ML20116E429
| ML20116E429 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 06/30/1992 |
| From: | Chicots J, Medeyski A, Perock J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20116E417 | List: |
| References | |
| WCAP-13365, NUDOCS 9211090169 | |
| Download: ML20116E429 (172) | |
Text
WESTINGHOUSE CLASS 3 WCAP-13365 E
ANALYSIS OF CAPSULE Y FROM THE WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK REACTOR VESSEL RACIATION SURVEILLANCE PROGRAM J. M. Chicots J. Perock A. Madeyski June 1992 Work Performed Under Shop Order KZ0P-106 Prepared by Westinghouse Electric Corporation for the Wolf Creek Nuclear Ope: ating Corporation M
Approved by:
T. A. Meyer, Managkr-Structural Reliability and Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box.355 Pittsburgh, Pennsylvania-15230-0355 C 1992 Westinghouse Electric Corp.
All Rights Reserved jn2 EsM Ziasjjp
PREFACE This report has been technically reviewed and verified.
Reviewer:
Sections 1 through 5, 7, 8 and E. Terek Appendix A Section 6 S. L. Anderson DL 3
['d d i A-Appendix B M. A. Ramirez O
i
TABLE OF CONTENTS Section Title East 1.0
SUMMARY
OF RESULTS 1-1
2.0 INTRODUCTION
2 0
3.0 BACKGROUND
3-1 i
4.0 DESCRIPTION
OF PROGRAM 4-1 4
5.0 TESTING OF SPECIMENS FROM CAPSULE Y 5-1 5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Rcsults 5-4 5.3 Tension Test Results 5-6 5.4 Compact Tension Tests 5-7 6.0 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1
8.0 REFERENCES
8-1 Ap>ENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS
~
APPENDIX B - HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION APPENDIX C - CALIBRATION REPORTS L
11 a...-
o
- )
)
LIST-OF-TABLES
. Ilkl.a lill.it ELift; 4-1 Chemical _ Composition and _ Heat Treatmstt of the' Wolf Creek' 4 -i Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data-for the Wolf._ Creek 5-8 Lower Shell Plate R2508-3 Irradiated at 550'F, Fluence 1.33 x 1019 n/cm2 (E > 1.0 MeV)-
5-2 Charpy V-Notch Impact Data for'the Wolf Creek' 5 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 1.33 x 10 n/cm2 (E->~1.0'MeV) 19 5-3 Instrumented Charpy impact TestLResults for the Wolf.-Creek 5-10 Lower Shell Plate R2508-3 Irradiated at 550*F, Fluence 1.33' x -
1019 n/cm2 (E > 1.0 MeV)-
5-4 Instrumented'Charpy Impact Test Results for the Wolf-Creek' 5-11E Weld Metal and Heat-Affected-Zone- (HAZ)' Metal, Irradiated tj 19 2
at-550'F, Fluence 1.33 x 10 n/cm (E >L1.0 MeV) 5-5 Effect of:5b0*F IrradiationLto 1.33 x 1019ln/cm' 5-127 2
(E > 1.0 MeV) on the Notch Toughness Properties.ofIthe Wolf Creek Reactor Vessel-Surveillance Materials t
0 iii r
's u
c w
lv m,
g-.
i LIST OF TABLES (Continued)
Table Title f_agt a
5-6 Comparison of the Wolf Creek Surveillance Material 5-13 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-7 Tensile Properties for the Wolf Creek Reactor Vessel 5-14 19 Surveillance Materials Irradiated at 550*F to 1.33 x 10 n/cm2 (E > 1.0 MeV) l 6-1 Calculated Fast Neutron Exposure Parameters at the 6-14 l
Surveillance Capsule Center 6-2 Calculated Fast Neutron Exposure Rates at the Pressure 6-15 Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux (E > 1.0 MeV) 6-16 within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux (E > 0.1 MeV) 6-17 within the Pressure Vessel Wall 6-5 Relative Radial Distributions of Iron Displacement Rate 6-18 (dpa) within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-19 6-7 Monthly Thermal Generation During the First Fuel Cycle of 6-20 the Wolf Creek Unit 1 Reactor 8 Measured Sensor Activities and Reactions Rates 6-21 j
iv i
LIST OF TABLES (Continued)
Tahle Title Eitqg 6-9 Summary of Neutron Dosimetry Results 6-23 6-10 Comparison of Measured and FERRET Calculated Reaction 6-24 Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Ener'gy Spectrum at the Surveillance 6-25 Capsule Center 6-12 Comparison of Calculated and Measured Exposure Levels 6-26 for Capsule Y 6-13 Neutron Exposure Projections at Key locations on the 6-27 Pressure Vessel Clad / Base Metal Interface 6-14 Neutron Exposure Values for Use in the Generation of 6-28 Heatup/Cooldown Curves i
6-15 Updated Lead Factors for Wolf Creek Unit 1 Surveillance 6-29 Capsules l
l v
l
LIST OF ILLUSTRATIONS Fiaure Title P122 i.
4-1 Arrangement of Surveillance Capsules in the Wolf Creek 4-4 Reactor Vessel 4-2 Capsule Y Diagram Showing Location of Specimens, Thermal 4-5 Monitors and Dosimeters
+
5-1 Charpy V-Notch Impact Properties for Wolf Creek Reactor 5-15 Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-2 Charpy V-Notch Impact Properties for Wolf Creek Reactor 5-16 Vessel Lower Shell Plate R2508-3 (Transverse Orientation) 5-3 Charpy V-Notch Impact Properties for Wolf Creek Reactor 5-17 Vessel Surveillance Weld Metal 5-4 Charpy V-Notch Impact Properties for Wolf Creek Resctor 5-18 Vessel Weld Heat-Affected-Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for Wolf Crcek 5-19 Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-6 Charpy impact Specimen Fracture Surfaces for Wolf Creek 5 Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation)
{
vi 1
.. _ ~ ~. _ _
i LIST OF ILLUSTRATIONS (Contihued)
Fioure lilljt ELqt I
5-7 Charpy Impact Specimen Fracture Surfaces for Wolf Creek-5-21' Reactor Vessel Surveillance Weld Metal 5-8 Charpy Impact Specimen Fracture Surfaces for Wolf ' Creek 5-22 Reactor Vessel Weld Heat-Affected-Zone Metal-5-9 Tensile Properties for Wolf Creek Reactor Vessel 5-23 Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-10 Tensile Properties for Wolf Creek Reactor Vessel 5-24'-
]
Lower Shell _ Plate R2508-3 (Transverse Orientation) j 5-11 Tensile Properties for Wolf Creek Reactor Vessel-l5-25 Surveillance Weld Metal 5-12 Fractured. Tensile Specimens from Wolf Creek-Reactor 5-26 l
L Vessel Lower Shell Plate R2508-3 (Longitudinal. Orientation) l 5-13 Fractured Tensile _ Specimens from Wolf Creek Reactor
'5 Vessel Lower Shell Plate-R2508-3-(Transverse Orientation)-
1 i
5 Fractured Tensile-Specimens from Wolf Creek Reactor.
5-28.
l
. Vessel Surveillance Weld Metal t
5-15 Engineering Stress-Strain Curves for Plate R2500 5-29 Tensile Specimens All3 and All4 (Longitudinal Orientation)1 t
vii Y
LIST OF ILLUSTRATIONS (Continued)
Fiaure Title Eage 5-16 Engineering-Stress-Strain Curve for Plate R 508-3 5-30 Tensile Specimens AL15 (longitudinal Orientation) 5-17 Engineering Stress-Strain Curves for Plate R2508-3 5-31 Tensile Specimens AT13 and AT14 (Transverse Orientation) 5-18 Engineering Stress-Strain Curve for Lower Shell Plate 5-32 R2508-3 Tensile Specimen ATIS (Transverse Orientation) 5-19 Engineering Stress-Strain Curves for Weld Metal 5-33 Tensile Specimens AW:
and AW14 5-20 Engineering Stress-Strain Curve for Veld Metal 5-34 Tensile Specimen AW15 6-1 Plan View of a Oual Reactor Vessel Surveillance Capsule 6-13 l
viii
^l SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule Y, the second capsule to be removed from the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor pressure vessel, led to the following conclusions:
The capsule received an average fast neutron fluence (E > 1.0 MeV) of o
1.33 x 1019 2
n/cm after 4.79 EFPY of plant operation.
Irradiation of the reactor vessel lower shell plate R2508-3 Charpy o
specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction (longitudinal orientation), to 1.33 x 1019 n/cm2 (E > 1.0 MeV)-resulted in a 30 ft-lb transition temperature increase of 30*F and a 50 ft-lb transition temperature increase of 40*F.
This results in a 30 ft-lb transition temperature of 10*F and a 50 ft-lb transition temperature of 40*F for longitudinally oriented specimens.
Irradiation of the reactor vessel 1:wer shell plate R2508-3 Charpy o
specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction (transverse orientation), to 1.33 x 19 10 n/cm2 (E > 1.0 MeV) resulted in_a 30 ft-lb transition temperature increase of 40*F and a 50 ft-lb transition temperature increase of 45'F.
This results in a 30 ft-lb' transition temperature of 40*F and a 50 ft-lb transition temperature of 85*F for transversely oriented specimens, o
The weld metal Charpy specimens irradiated to 1,33 x 1019 2
n/cm (E > i.0 MeV) resulted in a 30 ft-lb transition temperature increase of 50*F and a 50 ft-lb transition temperature increase of 43*F.
This results in a 30 ft-lb transition temperature of 0*F and a 50 ft-lb transition temperature of 30*F for the weld metal.
I l-1 l
l 4
o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal l9 Charpy specimens to 1.33 x 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lh transition teraperature increase of 50*F and a 50 ft-lb transition temperature increase of 40*F.
This results in a 30 ft-lb tiansition temperature of -95'F at.d a 50 ft-lb transition temperature of -70*F f0r the weld HAZ metal.
o Irradiation of lower shell plate R2508-3 (longitudinal orientation) to 1.33 x 10 n/cm2 (E > 1.0 MeV) resulted in an average upp r l9 shelf energy decrease of 27 ft-lbs, resulting in an upper shelf energy of 121 ft-lbs.
o Irradiation of lower shell plate R2508-3 (transverse orientation) to 1.33 x 1019 n/cm2 (E > 1.0 MeV) resulted in an average upper shelf energy increase of I ft-lb, resulting in an upper shelf energy of 94 ft-lbs, The average upper shelf energy of the weld metal decreased 6 f t-lb o
after irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV). This 19 results in an upper shelf energy of 94 ft-lb for the weld metal, o
The average upper shelf energy of the weld HAZ metal increased 19 19 ft-ib after irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV). This results in an upper shelf energy of 180 ft-lb for the weld H/.Z metal.
o The surveillance capsule Y test results indicate that the lower shell plate R2508-3 30 ft-lb transition temperature shift is less than the Regulatory Guide 1.99 Revision 2 predictions.
However, comparison of the 30 ft-lb transition temperature increase for the surveillance weld material is 15'F greater than the Regulatory Guide 1.99 Revision 2 predictions.
Regulatory Guide 1.99 Revision 2 requires a 2 sigma allowance of 56*F for weld metal be added to the predicted reference transition temperatu.re to obtain a conservative upper bound value. Thus, the reference transition temperature 1-2
increase for the surveillance weld metal is bounded by the 2 sigma-allowance for shift prediction.
o The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are -
expected to maintain an upper shelf energy of'no less than' 50 ft-lb throughout the life (32 EFPY) of the vessel as required by 10CFR50, Append!x G.
o The calculated end-of-life (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the Wolf Creek reactor vessel is as follows:
Vessel inner radius * - 2.50 x 1019 2
n/cm Vessel 1/4 thickness - 1.36 x 1019 2
n/cm Vessel 3/4 thickness - 2.93 x 1018 2
n/cm
- Clad / base metal interface 1-3
m SECTION
2.0 INTRODUCTION
This report prese-ts the results of the examination of Capsule Y, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Wolf Creek reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation.
A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-100lS, entitled
" Kansas Gas and Electric Company Wolf Creek Generation Station Unit No.1 Reactor Vessel Radiation Surveillance Program" by L. R. Singerill.
The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-79, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels".
Westinghouse Power Systems personnel were contracted to aid in the I
preparation of procedures for removing capsule "Y" from the reactor and its shipment to the Westinghouse Science and Technology Center Hot Cell Facility, where, the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This. report summarizes the testing of and the postirradiation data obtained from surveillance capsule "Y" removed from the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor vessel and discusses the analysis of the data.
l l-l l
l 2-1
SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry.
The beltijne region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment.
The overall effects of fast neutron irradiation on the mecnanical properties of low alloy, ferritic pressure vessel steels such as SA 533 Grade B Class 1 (base material of the Wolf Creek reactor pressure vessel lower shell plate R2508-3) are well documented in the literature.
Generally, low alloy ferritic materials snow 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 ft.st fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure,"
Appendix G to Section III of the ASME Boiler and Pressure Vessel Code [5],
The method uses fracture mechanics concepts and is based on the-reference nil-ductility temperature (RTNDT)-
RTNDT is defir.ed as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208)[6] or the temperature 60*F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the plate.
The RTNDT of a given material is used_to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G to the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature.
Allowable operating limits can then be determined using these allowaole stress intensity factors.
3-1
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted tu account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, such as the Wolf Creek Reactor Vessel Radiation Surveillance Programill, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested.
The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial +
ARTNDT) is used to index the material to the KIR 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.
3-2
SECTION
4.0 DESCRIPTION
OF PROGRAM Six surveillanco capsules for monitoring the effects of neutron exposure on the Wolf Creek reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant start-up. The six capsules were positioned in the reactor vessel between the neutron shielding pads and the vessel wall as shown in Figure 4-1.
The vertical center of the capsules is opposite the vertical center of the con e.
Capsule Y was removed after 4.79 Effective Full Power Years (EFPY) of plant operation.
This capsule contained Charpy V-notch, tensile, and 1/2 T compact tension (CT) specimens (Figure 4-2) from the lower shell plate R2508-3 and' submerged arc weld metal representative-of the intermediate to lower shell i
beltline weld seam of the reactor vessel.
Capsule Y also chittained Charpy V-notch specimens from weld Heat-Affected-Zone (HAZ) material. All heat-affected zone specimens were obtained from within the HAZ of plate R2508-3 of the representative weld.
All test specimens were machined from the 1/4 thickness location of the plate.
