR VESSEL SHELL PLATE W10201-6 (IRRADIATED AT 4.11 x 1019 n/cm2)
Sample Temperature impact Energy Lateral Expansion Shear No. *C F J ft Ib mm mils (%)
L48 26 78 30.0 22.0 0.55 21.5 33 L41 38 100 46.0 34.0 0.72 28.5 40 L46 66 150 65.0 48.0 1.09 43.0 60 L43 79 175 106.5 78.5 1.63 64 0 80 L47 93 200 116.0 85.5 1.99 78.5 92 L45 121 250 148.0 109.0 1.99 78.5 100 L44 177 350 139.0 102.5 1.88 74.0 100 TABLE 5-2 CHARPY IMPACT DATA FOR H. B. ROBINSON UNIT 2 REACTOR VESSEL WELD METAL AND HAZ MATERIAL (IRRADIATED AT 4.11 x 1019 n/cm2)
Sample Temperature impact Energy Lateral Expansion Shear No. *C l *F J l ft Ib mm l mils (%)
Weld Metal W17 79 175 19.0 14.0 0.24 9.5 15 W22 93 200 32.0 23.5 0.46 18.0 35 W18 93 200 23.0 17.0 0.38 15.0 15 W20 107 225 87.0 64.0 1.22 48.0 92 W19 121 250 52.0 38.5 0.75 29.5 40 W23 135 275 70.0 51.5 1.00 39.5 90 W24 149 300 82.0 60.5 1.26 49 5 98 HAZ Metal H17 93 200 117.5 86.5 1.61 63.5 98 H24 149 300 132.0 97.5 1.96 77.0 100 Irradiation of Charpy V-notch specimens from the intermediate shell course plate W10201-6 to 4.11 x 1019 n/cm2 resulted in a 41-joule (30 f t 15) and 68-joule (50 f t Ib) transition temperature increase of 42 C (75"F) and 44 C (80 F), respectively, when compared with unirradiated values as shown in figure 5-1. The upper shelf energy of plate W10201-6 decreased by 12 joules (9 f t Ib) 53
TABLE 5-3 CHARPY IMPACT DATA FOR THE ASTM A302B CORRELATION MONITOR MATERIAL (IRRADIATED AT 4.11 x 1019n/cm2)
Sample Temperature impact Energy Lateral Expansion Shear No. *C *F J f t Ib mm mils (%)
R63 26 78 37.5 27.5 0.56 22.0 15 R59 26 78 17.0 12.5 0.25 10.0 10 R62 66 150 20.5 15.0 0.28 11.0 35 R64 93 200 32.5 24.0 0.53 21.0 45 R58 107 225 46.0 34.0 0.81 32.0 96 R60 121 250 49.5 36.5 0 88 34.5 99 R61 149 300 46.0 34 0 0.81 32.0 100 R57 177 350 55.0 40.5 0.93 36.5 100 af ter irradiation to 4.11 x 1019 nicm 2. A comparison of these results with prior results af ter irradi-ation to 3.69 x 10 18 n/cm 2, shown in table 5-8, indicates that the increased neutron fluence pro-duced an additional 30' to 34*C increase in transition temperature and an additional redaction in the upper shelf energy level of 4 joules.
Weld metal specimens irradiated to 4.11 x 1019 n/cm2 as shown in figure 5-2 resulted in a 41-and 68- joule transition temperature increase of 158 C. A comparison of these results with prior test results on the weld metalirradiated to 5.84 x 1018 n/cm2 is shown in table 5-8. This compari-son shows that the 41- and 68-joule transition temperature increases for the 4.11 x 1019 n/cm2 irradiation were 61 C and 39 C higher, respectively, than the results for irradiation at 5.84 x 10 18 n/cm 2. Two Charpy impact tests were performed on the weld HAZ metal specimens as shown in figure 5-3 in the vicinity of the upper shelf to confirm that the HAZ metal exhibits an upper shelf energy in excess of 68 joules. The remaining HAZ specimens are being held in abeyance at the re-quest of the Carolina Power and Light Company. Reconstituting the weld portion of the unbroken specimens and performing additional Charpy tests on the weld metal are currently under consideration.
ASTM A302B correlation monitor materialirradiated to 4.11 x 1019 n/cm2 showed a 41-joule transition temperature increase of 84 C when compared with unirradiated results as shown in figure 5-4. Specimens from this material were oriented transverse to the major rolling direction of the plate and exhibited upper shelf energy levels less than 68 joules in both the unirradiated and ir-radiated conditions. A 4-joule decrease in sheif energy resulted from irradiation to 4.11 x 10 19 n/cm2. A comparison of the 41-joule transition temperature increase with prior results from irradi-ations to 3.69 x 1018 and 5.84 x 1018 n/cm2 as shown in table 5-8 indicates that irradiation to 4.11 x 1019 n/cm2 produced an additionalincrease in transition temperature of 45'C.
