ML20235J194
| ML20235J194 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 11/30/1988 |
| From: | Albertin L, Shaun Anderson, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML111660849 | List: |
| References | |
| WCAP-12020, NUDOCS 8902240208 | |
| Download: ML20235J194 (86) | |
Text
- _-___ __ __
i ' WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION WCAP-12020 ANALYSIS OF CAPSULE P FROM THE WISCONSIN PUBLIC SERVICE CORPORATION KEWAUNEE NUCLEAR PLANT REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko S. L. Anderson L. Albertin November 1988 Work performed under Shop Order No. KYFJ-106 APPROVED: -
k T.A.Meyer,Mdnager Structural Materials and Reliability Technology Prepared by Westinghouse for the Wisconsin Public Service Corporation Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.
- WESTINGHOUSE ELECTRIC CORPORATION Nuclear Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230 3342s.120664-t o 890224o20s 890217 PDR ADOCK 05000305 p PDC
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PREFACE
,. c This report has been technically reviewed and verified..
Reviewer Sections 1- through 5, 7 and 8 N. K. Ray ~~~d? t wf'7 SEction6 E. P. Lippincott YM '
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Section Title 'Page-1
SUMMARY
0F RESULTS 'l-1 l
2 INTRODUCTION' 2-1 3 BACKGROUND 3-1 4 DESCRIPTION OF PROGRAM 4 5 TESTING OF SPECIMENS FROM CAPSULE P 5-1 5-1. Overview 5-1 5-2. Charpy V-Notch Impact Test Results 5-3 5-3. Tension Test Results 5-5 .
5-4. Wedge Opening Loading. Tests 5-5 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1 6-1. Introduction 6-1 6-2. Discretc Ordinates Analysis 6-2 6-3. Neutron Dosimetry 6-7 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 1 REFERENCES 8-1 3342s-120068:10 jy
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' LIST OF ILLUSTRATIONS Figure-Title Page c 4-1 Arrangement of Surveillance Capsules in the 4-5
, . Reactor Vessel
'4 Capsule ~P Diagram Showing Location of Specimens,. 4 -
Thermal' Monitors,-and Dosimeters 5-1 Charpy V-Notch Impact Properties for Kewaunee 15 Reactor Vessel Shell Forging 122X208 val-(Tangential Orientation) 5-2' Charpy V-Notch Impact Properties for Kewaunee 5 .<
Reactor Vessel Shell Forging 123X167 VA1
- (Tangential Orientation) a 5-3 .Charpy V-Notch Impact Properties for Kewaunee 5-17 Reactor Vessel Weld Metal 5-4 Charpy V-Notch Impact Properties for Kewaunee 5-18 l Reactor Vessel. Weld HAZ Metal 5-5 Charpy V-Notch Impact Properties for Kewaunee 5-19 ASTM Correlation Monitor Material (HSST Plate 02) 5-6 Charpy Impact Specican Fracture Surfaces for Kewaunee 5-20 Reactor Vessel Sh011 Forging 122X208 VA1 (Tangential l
Orientation) 5-7 Charpy Impact Specimen Fracture Surfaces for Kewaunee 5-21 Reactor. Vessel Shell Forging 123X167 VA1 (Tangential. Orientation) 3342s-120044.10 y LL _ _ _ _ _ _ . _. _.__________..___._ ___ _.__ _ ____ _ _ Q
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LIST OF ILLUSTRATIONS (Cont)
Figure Title Page 5-8 Charpy Impact Specimen Fracture Surfaces for 5-22 Kewaunee Reactor Vessel Weld Metal 5-9 Charpy Impact Specimen Fracture Surfaces for 5-23 i Kewaunee Reactor Vessel Weld HAZ Metal l
5 Charpy Impact Specimen Fracture Surfaces for 5-24 Kewaunee ASTM Correlation Monitor Material (HSST Plate 02) !
5-11 Tensile Properties for Kewaunee Reactor Vessel 5-25 I c Shell Forging 122X208 VA1 (Tangential Orientation) ,
5-12 Tensile Properties for Kewaunee Reactor Vessel 5-26 Shell Forging 123X167 VA1 (Tangential Orientation) 5-13 Fractured Tensile Specimens from the Kewaunee 5-27 Reactor' Vessel Shell Forging 122X208 VA1 i (Tangential Orientation) 5-14 Fractured Tensile Specinens from the Kewaunee 5-29 Reactor Vessel Shell Forging 123X167 VA1 (Tangential Orientation) 5-15 Typical Stress-Strain Curve for Tension Specimens 5-30 6-1 Surveillance Capsule Geometry 6-12 e
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. LIST OF TABLES'
' Table Title Page. -
3-1 ReactorVesselToughnessData(Unirradiited) 3-3 c
4-1 Chemical Composition and Heat Treatment of the 4 Kewaunee Reactor Vessel Surveillance Materials 4-2 Chemistry..and Heat Treatment of'A533 Grade B Class 1 4-4 ASTM Correlation Monitor Material (HSST Plate 02) 5-l' 'Charpy V-Notch Impact Data for the Kewaunee 5-6
-Reactor Vassel Shell Forgings Irrediated at 550*F, Fluence 2.89 x 10 19 n/cm2 (E > 1.0 MeV) 5-2 Charpy'V-Notch Impact Data for the.Kewaunee. 5-7 Reactor Vessel Weld Metal and.HAZ. Metal Irradiated at 550*F, Fluence 2.89 x 10 19 n/cm2 (E > 1.0'MeV) 5-3 Charpy V-Notch Impact Data for.the Kewaunee ASTM 5-8 Correlation Monitor Material Irradiated at 550'F, 19 Fluence 2.89 x 10 n/cm2 (E > 1.0 MeV) 5-4 Irradiated Instrumented Charpy Impact Test Results for 5 Kewaunee Reactor Vessel Shell Forgings 5-5 Irradiated Instrumented Charpy Impact Test Results for 5-10 Kewaunee Reactor Vessel Weld Metal and HAZ Metal 5-6 Irradiated Instrumented Charpy Impact Test Results for 5-11 !
Kewaunee ASTM Correlation Monitor Material 2342s-120ess 10 yjj
LIST OF TABLES (Cont)
Table Title Page 19 2 5-7 The Effect of 650*F Irradiation at 2.89 x 10 n/cm 5-12
{E > 1.0 MeV) on the Notch Toughness Properties of Kewauneo Reactor Vessel Materials 5-8 Comparison of Kewaunee Reactor Vessei Surveillance 5-13 Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions 5-9 Tensile Properties for Kewaunee Reactor Vessel Material 5-14 Irradiated to 2.89 x 10 19 n/cm2 (E > 1.0 MeV)
~
6-1 Calculaced Fast Neutron Exposure Parameters at the 6-13 l- Center of Capsule P .
6-2 Calculated Fast Neutron Exposure Parameters at the 6-14 Pressure Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux 6-15 (E>1.0 MeV) Within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-16 (E>0.1 MeV) Within the Pressure Vessel Wall 6-5 Relative Radial Distribution of Iron Displacement 6-17 Rate (dpa) Within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-18 1
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LIST OF TABLES (Cont)
Table Title Page -
~6-7 Irradiation History of Neutron Sensors Contained 6-19 in Capsule P 6-8 Measured Sensor Activities and Reaction Rates 6-23 6-9 Summary of Neutron Dosimetry Results 6-25 6-10 Compariso:n of Measured and FERRET Calculated 6-26 Reaction Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-27 Capsule Center 6 -12 Comparison of Calculated and Measured Exposure 6-28 Levels for Capsule P 6-13 Neutron Exposure Projections at Locations on the 6-29 Pressure Vessel Clad / Base Metal Interface 6-14 Vessel Neutron Exposure Values (E > 1 MeV) for Use in the 6-30 Generation of Heatup/Cooldown Curves 6-15 Updated Lead Factors for Kewaunee Surveillance 6-31 Capsules 0
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SECTION 1
SUMMARY
OF RESULTS I
l' ine analysis of the reactor vessel material contained in Capsule P, the third surveillance capsule te be removed from the Wisconsin Public Service Corporation Kewaunee reactor pressure vessel, led to the following conclusions:
o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 2. 89 x 10 19 n/cm2 ,
c Irradiation of Charpy V-notch impact specimens from the reactor 19 vessel intermediate shell forging 122X208 VA1, to 2.69 x 10 n/cm 2 , resulted in 30 and 50 ft-lb transition temperature increases of 25'F and 10'F respectively, for specimens oriented parallel to the major working direction (tangential orientation). !
Irradiation of Charpy V-Notch Impact specimens from the vessel lower shell forging 123X167 VA1 resulted in 30 and 50 ft-ib transition temperature increases of 20*F for tangentially oriented specimens.
19 2 o Weld metal impact speciniens irradiated to 2.89 X 10 n/cm resulted in 30 and 50 ft-lb transition temperature increases of 230'F ano 235'F respectively.
19 2 o Irradiation to 2.89 x 10 n/cm resulted in a 3 ft-lb decrease !
in the average upper shelf energy of forging 122X208 VA1 and ne decrease in the upper shelf energy of forging 123X167 VA1. The weld metal decreased by 50 ft-lb from 126 to 76 ft-lbs. All materials tested exhibit a more than adequate shelf energy level for continued safe plant operation, o Comparison of the 30 ft-lb transition temperature increases for the Kowaunee surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2, shows that the forging material and weld metal transition temperature increase were less than predicted.
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SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule 9, the third capsule to be removed from the reactor in the contiruing surveillance program which monitors the effects of neutron irradiation on the Kewaunee reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Kewaunee reactor pressure vessel materials was desigr.ed and recommended by the Westinghouse Electric Corporation. A' description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by Yanichko. [1] The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based on ASTM E-185-70, " Recommended Practice for Surveillance Tests on Nuclear Reactors Vessels". Westinghouse Energy
^
Systems personnel were contracted to aid in the preparation of proceduree, for removing the capsule from the reactor and its shipnent 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 L
surveillance Capsule P removed from the Kewaunee reactor vessel and discusses the analysis of the data. The data are also compared to results of the l
previously removed Kewaunee surveillance Capsule V [2] and Capsule R [3].