Test specimens represent mate ul taken at least one plate thickness from the l
quenched end of the plate.
Base metal Charpy V-notch impact specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction uf the plate (longitudinal orientation) and also normal to the najor working direction (transverse orientation).
Charpy V-notch and tensile specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the weld direction.
I l
The CT specimens were machined such that the simulated crack in the specimen would propagate normal and parallel to the major working direction for the l
plate specimen and parallel to the weld direction.
l l
4-1
The chemical composition and heat treatment of the surveillance material is presented in Table 4-1.
Capsule Y contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt wire (cadmium-shielded and unshielde j.
In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were placed in the capsule to measure the integrated flux at specific neutron energy levels.
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two 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)
The location of the surveillance capsules within the reactor vessel is shown in Figure 4-1.
The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in capsule Y is shown in Figure 4-2.
-1 4-2
TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE WOLF-CREEK REACTOR VESSEL SURVEILLANCE MATERIALS L
Chemical-Composition (wt%)'
l Lower Shell Element Plate R2508-3 Weld Metal
[al C
0.20.
0.11-Mn 1.45 1.46 P
0.008 0.005 S
0.010
-0.011 Si 0.20 0.48 Ni 0.62 0.09 Mo 0.55 0.56 Cr 0.05 0.09.
Cu 0.07
-0.04?
Al 0.032
-0.009 Co 0.014 10.010 Pb
<0.0014
<0.001 W
< 0. 01 -
<0.01:
Ti
<0.01.
<<0.01 Zr
<0.001
<0.001 V
0.003 0.005-Sn' O.002 0.003.
As 0.007 20.304 Cb.
<0. 01.
<0.01-N 0.007:
0.006 i-B
' 0.001
<0.001
- Heat Treatment-History Material Temocrature-(*F)
JTime(Hri-
- Coolant Lower Shell Plate, Austenitizing 1575-1625 4
Water quenched-
~R2508 Tempered 1200-1250 4'
- Air cooled-Stress Relief 1100-1200 8.5 Farnace cooled Weld'Hetal Stress Relief 1100-1200 10.25 Furnace cooled 4
This weldment was fabricated by Combustion Engineering, Inc., using 3/16.
a.
inch Mil-B-4 weld filler' wire,- heat number 90146 and Linde 124 flux, lot number ~ 1061-and is. identical sto that L3ed -in the ' actual fabrication of-the
- reactor vessel intermediate to lower ill girth weld.
4-3
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en
O' CORE BARREL (3.05) Z
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NE M ON PAD CAPGULE U (3.85)
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(3.65) Y
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(3.85) X
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210' 180' PLAN VIEW Figure 4-1.
Armnament of Surveillv:e Capsules in the Wolf C eek Reactor Vessel 4-4 L:
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SECTION 5.0 TESilWG OF SPECIMENS FROM CAPSULE Y 5.1 Overview The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse Energy Systems personnel.
Testing was I
performed in accordance with 10CFR50, Appendices C and HI33, ASTM I73, and Westin9 ouse Remote Metallographic Facility Specification E185-82 h
(RMF) Procedure Mhl 8402, Revision 1 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blxks were carefully removed, inspected for identification number, and checked against the master list in WCAP-10015III.
No discrepancies were found.
Examination of the two low-melting point 579'F (30.;'C) and 590*F (310'C) outectic alloys indicated no melting of either type of thermal monitor.
Based on this examinatio,1, the maximum temperature to which the test specimens were exposed was less than 579'F (304'C).
The Charpy impact tests were performed per ASTM Specification E23-88[8] and RMF Procedure 3103, Revision 1 on a Tinius-Olsen Model 74, 358J machine.
The tup (striker) of the Cliarpy machine is instrumented with a GRC 6301 instrumentation system, feeding information into an IBM XT computer.
With this system, load-time and energy-time signals can be recorded in addition to the
,tandard measurement of Charpy energy (E ).
From the load-time curve D
( Appendix A), the load of general yielding (Pgy), the time to general yleiding (tgy), the maximum load (P ), and the time to maximum load (tg) g 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 (P ).
g 5-1 o
The energy at maximum load (Eg) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is roughly equivalent to the anergy required to initiate a crack in the specimen.
Therefore, the propagation energy for the crack (E ) is the difference p
between the total energy to fracture (E ) and the en.rgy at maximum load.
D The yield stress (oy) was calculated from the three-point bend formula having the following expression:
(1) ay - Pgy * (L/(B*(V-a)2*C))
where L distance between the specimen supports in the impact testing machine;
~
B the width of the specimen measured parallel to the notch; W = height of the The constant specimen, measured perpendicularly to the notch; a notch depth.
C is dependent on the notch flank angle (d), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending).
In three-point bending a Charpy specimen in which 4 - 45' and p -
0.010", Equation 1 is valid with C = 1.21.
Therefore (for L 4W),
oy - Pgy * (L/(B*(W-a)2*1.21]) - (3.3Pgy ]/(B(W-a)2)
W (2)
For the Charpy specimens, B - 0.394 in., W - 0.394 in., and a - 0.079 in.
Equation 2 then reduces to:
(3) oy - 33.3 x Pay where oy is in units of psi and Pgy is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.
Percent shear was determined from post-fracture photographs using the ratio-of-arras methods in compliance with ASTM Specification A370-89E93 The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
S-2
Tension tests were performed on a 20,000-pound Instron Model 1115, split-console test machine, per ASTM Specificatien E8-89b(10) and E21-79 (1988)III), and RMF Procedure 8107, Revision 1.
All pull rods, grips, and pins were made of Inconel 718 hardened to HRC45. The upper pull rod was connected through a universal joint to improve axiality of loading.
The tests were conducted at a constant e.rosshead speed of 0.05 inches per minute throughout the test.
Extension 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 E83-85[12),
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 J.a specimen, the following procedure was used to monitor specimen temperature.
Chromel-alurnel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip.
In the 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 of room temperature to 550'F (288'C). The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired spncimen temperatures.
Experiments indicated t!.at this method is accurate to 12*F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve.
The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were j
determined from post-fracture photographs.
The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was j
l computed using the final diameter measurement.
5-3 l
i 5.2 Charov V-Notch Impact Test Results The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule Y, which was irradiated to 1.33 x 1019 2
n/cm (E > 1.0 MeV), are presented in Tables 5-1 through 5-4 and are compared with unirradiated resultsill as shown in Figures 5-1 through 5-4.
The transition temperature increases and upper shelf energy decreases for the Capsule Y materials are summarized in Table 5-5.
Irradiation of the reactor vessel lower shell plate R2503 3 Charpy specihtens oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation) to 1.33 x 1019 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-1) resulted in a 30 ft-lb transition temperature increase of 30'F and in a 50 ft-1b transition temperature increase of 40'F.
This resulted in the 30 ft-lb transition temperature of 10*F and a 50 ft-lb transition temperature of 40*F (longitudinal orientation).
The average Upper Shelf Energy (USE) of the lower shell plate R2508-3 Charpy -
specimens (longitudinal crientation) resulted in a energy decrease of 27 ft-lb after irradiation to 1.33 x 1019 n/cm2 (E > 1.0 MeV) at 550*F.
This results in an average USE of 121 ft-lb (Figure 5-1).
1 Irradiation of the reactor vessel lower shell plate R2508-3 Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolling 19 direction of the plate (transverse orientation) to 1.33 x 10 n/cm2 (E >
1.0 MeV) at 550'F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 40'F and in a 50 ft-lb transition temperature increase of 45'F.
This resulted in the 30 ft-lb transition temperature of 40*F and a 50 ft-lb transition temperature of 85'F (transverse orientation).
5-4
i The average USE of the lower shell plate R2508-3 Charpy specimens (transverse orientation) resulted in an energy increase of I ft-lb after irradiation to n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average 1.33 x 1019 USE of 94 ft-lb (Figure 5-2).
Irradiation of the reactor vessel core region weld metal Charpy specimens to 1.33 x 10l9 n/cm2 (E > 1.0 MeV) at 550'F (Figura 5-3) resulted in a 50'F increase in 30 ft-lb transition temperature and a 50 ft-lb transition I
temperature increase of 45'F. This resulted in a 30 ft-lb transition temperature of O'F and the 50 ft-lb transition temperature of 30*F.
The average USE of the reactor vessel core region weld metal resulted in an 19 energy decrease of 6 ft-lb after irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average USE of 94 ft-lb (Figure 5-3).
Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 1.33 x 10I9 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-4) resulted in a 30 ft-lb transition temperature increase of 50*F and a 50 ft-lb transition temperature increase of 40'F.
This resulted in a 30 ft-lb transition temperature of -95'F and the 50 ft-lb transition temperature of -70'F.
The average USE of the reactor vessel weld HAZ metal experienced an energy increase of 19 ft-lb after irradiation to 1.33 x 1019 n/cm2 (E > 1.0 MeV) at 550'F.
This resulted in an average USE of 180 ft-lb (Figure 5-4).
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile or tougher appearance with increasing test temperature.
5-5
+ - - -
m m
g t
A comparison of the 30 ft-1b transition temperature increases and upper shelf energy decreases for the various Wolf Creek surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2I43 is presented in Table 5-6.
This comparison indicates that the transition temperature increases and the upper shelf energy decreases of the 19 2
lower shell plate R2508-3 resulting from irradiation to 1.33 x 10 n/cm (E > 1.0 MeV) nre less than the Regulatory Guide predictions.
This comparison also indicates that the upper shelf energy decrease of the weld metal resulting from irradiation to 1.33 x 10I9 n/cm2 (E > 1.0 MeV) is less than the Regulatory Guido prediction. Comparison of the 30 ft-lb transition temperature increase for the surveillance weld material is 15'F greater than the Regulatory Guide prediction.
However, the NRC Regulatory Guide 1.99, Revision 2 requires a 2 sigma allowance of 56*F for weld metal be added to the predicted reference transition temperature to obtain a conservative upper bound value.
Thus, the reference transition temperature increase for the surveillance weld metal is bounded by the 2 sigma allowance for shift prediction.
Table 5-6 also presents the E0L (32 EFPY) calculated upper shelf energy values for the surveillance materials.
The load-time records for the individual instrumented Charpy specimen tests are shown in Appendix A.
The calibration certification for all the equipment used in charpy and tensile tests are given in Appendix C.
5.3 Tension Test Results The results of the tension tests performed on the various materials. contained in capsule Y irradiated to 1.33 x 10 n/cm2 (E > 1.0 MeV) are presented in-19 Table 5-7 and are compared with unirradiated resultsIll as shown in Figures 5-9 through 5-11.
The results of the tension tests performed on the lower shell plate R2508-3 I9 2
(longitudinal orientation) indicated that irradiation to 1.33 x 10 n/cm (E > 1.0 MeV) at 550'F.. caused less than a 6 ksi increase in the 0.2 percent 4
5-6
1 offset yield strength and less than a 6 ksi increase in the ultimate tensile strength when compared to unirradiated dataIII (Figure 5-9).
The results of the tension tests performed on the lower shell plate R2508-3 (transverse orientation) indicated that irradiation to 1.33 x 1019 n/cm2 (E
> 1.0 MeV) at 550*F caused less than a 6 ksi increase in the 0.2 percent offset yield strength and less than a 9 ksi increase in the ultimate tensile strength when compared to unirradiated dataill (Figure 5-10).
The results of the tension tests performed on the reactor vessel core region weld metal indicated that irradiation to 1.33 x 1019 n/cm2 (E > 1.0 MeV) at 550*F caused less than a 5 ksi increase in the 0.2 percent offset yield strength and less than a 5 ksi increase in the ultimate tensile strength when compared to unirradiated dataill (Figure 5-11).
The small increases in 0.2%' yield strength and tensile strength exhibited by the lower shell plate R2508-3 and the weld metal indicate that this material is not highly sensitive to irradiation to 1.33 x 1019 n/cm2 (E > 1.0 MeV), as is also indicated by the Charpy impact test results.
The fractured tension specimens for the lower shell plate R2508-3 material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14.
The engineering stress-strain curves for the tension tests are shown in Figures-5-15 through 5-19.
5.4 Comoact Tension Tests Per the surveillance capsule testing program with the Wolf Creek Nuclear Operating Corporation, che 1/2-T compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and l.
Technology Center.
5-7 l
l
l TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE WOLF CREEK
~
LOWER SHELL PLATE R2508-3 IRRADIATED AT 550*F, FLUENCE 1.33 x 1019 n/cm2 (E_> 1.0 MeV) i Temperature Impact Energy Lateral Expansion Shear Sample No.
JJ'F
(*C)
(ft-lbl 121 (alls)
(mm)
(%)
onsitudinal 0 entation
~
17 15 0.38 10 ALS3
-50 22 17 0.43 15 AL67
-35 23 15 0.38 20' AL68 0
60 45 1.14 40 AL61 10 25 25 0.64 25 AL72 25 AL75 35 65 45 1.14 45 l
AL62 50 1
64
- 45 1.14 45 l
AL64 75 2
75
. 44-1.12 50 AL71 100
=
53 43 1.09 60 AL74 125 116 67 1.70 100 AL70 150 109 72 1.83 100 AL66 175 134 75 1.91 100 AL69 250 145 78 1.98 100 AL65 275 88 78 1.98 100 i
AL73 300 136 84 2.13) 100 Transverse Orientation 12 8
J.20 -
10 AT71
-35 12 10 0._.
10 AT63
-10 20 17 0.43 15 AT69 10 23 20 0.51 20 -
AT68 25
- 30 30 0.76 30 AT64 30 AT70 50 35' 28 0.71 30 AT65 60 39 34 0.86 35 AT75 75 62 42 1.07 45 AT61 85 47 39 0.99 45 AT68 100 51 43 1.09 50 AT67 125 64-51 1.30 65 AT62 150 74 56 1.42 80 AT73 175 39-69 1.75 100 AT72 225 1
100 78 1.98 100 AT74 275 1
94 68 1.73 100-5 l G
y
.N -a- -,, - -m-e-,n,.,e m.,n,.
n w
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s
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TABLE 5-2 t
CHARPY V-NOTCH IMPACT DATA FOR THE WOLF CREEK t
REACTOR VESSEL WELD METAL AND HAZ METAL 1RRADIATED AT 550'F, FLUENCE 1.33 x 1019 n/cm2 (E > 1.0 MeV)
Temperature Impact Energy Lateral Expansion Shear
+
Sample No.