5-4
TEMPERATURE (OC)
-50 0 50 100 150 200 250 l l l 3 3 100 -
O O O e
O g 80 -
9 E O l
< 60 -
m 40 -
2 0 20 -
0 100 2.5 g 80 -
g -
2.0 60 - O -
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nL$ m og 1.0 47 C 0.5 20 - -
0 _ 0.0 140 120 -
160 0
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UNIRR ADI ATE O g
[ -
120 80 -
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IRRADI ATED AT 288 C- 80
$ o O 4.11 x 1019 n/cm2
- 750F -
40 20 -
O 420C 0 - 0
-100 0 100 200 300 400 500 TEMPERATURE (OF)
Figure 5-1. Charpy V-Notch impact Data for H.B. Robinson Unit 2 Reactor Vessel Shell Plate W10201-6 5-5
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Figure 5-2. Charpy V-Notch impact Data for H.B. Robinson Unit 2 Reactor Vessel Weld Metal 5-6
TEMPER ATURE (OC)
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.s TEMPERATURE (OF)
Figure 5-3. Charpy V-Notch Impact Data for H.B. Robinson Unit 2 Reactor Vessel Weld HAZ Material 5-7
l l
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Figure 54. Charpy V. Notch Impact Data for H.B. Robinson Unit 2 ASTM A302B Correlation Monitor Material (Transverse Orientation) 5-8
l i
TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR H. B. ROBINSON UNIT 2 REACTOR VESSEL SHELL PLATE W10201-6 Normalized Energies Test Charpy Charpy Ma ximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress No. ("Cl (J) (kJ/m2) (kJ/m2) (kJ/m2) (N) tus) (N) (gs) (N) (N) (MPa) (MPa)
Y3
- L48 26 30 0 373 L41 38 46 0 576 350 226 12300 100 16000 450 16000 5400 630 728 L46 66 65 0 813 530 284 11700 95 16000 670 16000 7500 604 713 L43 79 106 5 1330 468 862 11600 125 15800 605 15700 12500 595 703 L47 93 116 0 1449 532 917 12600 165 16300 660 650 745 L45 121 148 0 1847 667 1180 11400 100 16800 810 585 725 L44 177 139 0 1737 603 1134 9100 100 14200 860 466 587
TABLE 5-5 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR H. B. ROBINSON UNIT 2 WELD METAL AND HAZ MATERIAL Normalized Energies l l
Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Ma ximum Load Load Stress Stress No. (*C) (J) (kJ/m2) (NJ/m2) (kJ/m2) (N) ( s) (N) ( s) (N) (N) (MPa) (MPa)
Weld Metal W17 79 19 0 237 153 84 15200 105 17000 205 17300 100 781 828 W22 93 32 0 398 308 90 14700 100 17900 360 17900 0 757 840 O W18 93 23 0 288 216 72 15200 105 17900 265 17900 100 781 850 W20 107 87 0 1085 448 636 14200 105 18500 500 732 841 W19 121 52 0 652 398 254 15000 105 18600 440 17700 3000 770 863 W23 135 70 0 873 387 486 15800 115 18800 430 813 891 W24 149 82 0 1025 356 669 13900 105 18100 415 717 824 HA2 Metal H17 93 117 5 1466 542 924 14800 105 18700 590 763 863 H24 149 132 0 1652 531 1121 13100 125 17900 605 676 799 l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - - --.
TABLE 5-6 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR ASTM A302B CORRELATION MONITOR MATERIAL I
Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Lead Maximum Load Load Stress Stress N o. (*C) (J) (kJ/m2) (kJ/m2) (kJ/m2) (N) (gs) (N) (p s) (N) (N) (MPa) (MPa)
U1 L H59 26 170 212 174 38 15500 110 16600 230 797 825 R63 26 375 466 223 243 16200 100 17800 260 17800 0 833 875 R62 66 20 5 254 157 97 15000 100 16600 210 16600 1700 770 811 R64 93 32 5 407 227 179 14700 125 17200 280 16900 6000 758 822 R58 107 46 0 576 206 370 13300 100 16100 275 684 755 R60 121 49 5 619 221 398 14400 105 17200 275 739 813 R61 149 46 0 576 198 378 13700 105 16600 265 705 780 R57 177 55 0 686 192 494 12800 105 16200 265 661 748 l
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THE EFFECT OF 2880C IRRADIATION AT 4.11 x 1019 n/cm2 (E > 1.0 MeV) ON THE NOTCH ,
t i TOUGHNESS PROPERTIES OF H. B. ROBINSON UNIT 2 REACTOR VESSEL MATERIALS i 1
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5-15
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TABLE 5-8
SUMMARY
OF H. B. ROBINSON UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE CH ARPY IMPACT TEST RESULTS 68 J 41 J 1 1
50 ft Ib 30 ft Ib Decrease in '
Trans. Temp Trans. Temp Upper Shelf Fluence Increase increase - Energy Material (1019 n/cm2) (*C) (*F) (*C) (* F) (J) (ft Ib)
Plate W10201 -4 0.369 25 45 17 30 11 8 Plate W10201 -5 0.369 22 40 11 20 18 13 Plate W10201-5 0.584 28 50 25 45 0 0 Plate W10201 -6 0.369 14 25 8 15 8 6 Plate W10201 -6 4.11 44 80 42 75 12 9 Weld Metal 0.584 119 215 97 175 57 42 Weld Metal 4.11 158 285 158 285 - -
HAZ Me'tal 0.584 47 85 36 65 47 35 HAZ Metal 4.11 - - - -
46 34 Correlation Monitor O.369 - -
39 70 4 3 Correlation Monitor 0.584 - -
39 70 4 3 Correlation Monitor 4.11 - -
84 150 4 3 The fracture appearance of each irradiated Charpy specimen from the various materials is shown in figures 5-5 through 5-7. An increasing ductile or tougher appearance with increasing tempera-ture can be noted for cach of the materials.
Figure 5-8 shows a comparison of the 41-joule transition temperature increase for the H. B. Robin-son Unit 2 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99 Revision 1. This comparison shows that all the materials exhibited 41-joule transition temperature increases lower than would be predicted by the Guide. This behavior is consistent with the results from other reactor vessel surveillance capsules evaluated as part of this EPRI program. Based on the surveillance capsule test results to date,it can be concluded that the H. B.
Robinson Unit 2 surveillance materials are not as sensitive to radiation as predicted by the Guide.
5-16
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Figure 5-8. Comparison of Actual Versus Predicted 41-Joule Transition Temperature increase for H.B. Robinson Unit 2 Reactor Vessel Materials Using the Prediction Methods of Regulatory Guide 1.99 Revision 1 5-3. TENSION TEST RESULTS The results of tension tests performed on shell plate W10201-6 and weld metal are shown in table 5-9 and figures 5-9 and 5-10, respectively. An increase in 0.2 percent yield strength of approxi-mately 12 and 27 ksi (82.7 and 186.2 MPa) was exhibited by the plate and weld metal, respectively. These increases ;ndicate that the weld metalis more sensitive to radiation than the base metal plate which is consistent with transition temperature increases observed for these materials. Photographs of the fractured tension specimens for each material are shown in figure 5-11. A typical stress-strain curve for the tension specimens is sho en in figure 5-12.