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_ _ _ _ _ _ . .____-________-__-___a
Q SECTION 3
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BACKGROUND The ability of the large steel pressare vessel containing the reactor core (,nd its primary coolant to resist fracture: constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron exposure. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel eteels such as SA 508 Class 2 (base materici of the Kewaunee reactor pressure vessel beltline) are well documented in the literature.
Generally, low alloy ferritic materials show an increase in hardness and-tensile properties and a decrease in ductility and toughness under certain
- conditions of irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Ncn-ductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the t
referencenil-ductilitytemperature(RTNDT)*
RT is defined as the greater of either the drop weight nil-ductility NDT transition temperature (NDTT per ASTM E-208) or the temperature 60'F less than the 50 ft Ib (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction i
of the material. The RT NDT f a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The K IR curve is a lower bound of i dynamic, crack arrest, and static fracture toughness results ootained from several heats of pressure vessel steel. When a given material is indexed to
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F i the KIR cune,-allowable stress intensity factors can be obtained .for this
. material ~as a function of temperature. Allowable operating limits can then be- ..
, determined utilizing these allowable. stress intensity factors.
1 RT and, in turn,- the operating limits of nuclear power plants can' be NDT adjusted to: account. for the. effects of radiation on the reactor vessel materialfproperties. The radiation-embrittlement or changes in mechanical properties-of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Kewaunee Reactor Vessel Radiation Surveillance Program,. [1] in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are
. tested. The increase in the average Charpy V-notch 30 ft 15 temperature-(ARTNDT) due to irradiation is added to the original RTNDT to adjust the RT NDT f r radiation embrittlement.. This adjusted RTNDT (RTNOT initial +
curve and, in' turn, to i
ARTNDT) is u ed to index the material to the KIR set operating limits for the nuclear power plant which take into' account'the effects of irradiation on the reactor vessel materials.
l The unirradiated fracture toughness properties of the Kewaunee reactor vessel material are identified in table 3-1.
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SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on .
the Kowaunee reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the neutron shield pads and the vessel wall at locations shown in'figura 4-1. The vertical center of the capsules is opposite the vertical center of the core.
Capsule P (Figure 4-2) was removed after 11.08 effective full power years of plant operation. This capsule contained Charpy V-notch impact, tensile, and IX Wedge Opening Loading.(WOL) fracture mechanics specimens from the reactor vessel intermediate shell ring forging 122X208 VA1 and lower shell ring forging 123X167 VA1, and Charpy V-notch specimens from submerged arc weld metal identical to the beltline region girth weld seam of the reactor vessel and weld heat-affected zone (HAZ) material. All heat-affected zone specimens ,
were obtained from within the HAZ of forging 122X208 VA1. The capsule also contained Charpy V-notch specimens from the 12-inch thick ASTM correlation .
i monitor _ material (HSST plate 02).
l The chemistry and heat treatment of the surveillance material are presented in table 4-1 and 4-2. The chemical analyses reported in table 4-1 were obtained from unirradiated material used in the surveillance program.
l All test specimens were machined from the 1/4 thickness location of the l forgings. Test specimens represent material taken at least one forgings j
) thickness from the quenched end of the forgings. All base metal Charpy V-notch impact and tensile specimens were oriented with the longitudinal axis ,
of the specimen parallel to (tangential orientation) the principal working i-direction of the forgings. Charpy V-notch specimens from the weld metal were e
3342sd O198810 4.}
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~ _ . _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
oriented with the longitudinal axis of the specimens transverse to the weld direction. The Wedge Opening Loading (WOL) test specimens in Capsule P were nachined parallel to the major working direction (tangential orientation).
~
All specimens were fatigue precracked per ASTM E399-70T.
Capsule P contained dosimeter wires of pure iron, copper, nickel, and unshielded aluminum-cobalt. In addition, cadmium-shielded dosimeters of Neptunium (Np237) and Uranium (U238) were contained in the capsule.
Thermal monitors made from two low melting eutectic alloys and sealed in Pyrox tubes were included in the capsule and were located as shown in figure 4-2.
The two eutectic alloys and their melting points are:
2.5% Ag, 97.5% Pb Melting Point 579'F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Poir.t 590*F (310*C)
The arrangement of the various mechanical test specimans, dosimeters and thermal monitors contained in Capsule P are shown in figure 4-2, 9
9
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. TABLE 4 -CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE- .
~KEWAUNEE-REACTOR VESSEL SURVEILLANCE MATERIALS CHEMICAL. ANALYSIS (k %) .-.
Forging Forging' Weld-
- Element- 122X208VA1 . 123X167VA1 -Metal.
C '0.21 0.20 0.12 Si O.25 0.28 0.20 Mo. 0.58: 0.58 0.48
. Cu 0.06 0.06 0.20 Ni- 0.71 0.75' -0.77 Mn .0.69' O.79 1.37 Cr. 0.40 0.35 0.090 V <0.02 <0.02 0.002 Co .0.011 0.012 0.001 i
. Sn 0.01 0.01' O.004' l
- Ti <0.001 <0.001 <0.001 Zr 0.001 0.001 <0.001
- As 0.001 0.004 0.004
- Sb~ <0.001 0.00* 0.001 S- 0.011 0.005 0.011 P 0.010 0.010 0.016 m.
Al' O.004 -0.305 0.010 B <0.003 <0.003 <0.003
-N 0.006 0.010 0.012 .
. Z$ - -
<0.001 HEAT TREATMENT
, Intermediate Shell Forging Heated at 1550*F'for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, water quenched Heat 122X208VA1 Tempered at 1230*f. for 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, air-cooled Stress-relieved at 1150*F for 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, furnace-cooled Low'er Shell Forging Heated at 1550*F for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, water quenched Heat 123X167VA1 Tempered at 1220*F for 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, f air-cooled Stress-relieved'at 1150*F for 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, i furnace-cooled Submerged Arc Weldment Stress-relieved at 1150*F for 19-1/4 hours, furnace-cooled The weldment was f abricated by Combustion Engineering. Inc., using 3/16 inch Mil B-4 modified weld filler wire, heat number Ifs!71 and Linde 1092 flux, lot number 3958 and is identical to that used ir, the actual fabrication of the reactor vessel intermediate to lower shell girth weld -
seam.
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l TABLE 4-2 h CHEMISTRY AND HEAT TREATMENT OF A533 GRADE,B CLASS 1 ASTM CORRELATION MONITOR MATERIAL (HSST PLATE 02)
Chemical Analysis (wt%)
P S Si Hi Mo Cu C Mn 1.48 0.012 0.018 0.25 0.68 0.52 0.14 0.22 .
. Heat Treatment 1
Heated at 1675 1 25'F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - air cooled Heated at 1600 1 25'F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - water quenched
. Tempered at 1225 1 25'F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - furnace-cooled Stress-relieved at 1150 1 25'F - 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> - furnace-cooled to 600'F g
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SECTION 5' TESTING OF SPECIMENS FROM CAPSULE P
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. 5-1. OVERVIEW-The postirradiation mechanical testing of.the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Energy Systems personnel.
Testing was performed in accordance with 10CFR50, Appendices G and HI43, ASTM Specification E185-82 a.nd Westinghouse Procedure MHL 8402, Revision 0 as modified by RMF Procedures 8102 and 8103.
Upyn receipt of the capsule at the laboratory, the specimens and spacer blocks were rarefu;1y removed, inspected for identification number, and checked against the master list in WCAP-8908 #1]. 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).
f
< The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 !
instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED ), From the load-time curve, the load of general yielding (Pgy),the time to general yielding (tGY), the maximum load (PM ), and the time to maximum load (t ) can be determined. Under some test conditions, a sharp M
l 3342s/101968 10 5-1 l
1
. .. 8 I
1 4
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 t!n
- load at wh'.ch fast fracture terminated is identified as the arrest load (Pg ).
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 approximately equi' valent to the energy required to initiate a crack in the specimen.
Therefore, the propagation energy for the crack (E p
) is the difference between the total energy to fracture (E ).and D
the energy at maximum load.
The yield stress (oy) is calculated from the three point bend formula.
The flow stress is calculated "from the average of the yield and maximum loads, also using the three point bend formula. ,
Percentage shear was determined from postfracture p'notographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77. The
- lateral expansion was measured using a dial gage rig similar to that shown in t the same specification.
4 Tension tests were performed on a 20,000 pound Instron, split-console test l r
machine (Model 1115) per ASTM Specifications i8-83 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test.
l Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-l loaded to the specimen and operated through specimen failure. The extenso-meter gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67.
l
- Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were condu:ted in air.
nu./io" " " 5-2
i Because of the difficulty in remotely attaching a thermocouple directly to the ]
j specimen. .the following procedure was used to monitor specimen tempr.rature.
Chromel-alur$1 thermocouple were inserted in shallow holes in the center an'd "f each end efithe gage section of a dummy specimen and in'each grip. In the .
test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip _and controller tanoeratures was developed over the range room temperature to 550*F (288'C). The upper grip was used to control . the furnace temperature. During the actual testing the grip temocratures were used to obtain desired specimen temperatures.
Experiments indicated that this method is accurate to~plus or minus 2*F.
The yield load, ultimate . load, fracture load, total elongatien, and un-iform elongation were determined directly fram the load-extension curve. The yield strength, ultimate strength, anc' fracture strength were calculated usir.g the original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used to calcelate the fracture stress (true stress at fracture) and percent reduction in area ,
was computed using the final diameter measurement. <
.(
5 2. CriARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy' V-notch impect tests performed on the various materials -
19 contained in Capsule P irradiated to approximately 550*F at 2.89 x 10 n/cm2 are prasented in tables 5-1 through 5-6 and figures 5-1 through 5-5. l The transition temperature increases and upper shelf energy decreases for the Capsule P material are shown in table 5-7.