(*F)
(*C)
(ft-lb) ill Imils)
(mm)
(%)
Weld Metal I
87 4
5 2
0.05 10 AW67
-125 73 NA NA NA NA NA-AW72
-100 68 5
3 0.08 15 AW63
- 90 51 8
6 0.15 15 AW62
- 60 37 23 14 0.36 45 AW69
- 35 23 29 22 0.56 50 l=
AW71
- 10 18 47 35 0.89 65 AW68 0
9 30 26 0.66 75 AW65 15 4
NA NA 0.00 NA AW75 25 AW73 35 2
70 51 1.30 85 AW66 65
-18 73 49 1.24 90 AW64 100 38 82 1
60 1.52 95 AW61 175 79 88 64 1.63 100 AW74 250 121 97 72 1.83 100 AW70 300 149 97 74 1.88 100 HAZ Vetal AH69
-200
-129 8
1 2
0.05 5
AH70
-160
-107 7
3 0.08 5
AH61
-125
- 87 18 9
0.23 10 AH74
-100-
- 73 33 18 0.46 15 AH67
- 75
- 59 35 11 8 0.46 20 AH64
- 50
- 46 107 1
' 66 1.68 70-AH71
- 50
- 46 120 1
66 1.68 75 l
AH68
- 25
- 32 55 31 0.79 40 l
AH65 0
- 18 95 49 1.24 70 9
114 56 1.42 75-AH72 15 AH62 50 10 153 80 2.03 100 AH73 125 52 147 79 2.01 90 AH63 200 93 180 90 2.29
- 100 AH66 250 121 195 98 2.49 100.
l AH75 250 121 226 306) 107 2.72 100 h
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TABLE 5-5 EFFECT OF 550*F IRRADIATION 101.33 x 10 n/cm (E > I,0 Mev) 04 THE hoiCH TOLGHMESS PROPERTIES OF THE VJLF CREEK REACTOR VESSEL 5tRVEILLAMCE MATERIALS UI III III III Average 30 ft-1b Average 35 mi1 Average 50 ft-tb Avengs Energy l
i.
t
~
Transitton lateral Espansion Transition Abscrptton at Temperature (*F)
Temperature (*F)
TSperature (*F)
Full $ hear (ft-lb)
Material Lbirradiated irradiated di Unitradiated Irradiated At uairradiated treadiated At tmirradiated irradiated A(f t-td) j i
I Plate R2508-3
- 20 13 30
- 10 45 55 0
40 40 148 121
- 27
' (Longitudinal) f i
Plate RZ508-3
.0-to 40 25 45 20 43 85 45 93 94
+ 1 w
i t
g (Transverse)
Weld Metal
- 50 0
50
- 25 20 45
- 15 30 45 100 94
- 6 i
j..
HAZ Metal
-145
- 95 50
- 90
- 25 65
-110
- TO 40 161 180
+ 13 s.
a 4
(1) " AVERAGE" is defined as the value read from the curse fitted through the data poierts of the Charry tests (Figures 5-1 throua% 5-4).
(
t l
i '
t' 3~-
{
i
.n,
m TABLE 5-6 4
COMPARISON OF THE WOLF CREEK SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS 30 ft-lb Transition Temo. Shift Upper Shelf Eneruy Decrease End-of-Li fe i
Upper Shelf Fluence Predictsd (a) Measured Predicted (a)
He'asured Eneruy (a) 19 2
Material Capsule 10 n/cm
(.F)
(*F)
(%)
(%)
(ft-lb) i l'
I Plate R2508-3 U
0.339 31 30 15 2-112.5 (Longitudinal)
Y 1.33 47 30 20 18 Plate R2508-3 U
0.339 31 25 15 0
112.5 (n
d;
{ Transverse)
Y 1.33 47 40 20 0
Weld Metal U
0.339 23 20 15 8
76.0 Y
1.33 35-50 20 6
t HAZ Metal 0
0.339-65 13 Y
1.33 50 0
Note:
(a) Based on Regulatory Guide 1.99, Revision 2 1-s-.
y a.
m
TABLE 5-7 TENSILE PROPERTIES FOR THE WOLF CREEK REACTOR VESSEL SURVEILLANCE i
MATERIALS IRRADIATED AT 550*F TO 1.33 X 10 n/cm2 (E > i.0 MeV)
I9 Test 0.25 Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp.
Strometh 14treesth Load Stree.
Strength Elongation Elongation in Area Mater 1 1 Nuober
(*F)
(keli (keli (k l o1 --
(kell (keil (51 (M) 2%)
Plate R360s-3 ALis so 08.7 87.s 2.e5 153.9 53.9 16.0 31.4 es (Longitudinal) AL14 115 63.2 83.3 2.43 130.9 50.4 12.0 20.s 72 ALIE ssO 57.e 34.s -
2.75 157.s 56.0 11.7 27.7 e6 FInte R260s-3 AT13 75 se.2 se.O s.10 136.8 83.2 13.5 26.1 54 (Transwere.)
AT14 140 84.2 83.0 2.80 147.8 67.0 12.0 23.0 et AT15 560 58.8 85.6 3.00 118.8 61.1 12.0 20.3 4e Wold ATIS O
S2.0 95.s 3.30 176.3 es.2 12.0 25.2 63 i
AW14 150 75.4 39.8 3.06 176.1 62.1 10.5 21.8 e6 AW15 660 71.3 91.7 3.03 169.7 et.s 10.5 22.4 et i
a
('C)
-150
-100
-50 0
50 100
!50 200 g --.
I i
i l
i ice ;-
t ys
-i-t
.Q BD ;-
he
!f 4.,0
/
w
& 40
. e?,' e o t< '
[0
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I I
I 0
300 25 st-n
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,,-e;--
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-e 60 g/'
13 a.
O 40 o.* *
- I '
10 g,r - e, e,
3a 05
- N;
,I I
~'
I I
I 0
0 o (H3TADIAI(D 9 IPCADlif(D t$50*f) ig(([ l}) a 10 n/ cal 160
' '*o 200 V
I, 140
./ '
7 120 b
160
,e i e e 100 o
C 0
e 120 -
80 o
?
e S
y 60 o *h sl 80
/
i~
/
- rr 40 e
20 t f, N jr, 40 l
l d l
l l
9 0
-200
-100 0
100 200 300 400 TEMPERATURE (*h l
l l
l Figure 5-1.
Charpy V-Notch Impact Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-15
l
('C)
-150
-100
-50 0
50 100 150 200 l
I I
I eq l
I I
8 80 ff 60 W
2 3
Ci 40 20 t
i 0
100 2.5 E
80 2.0 e
3 a--
_. m.
n e 60
[* W 1.5 U
40 1.0 e)
. Eberr ti 20 s'
0.5 I
'A I
0 0
o LMRRouTO e 3Routo es'n nuoc Ln n IEvat 120 160 100 o o e o
o' 120 g
@ 80 o
60 b
80 g
vi
~
@ 40 t
a
%gr 40 20 f
.M l
f i
i i
I I
0 0
twe
-200
-100 0
100 200 300 400 500 amwc TEMPERATURE (*F)
Figure ! 2.
Charpy V-Notch Impact Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) 5-16
J l
.i
('C)
-150
-100
-50 0
50 100 150 200 1
I I
I I
~l l
I IU0
~
f~ %.tMr o
8 83 8
SE 60 w
o M 40
- Y
- 1 1
1 1
t 0
100 2.5 h" 80 8
2.0 Ca#'--t '
60 1.5 o
b b
40 1.0 d
err -
g i
f I
I I
0 oUggeAn e asuAn cWn ruso: 1.n : d%d 120 160-100 7
120 t'
2 80 o
T b
60 80 5
t
@ 40 dr a
s'r o
e 40 20 o
I l-I
-l 1-l 0
-100 0
100 200 300-400
-500 0
-200 wu sans TEMPERATURE ('F) l.
I i
Figure 5-3.
Charpy V-Notch Impact Properties for Wolf Creek Reactor Vessel-Surveillance Weld Metal 5-17
.m
~-
I r-
'?C)
-150 -100
-50 0
50 100 150 200 Ti 12 1
't i
I I
100 e Lelt 8 80
/,
4 60 40 20 M*
i i
l I
I c
0 100 2.5 23 g 80 t
u,-
2.0 N 60 1.5 o
t
" 40 c't 1.0 I
I I
I I
O o tuuwus e muss es*n rutur Ln i18vo#
200 ra 100 240 160
,o e
o 200 2 140 3120 160 D
b 100 o.,
25 120 g 80
" 60
.o-e 80
- 't 40 o
40 20
.f 0'
0 l
- 200
-100 0
100 200 300 400 TEMPERATURE ('F) m Jose P
l Figure 5-4.
Charpy V-Notch Impact Properties for Wolf' Creek Reactor Vessel Weld Heat-Affected-Zone Metal 5-18
_.-.,,..,,-._.-m
i d
~
' 4' l'
a IN -
. : ifj
. Lij
'f'kis Q)
~
Ml
!' <~;4
, ',?. '(
- c. *.
k
<.:n ' D 19 a.-
ALO3 AL67 ALG8 AL61 AL72 4
I avh ~
n._.
]
AL75 AL62 AL64 AL71 AL74 y
s
.'92-
- 4,
- 1. 4 - :,,
1) d n-A,y.
AL70 AL66 AL69 AL65 AL73 Figure 5-5.
Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-19
f kl,,/.
?.lY[,Y
?
- kh.
~
,3;M 3:
- L9ry.,r$
qfeg;y',
g; w-
-c 9@i;?<
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SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral' part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis. The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the. l 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 dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of Light Water Reactor 6-1
Surveillance Results,.[25] recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom..[23] The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99, " Radiation Camage to Reactor Vessel Materials." This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance Capsule Y. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided. 6.2 Discrete Ordinates Analysis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the rnactor vessel surveillance program. The capsules are located at a:imuthal angles of 58.5, 61.0', 121.5*, 238.5*, 241.0', and 301.5* relative to the core cardinal axes as shown in Figure'4-1. A plan view of a dual surveillance :apsule holder attached to the neutron pad is shown in Figure 6-1. The stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in neight. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core. 6-2
From a neutron transport standpoint, the surveillance capsule structures are -~ignificant. They have a marked effect on both the distribution of' neutron s flux and the neutron energy spectrum in the water _ annulus between the. neutron pad and the reactor-vessel. In order to properly determine' the: neutron environment at the test specimen locations, the_ capsules themselves must bc included in the analytical model. In performing the fast neutron exposure evaluations for the surveillance - capsules and reactor vessel, two distinct sets of trans;. ort calculations were carried out. The first, a, single computation in the conventional _ forward _ mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {p(E > 1.0 Mev), d(E > 0.1 Mev), and dpa) through y the vessel wall. The neutron spectral. information was -required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios;-i.e., dpa/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 locations interior to the pressure _ vessel wall; i.e., the 1/4T, 1/2T, av 3/4T locations. The second set of calculations consisted of a series of-adjoint analyses-relating-the fast neutron flux-(E > 1.0 MeV) at surveillance-capsule positions, ~ and several azimuthal locations on the pressure vessel = inner radius.to neutron- -source distributions within the reactor core. The importance functions-generated.from these adjoint analyses provided the basis for all' absolute. exposure projections and comparison with measurement.- These importance functions, when combined with cycle _ specific neutron. source _ distributions, yielded absolute predictions of. neutron exposure at the locations of interest for each cycle of irradiation; and established the means to: perform similar predictions and dosimetry evaluations for all subsequent fuel cyclese It is important. to. note.that. the. cycle specific neutron source distributions' utilized, in these analyses _ included.not only spatial variations of fission rates within - the reactor core; but, also accounted for the' effects-of varying neutran yield'- 6-3 ,a
l. per fission and fission spectrum introduced by the build-up of plutonium'as the - burnup of individual fuel-assemblies increased. The absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the forward calculation provided the means to: 1. Evaluate neutron dosimetry obtained from surveillance capsule locations. 2. Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall. 3. Enable a direct comparison of analytical prediction with measurement. 4 4. Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves. The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R, 0 geometry using the DOT two-dimensional discrete ordinates codell33 and the SAILOR cross-section librarylI43 The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering was s treated with a P3 expansion of the cross-sections and the angular discretizat. ion was modeled with an S8 order of angular quadrature. The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inherent in-the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 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 4 single reactor would have a power distribution at the nominal +2a 6-4 i
t level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results. All adjoint analyses were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library. Adjoint source locations were chosen at several azimuthal locations along the - pressure vessel inner radius as well as the geometric cent,er of each surveillance capsule. Again, these calculations were run in R, 0 geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, p (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as: R(r,0)-fr[0IE 1(r, 0, E) S (r, 0, E) r dr de dE where: R (r, 0) - p (E > 1.0 MeV) at radius r and azimuthal angle 0 1 (r, 0, E) - Adjoint importance function at radius, r, azimuthal angle 0, and neutron source energy E. S (r, 0, E) - Neutron source strength at core location r, 0 and energy E. Although the adjoint importance functions used in the analysis were based on a response functica defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the' relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/p (E > 1.0 MeV) is insensitive to changing core source distributions. -In the application of'these adjoint importance functions to the Wolf Creek Unit I reactor, therefore, the iron displacement rates (dpa) and the neutron flux (E > 0.1 MeV) were computed on a cycle. specific basis by using dpa/p (E > 1.0 MeV) and d (E > 0.1 MeV)/p (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific & (E > 1.0 MeV) solutions from the individual adjoint evaluations. 6-5
The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design reports for the first five operating cycles of Wolf Creek Unit 1[15 through 20) Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measure'nent 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 [p (E > 1.0 MeV), l G(E > 0.1 MeV), and dpa] are given at the geometric center of the two surveillance capsule positions for both the design basis and the plant specific l core power distributions. The plant specific data, based on the adjoint transport analysis, at ; meant to establish the absolute comparison of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the cycles 1 through 5 plant specific power distribution. 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), aeutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tablet 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 through 6-5, 6-6