5-17
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5-4. WEDGE OPENING LOADING TESTS Wedge' opening loading (WOL) fracture mechanics specimens which were contained in the surveil-lance capsule have been stored at the Westinghouse Research Laboratory; they will be tested and reported on later.
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TABLE 5-9 TENSILE PROPERTIES FOR H. B. ROBINSON UNIT 2 REACTOR VESSEL MATERIALS 1RRADIATED TO 4.11 x 1019 n/cm2 Test 0.2% Yield Ultimate Fracture Fracture Fracture Uneform Sample Total Reductuwe Temp Strength Strength Load Stress Strength Elongation Elongation in Area Material No. ' C ('F) MPa (k si) M Pa (ksi) N (kip) MPa (kso) MPa (kss) (*4) (%) (%)
L7 Plate 93 389 527 12.800 980 405 13 8 23 7 59 P W10201-6 (200) (564) (76 el (2 88) (141 5) (58 76 L6 Plate 288 337 530 13.900 850 438 13 2 21 5 49 i
W10201 6 (550) (48 98 (76 8) (3 12) (124 0) (63 6)
W5 Weld 135 611 632 15.800 1370 499 87 178 64 Metal (275) (88 6) '(100 4) (3 55) (1982) (72 31 W6 Weld 288 597 702 18.700 1290 590 84 16 2 54 Met 4' (550) 186 6p (101 9) (4 20) (1872) (85 6) w
TEMPERATURE (OC) 0 50 100 150 200 250 300 350 110 l l l l l l l l
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Figure 5-9. Tensile Properties for H.B. Robinson Unit 2 Reactor Vessel Shell Plate W102016 5-20
TEMPERATURE (OC) 0 50 100 150 200 250 300 350 I I I I I I I I 120 -
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4676A 21 11%sg M i l l l l l 1 l l 1 2 3 4 6 7 8 9 0 5 1 TENTHS OF AN INCH SPECIMEN L-7 TESTED AT 2000 F 0
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1 TENTHS OF AN INCH SPECIMEN L 6 TESTED AT 5500F bb" Q h s h g; I I I I 1 I I I 1 2 3 4 l 6 7 8 9 0 5 1 TENTHS OF AN INCH SPECIMEN W-5 TESTED AT 2750F I I I I l l 1 I i 1 2 3 4 6 7 8 9 0 5 1 TENTHS OF AN INCH SPECIMEN W-6 TESTED AT 5500 F Figure 5-11. Fractured Tension Specimens from H.B. Robinson Unit 2 Shell Plate W10201-6 and Weld Metal 5 22
120 800 100 -
600 80 - -
=. %
5 a.
60 - - E.
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200 20 -
SPECIMEN W5 0
0 0 0.025 0.05 0.075 0.10 0.125 0.15 0.175 0.20 0.225 0.25 STRAIN (in./in.)
Figure 5-12. Typical Stress - Strain Curve for Tension Specimens
l SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY l
6-1. INTRODUCTION Knowledge of the neutron environment within the pressure vessel- surveillance capsule geometry is required as an integral part of LWFt pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known.
Second, in relating the changes observed in the test specimens to the present and future condition of the reactor 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 re-quirement is normally met by employing a combination of rigorous analytical techniques and mea-surements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis.
Tnis section describes a discrete ordinates Sn transport analysis performed for the H. B. Robinson Unit 2 reactor to determine the fast neutron (E > 1.0 MeV) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules and, in turn, to develop lead fac-tors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules.
Based on tpectrum-everaged reaction cross sections derived from this calculation, the anaiysis of the neutron dosimetry contained in Caosute l'is discussed.
7 6-2. DISCRETE ORDINATES ANALYSIS A plan view of the H. B. Robinson Unit 2 reactor gecmetry at the core midplane is shown in figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a 45-degree sector is depicted. Eight ir-radiaton capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance program. The location of each of the eight surveillance capsules relative to the reactor core cardinal axes is also depicted in figure 6-1.
A plan view of a single surveillance capsule attached to the thermal shield is shown in figure 6-2.
The stainless steel specimen container is 1-inch square and approximately 63 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5.25 feet of the 12-foot-high reactor core. "
6-1
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j Figure 6-1. R, Theta Reactor Geometry i
j From a neutronic standpoint, the surveillance capsule structures are significant. In fact, as is shown later, they have a marked impact on the distributions of neutron flux and energy spectra in the water annulus between the thermal shield and the reactor vessel. Thus, in order to properly ascer-tain 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. l 6-2
(00 ,100 ,200 ,30 0,40 0)
CHARPY SPECIMEN fHHHH, l _
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/l4 THERMAL SHIELD Figure 6-2. Plan View of a Reactor Vessel Surveillance Capsule in the analysis of the neutron environment within the H. B. Robinson Unit 2 reactor geometry, pre-dictions of neutron flux magnitude and energy spectra were made with the DOT [5] two-dimensional discrete ordinates code. The radial and azimuthal distributions were obtained from an R, & computation wherein the geometry shown in figures 6-1 and 6-2 was described in the analyti-cal model. In addition to the R, & computation, a second calculation in R, Z geometry was also car-ried out to obtain relative axial variations of neutron flux throughout the geometry of interest. In the R, Z analysis the reactor core was treat >3d as an equivalent volume cylinder and, of course, the sur-veillance capsules were not included in the model.
Both the R, O and the R,7 analyses employed 21 neutron energy groups, an S8 angular quadrature.
and a P1 cross-section expansion. The cross sections were generated via the Westinghouse GAMB1T[6] code system with broad group processing by the APPROPOS[7} and ANISN{8] codes.
The energy group structure used in the analysis is listed in table 6-l.