Irradiation of the vessel intctmediate shell forging 122X203 VA1 material I9 (tangewal orientation) specimens to 2 89 x 10 n/cm2 (figure 5-1) l resulted in 30 and 50 ft-lb transition temperature increases of 25'F and 10'F respectively, an:] an upper shelf energy decrease of 3 ft-lb when compared to the unirradiated data from reference (1).
1 3342a/10198810 6-3 a
-j, ; I h f
irradiation of the vassel lower chell . forging 123X167 VA1 material (tangential q 19 j
- g. orientation) r.pecimens to 2.39 x,10 m/cm2 (figurz 5-2) resulted in both L' 30 or 50 ft-lb transition tuparature increases of 20*F and no upper shelf energy decrease of when co'apared to the unirradiated data. .
Weld nvatal irradiated to 2.f;9 x 10 19 n/cm2'(11gure 5-3)- esulted in a 30 and 50 ft-lb transition temperature increase of 230*F and 235*F respectively, and an upper shelf energy decrease of 50 ft-lb from 125 to 76 ft-lbs..
19 Weld HAZ metal irradiated to 2.89 x 10 n/cn2 (figure 5-4) resulted in 30 and 50 ft-lb transitten temperature increases of 220*F and an upper shelf energy decrease of 44 f t-lb.
ASTM correlation r.onitor material (HSST Plate j02) irradiated to 2.89 x 10 1t v./cm2 (figure 3-5) showed a 30 and 50 ft-lb transition temperature increase of 155'F and an upper shelf decrease of 22 ft-lb. These value are similar to i
. those obtair.ed from other surveillance capsule programs. l I
- The fracture appearance of each irradiated Charpy specimen from the various L materials is shown in figures 5-6 through 5-10 and show an increasing ductile or tougher appearance with increasing test temperature.
i Table C-7 'shows a comparison of the 30 ft-lb transition temperature i f (ARTNDI) ' increases for the various Kewaunee surveillance materials with '
i predicted increases using the methods of proposed NRC Regulatory Guide 1.99, Revision. 2. [5] This comparison shows that the transition temperature increase resulting from irradiation to 2.89 x 10 19 n/cm is approdmately 2
25'T less than predicted by the Guide f or the shell forgings. The weld metal 19 transition temperature increase resulting from irradiation to 2.89 x 10 i n/cm2 is 12*F less than the Guide prediction. f p
l 1
m -ame io 5-4 i - -- ..__ ____ _ __ __ _ ___________
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' 5-3. JENSION TEST RESULTS f [l,IL L .l
- The resultstot tension' tests performed on the shall forgings.122Xf.08 VA1 and
- 19 2
.12LX167 VA1 (tangential orientation) irradiated lto 2.89.x 10 n/cm are
((
' ~
j.
.shown(in table 5-9 and figures 5-11, and 5-12, respectively. . Thissr, results~
show that irradiation produced approximately a 10 Ksi increane' in 0.2 percent
= i- yield; strength for the shell forgings. Fractured tensien speciraens'for each H of the materials are shown -in figures 5-13, and 5-14. A typical stress-strain curve for the tension specimens is shown in figure 5-15.
5-4. WEDGE OPENING LOADING TESTS i.
i Per the surveillance capsule testing contract with the Wisconsiti Public Service Corporation, 1X - Wedge Opening Loading fracture mechanics specimens will not be tested and will be stored at the Hot Cell at the Westinghouse R&D q Center.
I i
n u.no m a,o 5-5 l
U- - - - - - - _ _ - - _ - - - - _ -- _ _ _ _ _ _ _ ________________ _ _ _ l
g {g
-.= , ,
/
I TABLE'5-l' l
CHARPY V-NOTCH IMPACT DATA FOR THE KEWAUNEE
~
REACTOR VESSEL S'clELL FORGU#iS IRRADIATED AT 550*F, FLUENCE 2.89 x 10 0 ' n/cm2 (E >'1.0 MeV)
I Tempers.ture Impact Energy Lateral Expansion Shear Sample No. (*F) ,(' 0), (ft-lb) JJ1 ,(pils) ,(gal (%)
FORGING 12f!I208VA1 Tanzential Orientatio3 P62 - 90 (-68) 5.0 ('/.0) 6.5 (0.17)- 5 !
P63 - 50 ( 46)- 32.0 (43.5) 24.5 (0.02) 20 '
P64 0 (-18) 20.2 (27.0) 17.5 (0.44) 15 P70 0 (-18) 26.0 -(35.5) 20.0 (0.51) 20. ;
P61 25-(-4) 98.0 (133.0) 77.5 (1.97) 65 P66 50 (10) 18.0 (24.5) 19.5 -(0.50) 20 P65 50 (10) 95.0 (129.0) 71.0 -(1.80) 85
-P71 76 (24) -115.0 (156.0) 77.0 (1.96) 100 P68 125 (52) 153.0 (207.5) 91.5 (2.32) 100 P69 200 ( 93) ' 160.0 (217.0) 91.0 (2.31) 100 P67 300 (149) 157.0 (213.0) 95.0 100 (2.41)
P72 MACHINE MALFUNCTION FORGING 123X167VA1 fr.nrential Orientation
~
S70 - 50 (-46) 28.0 ( 38.0) 21.0 (0.53) 20 S68 - 25 (-32) 30.0 (40.5) 26.0 (0.66) 20 S69 - 25 ( -32) 8.0 (11.0) 8.0 (0.20) 5 863 0 (-18) 77.0 104.5) 61.0 (1.55) 60 S71 0 (-18) 44.0 59.5) 42.0 (1.07) 35 S67 25 (-4) 80.0 108.5) 64.0 (1.63) 70 S65 50 g10) 95.0 129.0) 73.0 (1.85) 95 SB2 76 ( 24) 107 0 (145.0) 71.0 (1.80) 100 SS6 125 (52) 155.0 (210.0) 92.0 (2.34) 100 S61 200 (93) 159.0 (215.5) 89.0 (2.26) 100 S64 300 (149) 170.0 (230.5) 83.0 (2.11)' 100 l
S72 400 (204) 152.0 (206.0) 82.0 (2.08) 100 l
r .
1uw'onss 'o 5-6 L - _---_----_-__--________o
I? : U U i [ ..
o i,,
4;
"/- TABLE 5-2 :
- p L i
- - CHARPY V-NOTCH IMPACT D/.TA FOR THE KEWAUNEE
- i. REACTOR VESSEL WELD METAL AMD MAZ METAL IRRADIATED AT 550'F 19 FLUENCE 2.89~x 10 nhm2 (E > 1.0 MeV) l l
,1 Temperature Impact Energy Lateral Expansion Shear f inja SampleNo.1*F), G), (ft-lb) Q ;(mils) (%)
Weld Metal W44 100. (38) 6.0 8.6J 14.0 (0.36) 10 W42 150 (66) 26.0 35.5) f5.0 (0.64) 25 l W46 175 (79) 22.0 30.0) 24.0 (0.61) 25 W48 200 ( 93) 37.0 (50.0) 32.0 (0.81) 35 W41 250 (121) 63.0 (85.5) 55.0 (1.40) 60 W43 350 (177) 73.0 (99.0) 60.0 (1.52) 100 4 W45 400 (204) 73.0 (99.0) 61.0 (1.55) 100 '!
W47 ,450 (232) 83.0 (113.5) 67.0 (1.70) 100 HAZ Metal H43 0 -18) 12.0 ( 16.5) 14.0 (0.36) 10 H44 76 24) 24.0 (32.5) 19.0 (0.48) 20 -
H45 150 66) 133.0 (180.5) 85.0 (2.16) 100 H47 150 66) 63.0 ( 85.5) 45.0 (1.14) 60 H41 200 ( 93) 111.0 (150.5) 82.5 (2.10) 100 H42 250 (121) 103.0- (139.5) 73.0 (1.85) 100 H48 350 (177) 134.0 (181.5) 91.0 (2.31) 100 H46 450 (232) 157.0 (185.5) 83.0 (2.11) 100 l
4 no. mn,-io 5-7
TAF,LE 5-3 CHARPY V-NOTCH IMPACT DATA FOR TdE KEWAUNEE
. ASTM CORRELATION MONITOR MATERIAL IRRADIATED AT 550*F' 19 FLUENCE 2.89 x 10 n/cm2 (E > 1.0 MeV)
Temperature Impact Energy Lateral Expansion Shear ,
Sample No. (*F) ('C) (ft-lb) (mils) (mm) (%)
R41 100 ( 38) 9.0 (IJ1 12.0) 9.0 (0.23) 10 20 R45 150 ( 66) 23.0 (31.0) 21.0 (0.53) 29.0 ( 39.5) 24.0 (0.61) 30 R46 200 (93) 25 R42 200 (93) 25.0 ( 34.0) 24.0 (0.61) 250 60.0 (81.5) 48.0 (1.22) 65 R47 (121)
R44 300 96.0 (130.0) 79.0 (2.01) 100 (149)
R48 350 108.0 (146.5) 82.0 (2.08) 100 (177) 450 98.0 (133.0) 89.0 (2.26) 100 R43 (232) 9 0
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O i i i i i i _t 0
- 200 -100 0 100 200 300 400 500 Temperature ( F)
Figure 5-1 Charpy V-notch Impact Properties for Kewaunee Reactor Vessel Shell .
Forging 122X208VA1(TangentiaiOrientation) nu.-inies >o 5-15
( 'C) !
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- 200 -100 0 100 200 300 40 500
- Temperature ( F)
Figure 5-2 Charpy V-notch Impact Properties for Kewaunee F:eactor Vessel Shell Forging 123X167 VA1 (Tangential Orientation) nu.-isma is 5-16
v ,
-150 -100 -50 0 !50 100 150 200 250 l I I l I 1 I3 7'
1 I
,j ~ 100 -
- 80 - 2% _
$ 60 - -
di 40 - o -
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l 0
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= 80 -
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~
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~
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3 -
o .
0 i i +/., i i i i 0
--200 -100 0 100 200 300 400 500 ,
Temperature ( F) .
Figure 5-3 Charpy V-notch Impact Properties for Kewaunee Reactor Vessel Weld -
Metal 33dh 16218810 5-U
. _ _ . _________-_-__--._--______--Q
(
- C)
-150 -100 -50 0 50 100 IS) 200 250
- i. i e i ' ' '
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$M
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t
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8 .