For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' azimuth is given by: 41/4T(45') - p(220.27, 45') F (225.75, 45') where: pl/4T(45') - Projected neutron flux at the 1/4T position on the 45' azimuth ( (220.27,45') - Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth. F (225.75, 45') - Relative radial distribution function from Table 6-3. Similar expressions apply for exposure parameters in terms of p (E > 0.1 MeV) and dpa/sec. The DOT calculations were carried out for a typical octant of the reactor. However, for the neutron pad arrangement in Wolf Creek Unit 1, the pad extent for all octants is not the same. For the analysis of the flux to the pressure vessel, an octant was chosen with the neutron pad extending from 32.5 - 45.0-degrees which produces the maximum flux. Other octants have neutron pads . spanning larger azimuthal sectors which provide more shiciding. For the octant with the 12.5 degree pad, the maximum flux to the vessel occurs near 25 degrees and the values in the tables for the 25 degree angle are vessel maximum values. Exposure values for 0,15, and 45 degrees can be used for all octants; values in the tables for 25 and 35 degrees are maximum values and only apply to octants with a 12.5 degree neutron pad. 6.3 Neutron Dosimet_r_y The passive neutron sensors included in the Wolf Creek Unit I surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the 6-7 e
evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure paran.eters of interest [p (E > 1.0 Mev), p (E > 0.1 MeV), dpa]. The relative locations of the neutron sensors within 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 fission monitors were accommodated within the dosimete" block located near the center of the capsule. The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest: o The specific activity of each monitor. o The operating history of the reactor. o The energy response of the monitor, o The neutron energy spectrum at the monitor location. o lhe physical characteristics of the monitor. The specific activity of each of the neutron monitors was determined using established ASTM procedures [21 through 34] Followinc sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the Wolf Creek Unit I reactor during cycles 1 through 5 was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable period. The irradiation history applicable to Capsule Y is given in Table 6-7. Measured and saturated reaction product specific activities as well as measured 6-8
full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6-6 and-6-7. Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code [35), TheFERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra, in the FERRET evaluations, a log-normal least-squares algorithm weights both the a priori values and the measured data in accordance~with the assigned uncertainties and correlations. In general, the measured values f are linearly related to the flux & by some response matrix A: (s,a) (s) (a) f -I A p 9 19 g where i indexes the measured values belonging to a single data set s, g designates the energy group and a delineates spectra that may be simultaneously adjusted. For example, R -I o p i 9 19 9 relates a set of measured reaction rates Rj to a single rpectrum p _by g-the multigroup cross section ajg. (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. 6-9
In the FERRET analysis of the dosimetry data, the continuous quantities (i.e., fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code (36). This procedure was carried out by first expanding the a priori spectrum into the SAND-Il 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 :ollapsed into the 53 energy-group structure using SAND 11 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 ur a pricri 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: gg,-Rh+R R,P M g g gg, where RN specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional uncertainties R specify additional random uncertainties for group g that are g correlated with a correlation matrix: 1 Pgg, - (1 - 0) 6gg, + 0 exp [~ ] The first term specifies purely random uncertainties while the second term describes short-range correlations over a range a (6 specifies the strength of the latter term). l 6-10 i l> .i
For the a priori calculated fluxes, a short-range correlation of a = 6 ~ groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E. Maerkr,[37]. Maerker's results are closely duplicated when a - 6. For the 19ral reaction rate covariances, simple normalization and random-uncertainties were combined as deduced from experimental uncertainties. Results of the FERRET evaluation of the Capsule Y dosimetry are given in Table 1 6-9. The data summarized in Table 6-9 indicated that the capsule received an 19 integrated exposure of 1.33 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 tabulated in Table 6-11 for the FERRET 53 energy group structure. A summary of the measured and calculated neutron exposure of Capsule Y is presented in Table 6-12. The agreement between calculation and measurement falls within 14% for all fast neutron exposure parameters listed. The thermal neutron exposure calculated for the exposure period under predicted the measured value by 56 percent. Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (4.79 EFPY) exposure derived from the Capsule Y measurements, projections are also provided for an i exposure period of 15 EFPY and to end of vessel design life (32 EFPY). l In the evaluation of the future exposure of tS reactor pressure vessel the average exposure rates derived from cycles 1 through 5 were employed. In i computing these' average exposure rates, the calculated averages were also scaled by the average measurement / calculation ratios observed from evaluations of dosimetry from Capsules Y and U. This procedure resulted in the following 6-11
1 bias factors being applied to the analytical results:. Flux (E > 1.0 MeV) Bias = 1.182 Flux (E > 0.1 MeV) Bias = 1.142 i dpa/sec Bias = 1.149 In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the Wolf Creek Unit I reactor coolant system, exposure I projections to 15 EFPY and 32 EFPY were also employed. Data based on both a 1 fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vess_el wall are provided in Table 6-14. In order to access RTNDT vs. fluence trend-- ~ curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations: d' (1/4T) = p (Surface) (d (S f e}} l p' (3/4T) - p (Surface) {dp (Su f e)} 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 Wolf Creek Unit 1 surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules. l t l l 6-12 l l
L F (TYPICAL) 0 - 61.00 ,. x - 58.5 - 81.625 IN. r/ w, A a h t aw sn sN i Figure 6-1. Plan View of a Dual-Reactor Vessel Surveillance Capsule 6-13 i
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE' SURVEILLANCE CAPSULE CENTER 4(E > 1.0MeV) d(E > 0.lMev) Iron Displacement Rate 2 2 In/cm -secl in/cm -secl Idea /sec1 29.0 - 31.5* 29.0* 31.5* 29.0* 31.5* DESIGN BASIS 1.13 X 1011 ll 11 11 2.21 X 10-10 2.36 X 10 1.21 X 10 5.07 X 10 5.42 X 10 CYCt.E-1 8.31 X 10 8.86 X 1010 ll Il 1.63 X 10-10 1.74 X 10-10 10 3.73 X 10 3.98 X 10 10 10 ll ai CYCLC 2 8.84 X 10 9.76 X 10 3.96 X 10 4.39 X 101I 1.73 X 10-10 1.91 X 10-10 b 10 10 Il CYCLE 3 7.18 X 10 7.78 X 10 3.22 X 10 3.50 X 1011 1.40 X 10-10 1.52 X 10-10 10 10 CYCLE 4 6.99 X 10 7.68 X 10 3.14 X 1011 11 1.37 X 10-10 1.50 X 10-10 3.45 X 10 CYCLE 5 7.05 X 1010. 7.50 X 10 3.16 X 10 3.37 X 10 10 11 ll 1.38 X 10-10 1.47 X 10-10
TABLE J-2 CALCVLATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 2 6(E > 1.0MeV) In/cm -sec) 0.0* 15.d' 21d'. 35,0* 45.0' 10 2.45 X 10 2.81 X 1010 10 10 3.01 X 10 10 2.66 X 10 DESIGN BASIS 1.78 X 10 10 10 10 CYCLE 1 1.31 X 1010 2.93 X 1010 2.21 X 10 1.79 X 10 2.05 X 10 10 10 2.51 X 1010 10 2.31 X 10 2.06 X 10 10 2.06 X 10 CYCLE 2 1.44 X 10 10 10 1.55 X 1020 1.71 X 1010 10 1.69 X 10 1.89 X 10 CYCLE 3 1.18 X 10 10 10 10 10 1.87 X 10 1.58 X 10 1,82 X 10 10 1.78 X 10 CYCLE 4 1.31 X 10 10 10 1.69 X 10 10 1.51 X 10 10 1.93 X 10 10 1.32 X 10 CYCLE 5 1.27 X 10 2 in/cm -seci 6(E > 0.lMeVi 0.0* 15.0' 25.0* 35.0* 45.0* 10 7.05 X 10M 10 8.22 X 1010 6.95 X 10 DESIGN BASIS 3.71 X 1010 5.61 X 10 10 10 5.14 X 10 10 6.04 X 1010 5.10 X 10 CYCLE 1 2.72 X 1010 4.06 X 10 10 10 6.28 X 10 10 5.84 X 10 10 6.31 X 10 CYCLE 2 3.00 X 1010 4.34 X 10 10 10 10 4.41 X 10 4.29 X 10 10 3.55 X 1010 5.15 X 10 CYCLE 3 2.45 X 10 10 10 4.49 X 1010 4.56 X 10 10 3.74 X 1010 5.09 X 10 CYCLE 4 2.71 X 10 10 10 4.29 X 1010 4.23 X 10 10 3.84 X 1010 5.26 X 10 CYCLE 5 2.64 X 10 Iron Atom Disolacement Rate Idoa/sec) 0.0' 15.0* 25 0' 35.0*
- 45.0*
DESIGN BASIS 2.78 X 10-11 4.13 X 10-11 5.04 X 10-11 4.14 X 10-11 4.48 X 10-11 CYCLE 1 2.04 X 10-11 2.99 X 10-11 3.70 X 10-11 3.04 X 10-11 3.27 X 10-11 CYCLE 2 2.24 X 10-II 3.19 X 10-11 3.87 X 10-11 3.48 X 10-11 4.00 X 10-11 CYCLE 3 1.84 X 10-11 2.61 X 10-11 3.16 X 10-11 2.63 X 10-11 2.73 X 10-11 CYCLE 4 2.03 X 10-31 2.75 X 10-11 3.12 X 10-11 58 X 10-Il 2.90 X 10-11 6 CYCLE 5 1.98 X 10-11 2.82 X 10-11 3.23 X 10-11 2.56 X 10-11 2.69 X 10-11 6-15 l
TABLE 6-3 RELATJVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSl'RE VESSEL WALL Radius IEm1_ O' 15* 25' 35' 45' 220.27(l) 1.00 1.00 1.00 1.00 1.00 220.64 0.976 0.979 0.980 0.977 0.979 221.66 0.888 0.891 0.893 0.891 0.889 222.99 0.768 0.770 0.772 0.770 0.766 224.31 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.550 0.554 0.552 0.543 226.95 0.462 0.460 0.465 0.463 0.452 228.28 0.386 0.384 0.388 0.386 0.375 229.60 0.321 0.319 0.324 0.321 0.311 230.92 0.267 0.263 0.275 0.267 0.257 232.25 0.221 0.219 0.225 0.221 0.211 233.57 0.183 0.181 0.185 0.183 0.174. 234.89 0.151 0.149 0.153 0.151 0.142 236.22 0.124 0.122 0.126 0.124 0.116 237.54 0.102 0.100 0.104 0.102 0.0945 238.86 0.0828 0.0817 0.0846 0.0835 0.0762 240.19 0.0671 0.0660 0.0689 0.0679 0.0608 241.51 0.0538 0.0s22 0.0550 0.0545 0.0471 242.17(2) 0.0506 0.0488 0.0518 0.0521-0.0438 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-16
i TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 MeV) WITHIN THE PRESSURE VESSEL WALL Radius Ism)_ O' 15' 25' 35' 45' 220.27(l) 1.00 1.00 1.00 1.00. 1.00 220.64 1.00 1.00 1.00 1.00 1.00 221.66 1.00 1.00 1.00 0.999 0.995 222.99 0.974 0.969 0.974 0.959 0.956 224.31 0.927 0.920 0.927 0.907 0.901 225.63 0.874 0.865 0.874 0.850 0.842 226.95 0.818 0.808 0.818 0.792 ~0.782 228.28 0.761 0.750 0.716 0.734 0.721 229.60 0.705 0.693 0.704 0.677 0.662 230.92 0.649 0.637 0.649 0.621 0.605 232.25 0.594 0.582 0.594 0.567 0.549 233.57 0.540 0.529 0.542 -0.515 0.495 234.89 0.487 0.478 0.490 0.465 0.443-236.22 0.436 '0.428 0.440 0.416 0.392 237.54 0.386 0.380 0.392 0.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 240.19 0.289 0.287 0.298 0.279 0.248 241.51 0.244 0.238 0.249 0.233 0.201. 242.17(2) 0.233 0.226 0.237 0.223 0.188 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-17
J F: TABLE 6-5' RELATIVE RADIAL DISTP.IBUTIONS OF: IRON DISPLACEMENT RATE = (dpa)- d WITHIN-THE PRESSURE----VESSEL WALL - Radius Lant_. 0 ___ 15' 25' 35' 45' 220.27(I) 1.00 1.00 -1.00 1.00 1.00 220.64 0.984 -0.981 0.984 0.983 0.9841 221.66-0.912 0.909 0.917 t'.921 -.0.915 222.99 0.815 0.812 0.826 0.833 - 0.821' 224.31-0.722 0.719 0.737-0.747 0.730. L 225.63 0.638 0.634 '0.656-0.668: 0.647 226.95 0.563 0.559 0.584-0.597 0.572: 228.28 0;497 0.493 0.519-0.533 0.506-229.60 0.439 0.435 0 462 0.475L 10.447f 230.92 0.387 0.383 0.410 0.423 0.394-- 232.25-0.341 0.338 0.364-- 0.376' O.347 H '233.57 0.300 0.297 0.322 0.334 0.305-234.89. 0.263 - 0.261- -0.285- -0.295' -0. 26LL 236.22 0.230 0.228 0.250 10.260 0.231L 237.54 0.199 .0.198 0.218 0.227 0.199 238.86 0.171-0.170-0.189-0.196 10.169: 4' 240.19-0.-145 0.144 0.161 0.167 0.140 -241.51-0.121
- 0.119-
.0.135 0.139 -0.113: ~ 242.17(2) 0.116-0.113 0.128 0.134 0.1061 NOTES:
- 1) Base Metal : Inner Radius --
- 2) Base Metal Outer: Radius-L I
~6-18
TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Weight
Response
Product Yield Material Interest Fraction Rance Hai f-Life (%) Copper Cu63(n.o)Co60 - 0.6917 E > 4.7 MeV 5.272 yrs Iron 'Fe54(n.p)Mn54 0.0582 E > 1.0 MeV 312.2 days Nickel NiS8(n,p)CoS8 0.6830 E > 1.0 MeV 70.90 days ?' Uranium-238* ' U238(n,f)Csl37-1.0 E > 0.4 MeV 30.12 yrs 5.99 G Neptunium-237* LNp237(n,f)Csl37 1.^ E > 0.08 MeV 30.12 yrs 6.50 . Cobalt-Aluminum
- CoS9(n,a)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs Cobal t-Aluminum Co59(n,0)Co60 -
0.0015 E > 0.015 MeV 5.272 yrs
- Denotes that monitor is cadmium shielded.