S A key input parameter in the analysis of the integrated fast neutron exposure of the reactor vessel is the core power distribution. For this analysis, power distributions representative of time-averaged conditions derived from statistical studies of long-term operation of Westinghouse 3-loop plants were employed. Tnese input distributions include rod-by-rod spatial variations for all peripheral fuel assemblies.
It should be noted that this particular power distribution is intended to produce accurate end-of-life neutron exposure levels for the pressure vessel. As such, the calculation is indeed representative of an average neutron flux and small (plus or minus 15 to 20 percent) deviations from cycle to cycle are to be expected. -
6-3
l l
l TABLE 6-1 21 GROUP ENERGY STRUCTURE Group Lower Energy (MeV) 1 7.79[a]
2 6.07 3 4.72 4 3.68 5 2.87 6 2.23 7 1.74 8 1.35 9 1.05 10 0.821 11 0.388 j 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 l 19 1.86 x 10-6 !
20 3.00 x 10-7 21 0.0 i
- a. Upper energy of group 1 is 10.0 MeV l
6-4
-- - 2 l \
t 1 Having the results of the R,9 and R,Z calculations, three-dimensional variations of neutron flux may )
l l be approximated by assuming that the following relation holds for the applicable regions of the j reactor, d(R,Z,0,Eg) = 4(R,9,Eg) F(Z,Eg) (6-1) l where 4(R,Z,0,Eg) = neutron flux at point R,Z,0 within energy group g 6(R,9,Eg) = neutron flux at point R,0 within energy group g obtained from the R,9 calculation 6-3. NEUTRON DOSIMETRY The passive neutron flux monitors included in the H. B. Robinson Unit 2 surveillance program are listed in table 6-2. The first five reactions in table 6-2 are used as fast neutron monitors to relate neutron fluence (E > 1.0 MeV) to measured materials properties 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 c )balt-aluminum monitors were also included.
The relative locations of the various monitors within the surveillance capsules are shown in figure 4-2. The nickel, copper, and cobalt-aluminum monitors, in wire form, ere placed in holes drilled m spacers at several axiallevels within the capsules. The iron monitors are obtained by drilling sam-f ples from selected Charpy test specimens. The cadmium-shielded neptunium 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 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 6-5
I 1 1
1 TABLE 6-2 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Monitor Material ' " "
Reaction of Interest eight i'ld R nge H f ife Fraction (%)
Copper Cu63 (n,a) Co60 0.6917 E > 4.7 MeV 5.27 yr fron Fe54(n.p)Mn54 0.0585 E > 1.0 MeV 314 d Nickel NiS8(n.p)Co68 0.6777 E > 1.0 MeV 71.4 d Uranium-238[a] U238(n.f)Cs1 37 1.0 E > 0.4 MeV 30.2 yr 6.3 Neptunium-237[a] Np237(n,f)Cs1 37 1.0 E > 0.08 MeV 30.2 yr 6.5 Cobalt-Aluminum [a} CoS9(n,Y)Co60 0.0015 0.4 eV < 0.015 MeV 5.27 yr Cobalt-Aluminum Co69(n,Y)Co60 0.0015 E < 0.0015 MeV 5.27 yr
- a. Denotes that monitor is cadmium shielded neutron flux levelincident on the various monitors may be derived from the activation measure-ments only if the irradiation parameters are well known. In particular, the following variables are of interest:
a The operating history of the reactor a The energy response of the monitor a The neutron energy spectrum at the monitor location a The physical characteristics of the monitor The analysis of the passive monitors and subsequent derivation of the average neutron flux re-quires 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 I cross section, the neutron energy spectrum at the monitor location must be calculated. i
. i The specific activity of cach of the monitors is determined using established ASTM i procedures.[9,10,11,12,13] Following sample preparation, the activity of each monitor is deter-mined by means of a lithium-drifted germanium, Gedli), gamma spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed from H. B.
Robinson Unit 2, the overall 20 deviation in the measured data is determmed to be plus or minus 10 percent. The neutron energy spectra 6, e determined analytically using the method described in paragraph 6-1.
6-6
I Hsving tho me:sur:d activity of tha monitors and the neutron energy spectra st the locations of interest, the calculation of the neutron flux proceeds as follows.
The reactor product activity in the monit r is expressed as 1
' i NO N P 1 i R = 7 f; y o(E) p (E) dE [ p (1 - e -At-) 1 d e-At !
max (6-2)
E j=1 where R = induced product activity No = Avogadro's number A = atomic weight of the target isotope fi = weight fraction of the target isotope in the target material y = number of product atoms produced per reaction e(E) = energy-dependent reaction cross section 4(E) = energy-dependent neutron flux at the monitor location with the reactor at full power Pj = average core power level during irradiation period j P max = maximum or reference core power level i
A = decay constant of the product isotcpe tj = length of irradiation period j
= decay time following irradiation period j td l
l Because neutron flux distributions are calculated using multigroup transport methods and, ft.rther, because the prime interest is in the fast neutron flux above 1.0 MeV, spectrum-avoraged reaction cross sections are defined such that the integral term in equation (6-2) is repiaced by the following relation.