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. 2.89 x 10 n/cm
-4 -
220 F 4 8 .
1 20 -
0 i i i- i i i i 0 ,
- 200 -100 0 100 200 300 00 500 Temperature ( F)
Figure 5-4 Charpy V-notch Impact Properties for Kewaunee Reactor Vessel Weld HAZ Metal 1
mr..iema io 5-18
( 80)
- 150 -100 -50 0 50 100 150 200 250
- I I I I I l- 1 I I 100 -
>3 e : -
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Iii 60 - -
g#
m - -
3 2
20 -
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i 0
100 , , , , , , i. , 2.5 i
= 80 10
. E o M - -
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=
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2
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~
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20 .
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j 200 m
le -
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R 100 -
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w 60 -
155* F 80 Irradiated at 550*F-4 -
8 155 F_ 2
- 2. 89 x 10 '9 n/cm @ -
0 i i i i i __
0
- 200 -100 0 100 200 300 40 500 Temperature (*F) ,f Figure 5-5 Charpy V-Notch Impact Properties for Kewaunee ASTM Correlation ,
Monitor Material (HSST Plate 02) n u.-: m se:io 5-19
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l P68 P69 P67 Figure 5-6 Charpy Impact Specimen Fracture Surfaces for Kewaunee Reactor Vessel Shell Forging 122X208 VA1 (Tangential Orientation) nu. aema io 5-20 101-18097
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S66 SS1 SS4 S72 Figure 5-7 Charpy Impact Specimen Fracture Surfaces for Kewaunee Reactor
~
l Vessel Shell Forging 123X167 VII (Tangential Orientation) 1 1
1 .
3342s-162188 10 5n RM-18098 1
l l
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Figure 5-8 Charpy Impact Specimen Fracture Surfaces for Kewaunee Reactor
< Vessel Weld Metal aus.-issies in 5-22 101-18100
a s 6
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.u ? e H41 H42 H48 H46 Figure 5-9 Charpy Impact Specimen Fracture Surfaces for Kewaunee Reactor Vessel HAZ Metal .
no. inies to 5-23 IW-18101
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1 RM-18097
l
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~
Vessel Shell Forging 123X167 VA1 (Tangentisl Orientation) nn. ,wse so 5.g3 i
l l$1-18098
I-e.
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4 m w i m u io 5-22 RM-18100
., */
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2xu-ima io 5-23 I
RM-18101
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R47 E.44 R48 R43 Figure 5-10 Charpy Impact Specimen Fracture Surfaces For Kewaunee ASTM j
- Correlation Monitor Material (HSST Plate 02; m 2.-isussic 5-24 RM-llS99
a Curve 756428-A og .
- 50 '0 50 100' 150 200 250. 300 120 i i i i .i i i ' -
800' 110 -
100 -
700 ;L
$90 -
600 "-
,, 80 ..
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0 E
E- N ^ Tensile Strength -
g 70 -
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- _ ago 50 0 2 % Yleid Strength 40 I I I i _ i i l' - 300
+
Code :
Open Points-Unirradiated -
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- 3 -- -
9 :
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E
_g _. _
E 40 3
a 30 -
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20 -
' Ik: ==~- :#~# _
10 - 8 5: ::-e- -
y-e _
0 ___L___ i I ____ , Uniform Elongation g___
-100 0 100 200 300 400 500 600 ,
Temperature ( *F)
Figure 5-11 Tensile Properties for Kowaunee Reactor Vessel Shell Forging 122X208 VA1 (Tangential Orientation) no.-im m o 5-25
= - - - _ _ - _ _ - - - _ . - _ - _ _ _ - _ _ _ - _ _ _ _ _ _ - . .
f
- n .
curve 7564U -A o
, *C i
~
- 50 'O 50 100 150 200 250 300
- 120 .i i i i i i i ' -
800 110 -
700 4 100 -
- N l 25 Q +
.- 600 o-2 P . m gg _ :. 3 j Tensile Strength -
500 v, 70 -
% w . i 60 :- 2 -
0- 400 2
50 0.2 % Yield Strength - i i i i i i i i 300 40 Code:
Open Points -Unirradiated 19 2 Closed Points -Irradiated at 2.89 x 10 n/cm 80 i L i i i 70 - s .
2 _- _
0- 1 j
60 - Reduction Area _
s%
$M -
Total Elongation 1,j 30 -
c &
h ,
^
e
= -
20 -
8 10 - +- ? --
2 - -
I I juniform Elongation ,
0 I
-100 0 100 200 300 4tX) 500 600 Temperature ( *F)
Figure 5-12 Tensile Properties for Kewaunee Reactor Vessel Shell Forging
- 123X167 VA1 (Tangential Orientation)
I uu. $nin 'a 5-26
i l
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. Specimen P27 300*F l
k s.yw s l' ? 4 * -. j k $ >b ' .h t ' T - $ ; 'I' 4 i, $' ' kW#,y,kt*7Ik- T.#~ ' ;
67 ' L L2 ,2-3 ,'N ^ ge? Q h T :_W,ITl f% ' % Ql1 D ptittas; ::o yp..y .. '
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Specimen P24 550*F l Figure 5-13 (Continued) 5-28 RM-18103
Curve 75%27 -A
. C
- 50 0 50 100 150 200 250 300
. 120 i i i i i i i I -
800 110 -
700 100 -
600g 3
. v, 80 -
. g M 8 0 -
500 -
m 70 W
= -
60 - 2 4 3- 400 2
50 0.2 % Yield Strength i i i i i i ir 300 40 Code:
Open Points -Unirradiated 19 2 Closed Points -Irradiated at 2.89 x 10 n/cm M i j i i i
^
^
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g!_ Reduction Area -
3
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s 40 -
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o 'e .
= ^ -e 20 -
^ 9 10 - : 9 :
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0
-100 0 100 200 300 400 500 600 l
Temperature ( F)
Figure 5-12 Tensile Properties for Kewaunee Reactor Vessel Shell Forging
- 123X167 VA1 (Tangential Orientation) su2.-mw o 5.gg L_
'" ' * ..,>' i ' h,- h.
6 3 ,. -t !icM L
'~
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l
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' Tg'iUf4)Mmml/nt 9l>aw yJ n-rab am q 4 .,a.
ea 6:1#; -shd Specimen P23 150*F Figure 5-13 Fractured Tensile Specimens from the Kewaunee Reactor Vessel Shell ,
Forging 122X208 VA1 (Tangential Orientation) nn.-,u,n s o 5-27 RM-18102
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Figure 5-13 (Continued) 5-28 RS 18103
1
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,pg
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Specimen S18 550*F Figure 5-14 Fractured Tensile Specimens from the Kewaunee Reactor Vessel Shell Forging 123X167 VA1 (Tangential' Orientation) mu-imen io 5-29 RM-18104 a
1 l
1 i
l 120 i i i l 100 - -
80 - -
s
- M . _
- N E i
$ 40 - - -
Specimen P-24 -
(550*F) 4 0 ' ' i i 0 0.05 0.10 0.15 0.20 0.25 Strain, in/in Figure 5-15. Typical Stress-Strain Curve for Tension Specimens l-l no.-isma io 5-30
SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY O
~
6.1 INTRODUCTION
Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor >
pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally mot by employing a combination of rigorous analytical techniques and j measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from .
analysis.
I The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy
'i l dependent damage function for data correlation, ASTM Standard Practice E853,
" Analysis and Interpretation of Light Water Reactor Surveillance Results," .
3342s-162168 10 g.}
recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference [6]. The energy dependent dpa function to be used for this evaluation is specified in ASTH q Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels l
~
in Terms of Displacements per Atom." The application of the dpt. parameter to f the assessment of embrittlement gradients through the thickness of the I pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials."
This section provides the results of the neutron dosimetry evaluations i performed in conjunction with the analysis of test specimens contained in surveillance capsule P. Fast neutron exposure parameters in terms of fast -
neutron fluence (C > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and i iron atom displacements (dpa) are established for the capsule irradiation history. The analytical fornalism re~ating the measured capsule exposure to the exposure of the vessel wall is described and used to ; oject the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at tha surveillance capsule and with the
, projected exposure of the pressure vessel are provided.
6.2 DISCRETE ORDINATES ANALYSIS A plan view of the reactor geometry 4t the core midplane is shown in Figure 4-1. Six irradiation capsules attached to the thermal shield are included in the react r design te constitute the reactor vessel " surveillance program. The capsules are located at azimuthal angles of 77*, 67*, 57*, 257*, 247*, and 237' relative to the core cardinal area as shown in figure 4-1.
A plan view of a surveillance capsule holder attached to the thermal shield is shown in figure 6-1. Tha stainless steel specimen containers are 1-inch square and approximately 64 inches in height. The containers are positioned j axially such that the specimens are centered on the core midplane, thus spanning the central 5.33 feet of the 12-foot high reactor core.
I 3342s.I20688 10 g_g
.__________________________________________m
F From a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect en both the distribution of noutron flux and the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel. In order to properly determine the neutron ~ '
environment at the test specimen lo:atiors, the capsules themselves must be included in- the analytical model.
In performing the fast neutron exposure evaluations for the surveillance
, capsuies and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single coaoutation in the conventional forward modo, was used prirr.arily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radia'l distributions of exposure parameter's (#(E > 1.0 Mev,) $(E > 0.1 Mev),
[
I and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosinetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter i ratios; i.e., dpa/d(E > 1.0 MeV), within the pressure vessel geometry. ,
The relative radial gradient information was required to permit the projection of measured exposure parameters to lece,tions interior to the pressure vessel ,
l wall; i.e., the 1/4T,1/2T, and 3/4T locations.
I The second set of calculations consitted of e. series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from tSese adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with plant specific neutron source distributions, yielded absolute predictions of neutron exposure at the
! locations of interest and established the means to perform similar predictions and dosimetry evaluations for future exposure.