TABLE 6-7 MONTHLY THERMAL GENERATION DURING THE FIRST FIVE FUEL CYCLES 0F THE WOLF CREEK UNIT 1 REACTOR THERMAL THERMAL THERMAL THERMAL GENERATION GENERATION GENERATION GENERATION MONTH (MW-bri MONTH (MW-br) MONTH (MW-hr) MONTH (MW-hr) 5/85 0 1/87 1533313 9/88 245016' 5/90 1003923 6/85 356676 2/P7 2192444 10/88 492163 6/90 2442569 7/85 1025780 3/87 2471746 11/88 0 7/90 2515109 8/85 1643803 4/87 2247475 12/88 0 8/90 2534494 9/85 2053023 5/87 2436662 1/89 2095086 9/90 2453417 10/85 2086772 6/87 2250313 2/89 2113705 10/90 2533710 11/85 2366472 7/87 2066874 3/89 2535552 11/90 2421081 12/85 2368666 8/87 2527262 4/89 2454150 12/90 2531359 1/86 2480479 9/87 1954923 5/89 2498149 1/91 -2363291 2/86 2005668 10/87 0 6/89 2448863 2/91 1840498 3/86 2513225 11/87 0 7/89 2493515 3/91 1969185 l 4/86 933250 12/87 0 8/89 2534633 4/91 1506284 5/86 2341310 1/88 1216547 9/89 2453774 5/91 1692964 6/86 1670026 2/88 956585 10/89 2516573 6/91 2434282 7/86 2210358 3/88 2526972 11/89 2450503 7/91 2534580 8/86 2439547 4/88 2452604 12/89 2536033 8/91 2466385 9/86 2406802 5/88 2533966 1/90 2534772 9/91 1221097 10/86 1219774 6/88 2451743 2/90 2017613 11/86 0 7/88 2531412 3/90 599723 12/86 65'0000 8/88 2533606 4/90 0 t l i= l-6-20 l l
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES ) Measured Saturated Reaction Monitor and Activity Activity Rate Axial location (dis / set-ami (dis /sec-om) 1RPS/ NUCLEUS) Cu-63 (n.a) Co-60 Top 1.37.x 105 3.41 x 105 i Middle 1.2C x 10 2.99 x 105 5 5 3.01 x 105 Bottom 1.21 x 10 Average 1.26 x 105 3.14 x 105 4.78 x 10'I7 Fe-54(n.p,*tc.-54 Top 1 56 x 106 3.03 x 106 Middle 1.49 x 106 2.72 x 106 6 6 Bottom 1.48 x 10 2.71 x 10 Average 1.54 x 10 2.8?. x 106 4.49 x 10-15 6 Ni-58 (n.p) C0-58 l Top 8.04 x 106 I 4.45 x 10 Middle 7.38 x 106 7 4.08 x 10 6 Bottom 7.33 x 10 4.06 x 107 Average 7.58 x 106 4.20 x 107 5.99 x 10-15 U-238 (n,f) Cs-137 (Cd) l 5 Middle _5.43 x 10 5.40 x 106 3.00 x 10~l4 6-21 l l. 2 ..,., ~ , ~.. .,..,.. _ ~..., ~. -..,
... = - _. _ _ _ _ _ _ TABLE 6-8 HEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /sec-am) (dis /sec-am) (RPS/ NUCLEUS) i Np-237(n,f) Cs-137 (Cd) P Middle 4.40 x 10 4.38 x 107 2.65 x 10~I3' 6 Co-59 (n,6) Co-60 Top 2.59 x 107 6.45 x 10 7 BJ.com 2.57 x 107 7 6.40'x 30 Average 2.58 x 107 6.42 x 107 4.19 x 10-12 C0-59 (n,6) Co-60 (Cd) 7 7 Top 1.30 x 10 3.24 x 10 7 7 Middle 1.36 x 10 3.39 x 10 7 Bottom 1.39 x 10 3.46 x 107 Avera9e 1.35 x 10 3.3b x 107 .2.19 x 10-12 7 { 6-22 .~
TABLE 6-9
SUMMARY
OF NEUTRON DOSIMETRY RESULTS TIME AVERAGED EXPOSURE RATES d (E > 1.0 MeV) (n/cm -sec} 8.77 x 1010 2 8% 4 (E > 0.1 MeV) (n/cm -sec) 3.90 x 1011 2 15% dpa/sec 1.69 x 10-10 11% 2 p (E < 0.414 eV) (n/cm -sec) 8.29 x 1010 21% INTEGRATED CAPSVLE EXPOSURE 2 4 (E > 1.0 MeV) (n/cm } 1.33 x 1019 8% 2 + (E > 0.1 HeV) (n/cm ) 5.91 x 1020 15% dpa L.56 x 10-2 11% 2 + (E < 0.414 eV) (n/cm ) 1.25 x 1019 1 21P. NOTE: Total Irradiation Time = 4.79 EFPY 6-23 i: i usamms M
TABLE 6-10 COMPARISON,OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER Adjusted Englj.gn Measured Calculation Q Cu-63 (n,a) Co-60 5.50x10-17 5.64x10~I7 1.03 Fe-54 (n.p) Mn-54 5.92x10-15 5.80x10-15 0.98 Ni-58 (n,p) Co-58 7.88x10-15 7.87x10-15 1.00 U-238 (n.f) Cs-137 (Cd) 3.41x10-14 3.30x10~l4 0.97 Np-237 (n,f) Cs-137 (Cd) 3.llx10-13 3.23x10-13 1.04 Co-59 (n,a) Co-60 (Cd) 2.66x10-12 2.67x10-12 1.00 Co-59 (n.a) Co-60 5.30x10-12 5.26x10-12 o,gg 6-24
TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER Energy Adjus}cdFlux Energy AdjusgedFlux Group (Mev) (n/cm sec) Group (Mev) (n/cm -sec) I 6 6.95x10 28 9.12x10-3 1.7ex1010 I 1.73x10 2 1.49x10I 7 1.56x10 29 5.53x1'~3 2.22x1010 3 1.35x101 5.98x107 30 3.35x10-3 6.94x109 4 1.16x101 1.32x108 31 2.84x10-3 6.62x109 5 1.00x101 2.89x108 32 2.40x10-3 6.35x109 6 8.61x100 4,ggxin8 33 2.03x10-3 1.78x1010 7 7.41x100 1.lix109 34 1.23x10-3 1.62x1010 8 6.07x100 1.57x109 35 7.49x10-4 1.49x1010 0 9 4.97x10 3.31x109 36 4.54x10-4 1.41x1010' 10 3.68x10 4.42x109 0 37 2.75x10-4 1.50xl'010 0 9 11 2.87x10 9.44x10 38 1.67x10-4 1.55x1010 0 12 2.23x10 1.33x1010 39 1.0lx10-4 1.62x1010 13 1.74x100 1.90x1010 40 6.14x10-5 1.61x1010 0 14 1.35x10 2.12x1010 41 3.73x10-5 1.59x1010 r 15 1.11x100 3.90x1010 42 2.26x10-5 1.56x1010 16 8.21x10~l 4.48x1010 43 1.37x10-5 1.52x1010 17 6.39x10-I 4.66x1010 44 8.31x10-6 1.46x1010 18 4.98x10~I 3.38x1010 45 5.04x10-6 1.35x1010 19 3.88x10-1 4.74x1010 46 3.06x10-6 1.27x1010 20 3.02x10-1 4.86x1010 47 1.86x10-6 1.17x1010 21 1.83x10~l 4.79x1010 48 1.13x10-6 8.68x109 22 1 lix10'l 3.81x1010 49 6.83x10-7 1.09x1010 1 23 6.74x10-2 2.63x1010 50 4.'14x10'7 1.44x1010 24 4.09x10-2 1.48x1010 51' 2.51x10~7 1.41x1010 25 2.55x10-2 1.94x1010 52 1.52x10-7 1.32x1010 26 1.99x10-2 9.52x109 53 9.24x10-8 3.74x1010 27 1.50x10-2 1.20x1010 NOTE: Tabulated energy. levels represent the ' upper energy of each group. 6-25
TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE Y i Calculated Measured- [ld f(E > 1.0 MeV) {n/cm ) 1.15 x 1019 1.33 x 1019 0.87 2 2 19 5.91 x 1019 0.97 f(E > 0.1 MeV) {n/cm ) 5.16 x 10 dpa 2.; x 10-2 2.56 x 10-2 o,gg f(E < 0.414 eV) (n/cm ) % 37 y igl8 1.25 x 1019 0.44 i 2 t I I 6-26 l L.
TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS i ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 4.79 EFPY l 0* 15* 25* 30' 45* i 4 (E > 1.0 Mev) 2.32 X 1018 18 18 18 18 3.31 X 10 3.62 X 10 3.62 X 10 3.44 X 10 i [n/cm2] + (E > 0.1 MeV) 4.67 X 1018 6.73 X 10l8 18 9.55 X 10 9.56 X 1018 8.31 X 1018 [n/cm2] Iron Atom Displacemtnts 3.52 X 10-3 4.98 X 10-3 5.89 X 10-3 5.90 X 10-3 5.32 X 10-3 [dpa) 15.0 EFPY 0* 15* 25' 30* 45* 18 I9 I9 + (E > 1.0 Mev) 7.47 X 10 1.06 X 10 1.17 X 10 1.17 X 10I9 19 1.12 X 10 { 1 [n/cm2] b +. (E > 0.1 MeV) 1.50 X 10I9 2.17 X 10I9 I9 3.09 X 10 3.09 X 10I9 2.72 X 10I9 [n/cm2] 4 Iron Atom Olsplacements 1.13 X 10-2 1.60 X 10-2 1.91 X 10-2 1.91 X 10-2 1.74 X 10-2 l {dpa) i 32.0 EFPY 0* 15* 25* 30* 45' I9 I9 I9 I9 I9 f (E > 1.0 Mev) .1.59 X 10 2.27 X 10 2.50 X 10 2.50 X 10 2.39 X 10 [n/cm2] r 19 39 I9 I9 19 [ 4 (E > 0.1 MeV) 3.20 X 10 4.62 X 10 6.59 X 10 6.60 X 10 5.80 X 10 [n/cm2] Iron Atom Displacements 2.41 X 10-2 3.42 X 10-2 4.07 X 10-2 4.07 X 10-2 3.71 X 10-2 [dpa) Y l- ~_ ~_
TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES 1 i 15 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE I 2 2 (n/cm ) (equivalent n/cm ) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 18 18 17 18 18 l8 0* 7.47 x 10 4.06 x 10 8.67 x 10 7.47 x 10 4.72 x 10 1.63 x 10 t-18 I9 18 18 i; 18 1.24 x 10 1.06 x 10 6.72 x 10 2.33 x 10 15' l.'06 x 10l9 5.78 x 10 18 I9 18 18 j 19 18 1.36 x 10 1.17 x 10 7.39 x 10 2.56 x 10 25*(a) 1.17 x 10 6.36.x 10 I9 18 18 l9 18 18 [ 30*(a) 1.17 x 10 6.39 x 10 1.37 x 10 1.17 x 10 7.68 x 10 2.85 x 10 I9 18 18 f 19 18 18 - 1.12 x 10 7.08 x 10 2.46 x 10 45' l.12 x 10 6.09 x 10 1.30 x 10 t i' 32 EFPY [ ~ 6 NEUTRON FLUENCE (E > I.0 MeV) SLOPE doa SLOPE 2 2 (n/cm ) (equivalent n/cm ) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T I9 18 - I9 1.01 x 10 3.49 x 10 18 1.59 x 10 l9 18 1.85 x 10 0* 1.59 x 10 8.66 x 10 I9 18 19 I9 18 l 15* 2.27'x 10I9 1.23 x 10 2.64 x 10 2.27 x 10 1.43 x 10 4.97 x 10 18 l9 5.47 x 10 18 19 1.58 x 10 19 1.36'x 1019 2.90 x 10 2.50 x 10 25*(a) 2.50 x 10 { 18 l9 6.08 x 10 l9 19 18 l9 1.64 x 10 30*(a) 2.50.x.10 1.36 x 10 2.93.x 10 2.50 x 10 I9 18 [ l9-I9 18 2.39'x 10I9 1.51 x 10 5.24 x 10 45* 2.39 x 10 1.30 x 10 2.78 x 10 -[ (a) Maximum point.on the pressure vessel r r i 4.,
TABLE 6-15 UPDATED LEAD FACTORS FOR WOLF CREEK UNIT 1 SURVEILLANCE CAPSULES Caosule Lead Factor U Withdrawn Y 3.75(a) V 3.76(b) W 4.08(b) X 4.08(b) Z 4.08(b) (a) Plant specific evaluation based on end of cycle 5 calculated fluence. (b) Projection based on average flux through cycle 5. t l l 6-29
SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The follow!ng removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the Wolf Creek reactor vessel: Capsule Estimated Location Lead Fluence 2 Capsule (deg.) Factor Removal Time (a) (n/cm ) t U 58.5 3.85 1.08 (Removed) 3.39 x 1018 (Actual) Y 241.0 3.75 4.79 (Removed) 1.33 x 1019 (Actual) V 61.0 3.76 8.5 2.50 x 1019 (b) X 238.5 4.08 14.0 4.46 X 1019 W 121.5 4.08 Standby -Z 301.5 4.08 Standby (a) Effective Full Power Years (EFPY) from plant startup. (b) Approximate fluence at reactor vessel inner wall at end of life (32 EFPY). 7-1
SECTION
8.0 REFERENCES
1. L.R. Singer, et. al., " Kansas Gas and Electric Company Wolf Creek Generating Station Unit No. 1, Reactor Vessel Radiation Surveillance Program," WCAP-10015, June 1982, 2. S.E. Yanichko, et. al., " Analysis of Capsule U from the Wolf Creek Nuclear i Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program," WCAP-ll553, August 1987. 3. Code of Federal Regulations,10CFR50, Aopendix G, " Fracture Toughness Requirements", and Appendix H. " Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C. 4. Regulatory Guide 1.99, Proposed Revision 2, " Radiation Damage to Reactor Vessel Materials". U.S. Nuclear Regulatory Commission, February, 1986. 5. Section !!! of the ASME Boiler and Pressure Vessel Code, Appendix G, " Protection Against Nonductile Failure." 6. ASTM E208, " Standard Test Method for Conducting Drop-Weight Test to Detarmine Nil-Ductility Transition Temperature of Ferritic Steels." 7. ASTM E185-82, " Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)." 8. ASTM E23-88, " Standard Test Methods for Notched Bar Impact Testing of Metallic Materials." 9. ASTM A370-89, " Standard Test Methods and Definitions for Mechanical Testing of Steel Products." 10. ASTM E8-89b, " Standard Test Methods of Tension Testing of Metallic Materials." 8-1
1
- 11. ASTM E21-79(1988), " Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."
- 12. ASTM E83-85, " Standard Practice for Verification and Classification of Extensometers."