)
a(E) p(E) dE = a p (E > 1.0 MeV)
E where N
o(E) c(E) dE 89 4g
-,o ,g=1 8
m f p(E) dE 0g 1.0 MeV 9 91.0 MeV 6-7
~ . , . . - . ,
Thus, equation (6-2) is rewritten No P
_ i R = 7 f; y a 6 (E > 1.0 MeV) p (1 - e-At-)1 e-Ard j= max or, solving for the neutron flux, i
R (E > 1.0 MeV) = N (6-3) l f; y o (1 - e 5) e- td A j,3 Pmax The total fluence above 1.0 MeV is then given by N P-
1.0 MeV) = 4 (E > 1.0 max MeV) tj [ P i= 1 where N
[P I j"1 max tj = total effective full power seconds of reactor operation up to the time of capsule removal An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered CoS9(n,y)Co60 data by means of cadmium ratios and the use of a 37-barn 2,200 m/sec cross section. Thus, 1
D-11 Rbare D[
'hTh = N (6-4)
Ng P -At j j} * -Atd yf y er P (1
max j= 1 where D is defined as Rb are/ rcd covered-f 6-4. TRANSPORT ANALYSIS RESULTS Results of the Sn transport calculations for the H. B. Robinson Unit 2 reactor are summarized in fig-ures 6-3 through 6-7 and in tables 6-3 through 6-5. In figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vesselinner radius,1/4 thickness l location, and 3/4 thickness location are presented as a function of aximuthal angle. The influence of the surveillance capsules on the fast neutron flux distribution is clearly evident. In figures 6-4, the radial distribution of maximum fast neutron flux (E > 1.0 MeV) through the thickness of the 6-8
1012 8 -
6 -
4 -
2 -
10 11 -
8 6
{
4 4 -
N
$ 2 - SURVEILLANCE CAPSULES
}
X 3 10 10 m
-- PRESSURE 8 - VESSEL IR z -
o 6 c: - 1/4 T
$ 4 - LOCATION z -
2 -
109 --
3/4 T 8 -- LOCATION 9
6 -
4 -
2 -
l l l l l l 10 8 0 10 20 30 40 50 60 70 AZlMUTHAL ANGLE (DEG)
Figure 6-3. Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E < 1.0 MeV) Within the Pressure Vessel -
Surveillance Capsule Geometry 69 l
1012 9 -
8 -
7 -
6 -
5 -
4 -
3 -
2 -
10 7 397,49
^ 8 -
g 7 -
6 -
E 5 -
203.39 u
3 4 IR X
] 3 -
I 2
0 1/4 T 209.30 2 -
m 5
z 10 10 i
1/2 T 9 - 215.20 8 -
7 -
)
- l 5 -
l 4 - 3/4 T 3 -
221.10 $
l 2 -
I g l l \ l l l ! ! !
196 198 200 202 204 206 208 210 212 214 216 218 220 222 224 R ADIUS (cm)
Figure 6-4. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV)
Within the Pressure Vessel 6-10
10 0 _
8 -
6 -
4 -
2 -
10-1 --
x 8 -
S
- u. 6 -
2 -
O cc 4 -
m -
z w
> 2 -
5 w
C 10-2 _
8 -
6 -
9 4 -
CORE
, MIDPLANE
, 2 -
m TO VESSEL
' CLOSURE HEAD 10-3 l l l l
-300 -200 -100 0 100 200 300 DISTANCE FROM CORE MIDPLANE (cm)
Figure 6-5. Relative Axial Variation of Fast Neutron Flux (E > 1.0 MeV)
Within the Pressure Vessel 6-11 1
1012 9 -
8 -
7 -
6 -
5 -
4 -
3 -
2 -
CAPSULE T 11 -
10 -
9 -
CAPSULE S 8
191.69 m
6 -
E 5 -
c 4 ~ l
{ CAPSULF. V 3 3 -
N CAPSULEF
$ U+Y o 2 -
% % CAPSULES o CAPSULE X, W,
- 2 g CENTER 10 9 _~
8 -
l 7 -
l 6 -
5 -
1 4 -
3 - l 2 - THERMAL HOLDER CAN SHIELD r- CAN HOLDER j* - H O2 ---*.,' j' /:*-- TEST SPECIMENS ->; ,
' j gH 0-s , < x, , s< s 2 3 l 5 10 9 ' l l 188 189 190 191 192 193 194 R ADIUS (cm)
Figure 6-6. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV)
Within the Surveillance Capsules 6-12
10 9 8 Z 6 -
4 -
2 -
NiS8(n,p)CoS8 Np237(n,f)Cs137 6 -
4 -
191.69
[ 2 -
h U238(n,f)Cs137 y 107 --
p 54(n,p)Mn54 a 8 -
w -
> 6 -
o 4 -
V CAPSULE CENTER 2 -
106 _
~
8
) Z Cu63(n,a)Co60 6 -
4 -
-THERMAL HOLDER CAN 2 -
SHIELD CAN HOLDER
- -- H 2O + TEST SPECIMENS -> 4H 20-1 l c i l l g ; l 105 188 189 190 191 192 193 194 R ADIUS (cm)
Figure 6-7. Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule T 6-13
TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MeV) AND LEAD FACTORS FOR H. B. ROBINSON UNIT 2 SURVEILLANCE CAPSULES Azimuthal *(E > 1.0 MeV) Lead Location (deg) (n/cm2-sec) Factor 0 1.68 x 1011 2.63 10 1.20 x 1011 1.88 20 4.85 x 1010 0.76 30 3.71 x 1010 0.58 40 2.31 x 1010 0.36 TABLE 6-4 CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF THE H. B. ROBINSON UNIT 2 SURVEILLANCE CAPSULES Group Neutron Flux (n/cm2-sec)
No. O deg 10 deg 20 deg 30 deg 40 deg 1 1.24 x 109 8.66 x 108 5.04 x 108 3.73 x 108 2.73 x 108 2 4.12 x 109 2.89 x 109 1.71 x 109 1.27 x 109 9.41 x 108 3 6.79 x 109 4.71 x 109 2.49 x 109 1.86 x 109 1.30 x 109 4 7.54 x 109 5.23 x 109 2.45 x 109 1.85 x 109 1.22 x 109 5 1.20 x 1010 8.32 x 109 3.56 x 109 2.72 x 109 1.73 x 109 6 2.32 x 1010 1.62 x 1010 6.77 x 109 5.19'x 109 3.23 x 109 7 3.02 x 1010 2.15 x 1010 8.51 x 109 6.54 x 109 4.01 x 109 8 3.92 x 1010 2.83 x 1010 1.08 x 1010 8.26 x 109 4.99 x 109 9 4.38 x 1010 3.21 x 1010 1,17 x 1010 8.98 x 109 5.42 x 109 q 10 4.37 x 1010 3.21 x 1010 1.16 x 1010 8.89 x 109 5.33 x 109 l