O mwoime io 6-3
9
'i
.yll ' . J I
b n
'[ The: absolute plaitt' specific ~ data from the aSjoint evaluations together with relative. neutron energy spectra.and ra' dial: distribution information'from the.
. forward calculation provided.the means to:-
l'. '
'1. Evaluate neutron dosimetry Ebtained from surveillance, capsule locations.
- 2. Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.
a I
- 3. Enable a direct comparison of analytical prediction with measurement.
The forward transport calculation for the reactor model summarized in figures 4-1 and 6-1 was carried out'in R, e geometry using the DOT two-dimensional-discrete ordinates code [7] and the SAILOR cross-section library (8]. The
. SAILOR library--is a 47 group ENDFB-IV based data set produced specifically for
. light water reactor applications. In these analyses anisotopic scattering was treated with a P3. expansion of the cros;-sections and the angular
. -discretization was modeled with in S8 rder of angular quadrature.
Theireference core power distribution utilized in the forward analysis was derived'from statistical studies of long-term operation of Westinghouse 2-loop plants. . Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty' derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely
'that a single reactor would have a power distribution at the nominal +2a level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.
nu.-noma in 6-4
All adjoint analyses were also carried out using an S8 rder of angular quadrature and the P3 cross-section approximation from the SAILOR library. , l Adjoint source locations were chosen at several azimuthal locatiens along the pressure vessel inner radius as well as the geometric center of each ,
surveillance capsule. Again, these calculations were run in R, e geometry '
to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, 4 (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as:
R (r, 0) = /r #0 #E I(r, 0, E) S (r, 0, E) r dr de dE where: R(r,0) = $ (E > 1.0 MeV) at radius r and azimuthal angle 0 I (r, 0, E) = Adjoint importance function at radius, r, azimuthal angle 0, and neutron source energy E.
-1 S(r,0,E) = Neutron source strength at core location r, e and energy E. .;
Although the adjoint importance functions used in the analysis were based on a response function defined by the threshold neutror, flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading j patterns significantly impact the magnitude and the spatial distribution of I the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/$ (E > 1.0 MeV) is insensitive to changing core source distributions. In the application of I these adjoint important functions to the Kewaunee reactor, therefore, calculation of the iron displacement rates (dpa) and the neutron flux (E >
0.1 MeV) were computed c.,n a plant specific basis by using dpa/4 (E > 1.0 MeV) and e (E > 0.1 MeV)/# (E > 1.0 MeV) ratios from the forward analysis in conjunction with the plant specific 4 (E > 1.0 MeV) solutions from the individual adjoint evaluations. .
I 23u.-u m aio 6-5 l l
1 The reactor core power distribution used in the plant specific adjoint calculations represented a time weighted average over the first 13 fuel cycles.
Selected results from the neutron transport analyses performed for the Kewaunee reactor are provided in tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall.
In table 6-1, the calculated exposure parameters (# (E > 1.0 MeV),
- (E > 0.1 MeV), and dpa) are given at the geometric center of surveillance capsule P using plant specific core power distributions and averaging over cycles 1-13. These plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis and to provide an evaluation of the more recent loading patterns appropriate for projecting into the future. Similar data are given in table 6-2 for the pressure vessel inner radius. Again, the three
- pertinent exposure parameters are listed for the cycle 1-13 average plant t.pecific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself.
Radial gradient information for neutron flux (E > 1.0 MeV), neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in tables 6-3 through 6-5.
For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 0*
, azimuth is given by:
9
= 4(168.04, 0*) F (172.23, 0')
1/4T(0')
m2.o rma io 6-6
s .
1 where = Projected neutron flux at the'1/4T position
- 1/4T(0')
on the O' azimuth 4
d'
- (168.04,O*) = Projected or calculated neutron flux at the .
vessel inner radius on the O' azimuth.
F(172.23,O*) = Relative radial distribution function from
}
table 6-3.
Similar expressions apply for exposure parameters in terms of #(E > 0.1 MeV) and dpa/sec.
6.3 NEUTRON DOSIMETRY The passive neutron sensors included in the Kewaunee surveillance program are listed in table 6-6. Also given in table 6-S are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of ,
the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest (# (E > 1.0 .
Mev), # (E > 0.1 MeV), dpa).
The relative locations of the neutron sensors w'ithin the capsules are shown in figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium shielded neptunium and uranium fiss'.on monitors were accommodated within the dosimeter block iccated near the conter of the capsule.
The use of passive monitors such as those listed in table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest.
Rather, the activation or fission process is a measure of the ir,tegrated effect that the time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessnent of ,
the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well .
known. In particular, the following variables are of interest:
au,mme io 6-7
d
{
o The specific activity of each meniter. i o The operating history of the reacter.
o The energy response of the monitor. 1 o The neutron energy spectrum at the monitor location, o The physical characteristics of the monitor.
The specific activity of each of the neutron monitor-s was determined using j established ASTM procedures (6, 9-21]. Following sample preparation and weighing, the activity of each monitor was determined by means of a f lithium-drif ted germanium, Ge(Li), gamma spectrarreter. The irradiation history of the Kewaunee reactor during cycles ?.-13 was obtained frem NUREG-0020, " Licensed Operating Reactors Statt.s Summary Report" for the j applicable period. 4 The irradiation history applicable to capsule P is given in table 6-7.
Measured and saturated reaction product spacific activities as well as l
. measured full power reaction rates are listed in table 6-8. Reaction rate values were derived using the partinent data from te.bles 6-6 and 6-7, ,
Values of key fast neutron exposure paranvaters were derived frorr the mee.sured reaction rates n ing the FERRET least squares adjustment code (22]. The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrurr. at the centor of the surveillance capsule as input and pecceeded to a6just the a priori (calculated) group fluxes to produce a best fit (in a least squares sanse) to the reaction rate data. The exposure l parameters along with associated uncertainties were then obtained from the ,
adjusted spectra. 1 t
In the FERRET enluations, c. log nornal least-squares algorithm weights both the a priori values and the measured clata in accordance with the assigned uncertainties and correlations. In general, the measured values ' are linearly related to the flux 4 by same response matrix A; r (s,a) = I A. (s)
,9 (a) i g 1g m2.- io 6-8 l
i i
s I
where i 'irdexes the measured values belonging to a single data set s, g designates the energy group and a delineates spectra that may be .
simultaneously adjusted. For example,
~
R I
$=9 o$g p g relates a set of measured reaction rates R$ to a single spectrum #g by ,
} the multigroup cros's section o$g. (In this case, FERRET also adjusts the cross-sections.) The log normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.
1 In tha FERRET analysis of the dosimetry data, the continuous quantities (i.e.,
fluxes and cross-r.ections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code (23]. This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620 point spectrum was then easily collapsed to the group < scheme used in FERRET.
The cross-sections were also collapsed into the 53 energy group structure using SAND II with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/B-V dosimetry file.
Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.
For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight. In some cases, as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form is used:
M gg, = Rf + Rg R,P g gg, .
m2.-> 2eu io 6-9
g, '*-
1
'l where NR specifies an overall fractional normalization uncertainty (i.e., !
- complete correlation) for the corresponding set of values, ~ The fractional
~
uncertainties Rg specify' additional random uncertainties for group g that are' correlated with a correlation matrix: ,
o P
gg, = (1 - 0) 6gg + 0 exp [- (g-g')']
2r The first term specifies purely random uncertainties while the second term describes short-range correlations over a range r (0' specifies the strengthofthelatterterm.)
For the a priori calculated fluxes, a short-range correlation of r = 6 groups was used. This choice implies that neighboring groups are strongly correlated when e is close to 1. Strong long-range correlations (or
- anticorrelations) were justified based on information presented by R. E. I Maerker-[24]. Maerker's results are closely duplicated when r'= 6. For the
~
integral. reaction rate covariances, simple normalization.and random uncertainties were combined as deduced from experimental uncertainties.
Results of the FERRET evaluation of the capsule P dosimetry are given in table 6-9 .The data summarized in table 6-9 indicated that the capsule received an 19 integrated exposure of 2.89 x 10 n/cm2 (E > 1.0 MeV) with an associated uncertainty of + 8%. Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit.of the adjusted spectrum are provided in table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is q tabulated in table 6-11 for the FERRET 53 energy group structure.
(
e s #
..... . 3 10 l
l l j
A summary of the measured and calculated neutron exposure of capsule P is presented in table 6-12. The agreement between calculation and measurement '
falls within 7% for all fast neutron exposure parameters, whereas, the thermal neutron exposure calculated for capsule P was within 10% of the measured value. l
-l Neutron exposure projections at key locatione on the pressure vessel inner radius are given in table 6-13. Along with the current (11.0 EFPY) exposure derived from the capsule P measurements, projections are also provided for an exposure period to end of vessel design life (32 EFPY). The calculated exposurs rates given in table 6-2 were used to perform projections beyond the end of cycle 13.
In the calculation of exposure gradients for use in the development of heatup and tooidown curves for the Kewaunee reactor coolant system, exposure projections to 20 EFPY and 32 EFPY were employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in tab'e 6-14. In order to access RTN9T vs. ,
fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions wert defined by the relations ,
dps(1/4T) 6' 1/4T = + (Surface) { )
cpa (Surface) dpa (3/4T) t' 3/4T = 4 (Surface) { )
Tpa (Surface)
Using this approach results in the dpa equivalent fluence values listed in table 6-14.
In table 6-15 updated lead factors are listed for each of the Kewaunee surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules.
O an.o m e. io 6-11
^ - - - - - - __ _
PRESSURE VESSEL SURVEILL.ANCE CAPSULE 0'
13* (CAPSULES V,R)
/ ' 23* (CAPSULES T,P)
THERMAL 33* (CAPSULES S,N) .
SHIELD l Yllflll/
. YHHmuty ,'
4go F//////////////////
j
- JHHHHH/N
/ /
I,/
~ '
I / ,,,,,,,,,,,,
/
/ /
/ l/
c
,[/ / -REACTOR CORE l///'
///
/
t
- F'igure 6-1, Surveillance Capsule Geometry i
un.-tuiu in 6-12
~----- ----- m m
~,.-
7-kL YJ ",ic; s
,a,
+) s. .> ,i < ,
.... e a, -
- g. .. .q
-n- ..