13. 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. 14. "0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors". 15. P.C. Cook, et. al., "The Nuclear Design and Core Physics Characteristics of the Wolf Creek Generating Station Unit 1 - Cycle 1", WCAP-10483, February 1984. (Proprietary) 16. D.S. Leach, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 Cycle 2", WCAP-ll251, Revision 1, December 1986 (Proprietary) 17. D.S. Leach, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 - Cycle 3", WCAP-11543, September _1987_(Proprietary) 18. D.S. Leach, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 - Cycle 4", WCAP-11956, October 1988 (Proprietary) t 19. H.M. Baker, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 Cycle 5", WCAP-12530, April 1990 (Proprietary) 20. H.Q. Lam, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 - Cycle 6", WCAP-13079, November 1991 (Proprietary) l 8-2 ~
- 21. ASTM Designation E482-89, " Standard Guide for Application of Neutron-Transport Methods for_ Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 22. ASTM Designation E560-84, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials,- Philadelphia, PA,1991.
23. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991. 24. ASTM Designation E706-87, " Standard Master Matrix for Light-Water ~ Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 25. ASTM Designation E853-87, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASYM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 26. ASTM Designation E261-90, " Standard Method for Determining Neutron Flux, Fluerre, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
27. ASTM Designation E262-86, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 28. ASTM Designation E263-88, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society-for Testing and Materials, Philadelphia, PA,1991.
8-3
t l
- 29. ASTM Designation E264-87, " Standard Method for Determining fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards Section 12, American Society for Testing and Materials, Philadelphia, P%,1991.
30. AS1H Designation E481-86, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991. 31. ASTM Designation ES23-87, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991. ( 32. ASTM Designation E704-90, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section~12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 33. ASTM Designation E705-90, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
I 34. ASTM Designation E1005-84, " Standard Method for Application and Analysis l of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM i Standards, Section 12, American Seciety for Testing and Materials, Fhiladelphia, PA,1991, i j 35. F. A. Schmittroth, FERRET Data Anal / sis Core. HEDL-THE 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979. l 36. W. N. McElroy, S. Berg and T. Crocket, A_ Computer-Automated Iterativa Method of Neutron Flux Soectra Determined by Foil Activation, AFWL-TR-7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967. 37. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al.,1981. l l C-4 i E i
APPENDIX A Load-Time Records for Charpy Specimen Tests The load-time records for the individual instrumented Charpy specimen tests are shown in Figures A-1 through A-31. e As shown in Figures A-14, A-19, A-24, A-26 and A-31, some' instrumented test records exhibit a sharp drop in load during the early portions of the test. This is caused by resonances that are excited in either the specimen or the test machine by the initial shock of the tup striking the specimen. This phenomenon is generally referred to as " ringing". It does not adversely affect t the Charpy results. It is possible to apply smoothing techniques to the data l-to remove some of these effects, but resolution-in other portions of the curve would be sacrificed. The most pronounced effects of ringing usually occur before general yielding occurrs and the analysis of the data is not seriously affected. Therefore, the data has been presented without significant smoothing. l l' A-0
.Ji i j D D A A O O L L E T R S U E T. A ( A a A R = F p 4 w P P D j A O , [ I I I I L M dr U o M c s e x r A e M m [ M i = I t T g d P a o t g I I 1 I I s ' iI I 8 I 3 1 i l = d ez i l ae d I tg 1 A eru W, g i F t L A R D E 4 A tE O G L O L y I l I l l 4 l t f y gi P Y g = t 04Os-ya n (llllllllll, J llIllillj !lflflill r!l llllf{
r WM ~ n ns w e i - - = = - g l t I I* * = s !E ~ ~ b-A ^ x ,~ _ f 4
- t..
S., e.
- g..
ties t satt a ~,. .n w a i a r g t I 4 s' I8 1 A n .u-Figure A-2. Load-time records for Specimens AL63 and AL67 A-2 l
t , = = -~ a t I, 8 ~ \\ 1 I8 t lwnh A, 7.1. ~. f lag i aggt ) ~
- ~
1 T 3 \\ [g t I 18 A_ i.. u a. Figure A-3. Load-time records for Specimens AL68 and AL61 A-3
1 I i d 4 i n, a sn an l l f I.. \\'. 18 1 e a _v_--__ ?. ..a, l I ..a on i 1 8 s 18 i I . ~,-- -n ~~
- i..
1 Figure A-4. Load-time' records for Specimens AL72 and AL75 A-4
...-..q l ~'*
== 3 1 I, 8 t I8 i hw~.m _. _. T. m ~~ ^= = i i i I g .E s ;. I8 J 1 W_ ._z. t. = .a, Figure A-5. Load-time records for Specimens AL62 and AL64 A-6
~~ =" s I i 18 I n. .a. ~'* i i i g t I 3. .s' 18 ~ n .,a. Figure A-6. Load-time records for Specimens AL71 and AL74 i l A-6
- c. v.
I ' i' ~ I.t } i i g I: s" 18 I n. c. Figere A-7. Load-time records for Specimens AL70 and AL66 A-7.
I s.. ..a It t e. ~ aa .a I: .a I8 TIE 4 8tICL 3 Figure A-8. Load-time records for Specimens AL69 and AL65 l i i A-8
-e I .m e i f 1. ( s" It l t \\ i N ii. .a. ..m .m 8 I 6 g 1 I t s .o \\ s ";.l t Ii LA A -*x. 7. 114 4 NEC B Figure A-9. Load-time records for Specimens AL73 and AT71 A-9
i ~w ...u .,o I 5 [ a. = %_.u~~ i a i g I .E.. .i l8 i A_t-m m,< n..- x. vi. e, Figure A Load-time recorda Inr Specimens AT63 and AT89 A-10
-_.-- a s i g .I 3 e I *, f-1 M AA,*_- m,. _A .9
- 4. 0
- 4. 0
- 9. 9
- 4. 4
- t. e f14 4ea I
== i I I s ;- t: l s u d MAuu m ___ i.. s. n. Figure A-11. Load-time' records for Specimens AT66 a.nd AT64 l l< i A-11 (.
- M e
snL ef9 e,9e g i i i 3 I I l g i-w_ n.- -.
- 8...
- 6..
- e..
n. .a. ~= .a.,a 8 s I8 Figure A-12. Load-time records for Specimens AT70 and AT65 A-12
1eefW a of*e news I i s i g g' } ~ .I.. S . ~. .4
- 6. 4
- 4. 6
- 8. 0
- 4. 0 n.s n.
posity a et41
- f 88 i
i 1 3 I: s" 58 n. Figure A-13. Lord-time records for Specimens AT75 and AT61-A-13
i I i oww a.t I .~ e l I8
- - l 9-1 Manna _ _
- 3.. '
n. < =r, a afv efp g 3 'I .r.. 8 s I :. -3 s.... % _ _
- s..
ftM 4 WE 1 Figure A-14. Load-time records for Specimens AT68 and AT67 A-14
-+ OwetW .EL atu atu g i 1
- e 9
i .I 9 e., N e ,e l k { .4
- 1. 0 8.4
- 3. 8 4,8 54 flat i sulEC 3 Owetw h afPS afP3 s
i i g I .e I ea w o. S e 1 .4 4.a OS
- 3. 4 4.8 lo t itsE 4 suEE D Figure A-15.
Load-time records'for Specimens /.T62 and AT73 I A-15 l
l 1: 1 l l AT72 No record. Computer malfunction. .~ .a.n. .n. N I i I:.e. s g ;- i m ..a, Figure A-16. Load-time ret.rds for Specimens AT72 and AT74 A-16
,
~~ 3 IE. s" .l* l 1 f 14 ( RAC i 1 AW72 No record. Compu',er malfunction. I Figure A-17. Load-time records for Specimens AW67 and AW72 A-17
Gae U e4 enma mas g 3 I i r I .e 8 3* I
- L% L_
.8 5.. 8.. - L.
- 4..
g4 ,i. 3 I a s I n. e r A .AA-A t ' flE 4 ellKt i Figure A-18, Load-time records for Specimens AW87, and AWB2 A I -= = -.. => i g I .r e. I ea w 7 - - i e.
- 4..
- 8..
- 3..
- 4..
to tW t e Igg y . -i i g t f I 5" l i~
- -f 4
i.. s. TM t IIICC 3 Figure A-19. Load-time records for Specimens AW69 and AV71 I A-19
6 ~~ =- = i i g l I: I8 I fit t MEtt 3 I 8 s g i-TIFE ( WEEC 1 Figure A-20. Load-time records for Specimens AW68 and AW65 - A-20.
-I i AW5 No record. Computer malfunction. - l l ~ 1 g i I s i I S ;_ f n. <=> Figure A-21. Load-time records for Specimens AW 5-and AW3 A-21
PomM a eted med i i i 1 g 1 I I a.* I w It i' I l g .5
- 6. 6 4..
- 3. 6
- e. 4 S. 6 f14 t IEEEC a g
i 4 8 a g .s.. l 5* l :. L i i..
- e..
Tim t att: ) i Figure A-22. Load-time records for Specimens AW66 and AW64 A-22 J 'l
0* MAP & est met g I i i 2 3 .I.. 3 e. 18 -_ i i e .4 4.8 8.4
- 3. 0
- 9. 8
%. 4 ft4 ( FEEC 6 i an, a -. - -a i i I a g ) I .r.. s' ] :. t
- i..
flME 4 pHEC 3 Figure A-23. Load-time records for Specimens AW61 and A%74-A-23
~~ a a. a. i i g t I ! E ~ j fi8t ( RIEE ) t I 1 .a l. I A. n.. w. n. <,ar, Figure A-24. Load-time records for Specimens AW70 and AHS9 A-34
g s I .s.. d s I8 1 i.. u. ..e, I S* I8 h n.
- 3.. '
- s..
n. Figure A-25. Load-time records for Specimens AH70 and AHS1 l A-25
I f E WW & 894 .04 ,s 3 i t i ? I E .I ~ ~ I s; g t, h_- " e --- -.. e .I 4.9 .9
- 3. 0
- g. 0
- 9. 6 m
/ I I S I8 m____
- i..
fM ( 840; t I i 1 t Figure A-26. Load-time records for Specimens AH74 and AHS7-i A-26
~~ a~ ~ 1 1 3 1 ab l-I .E..
- 2 I8 t
= <=, =-i -i a f-I S ;. It .v - - = i=> Figure A-27. Load-time records for Specimens AH04 and AH71 A-27 w e rs,-
j ~~ an 4 i i i g V I , =. 2 e. N l..
- 4. 6
- 4..
- e..
m. g d I a s l I' _:__~ i,. s. TIM iEA l l l Figure A-28. Load-time records for Specimens AH68 and AH85 l I - A-2C
t 4 \\ I~ ~ I m .{ i;.' AH62 No record. ~ Computer malfunction. Figure A-29. Load-time records for Specimens AH72 and AH62 A-29
a
- ~
i 3 I l t.' i.. TIM ( reg & AH63 i No record. Computer malfunction. Figure A-30. Load-time records for Specimens..a73 and AH63 l ^~"
l . ~ ~ ,= ~ ~ E 3 ] :.- f' .I 4.4
- 8. 8
- 3. 0 e.e
- 3. s n<
a.as > .a m g I i I .a s* /~ j t. n, <,-x, j Figure A-31. Load-time i ords for Specimens ABS 6 and AH75 A-31 C e
l APPENDIX B I Heatup and Cooldown Limit Curves for Normal Operation 4 i B-O'
TABLE OF CONTENTS le1L1931 litlR PJLqt LIST OF ILL'JSTRATIONS B-2 LIST OF TABLES B-3 1 lh,AODUCTION B-4 2 FRACTURE TOUGHNESS PROPERTIES B-4 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS B-5 4 HEATUP AND C00LDOWN PRESSURE-TEMPERATURE LIMIT CURVES B-B 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE B-9 6 REFERENCES B-20 l ATTACHMENT 1: DATA POINTS FOR HEATUP AND C00LDOWN CURVES B-21 I B-1
9 LIST OF ILLUSTRATIONS Fiqure Title EiL91 B-1 Wolf Creek Reactor Coolant System Heatup Limitations 8-14 (Heat up rate up to 60'F/hr and 100'F/hr) Applicable for the First 12 EFPY (With No Margins For Instrumentation Errors) B-2 Wolf Creek Reactor Coolant System Cooldown Limitations B-15 (Cooldown Rates up to 100*F/hr) Limitations Applicable for the First 12 EFPY (With No Margins For Instrumentation Errors) B-3 Wolf Creek Reactor Coolant System Heatup Limitations B-16 (Heat up rate up to 60'F/hr and 100*F/hr) Applicable for the First 15 EFPY (With No Margins For Instrumentation Errors) B-4 Wolf Creek Reactor Coolant System Cooldown Limitations B-17 (Cooldown Rates up to 100*F/hr) Limitations Applicable for the First 15 EFPY (With No Margins For Instrumentation Errors) B-5 Wolf Creek Reactor Coolant System Heatup Limitations B-18 (Heat up rate up to 60*F/hr and 100*F/hr) Applicable for the First 18 EFPY (With No Margins for Instrumentation Errors) B-6 Wolf Creek Reactor Coolant System Cooldown Limitations B-19 (Cooldown Rates up to 100'F/hr) Limitations Applicable for l the First 18 EFPY (With No Margins For Instrumentation Errors) B-2 o
LIST OF TABLES Table Litle P.Agg B-1 Wolf Creek Reactor Vessel Toughness Table B-ll (Unirradtated) B-2 Sumary of Adjusted Reference Temperatures (ART's) at 1/4T B-12 and 3/4T Locations ~ ' B-3 Calculation of Adjusted Reference Temperatures for the B-13 Limiting Wolf Creek Reactor Vessel Haterial - Lower Shell Plate, R2508-3 B-3
- 1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNCT (reference nil-ductility temperature) for the reactor vessel. The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture toughness properties and estimating the radiation-induced ARTNDT' RTNDT is designated as the higher of either the drop weight nil-ductility-transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.
RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper-and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has puolished a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Rac.ation Embrittlement of Reactor Vessel Material s')Ill. Regulatory Guide 1.99, Revision 2 is used for the calculation of RTNOT values at 1/4T and 3/4T locations (T is the thickness of the vessel at the beltline region). 2. FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan [2] The pre-irradiation fracture-toughness properties of the Wolf Creek reactor vessel are presented in Table 1. B-4
3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR' is obtained from the reference for the metal temperature at that time. KIR fracture toughness curve, defined ia Appendix G to the ASME CodeI33 The KIR curve is giien by the following equation: KIR - 26.78 + 1.223 exp [0.s145 (T-RTNDT + 160)] (1) where KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT Therefore, the governing equation for the heatup-cooldown analysis is defined.- in Appendix G of the ASME Code [3] as follows: (2) C*KIM + KIT IKIR where Kyg - stress intensity factor caused by membrane (pressure) stress KIT - stress intensity factor caused by the thermal gradients Kyp = function of temperature relative to the RTNDT of the material C - 2.0 for Level A and Level B service limits C - 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical-B-5
- ~ - -- l "l l At any time during th'e heatup or cooldown transient, KIR is determined by the - metal temperature at the tip of.the postulated flaw, the appropriate valu'e for RTNDT, and the reference fracture toughness curve. The -thermal ~ stresses - resulting from the. temperature gradients through theLvessel wall are calculated and then the corresponding (thermal) stress-intensity factors,; KIT, for the-reference flaw are computed. From equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures'are calculated. For the calculat'on of the. allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wa During cooldown, the controlling location of the flaw is always at tSe inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasi_ng cooldown rates. Allowable pressure-temnerature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed-for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent.on the material temperature-at -the. tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results.in a higher value of KIRatthe1/4Tlocationlforfinitecooidownrates.thanforsteady-state operation. Furthermore, if conditions exist so that the increase in KIR cxceeds KIT, the calculated allowable pressure during cooldown will'be greater than the steady-state value. The above procedures are needed because there is no direct control on ~ temperature at the 1/4 T location and,-therefore, allowable pressures may unknowingly be violated if the rate of. cooling is decreased _at various B-6 ,rr- -we-- r,,,_---,-wr ,n-+e-- n ,c,, er,.- ~- ~ -m,--r-, ~
.~ intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIR for the 1/4 T crack during heatup is lower than the KIR for the 1/4 T crack during steady-state' conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower Kgg's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation.of the pressur e-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the therme.1 gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any Thes' thermal stresses are dependent on both the pressure stresses present. e rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature cur'ves for both the steady state and finite heatup rate situations,- the' final limit curves are produced by constructing a composite cure ba' sed on a point-by-point comparison of the v steady-state and finite heatup rate data. At any given temperature, the B-7
allowable pressure is taken to be the lesser of the three va'aes taken from the curves under consideration. The use of the comp: site Jrve. is necessary to set-conservative heatup limitations because it is possi%e for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion. Finally, the 1983 Amendment to 10CFR50I43 has a rule which addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RTNOT by at least 120*F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Wolf Creek). Table B-1 indicates that the ;nitial RTNDT of 20*F occurs in both the closure head flange and the vessel flange of Wolf Creek, so the minimum allowable temperature of this region is 140*F. These limits are shown in Figures 8-1 through B-6 whenever applicable. 4. HEATUp AND C00LDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of _the primary reactor pressure vessel have been calculated using the methods discussed in Section 3. Figures B-1, B-3 and B-5 contain the heatup curves for 60*F/hr and 100'F/hr for 12, 15 and 48 EFPY, respectively. Figures B-2, B-4 and B-6 contain the cooldown curves up to 100*F/hr_ a~pplicable for the first 12, 15 and 18 EFPY of operation, respectively. No margins for possible instrumentation errors are included in the dev-elopment of heatup and cooldown curves. Allowable combinations of temperature and pressure for specific temperature change rates are below and to-the right of the limit lines shown in Figures B-l' through B-6. This is in addition to other criteria which must be met before the reactor is made critical. The leak limit curve shown in Figure B-1, B-3 and B-5 represents-minimum temperature requirements at the leak test pressure specified by applicable- -codes [2,3], B-8
The leak test limit curve was determined by methods of References 2 and 4. The criticality limit curves shown in Figures B-1, B-3, and B-5, specify pressure-temperature lir4'.s for core operation to provide additional margin during actual power production as specified in Reference 4. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher thar the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum pressure-temperature curve for heatup and cooldown calculated as described in Section 3. The maximum temperature for the inservice hydrostatic test for the Wolf Creek reactor vessel at 12,15 and 13 EFPY is 217'F, 219'F and 222*F, respectively. A vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel. Figures B-1 through B-6 define limits for ensuring prevention of nonductile failure for the Wolf Creek reactor vessel. S. CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2[Il the adjusted reference temperature (ART) for each material in the beltline is given by the following expression: ART = Initial RTNDT + ARTNDT + Margin (3) Initial RT NDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of n.aterial may-be used if there are sufficient test results to establish a mean and standard deviation for the class. ARTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows: B-9
NDT - [CF)f(0.28-0.10 log f) (4) ART To calculate ARTHDT at any depth (e.g., at 1/4T or 3/47), the following formula must first be used to atte.nuate the fluence at the specific depth. f(depth X) " Isurface(e * ) (5) where x (in inches) is the depth into the vessel wall measured from the vessel clad / base metal interface. The resultant fluence is then put into equation (4) to calculate ART lDT at the specific depth. CF ('F) is the chemistry factor, obtained from Reference 1. All materiuls in the beltline region of Wolf Creek were considered in determining the limiting material. The results of the RTHDT at 1/4T and 3/4T are summarized in Table B-2. From Table B-2, it can be seen that the limiting material is the lower shell plate, RP'48-3, for heatup and cooldown curves applicable up to 12 and 15 EFPY, and the lower shell plate, R2508-1, for heatup ano cooldown curves applicable up to 18 EFPY, Sample calculations to determine the RTNDT values fo-12 EFPY are shown in Table B-3. s B-10 m
TABLE B-1 i WOLF CREEK REACTOR VESSEL TOUGHNESS TABLE (Unirradiated)[5] CU NI I-RTNDT (a) Haterial Description (%) (%) ('F) Closure Head Flange 20 (b) Vessel Flange 20 (b) Intermediate Shell, R2005-1 0.04 0.66 -20 Intermediate Shell, R2005-2 0.04 0.64 -20 Intermediate Shell, R2005-3 0.05 0.63 -20 Lower Shell, R2508-1 0.09 0.67 0 Lower Shell, R2508-2 0.06 0.64 10 i Lower Shell, R2508-3 0.07 0.62 40 Intermediate and Lower Shell 0.04 0.04 -50 Longitudinal Welds C1 cumferentir' Weld 0.05 0.05 -50 ~ (a.) The initial RTNDT (1) values for the plates and welds are measured values. (b.) To be used for considering flange requirements for heatup/cooldowncurvesI43 1 B-ll I ~ J-
TABLE 8-?.
SUMMARY
OF ADJUSTED REFERENCE TEMPERATURES (ART's) AT 1/4T and 3/4T LOCATION ('F) 12 EFPY 15 EFPY 18 EFPY MATERIAL DESCRIPTION 1/4-T 3/4-T 1/4-T 3/4-T 1/4-T 3/4-T Intermediate Shell Plate, R2005-1 36 29 37 30 39 31 Intermediate Shell Plate, R2005-2 36 29 37 30 39 31 Intermediate Shell Plate, R2005-3 40 31 42 33 43 35 Lower Shell Plate, R2508-1 82 67 86 70 89* 73 Lower Shell Plate, R2508-2 75 65 77 67 79 69 Lower Shell Plate, R2508-3 (84)* (75)* (86)* (77)* 88 (79)* Longitudinal Welds 29 22 31 23 32 25 Circumferential Weld (7) (-11) (9) (-7) (10) ( 4; RTNDT numbers within ( ) are based on chemistry factors calculated using siirveillance capsule data. These RTNDT numbers were used to generate heatup and cooldown curves. s B-12 w n.-~,.
TABLE B-3 CALCULATION OF ADJUSTED REFEREf4CE TEMPERATURES FOR THE LIMITitiG WOLF CREEK REACTOR VESSEL MATERIAL AT 12 EFPY - LOWER SHELL PLATE, R2508-3 -1 Reaulatory Guide 1.99 - Revision 2 12 EFPY Parameter 1/4 T 3/4 T Chemistry Factor, CF ('F) 44 (33) 44 (33) 19 n/cm )(a) .548 .195 2 Fluence, f (10 Fluence Factor, ff .832 .563
- c.********
i ARTNDT - CF x ff ('F) 37 (27) 25 (18) Initial RTNDT, I (*F) 40 40 Margin, M (*F) (b) 34 (17) 34 (17) Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 111 (84) 99 (75) ART - Initial RTNDT + ARTNDT + Margin b l9 2 (a) Fluence, f, is based upon fsurf (10 n/cm,-E>l Mev) = 0.92 at 12
- EFPY, The Wolf Creek reactor vessel wall thickness is 8.625 inches at the beltline region.
l (b) Margin is calculated as, M - 2 [ 012 + oA ) 6 The standard deviation l for the initial RTNDT margin term, oi, is assumed to be 0*F since the initial RTNDT is a measured value. The standar1 deviation for ARTNDT term, o, is 17'F for the plate, except that oA need not exceed 0.5 times the a mean value of ARTNDT* 8, is 8.5'F for the plate (half the value) when A L surveillance data is used. The numbers within ( ) are calculated using surveillance capsule data. l B-13
MATERIAL PROPERTY BASIS 4 LIMITING ART AT 12 EFPY: 1/4T, 84'F 3/41, 75'F t - - - idp.1 1 it' ,re i t i I, b._. I LEAK TEST LIMIT - I I 41: :I! ! I i -,sp / I ili i ! !/i s i i 4. i .i / ! i I, li i i !/ ! ! i i i. < i i I i 3 /Q fif!f* i i ' !/ i i i I J A c'. 11. I !I! ! i .I
- I.
/ ! i /_ i ! ,m,.. I: i l* i' !!! c i /! I i ;/ i i i i ' lIt /i f i /. I!,! g i 4 J !/i ! / ! i i i ' i ! ! i i i a 1i /! 1i ! /r ! i ! i mer- ../ li/;;';;.'i 1 i
- /./
I' !A i i > + ' UNACCEPTABLE /j!//!i/{jll;,l4 OPERATION /: )
- 1.,i.
u ~u r> ~ ,,,I tr' ACCEPTABLE,,,, o w i if ~, i f i i f 4 ., i i . /! !/ OPERATION i i . : /it /. i i i/! '/ i / !/ ! y 3.m n i t 1 1 ! + t V+ i !ii* I/! ! l !/i lii! ! ! I ' 6 1 8 i i> ++ 5 g
- i p
i y . /. /. /, 1 i, i i,, i O <i iie i ie i e #
- fi i / ' /! ! /!
i i i e
- e '
i i e i iii /!!/ /i if i ie i,,' /i ! V ! /i ! / t i ! ! i t i a j i ' i i ii x i i ,bnn.i i + + Q,, vv[ i <i . f i j jj / ! lf i i i i i e i j'i o ';HEATUP RATES !/i ' / 4 . t I t, i! ' ' t ' ./ /,
- f,
y 1 i i i. t a i i
== vr Iw _f ,4 i/ i 4 i i a t O ,,ni 60'F/HR /1 /- i! / a 'I I ~ -*i O '/' !' !': CRITICALITY LlHIT q('-l': S !l'l l d'[ i !, !ll BASED ON INSERVICE l ~.~ i;.~ii, ii HYDR 0 STATIC TEST SCC l; i l HEATUP RATES ' l l l TEMPERATURE (217'F) b,'lli, UP TO ii FOR THE SERVICE T'. < i .! i i - r[100*F/HR
- PERIOD UP TO 12 EFPY l,;
+ i i i ~ .. i .T iii i i 4 i ! i ii i i i i i ! i. i i f -=Y' i 6 i i i + j i j e ii j i e i i i i e ii I I i 4 i l,i i j i , 4 i ... t 3 l^ i ii ~ IIl i i ii j i > + i i i i i i "O SO '00 '50 000 250 300 350 400 45? 20 INQlCATED ituPERATURE (DEG.F) Figure B-1 Wolf Creek Reactor Coolant System Heatup Limitations (Heat up rates up to 60*F/hr and 100*F/hr) Applicable for the First 12 EFPY (With No Margins For Instrumentation Errors) B-14
MATERIAL PROPERTY BASij LlHITING ART AT 12 EFPY: 1/4T, 84*F 3/4T, 75'F u-- ww : p [ _413 *3 i a j i i
- ii r
- i i I i j i 1 e i i,
i I e i .I i ! /! i i i i 8 t 21 1 i i i i .3gC 3, i li i i i, i,i, i if!! i l i i i e i 1 e 4 e < i i i i ..j i/i, i i i ili i i i + . i I i fi i i j i 1 2000 ,i ,j 1 i, . i i 1 + i l! I i i ! ' I .I* ii~i+ i i i { i ' + + i ! i i i i i i j i, i j if ! e i i i 4 i ; 4 3 74 ! i i j,',* C i i j i e t i f 4 4 i j j j i 4 3 I i i ' ! 3 t/ i UNACCEPTABLE ' '_i !, !,! '.,,i - i ' i ~ OPERATION / i i i 4 15CCj ,, ; i ,i i ,f ! o i I i i i ~ i ,/! i g i i t i i i 4 i /i i ,ii;i i + i f, i, i i l,i,,,
- fi, : ^ i ! i ACCEPTABLEi i
i i ~ i W IN i 3 l :fi l.; lll OPERATION ! i l,, ii j ' i
- i
~ !/ .! i i i i, 4 /! i e t i i I. { i! i oc e 1
- VL-fi i e i
j j ! i i j {. . i iii,i jl + i i ~I f' i i e t i ! i t. , i fe S b C00LDOWN RATES " ii i !i i ii i
- F/HR -
i i ,' ' ' ' i I, ' ', I' u_ 7 ". C 0 i g 9 20 i ' i i i i iii i
- ! ' ',,,L
- 4. -.