11 1.36 x '011 1.01 x 1011 3.51 x 1010 2.68 x 1010 1.60 x 1010 l 12 1.64 x 1011 1.23 x 1011 4.14 x 1010 3.16 x 1010 1.86 x 1010
]
13 7.86 x 1010 5.86 x 1010 1.96 x 1010 1.50 x 1010 8.72 x 109 14 6.36 x 1010 4,74 x 1010 l 1.57 x 1010 1.21 x 1010 7.03 x 109 15 5.22 x 1010 3.87 x 1010 1.29 x 1010 9.92 x 109 5.75 x 109 16 1.21 x 1011 9.00 x 1010 2.96 x 1010 2.27 x 1010 1.32 x 1010 17 1.03 x 1011 7.64 x 1010 2.49 x 1010 1,91 x 1010 1,11 x 1010 18 1.06 x 1011 7.95 x 1010 2.56 x 1010 1.97 x 1010 1.13 x 1010 19 8.80 x 1010 6.51 x 1010 2.10 x 1010 1.62 x 1010 9.24 x 109 20 1.01 x 1011 7.40 x 1010 2 41 x 1010 1.85 x 1010 1.06 x 1010 21 2.75 x 1011 1.77 x 1011 6.72 x 1010 5.09 x 1010 2.93 x 1010 6-14
TABLE 6-5 SPECTRUM AVERAGED REACTION CROSS SECTIONS AT THE DOSIMETER BLOCK LOCATION FOR H. B. ROBINSON UNIT 2 SURVEILLANCE CAPSULES J(barns)
Reaction O deg 10 dog 20 dog 30 dog 40 deg Fe54(n p) Mn54 0.749 .0729 .0900 .0882 .0970 NiS8(n.p) CoS8 .100 .0979 .117 .115 .125 Cu63(n.o)Co60 .000533 .000522 .000753 .000730 .000861 U238(n,f) F.P. .365 .362 .382 .380 .389 Np23 7(n,f) F.P. 2.57 2.61 2.47 2.47 2.45 reactor pressure vesselis shown. The relative axial variation of neutron flux within the vesselis 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 valces from figure 6-5.
In figure 6-6 the radial variations of fast neutron flux within surveillance Capsule T are presented.
These data,in conjunction with the maximum vessel flux, are used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E .> 1.0 MeV) at the dosimeter block location (capsule center) to the maximum fast neutron flux at the pressure vesselinner radius. Updated lead factors for all of the H. B. Robinson Unit 2 surveillance capsules are listed in table 6-3.
Since the neutron flux monitors contained within the surveillance capsules are not alllocated at the same radiallocation, tha measured disintegration rates are analytically adjusted for the gradients that exist within the capsulee so that flux and fluence levels may be derived on a common basis at a common location. This point of comparison was chosen to be the capsule center. Analytically determined reaction rate gradients for use iri the adjustment procedures are shown in figure 6-7 2 for Capsule T. All of the applicable fast neutron reactions are included.
In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. The neutron energy spectrum calculated to exist at the center of each of the H. B. Robinson Unit 2 surveillance capsules is given in table 6-4. The associated spectrum-averaged cross sections for each of the five fast neutron reactions are given in table 6-5.
6-5. DOSIMETRY RESULTS The irradiation history of the H. B. Robinson Unit 2 reactor is given in table 6-6. Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsule T are listed in 6-15
TABLE 6-6 1RRADIATION HISTORY OF CAPSULE T Irradiation Decay Time Month Year Pj(MW) Pmax(MW) Pj/Pmax Time (Days) (Days) 7 1971 1541 2300 .670 31 4089 8 1971 1541 2300 .670 31 4058 9 1971 1593 2300 .692 30 4028 10 1971 1541 2300 .670 31 3997 11 1971 1903 2300 .827 30 3967 12 1971 2207 2300 .960 31 3936 1 1972 2044 2300 .889 31 3905 2 1972 2241 2300 .974 29 3876 3 1972 2235 2300 .972 31 3845 4 1972 2208 2300 .960 30 3815 5 1972 402 2300 .175 31 3784 6 1972 1357 2300 .590 30 3754 7 1972 1626 2300 .707 31 3723 8 1972 1967 2300 .855 31 3692 9 1972 2031 2300 .883 30 3662 10 1972 2099 2300 .912 31 3631 11 1972 902 2300 .827 30 3601 12 1972 1498 2300 .651 31 3570 1 1973 1213 2300 .529 31 3539
, 2 1973 1196 2300 .520 28 3511 3 1973 819 2300 .356 31 3480 4 1973 0 2300 .000 30 3450 5 1973 436 2300 .189
{
31 3419 6 1973 1821 2300 .792 30 3389 7 1973 1847 2300 .803 31 3358 l 8 1973 2006 2300 .872 31 3327 9 1973 2063 2300 .897 30 3297 10 1973 2031 2300 .883 31 3266 11 1973 1583 2300 .688 30 3236 12 1973 1952 2300 .849 31 3205 1 1974 2082 2300 .905 31 3174 2 1974 2162 2300 .940 28 3146 3 1974 2226 2300 .968 31 3115 4 1974 2179 2300 .947 30 3085 5 1974 315 2300 .137 31 3054 6-16
TABLE 6-6 (cont) lRRADIATION HISTORY OF CAPSULE T i