, . ;v ;
q l 1 s "
(.+ . ,
s ,
-.j j
TABLE 6-1: n. , ' >;
-v ',
s 4
3 1' -
- CALCULATED FAST NEUTRON EXPOSURE PARAMETERS-
- 'l s.
. i- AT THE CENTER OF CAPSULE P-
- 'l' , 'j .
- - i i
- Cycle'l-13(a)4
..p, 10.
,(E>il0MeV)(b)-
. '7.66 x 10 11
((E>"0.1'MeV)(b). 2.74 x 10 dpa/sec 1.33 x 10-10 4
j .-
[(a)~AveragedovertheplantspecificexposureforcapsuleP. ,
'I.
.e 2-.
(b) n/cm _ sec
=
me.-immo 6-13
a TABLE 6-2 l-CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT
,~
THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE-AVERAGED OVER CYCLES 1-13 0* 15' 30' 45' 10 1,75 x 10 10 10
- (E> 1.0Mev)(a) 3.76 x 10 10 2.35 x 10 1.58'x 10 10 6.24 x 10 10 4.65 x 10 10 4,1g x 10 10
- (E> 0.1Mev)(a) 9.98 x 10 dpa/sec. 6.22 x 10
-11 3.89 x 10
~11 2.90 x 10 -11 2.61 x 10 ,
2 (a). n/cm - see e
aus.-isme.io 6-14
f . . .
TABLE 6-3 1' RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX.(E > 1.0 MeV)
WITHIN THE PRESSURE VESSEL WALL ,
Radius (cm) 'O' 15' '30' 45' 168.04(I) 1.00 1.00 1.00 '1.00 168.71 0.935 0.938 0.936 0.937 170.12 0.816 0.817 0.814 0,818 171.53 0.680 0.689 0.683 0.691 172.94 0.563 0.573 0.566 0.574 174.35 0.462 0.473 0.465- 0.473 175.75 0.376 0.388 0.380 0.388 177.16 0.305 0.316 0.309 0.316 ..
178.57 0.246 0.256 0.250 0.256 '!
179.98 0.196 0.206 0.201 -0.206 ,
181.39 0.155 0.164 0.160 0.164 182.80 0.118 0.128 0.125 0.129 0.0946 0.104 0.103 0,,105
.183.83-184.80(2) 0.0857 0.0967 0.0956 0.0982 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius suri-irossaio 6-15
,. . o
!;.4 3 TABLE 6-4' L
RELATIVE RADIAL-DISTRIBUTIONS OF NEUTRON FLUX (E p 0.1.MeV)
'WITHIN THE PR. ESSURE' VESSEL WALL -
i Radius (cm) .
O' 15' . . .
30' ' 45' 168.04(1) 1.00 1.00 1.00 1.00:
168.71 1.00 1.00 1.00 1.00 170.12 0.964' -0.977 0.981 0.985-171.53 0.901 0.915 0,,922 0.930 ,
172.94 0.828 0.844 0,.852 'O.862 174.35- 0.752 0.770 0.779 0.790 175.75 0.675 0.696 0.705 0. ;'18
.. 177.16 0.600 0.622 0.632 0.645 178.57 0.526 0.550 0.561 ~ 0.574
- 179.98 0.454 0.479 0.491. 0.503 181.39 0.383 0.409 0.422 0.434 182.80 0.310 0.338 0.354 0.365 183.83 0.256 0.285 -0.303 0.314 184.80(2) 0.234 0.267 0.287 0.298 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius ,
e i
2m..uom no 6-16
x, a
f m 39 p. pm -l4 p,>e 9as. ..
w, ;
'. 49 g f 3b
,1, g3
.ng
- 3 gjp
"-Q Q:
TABLE 6-5' RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE.(dpa') l 1
- WITHIN THE PRESSURE VESSEL' WALL' .
.f
?i; V~~
L Radius p
, ' m.
, (cm) --
0* 15' 30*' 45*-
4 hi: 168.04 O) 1.00 1.00. 1.00 1.00 q 168.71- 0.944 0.947 0.945 0.946' IL
~170.12 0.832 0.833 0.830' O.834
,171.53 0.714 0.723 0.717 0.726 172.94 0.625 0.536 0.628 0.637 174'35-
. 0.545 0.558 0.549 0.558-175.75 0.466. 0.481 0.471 0.481 177.16 0.400 0.414 .0.405 0.414 ,
'178.57 0.344 0.358 0.350 0.358 l 179.98 0.290 0.305 0.297- 0.305- ..
181.39 0.243- 0.257 0.251 0.257 182.80 0.196 0.212 0.208 0.214 183.83 0.163 0.179 0.177 0.181 184.80(2) 0.154 0.174 0.172 0.177 l
NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius I
L sm.-noen io 6-17
,s ae:
i TABLE 6-6
~ NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS {
1 tv Reaction Target Monitor of Weight Response Product Material- Interest- Fraction Ranga Half-Life Copper CuS3(n,a)Co60 0.6917 E> 4.7 MeV 5.272 yrs-Iron Fe54(n,p)Mn54 0.056 E> 1.0 MeV 312.2 days Nickel NiS8(n,p)CoS8 0.6827 E> 1.0 MeV 70.91 days Uranium-238* U238(n,f)Cs137 1.0 E> 0.4 MeV 30.17 yrs 6.0 Neptunium-237* Np237(n,f)Cs137 1.0 E> 0.8 MeV 30.17 yrs 6.5 Cobalt-Aluminum
- CoS9(n,r)Co60 0.0015 0.4ev<E< 0.015 MeV 5.272 yrs Cobalt-Aluminum CoS9(n,r)Co60 0.0015 E< 0.015 MeV 5.272 yrs
- Denotes that monitor is cadmium shielded.
O nu.->uies ta 6-18
\
TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS '
CONTAINED IN CAPSULE P Irradiation P(a) p,(a) ,
d 3 Irradiation Decay (b)
Period Month Year (MWt ) P Time (days) Time (days)
Rd.
4 1974 297.6 0.1804 17 5126 5 1974 807.0 0.4891 31 5095 6 1974 1201.3 0.7280 30 5065 7 1974 1044.2 0.6328 31 5034 8 1974 1575.4 0.9548 31 5003 9 1974 910.9 0.5520 30 4973 10 1974 426.9 0.2587 31 4942 11 1974 1044.7 0.6332 30 4912 12 1974 1220.0 0.7394 31 4881 1 1975 1050.9 0.6369 31 4850 3
2 1975 1381.4 0.8372 28 4822 3 1975 1474.4 0.8936 31 4791 4 1975 1149.2 0.6965 30 4761 5 1975 1175.4 0.7124 31 4730 6 1975 1080.3 0.6547 30 4700 7 1975 1132.2 0.6862 31 4669 .
8 1975 1580.1 0.9576 31 4638 9 1975 839.1 0.5086 30 4608 10 1975 1232.5 0.7470 31 4577 -
11 1975 1143.5 0.6930 30 3547 12 1975 1574.7 0.9544 31 4516 1 1976 1486.8 0.9011 31 4485 2 1976 720.7 0.4368 29 4456 3 1976 0.0 0.0000 31 4425 4 1976 523.7 0.3174 30 4395 5 1976 920.9 0.5581 31 4354 6 1976 474.4 0.2875 30 4334 7 1976 1587.0 0.9618 31 4303 8 1976 1600.8 0.9702 31 4272 9 1976 1473.4 0.8930 30 4242 10 1976 1611.4 0.9766 31 4211 11 1976 1591.5 0.9645 30 4181 12 1976 1595.6 3.9670 31 4150 1 1977 815.6 0.4943 31 4119 2 1977 0.0 0.0000 28 4091 3 1977 241.6 0.1464 31 4060 4 1977 1388.5 0.8415 30 4030 5 1977 1597.5 0.9682 31 3999 6 1977 1606.2 0.9734 30 ?M9 7 1977 1614.6 0.9786 31 3338 -
8 1977 1410.7 0.8550 31 3907 9 1977 1632.0 0.9891 30 3877 10 1977 1622.4 0.9832 31 3846 -
11 1977 1632.9 0.9896 30 3816 12 1977 1605.0 0.9727 31 3785 3m.-isma ' 6-19
TABLE 6-7 (cont.) -
IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE P __
. Irradiation P.(a) p (a) d Period J Irradiation Decay (b)
Month Year (MW) Pp ,f, Time (days) Time (days) t i
1 1978 1637.7 0.0926 31 3754 2 1978 1634.5 0.9906 28 3726 3 1978 1618.1 0.9807 31 3695 4 1978 1127.6 0.6834 30 3665 5 1978 97.4 0.0590 31 3634 6 1978 1412.3 0.8560 31 3604 7 1978 1518.6 0.9203 31 3573 8 1978 1590.0 0.9636 31 3542 9 1978 1579.2 0.9571 30 3512 10 1978 1623.8 0.9841 31 3481 11 1978 1568.6 0.9507 30 3451 12 1978 1613.7 0.9780 31 3420 1 1979 1629.1 0.9873 31 3389 2 1979 1558.9 0.9448 28 3361 3 1979 1526.5 0.9252 31 3330 1599.4 0.9693 30 3300 4 1979 5 1979 1316.4 0.7978 31 3269 6 1979 0.0 0.0000 30 3239 7 1979 0.0 0.0000 31 3208 8 1979 994.8 0.6029 31 3177 9 1979 1564.8 0.9483 30 3147 10 1979 1592.2 0.9650 31 3116 11 1979 1621.5 0.9827 30 3086 12 1979 1616.5 0.9797 31 3055 1 1980 958.5 0.5809 31 3024 !