irradiation Decay Time Month Year Pj(MW) Pmax(MW) Pj/Pmax Time (Days) (Days) 6 1974 161 2300 .070 30 3024 l 7 1974 2028 2300 .882 31 2993 8 1974 1949 2300 .848 31 2962 9 1974 1994 2300 .867 30 2932 10 1974 1991 2300 .865 31 2901 11 1974 2239 2300 .973 30 2871 12 1974 2280 2300 .991 31 2840 1 1975 2090 2300 .909 31 2809 2 1975 2255 2300 .980 28 2761 3 1975 2167 2300 .942 31 2750 4 1975 795 2300 .346 30 2720 5 1975 298 2300 .130 31 2689 6 1975 1738 2300 .756 30 2659 7 1975 1976 2300 .859 31 2628 8 1975 2130 2300 .926 31 2597 9 1975 2051 2300 .892 30 2567 10 1975 2125 2300 .924 31 2536 11 1975 0 2300 .000 30 2506 12 1975 1133 2300 .493 31 2475 1 1976 1899 2300 .826 31 2444 2 1976 2267 2300 .986 29 2415 3 1976 2194 2300 .954 31 2384 4 1976 2132 2300 .927 30 2354 5 1976 1925 2300 .837 31 2323 6 1976 2166 2300 .943 30 2293 7 1976 1988 2300 .864 31 2262 8 1976 2066 2300 .898 31 2231 9 1976 2133 2300 .927 30 2201 10 1976 1991 2300 .866 31 2170 11 1976 0 2300 .000 30 2140 12 1976 1044 2300 .454 31 2109 1 1977 2181 2300 .948 31 2078 2 1977 1170 2300 .509 28 2050 3 1977 1987 2300 .864 31 2019 4 1977 1726 2300 .750 30 1989 6-17
s TABLE 6-6 (cont) lRRADIATION HISTORY OF CAPSULE T frradiation Decay Time Month Year Pj(MW) Pmax(MW) Pj/P max Time (Days) (Days) 5 1977 2138 2300 .930 31 1958 6 1977 2036 2300 .885 30 1928 7 1977 1890 2300 .822 31 1897 8 1977 1502 2300 .653 31 1866 9 1977 940 2300 .409 30 1836 10 1977 1047 2300 .455 31 1805 11 1977 275 2300 .120 30 1775 12 1977 2126 2300 .924 31 1744 1 1978 1778 2300 .773 31 1713 2 1978 0 2300 .000 28 1685 3 1978 0 2300 .000 31 1654 4 1978 200 2300 .087 30 1624 5 1978 2036 2300 .885 31 1593 6 1978 2106 2300 .916 30 1563 7 1978 1643 2300 .715 31 1532 8 1978 2067 2300 .899 31 1501 9 1978 1477 2300 .642 30 1471 10 1978 2176 2300 .947 31 1440 11 1978 2193 2300 .953 30 1410 12 1978 2120 2300 .922 31 1379 1 1979 2055 2300 .893 31 1348 2 1979 2007 2300 .872 28 1320 3 1979 2180 2300 .948 31 1289 4 1979 678 2300 .295 30 1259 5 1979 0 2300 .000 31 1226 6 1979 0 2300 .000 30 1198 7 1979 475 2300 .207 31 1167 8 1979 2162 2300 .940 31 1136 9 1979 2050 2300 .891 30 1106 10 1979 2104 2300 .915 31 1075 11 1979 2147 2300 .933 30 1045 12 1979 2157 2300 .938 31 1014 1 1980 2204 2300 .958 31 983 2 1980 2163 2300 .940 29 954 3 1980 1192 2300 .518 31 923 6-18
TABLE 6-6 (cont)
IRRADIATION HISTORY OF CAPSULE T ,
Irradiation Decay Time Month Year Pj(MW) Pmax(MW) Pj/Pmax Time (Days) (Days) 4 1980 899 2300 .391 30 893 5 1980 1367 2300 .594 31 862 6 1980 1994 2300 .867 30 832 7 1980 1093 2300 .475 31 801 8 1980 219 2300 .095 31 770 9 1980 0 2300 .000 30 740 10 1980 228 2300 .099 31 709 11 1980 1912 2300 .831 30 679 12 1980 401 2300 .174 31 648 1 1981 1595 2300 .693 31 617 2 1981 1930 2300 .839 28 589 3 1981 2196 2300 .955 31 558 4 1981 2141 2300 .931 30 528 5 1981 989 2300 .430 31 497 6 1961 1108 2300 .482 30 467 7 1981 1817 2300 .790 31 436 8 1981 0 2300 .000 31 405 9 1981 922 2300 .401 30 375 10 1981 1058 2300 .460 31 344 11 1981 239 2300 .104 30 314 12 1981 1391 2300 .605 31 283 1 1982 1825 2300 .793 31 252 2 1982 1677 2300 .729 28 224 table 6-7. The data are presented as measured at the actual monitor locations as well as adjusted to the capsule center. All 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 T are presented in table 6-8. The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in table 6-9. Due to the relatively low thermal neutron flux at the capsule locations, no burnup correc-tion 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 NiS8 (n,p) CoS8 reaction and even less significant for all of the other fast neutron reactions.
6-19 J
l TABLE 6 7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE T Saturated Activity Adjusted Saturated Activity Reaction Radial dis /s dis /s and Location 9 g Axial Location (cm) Capsule T Calculated Capsule V Calculated Fe54(n.p)Mn54 R-63 191.47 9.05 x 106 8.35 x 106 8.62 x 106 H-17 191.47 8.93 x 106 8 35 x 106 8 51 x 106 R-57 191.47 9.78 x 106 8.35 x 106 9.32 x 106 W-18 192 47 7.53 x 106 6.65 x 106 9 01 x 106 L-42 192.47 7.47 x.106 6 65 x 106 8 94 x 106 L-48 192.47 7.38 x 106 6 65 x 106 8 83 x 106 Average 8.87 x 106 7.96 x 106 Cu63(n.n)Co60 Top 191.47 7.90 x 105 5.98 x 105 7.53 x 105 Bottom 191 47 8 26 x 105 5.98 x 105 7 87 x 105 Average 7.70 x 105 5.70 x 105 N,5 8(n.p)CoS 8 Middle 191.47 1.31 x 108 1.22 x 108 1.24 x 108 1.16 x 108 Np237(n.f)Cs137 Middle 191.70 8.72 x 107 7.15 x 107 8.72 x 107 7.15 x 107
\
U238(n.f)Cs 37fa 1
1 Middle 191.70 132 x 107 9.65 x 106 1.32 x 107 9.65 x 106 a includes a 20 percent correction for U235 fissions and Pu239 produevon Using the iron data presented in table 6-8, along with the lead factors given in table 6-3, the fast neutron fluence (E > 1.0 MeV) for Capsule T as well as for the reactor vesselinner diameter is summarized in table 6-10. The agreement between calculation and measurement is excellent with a measured fluence level of 4.11 x 1019 n/cm2 compared to a calculated value of 3.81 x 1019 n/cm2 for Capsule T.
e h*20
l TABLE 6 8 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE T Adjusted Saturated Activity dis /s 6 (E > 1.0 MeV) <!> (E > 1.0 MeV) l Reaction T (n/cm2- sec) (n/cm2)
Measured Calculated Measured Calculated Measured Calculated Fe54(n.p)Mn54 8.87 x 106 7.96 x 106 1.81 x 1011 1.68 x 1011 4.11 x 1019 3 81 x 1019 l Cu63(n.a)Co60 7,70 x 105 5.70 x 105 2.19 x 1011 4 97 x 1019 N,58(n.p)CoS8 1.24 x 108 1.16 x 108 1.7 7 x 1011 4 02 x 1019 Np237(n.f)Cs137 8.72 x 107 7.15 x 107 2.05 x 1011 4.65 x 1019 U238(n.f)Cs137 1.32 x 107 9.65 x 106 2.27 x 1011 5.16 x 1019 Note. Irradiation time eauals 2.27 x 108 EFPS TABLE 6-9 RESULTS OF THERM AL NEUTRON DOSIMETRY FOR CAPSULE T Saturated Activity Axial dis /s 6Th Location g (n/cm2 sec)
Bare Cd Covered Top 1.24 x 108 5.62 x 107 1.20 x 1011 Middle Not Measured 6.04 x 107 1.34 x 1011[a]
Bottom 1.4S x 108 6.11 x 107 1.55 x 1011
- a. Theimal neutron flux at the middle position is based on a bare wire activity equal to the aver-age of the top and bottom measursments.
Also listed in table 6-10 is a comperison of calculated values with the average value of all dosime-ter measurements. Again, the agreement is good and within the assigned uncertainty of plus or minus 20 percent for the analytical predictions.
t 6-21
TABLE 6-10
SUMMARY
OF NEUTRON DOSIMETRY RESULTS FOR CAPSULE T frradiation 4 (E > 1.0 $ (E > 1.0 Vessel Vessel Basis Time Lead MeV) MeV) Fluence Fluence (calculcated )
y (n/cm2-sec) Factor (EFPS) (n/cm2) (n/cm2) (n/cm2)
N 2 L7 Fe54(n.p)Mn54) hx 108 1.81 x 1011 4.11 x 1019 2.63 1.56 x 1019 1.45 x 1019 Dosimeter Avg 2 27 x 109 2.02 x 1011 4.58 x 1019 2 S3 1.74 x 1019 1.45 x 1019 k
t
% jg f
/
d ,
- ~
m .-
A >
SEFERENCES D
I
, ~
't. Yanichko, S. E., " Carolina Power And Light Co., H. B. Robinson Unit No. 2 Reactor Vessel Radi-ation Surveillance Program," WCAP-7373, January,1970.
s
- 2. Yenichko, S. E., et al, " Analysis of Capsule S From Carolina Power and Light Co., H. B. Robin-
. son Uait 2 Reactor Vessel Radiation Surveillance Program," WC AP-8249, December 18, i 1973.
- 3. Norris, E. B. " Reactor Vessel Material Surveillance Program For H. B. Robinson Unit 2 Analysis
'of Cspsule V " Final Report - SWRI Project No. 02-4397, October 19,1978.
- 4. t.etter From E. B. Norris, Southwest Research institute To Talmadge Clements, Carolina Power and Light Co.,"H. D. Robinson Unit 2 Reactw Vesset Material Surveillance Program," April 4, 1977
- 5. Soltesz, R. G., et al., FinalProgress Report, Naclear Rocket Shielding Methods, Modification, Updating, andinput Data Preparation Vol. 5, Two-DimensionalDiscrete Ordinates Transport Technique,WANL-PR(LL)034, August,1970.
6. Collier, G., et al, Second Version of the GAMBI TCode, WANL-TME-1969, November 1969.
o.
, 7. Soltesz, R. G., et al, FinalProgress Report, Nuclear Rocket Shielding Methods, Modification, Updating, andInput Data Preparation, Vol. 3 Cross Section Generation andData Processing Techniques,WANL-PR(LL)034, August ,1970.
8 Soltesz, R. G., et al, FinalProgress Report, Nuclear Rocket Shielding Methods Modification, Updating, andInput Datb Prepsration, Vol. 4. One-DimensionalDiscrete Ordinates Transport
. Technique,WANL-PR(LL1034, August 1970.
S. ASTM E 261-77, "St'andard Practice for Measuring Neutron Flux, Fluence, and Spectra by Radioactivation Techniques." 198 t AnnualBook of ASTM Standards. Part 45 Nuclear Stan-dards pp 915 926 Philadelphia: American Society for Testing and Materials,1981 A-1
~
- 10. ASTM E 262-77, " Standard Test Method for Measuring Thermal Neutron Flux by Radioactivo-tion Techniques." 1981 AnnualBook of ASTMStandards. Part 45 Nuclear Standards, pp 927-935 Philadelphia: American Society for Testing and Materials,1981.
- 11. ASTM E 263-77, " Standard Test Method for Measuring Fast-Neutron Flux by Radioactivation of Iron," 1981 AnnualBook of AS TM Standards. Part 45 Nuclear Standards. pp 936-941 Philadelphia: American Society for Testing and Materials,1981.
- 12. ASTM E 481-78, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver," 19B1 Annua / Book o/ ASTM Standards. Part 45 Nuclear Standards. pp 1063-1070 Philadelphia: American Society for Testing and Materials,1981.
- 13. ASTM E 264-77," Standard Test Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel." 1981 AnnualBook ofASTM Standards. Part 45 Nuclear Standards. pp 942-946 Philadelphia: American Society for Testing and Materials,1981.
)
)
l A2
. _ _ _ _ _ .