2 1980 1616.1 0.9795 29 2995 3 1980 1643.9 0.9963 31 2964 4 1980 1632.6 0.9895 30 2934 5 1980 469.9 0.2848 31 2903 6 1980 191.6 0.1161 30 2873 7 1980 1595.2 0.9668 31 2842 8 1980 1512.6 0.9167 31 2811 9 1980 1314.4 0.7966 30 2781 10 1980 1587.5 0.9621 31 2750 11 1980 1644.4 0.9966 30 2720 12 1980 1597.4 0.9681 31 2689 1 1981 1644.4 0.9966 31 2658 2 1981 1604.1 0.9722 28 2630 3 1981 1528.1 0.9261 31 2599
- 4 1981 1079.4 0.6542 30 2569 5 1981 0.0 0.0000 31 2538 i 6 1981 1079.8 0.6544 30 2508 7 1981 1618.6 0.9810 31 2477 l 8 1981 1643.1 0.9958 31 2446 9 1981 1613.7 0.9780 30 2416 sm.osaise io 6-20 J
0 TA3LE 6-7 (cont.)
IRRADIATION HISTORY OF NEUTRON SENSORS
- CONTAINED IN CAPSULE P Irradiation P (a) p,(a) ,
Period 3 3 Irradiation Decay (b)
Month Year (MWt ) Pp ,f, Time (days) Time (days) 10 1981 1573.6 0.9537 31 2385 11 '1981 1594.7 0.9665 30 2355 12 1981 1638.,8 0.9932 31 2324 1 1982 1617.6 0.9803 31 2293 2 1982 1600.0 0.9697 28 2265 3 1982 1646.0 0.9976 31 2234 4 1982 452.0 0.2744 30 2204 5 .1982 280.4 0.1699 31 2173 6 1982 1615.8 0.9793 30 2143 7* 1982 1637.3 0.9923 31 2112 8 1982 1643.0 0.9958 31 2081 9 1982 1639.9 0.9939 30 2051 10 1982 1632.8 0.9896 31 2020 11 1982 1636.5 0.9918 30 1990 12 1982 1448.2 0.8777 31 1959 1 1983 1637.7 0.9926 31 1928 .
2 1983 1641.0 0.9946 28 1900 3 1983 873.0 0.5291 31 1869 4 1983 0.0 0.0000 30 1839 -
5 1983 707.2 0.4286 31 1808 6 1983 1631.4 0.9887 30 1778 7 1983 1557.8 0.9441 31 1747 8 1983 1645.1 0.9970 31 1716 9 1983 1644.7 0.9968 30 1686 10 1983 1645.3 0.9971 31 1655 11 1983 1644.2 0.9965 30 1625 12 1983 1645.8 0.9975 31 1594 i 1 1984 1647.7 0.9986 31 1563 2 1984 1637.0 0.9921 29 1534 3 1984 716.3 0.4342 31 1503 4 1984 0.0 0.0000 30 1473 5 1984 1125.6 0.6822 31 1442 6 1984 1589.7 0.9635 30 1412 7 1984 1612.0 0.9770 31 1381 8 1984 1642.9 0.9957 31 1350 9 1984 1643.7 0.9962 30 1320 10 1984 1619.6 0.9816 31 1289 11 1984 1643.8 0.9962 30 1259 12 1984 1642.8 0.9956 31 1228 1 1985 1645.5 0.9973 31 1197 -
2 1985 403.3 0.2445 28 1169 3 1985 0.0 0.0000 31 1138 4 1985 948.2 0.5747 30 1108 -
5 1985 1629.6 0.9876 31 1077 6 1985 1638.2 0.9928 30 1047 nn.-,wss ,o s.21
m R
--g
. .n a i'
t , [. 4 i 'O yy }f .
1 f f TABLE 6-7 (cont.).
' IRRADIATION HISTORY OF NEUTRON SENSORS-
' CONTAINED IN CAPSULE P Irradiation, P (a)' p (a) 3 Period d Irradiation. ' Decay (b)
Month Yoan (MW). P Time (days) Time-(days) t Ref.
7 1985 ~
1638.1 0.9928 31 1016 8 1985 '
1543.7 0.9356 31 '985 9 -1985 1636.2 0.9916 , 30 955 10 1985 1643.8 0.9962 31 924 11- 1985 1557.1 0.9437 30- 894-
'12 1985 1587.8 0.9623 31 863 1 .1986- 1636.9 0.9921 31' 832 2' 1986 1587.8 0.9623 28 804
.3 1986 0.0 0.0000' 31 773
-4 1986 372.1 0.2255 30 743 5 1986 1601.9 0.9709 31 712
- 6. 1986 1634.6 0.9907 30 682 7 1986 1637.3 0.9923 31 651 8 1986 1585.8 0.9611 31 620, 9 1986 -1639.2 0.9934 30 590 10 1986 1615.7 0.9792 31- 559 11 1986 1637.7 0.9925 30 529
^
-12 1986 1634.2 0.9904- 31 498 1 1987 1634.0 0.9903- 31 .467 2 1987 1376.0 0.8339 28 439 3 -1987 0.0 0.0000 31 408 4 1987 1332.4 0.8075 30 378 5 1987 1593.9 0.9660 31 347 6 1987 1592.8 0.9653 30 317 7 1987 1591.0 0.9643 31 286 8 1987 1638.5 0.9930 31 255 9 1987 1622.3 0.9832 30 225 10 1987 1639.5 0.9936 31 194 11' 1987 1634.7 0.9907 30 164 12 1987 1633.5 0.9900 31 133 1 1988 1636.5 0.9918 31 102 2 1988 1630.e. 0.9881 29 73 3 1988 1205.9 0.7309 2 71
(
(a) P 3 si average power in period j; Pref is 1650 MW.
~
(b) Decay time is relative to the dosimetry counting reference date, 5/12/88.
i um-ima io 6-22
1 t
TABLE 6-8=
MEASURED SOR ACTIVITIES AND REACTION RATES Measured: Reaction'.
- Monitor.and . Activity Saturated Activity (a). Rate (rps/ nucleus Axial Location iBq/cm Bq/gm _.
'Cu-63 (n,a) Co Top-Middle 2.12 x 10 5 2.98 x 10 5 5
Bottom-Middle. 2.44 x 10 5 3.43 x 10 Average- 2.28 x 10 5 3.20 x 10 5 4.89 x:10 ~17 -
Fe-54 (r.,p) Mn-54 6 6 Top 2.74 x 10 4.14 x 10 "
6 Top-Middle 1.95 x 10 6 2.95 x 10 Middle '2.11 x 10 6
3.19 x 10 6-Bottom-Middle 2.19 x 10 6
3.31 x 10 6 Bottom' 2.33 x 10 6 3.52.x 10 6
-15 Average 2.26 x 10 6
3.42 x'106 5.45 x 10 Ni-58 (n,p) Co-58 Hiddle 2.44 x 10 7
4.83 x 10 7
6.89 x 10 -15 6
m w uonein 6-23
, o TABLE 6-8 (Cont'd)
MEASURED SENSOR ACTIVITIES AND REACTION RATES Measured Reaction Saturated Activity (8) Rate Monitor and Activity Axial Location Bq/gm Bq/gm (rps/ nucleus)
Np-237 (n,f) Cs-137 6 7 -13 Middle 8.28 x 10 3.78 x 10 2.29 x 10 U-238 (n,f) Cs-137 6 6 -14 Middle 1.13 x 10 5.16 x 10 3.41 x 10 Co-59 (n,r) Co-60 7 7 Top 4.20 x 10 6.40 x 10 7 7 Bottom 5.01 x 10 7.63 x 10 Average 4.61 x 10 7
7.01 x 10 7
4.58 x 10 -12 Co-59 (n,r) Co-50 (Cd)
Top 1.79 x 10 7
2.73 x 10 7 7 7 Bottom 2.04 x 10 3.11 x 10 Average 1.92 x 10 7
2.92 x 10 7
1.92 x 10 -12 (a) Adjusted z.o center of surveillance capsule.
S nn.-maa in 6-24
. m
.. o 'c
-},
f
-TABLE 6-9
~
SUMMARY
OF NEUTRON, DOSIMETRY RESULTS
';; \
TIME AVERAGED EXPOSURE RATES-8.27 x 10 10 2
3 - #(E'>1.0'MeV)(n/cm-sec) ~ 8%
11 2
v'(E>0.1MeV)(n/cm-sec) 2.87 x.10 - 15%
-10 dpa/sec- 1.40 x 10 10%
10 p(E < 0.414 eV) (n/cm2-sec) 9.16 x 10 19%
INTEGRATED CAPSULE EXPOSURE 2.89 x 10 19 2
4 (E > 1.0 MeV) (n/cm ) 18%-
2 20 6-(E > 0.1 MeV) (n/cm ) 1.00 x 10 15%
-2 dpa 4.90 x 10 10%
)
2 19 t (E < 0.414 eV)'(n/cm ) 3.20 x 10 .1 19%
[
NOTE: Total Irradiation Time = 11.08 EFPY i
mamom io 6-25 c: -_ __ - - . - - - - - - - - - - - _ _ - _ _ _ - _ - - _ - _
j' .4
, .t l4;f" TABLE.6 10 1.' ~COMPARIS0N OF MEASURED AND' FERRET CALCULATEDLY
$' REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER L
. Adjusted-Reaction Measured ' Calculation C/M
-17 ~17
-Cu-63 (n,a) Co-60_ :4.88 x 10 5.03 x'10 1.03
-15 -15 Fe-54 (n.p) Mn-54' 5.45.x 10 5.26 x-10 0.96 Ni-58 (n,p) Co-58; 6.89 x 10 -15 6.92 x 10 1.01
-14 U-238 (n,f) Cs-137 (Cd) 2.69 x 10 2.71 x.10 ~14 1.01 Np-237 (n,f) Cs-137-(Cd) .2.29 x 10 -13. 2.32 x 10 -13 1.01 Co-59 (n,r) Co-60 4.44 x 10 -12 4,44 x.10 -12 1.00
-12 -13 Co-59 (n,r) Co-60 (Cd) 1.85 x 10 1.85 x 10 1.00 1
4 nu.-ieves to 6-26 l
k . .
l' L r TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT-q: THE SURVEILLANCE CAPSULE CENTER
~
Energy Adjusted Flux Energy Adjusted Flux 2 2 Group. (MeV) (n/cm -sec) Group (MeV). .(n/cm -sec) 1 1.733E+01 6.87E+06 28 9.119E-03 1.17E+10 2 1.492E+01 1.55E+07 29 5.531E-03 1.42E+10 3 1.350E+01 6.00E+07 30 3.355E-03 4.44E+09 4 1.162E+01 1.36E+08 31 2.839E-03 4.22E+09 5 1.000E+01 3.04E+08 32 2.404E-03 4.10E+09 6 8.607E+00 5.34E+08 33 2.035E-03 1.20E+10 7 7.408E+00 1.27E+09 34 1.234E-03 1.20E+10 8 6.065E+00 1.89E+09 35 7.485E-04 1.19E+10 9 4.966E+00 4.00E+09 36 4.540E-04 1.18E+10 10 3.679E+00 5.15E+09 37 2.754E-04 1.24E+10 11 2.865E+00 1.03E+10 38 1.670E-04 1.40E+10 12 2.231E+00 1.32E+10 39 1.013E-04 1.29E+10 ,
13 1.738E+00 1.74E+10 40 6.144E-05 1.27E+10 14 1.353E+00 1.79E+10 41 3.727E-05 1.23E+10 15 1.108E+00 3.11E+10 42 2.260E-05 1.18E+10 16 8.208E-01 3.275+10 43 1.371E-05 1.14E+10 17 6.393E-01 3.21E+10 44 8.315E-06 1.09E+10 18 4.979E-01 2.24E+10 45 5.043E-06 1.02E+10 19 3.877E-01 2.94E+10 46 3.059E-06 9.69E+09 20 3.020E-01 3.26E+10 47 1.855E-06 9.07E+09 21 1.832E-01 2.99E+10 48 1.125E-06 7.63E+09 22 1.111E-01 2.36E+10 49 6.826E-07 7.68E+09 23 6.738E-02 1.73E+10 50 4.140E-07 1.23E+10 24 4.087E-02 1.08E+10 51 2.511E-07 1.26E+10 25 2.554E-02 1.16E+10 52 1.523E-07 1.28E+10 26 1.989E-02 7.30E+09 53 9.237E-08 5.40E+10 ,
27 1.503E-02 1.00E+10 NOTE: Tabulated energy levels represent the upper energy of each group.
ww, wee so 6-27
.[t
-. TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED .i EXPOSURE LEVELS FOR CAPSULE'P !
\
Calculated Measured , C/M' .j 2
f(E> 1.0 MeV) (n/cm ) 2.68 x 10 19 2.89 x 10 19 0.93 2
f(E> 0.1 MeV). (n/cm ) 9.60 x 10 19 l'.00 x 10 20- 0.96
-2 dpa 4.67 x 10 -2 4.90 x 10 0.95 19 f(E<'O.414 eV) (n/cm )
2 2.87 x 10 19 3.20 x 10 0.90 4
I e
t .
sus.-sw ss ,o s.gg
._. __-___.-_____-______a
s, .- .
c i'. _
L --
TABLE 6-13
-NEUTRON EXPOSURE PROJECTIONS AT~ LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE A2IMUTHAL ANGLE; t 0*(*)- 15' 30' 45'-
11.08 EFPY 10 4 (E> 1.0 MeV)(b) 1.42 x-10 19 8.87 x 10 18 6.60 x 10 18 5.96 x 10
'3 (E> O.1 MeV)(b) 3.77 x 10 19 2.36 x 10 19 1.75 x 10 19 1.58 x 10 19
-2
- dpa 2.35 x 10 -2 1.47 x 10 1.09 x 10 9.87 x.0 -3 ,
4 32.0 EFPY 1.82 x 10 19' 19 19 19 t (E> 1.0 MeV)'(b) 3.90 x 10 2.44 x 10 1.64 x 10 19
+ (E> 0.1 WeV)(b) 1.04 x 10 20 6.48 x 10 19 4.83 x 10 19 4.36 x 10
-2 -2 -2 -2 dpa 6.46 x 10 4.04 x 10 3.01 x 10 2.72 x 10 (a) Maximum point on the pressura vessel ,
2 (b) n/cm nu. ismeaa 6-29
8 8 8 8 9 8 8 8 1 1 1 1 1 1 1 S 0 0 0 0 0 0 0 0 E T 1 1 1 1 T 1 1 1 1 V
R 4 x x x x 4 x x U / /
.- C 3 7 0 7 5 3 5 3 2 6 6 4 1 9 0 9 0 6 N
W 6 4 3 2 1 6 5 4 O -
D L
O O
C -
/ 9 9 8 8 9 9 9 9 P 1 I 1 1 1 1 1 1 U E 0 0 0 0 -
0 0 0 0 T P T 1 1 1 1 E T 1 1 1 1 A O P E L 4 x x v. x O 4 x x x x H S ) / L ) /
2 1 6 5 4 9 S2 1 1 6 2 2 F a p c m 6 0 7 0 a c m 6 6 2 1 O p /
d / 1 1 7 7 2 1 1 1 N n d n O ( (
I T
A R
E 9 9 9 9 9 9 9 9 N 1 1 1 1 1 1 1 1 E e 0 0 0 0 e 0 0 0 0 G c 1 1 1 1 c 1 1 1 1 a a E f x x x x f x x x H r r T u 8 5 5 4 u 0 4 2 4 S 4 5 1 0 S 9 4 8 6 4 N 1 I 2 1 1 1 3 2 1 1 6 E S
E U L
B R
. A O 8 8 8 8 9 9 8 8 T F 1 1 1 1 1 1 1 0 0 0 0 0 0 0 0
) T 1 1 1 1 i 1 1 1 1 V Y Y e P 4 x x x x P 4 x x x N F E / F E /
E P 3 1 0 0 4 E P 3 4 6 3 7 0 O 4 9 1 9 O 9 5 3 0
~
0 L 2 L 1 2 S 4 2 2 1 3 S 6 4 3 3 E ) )
( V V
- e e S M M E
U 0 9 8 8 8 0 9 9 9 9 L ) 1 1 1 1 ) I 1 I 1 A 12 0 0 0 0 12 0 0 0 0 V mT 1 1 1 1 mT 1 1 1 1
> c > c E E / 4 x x x x E / 4 x x x x R ( n / ( n /
U ( 1 4 8 9 8 ( 1 3 4 4 4 S E 5 7 1 5 E 4 5 1 O C C P N 1 9 7 6 N 2 1 1 1 X E E E U U L L N F F O
R N N T O 9 9 9 9 O 9 9 9 9 U R 1 1 1 1 R I 1 1 1 E T e 0 0 0 0 T e 0 0 0 0 N U c 1 1 1 1 U c 1 1 1 a E a
_ L E
E N f r
x x x x N f r
u x x x x S u 8 5 5 4 0 4 2 4 S S 4 5 1 0 S 9 4 8 6 E
V 2 1 1 1 3 2 1 1
) )
_ a a
( (
_ . * *
- e *
- 0 5 0 5 O 5 0 5 1 3 4 1 3
_ d
- v -- -= --------------- ----- - - - - -
q7p
- n. . .
.g.., s
,.\ -
1 i
< .g . ,..
s
-Q , , t .-: . c' ! 1
,e
- j <.
TABLE:6-15
- t. w '
..u f juPDITED'LEADFAC10RS'F0kKEWAUNEE+ -'-
. SURVEILLANCE CAPSULES'
'.6 1
P<, '
Capsule . Lead Factor. y o
V Removed (1.29 EFPY)-
R Removed (4.57EFPY)~
P' . Removed (11.08 EFPY) .
T' 2.04 S. 1.91 o N ' 1. 91.
.5
.4+
.i l-p l
a i
.i l.
l l swoisa io 6-31 l.
i .. ,
l' ." __ _ _ . _ . _ _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ . _ _ _ _____ _._.____ _ _____
a SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHED'ULE The following removal scheduie meets ASTM E185-82 and is recommended for future capsules.to be removed from the Kewaunee reactor vessel:
Estimated Capsule-Vessel Removal. Fluence Location Lead 2 Capsule (deg) Factor Time (*) (n/cm )
V. 77 -
1.29 5.99 x 10 18(b) 2.07 x 10 19(b)
R 257 -
4.57 P 247 -
11.08 2.89 x 10 19(b)'
T 67 2.04 16 4.00 x 10 19(c) 19 S 57 1.91 32 7.45 x 10 N 237 1.91 . Standby --
1 a) Effective full power years frem plant startup .,
b) Actual fluence c) Approximate fluence at vessel inner wall at end of design life (32 EFPY) i
- e t
3342s-t 0196f.:10 7.}
I. _ . _ _ _ - - - - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ . _ - _ _ -
J
a , <
SECTION 8 REFERENCES
- 1. Yanichko, S. E. et. al., " Wisconsin Public Service Corp. Kewaunee Nuclear .
Power Plant Reactor Vessel Radiation Surveillance Program, WCAP-8107, April 1973.
- 2. Yanichko, S. E. et. al. " Analysis of Capsule V from the Wisconsin Public Service Corporation Kewaunee Nuclear Plant Reactor Vessel Radiation
]
Surveillance Program," WCAP-8908, January 1977.
- 4. Code of Federal Regulations, 10CFR50, Appendix G, " Fracture Toughness Requirements" and Appendix H, " Reactor Vessel Material Surveillance ,
Program Requirements," U.S. Nuclear Regulatory Commission, Washington, DC.
- 5. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May, 1988.
I i
- 7. R. G. Soltesz, R. 'K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)-034, Vol. 5, August 1970.
7 3*42s*101988 to g_}
.. _ _ ___ -_-_________________________w
l .
- 8. "0RNL RSIC Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water L* Reactors".
- 11. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
e sm, noms ,o 9.g
i l
American Society for Testing and Materials, Philadelphia, PA,1984.
- 22. F. A. Schmittroth, FERRET Data Analysis Code, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
- 23. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation, AFWL-TR-67-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
l
- 24. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al.,1981.
"**'"'* 8-3
. . .