i i i,, i i 40 -_ 4v! !t 60 s ! i i ii i i iii! ,00 i'i i i ' i i ' ' ' ' 100 ' i i i ! ' i t-i 4 i ;i. i... i i i i, i, i r 4 ,ii,,,, i, i ! ! ' 1 i iiI t ! i i i i i ii e iii i i i e i i i i I i i i e i i j i i, , i 4 i i i i i i i i i i i i ij i 6 i i 4 l t { l i '. 1 i i i; i t 8 [ <; w,,. 4 ! ! i I ! i ; i_i_i i !iij } i : : ! ! .2 ! i 1 - i i, i i i, ,i 50 ' 00 50 200 250 300 353 400 450 soo INDICATED TEMPERATURE (OEG.F) Figure B-2 Wolf Creek Reactor Coolant System Cooldown liettations (Cooldown rates up to 100*F/hr) Applicable for the First 12 EFPY (With No Margins For Instrumentation Errors) B-15
MATERIAL PROPERTY BASIS LIMITING ART AT 15 EFPY: 1/4T, 86'T 3/4T, 77'F I 'k lQ ../ i i ? !. i1 i ;
- ! : /l i i
,1!i!; 4 i 't i , i f f i F -LEAK TEST LIMIT ;i /! t li i I il I! ! '* gi i 'l l I il : l -s c ' i ! I L 4 ' . i II. i i i / i ?/i / tIi ! i i i ! i ' i i ! !) il ii! I I i ;ITI il !, i 'i ! i 4 + i e i !/ .I
- /i1
/ !t ' + i + i i i ! f !/ i i il I! i + 6vggi i i i - + i j il tI I f. ., n i i a r i i i i , i i / if /, ifi ,i , ii i lii li. + L i > i i! !a u' i i i 1 ! ! /! I t /i / I ! _ 1 + 1 > t i i/1 / i/! _ f i i ! i i i ,,: e i i ) Ji I! J< il ' c 1 "i UNACCEPTABLE Ti .1i I! I i iz i i i> OPERATION - I I.I:/ I H. Ii H ! [: l/ ACCEPTABLE l' i !i i ^ t i / / OPERATION i u n'"I O ii! i e i ii ie iii ifi i ], i e i i i h I ~ / i /. / I ! f. ! i 4 a ii i i, / ) iie f fjfj /* I i i e i i i i* r i / f i/, j ! i i e i *. e i i/ ' I' r Y ' I! t t ' i ! u inc - !r i i.'! i a vi v;4 i, i i , j i o i i ! v .fvi / i -- i ! ! i i,, l i i ! ! il fi/ ! /i i l 4 i i i i 5 i i u i > 1 / Ji ! / 1 i 1 e i ! i 2 innn! i i! ii<iiii i/ fi / i !/ 1 i i i >iii>< 4 '"**I i I ii fi / /{ i/ i i I I i i i c a HEATUP RATESii! 4/ ! /i #'i i i i t i! UP TO / i f fi ! i e, e i i i ! 60'F/HR , v.<iii/<! ,5C 2 i i CRITICALITY LIMIT i i i ! i
- iii i
!L i _ BASED ON INSERVICE I ii lI --- li i ' l HYDROSTATIC TEST lll,'li' i TEMPERATURE (219'F) ii!. '-CC HEATUP RATES - FOR THE SERVICE i i H. '. UP TO i PERIOD UP TO 15 EFPY l i ii i i ! i i 100 F/HR .,ri !ri!. i i i ! i i i i, i i i i i i i i 2 i i .,, i i , i i i < i , ie s e i i i i i, r > 4 i l ! i i i ii. i i I i ' + i ! ! i + 1 ~~3 i I ,i ! j i i i i + i e 1 i i i i I 1 i ~ :' 50 '00 '50 200 250 300 350 &OD 45; i. INDICATED TEMPERATURE (DEC.F) Figure B-3 Wolf Creek Reactor Coolant System Heatup Limitations (Heat up rates up to 60*F/hr and 100*F/hr) Applicable for the First 15 EFPY (With No Margins For Instrumentation Errors) B-16
HATERf*1 PROPERTY BASIS LIMITING ART AT 15 EFPY: 1/4T, 86*F 3/4T, 77'F 'w- -
- J _ c-c 1_:
1 i 4 = . i i e i i i f 1 c a ~~ g: ! I: n 4 ,/ i i i < i J. i ; n t I e a I i / j I ~
- ggg[
l' I i / ! i t I i c n ii/ i i Ifi i i + t / ; , i .,._'"i UNACCEPTABLE f O j j fi i i i h l - / l. ACCE? TABLE, /i OPERATION t i w usc: i, a 1 ! I i, i,. i i gg 1 i i i i -/, 6 4 i e i 3 / j ! i i i,, i / i i i 4 9
- i y
/! CE iqnn: . -/: i i 4 i i t r i i c.
- f T
7 a M - C00LDOWN RATES F' l i,,'!l O. ,, C r- 'F/HR i O c i i i i ,i ,, i, z M 20 i ; j i,i; i ^ , ;4 ~ C 40W ' ' ' - ! i ! - 'i- '.i,' i 500 6 60'p' i !,' i i ', ' ',' i ' ' ' ' i y 4... ..i 100 i 4 - 4 i . i i i i i ! i ii..i i i i i i, i gf I i , i ! ! I i ij i i i + y,,, i 1 3 i 6 l ii i i,, i s i !, t i j i e i 6 i! i. e i_ i r 4 i i i j ~ : 50 ^;0 ~5; 200 250 300 353 c03 25; 5:: INDICATED TEMPERATURE (OEG.F) Figure B-4 Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100*F/hr) Applicable for the First 15 EFPY (With No Margins For Instrumentation Errors) B-17 m- . - +,,. - e. - + -, - - - gw
~ MATERIAL PROPERTY BASIS i LIMITING ART AT 18 EFPY: 1/4T, 89'T 3/4T, 79'T u-- .-i, r s ! il i 1 i, i e ss ' i it i I,i! 1, LEAK TEST LIMIT ~ - I! I! I I!I' I i I_ ' fif J! 1 i i 1 il 'J !I _I i "g: +
- st iii i I i i i
iI i
- l i i l! I ih i
! i i ' i.' i i .,,i i I i i i i/ I i !Ii f ili i i i i i + ! e il i ) ! I /ii Ii i 'I i II ! f II i' t + i I il ! I i/! Ii i 'g3 : 4 i ' i i ' ' ' _!I ! Ii I! /> i i i i !! !
- i i+if ili fi I
! i + i ! I /!/ i I ! !t ! i i ' 4 i UNACCEPTABLE ! ! il I !Ii i i ! a i i ! + 175 OPERATION l'/ /> !! i / ' ' , i ,1 ,I ACCEPTABLE i 4 4 l,' l l
- f !
7-,0PERAT10N !ll'lj i ii i I ^ 1500-o i ii i
- i i
! /, +(. /! ! (! t i 1 i 7 ~ i 7i / /i./ i i + ' i i a. i i I! !/i / ! i/! ! i! i + ' +i i ii /! ! (!/ ! ( i i w i i i i e i e i. !9 i i j 1 /i ! z ;/ ir i i 1 1 ! i, i, ! 1, g3c ! <i a i ifI 7it i/ ! ! i ! l ! I i i a D i r i ! i * '/ ! I !/ Ie i ! i ! ! i, i i j, / ifj f f4 i j t i ! j i e i t i i i i i < > i i+ ! ;/ / !/! /'ii! i i + i w i i i i if rocc!l ! ! : HEATUP RATES iiI#, f l/ l [ ! l l ! i l,' ' 1 l l f UP TO i i ; / i./ ! ! !/i ii i, i i i. ! o l 'l 60'F/HR I / ,1 , v' i i i i w i 0
- r. i
~ iTiA vi / i i i i ! 4
- l
i i i i i I i 'I i i i ' i i i r! i i i J i i -- g( l g!ff! I I I i I i i i --e
- /.
1 1 i i i. 1 i 1 . 1 1 I i f i ie + 1 , i i i , i i i /j i i i i 1 , i i Ji i i e i i Sncr ! ! i 1
- i !
/ ! I t ! i i i ---wi i i ii i i i ~ ~ ' j I ' i fi i 1 i i i i ! t 8 i fI i ! ! I i ( I i
- I i
i j i i e i i _f+ i l j i i_ i i a 1 i i, i i I ,, g 31 } > > i i i /i i i i i + i i ' ** f , i .f i e i t i: ~ i < ii y e i F-UNACCEPTABLE i i 4 F+ i OPERATION W/rri ; ; !,iii1 i !! !i' i r i 4 i i ; , S c,. u o i i f, i i 4 3 / i g ,i i !.. i, ,i i, j i i ![ i I 5 i r j 1 ) ', 4 g ' l i ji ! ! Ii i ! i I. 1/ i ! ! ! ' ! i ).1 ! 1 i t y g - i, ,,,iii! i i ! i /!! ! i ! ! i i i i i i i 23 i + i i! i i i i/ ! i i i i + i 4 i<!i: ! ACCEPTABLE-i! i'!./ii i w ^: ! OPERATION E 10CC i ll /. lllI i i1 i j' ' : i o i ' i -. !. i p-C00LDOWN RATES l !ql,!ll l! !l l,',, l. 2 i i O 750p F/HRt!ii i t i i i! i i i i i ! i i ii, , i i - + i i 0 -o f E 9 20 ~ 1 ~. i i r t t i I e i ' i ' i i i i i r i. i. i i. i 40 e i ! i t '
- CC r t 60hpq i.
i i iI i ! i i ! i i i., , j,;;;,,,, ,i pt 10V !ii i i ! i ! i i ' i ' i i ' ' ' - i i i ! ! i i i a ! i ! i i i i i ? ! 1 I '! i ! i i I II i! ! ' i ! i ! ' i -c.i 4 I i i t i i i i i i I i i i i! i i 6 i > 1 I ! - i 1 , i i ! _ 4 . i.,< , 1 1 i e , j i i i i i jj# i i t il [ i e i i ! I ' i i a e e i ! j i i, i ! i [ } e e i i i 4 e j i i e ij t i i I ! ! ! I i I ! i i ! i i ' t i I i f' i g p l j - a i i e. I i i t i i !. I t 4 s 50 '00 '50 200 250 300 350 400 450 SCC lt:DICATED TEWPERATURE (DEG.F) Figure B-6 Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100*F/hr) Applicable for the First 18 EFPY (With No Margins For Instrumentation Errors) B-19
o 6. REFERENCES 1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Comission, May,1988, 2 " Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981. 3 ASME Boiler and Pressure Vessel Code, Section !!!, Division 1 - Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure", pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York, 1986. 4 Code of Federal Regulations, 10CFR50, Appendix G, " Fracture Toughness Requireme'.ts", U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Vol. 48 No. 104, May 27, 1983. 5 h*AP-ll553, " Analysis of Capsule U from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program". S. E. Yanichko, et al., August 1987. B-20
m i i i l i T t i l l l ATTACHMENT 1 DATA POINTS FOR HEATUP AND C00LDOWN CURVES (With No Margins for Instrumentation Errors) I B-21 l, -
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SAP COOLDOwN CURVES REG. GUIDE 1.99.REV.2 WITHOUT MARGIN 05/19/92 THE F OL LOwlNG D A T A WERE PtOffED FOR COOLDOwN PROFILE 2 6 20 DEG-F / HR COOLDOWN I 1RRADIAi!ON PERIOD = 15 000 E f P VEARS FLAW DEPTH = ADWIN i INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED T E MP E R A TURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG F) (PSI) (DEG. F ) (PSI) ( DEG. F ) (PSI) 7 115.000
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85.000 4HP-fCj 6 8 120.000 7te-*$ 13 145,000 936.97 2 90.000 14 150.000 976 71 j 3 95.000 frfr5-59 ' (,21 $ 9 125.000 eet-*1 le 2.1 + 15 t55.000 tO19.30 4 100.000 N 10 130.000 m' 16 tGO.OOO 1065 03 6 110.000
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SAP COOLDOWN CURVES REG. GulDE t.99.REV.2 wlTHOui MARGIts 05/19/92 THE FOLLOWING DATA WERE PL O l i E D F OR COOT DOWN PROF I L E 3 4 40 OEG-F / HR EOOLDOwn I i IRRADIAflON PERIOD = 15 000 E T P V E ARS FLAW DEPfH = AOWIN i I INDICATED INDIC A I E D ltolCATED INDICAff0 IPOICAIED INDICATED L TEMPERAiuRE PRE S5URE T E MPE R A TURE - PRES 50RE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG. f i (PSII (DEG fp (PSID 1 85.000 597.38 7 115 000 M' 13 145.000 925 03 2 90.000 615.02 8 120.000 6 14 150 000 967 Gt t [ b28 4 15 155 000 1013 45 3 95.000 69+-097 9 125 000 r' 4 100.000 t N 130.000 e++-99 } 16 160 000 1062 66 h lh Qg sr 10 5 805.000 It 135.000 ete-et' 17 965.000 tt15 59 6 't10.000 '90 0' 12 140.000 885.27 1 \\M L fd (A C M R* r r .. e Lab y N I (. t E -I = q ,n.,, e eyw.e - m a we,,
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4 SAP 60 & 100 DEG-f/HR aftATGPS REG.GulOE f.99.REV. 2 ttI TH0sli CARGIN 05/19/92 THE fOLLOWING DATA bERE CALCULATEOFOR THE INSERVICE HVDROSTAilC LEAM TEST. ~ MINIlsOM INSERVICE LEAK TEST TEMPERATURE ( 15.000 EfPV) I PRESSURE (PSI) IfMPERATURE (DEG.F) 2000 198 f 2485 219
- RESSURE PRESSURE STRESS t S K1M (PSI)
(PSI) (PSI SQ.RT.IN ) 2000 21130 89059 2485 26254 891616 u .. m '5 -g =. V er
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SAP 60 & 100 DEG F/HR HEAIUPS REG. GUIDE e.99.REV.2 WtiHoui MARGIN 05/19/92 COMPOSITE CURVE PLOliED FOR HEAfuP FROFIEE 2 HEATUP Rait(5) (DEG F/HR) 60.0 = IRRADIATION PERIDO 95.000 Efe VEARS = FLAW OEPTH = (t-aOWINIT INDICATED INDICATED INDICATED INDICA 1ED IND IC A T E D INDICATED TEMPERATURE PRESSURE T E MPE R A TURE PRESSURE T E MPE R A t uR E PRESSURE (DEG.F) (PSI) (DEG.F) (PSI) (DEG.F ) (PSI) 1 85.000 9tt-*S[ gg 14 150.000 750.34 27 215.000 1324.97 2 90.000 MI 15 155.000 777.13 28 220.000 1395.89 3 95.000 603.43 16 160 000 806.09 29 225.000 1471 73 4 100.000 604.48 17 165.000 837.53 30 230.000 1553.08 5 105.000 608.64 18 170.000 87 76 31 235.000 1640.25 6 110.000 615.36 19 175.000 908.62 32 240.000 1733.49 7 115.000 h 20 180.000 948.32 33 245.000 1833.30 8 120.000 6dFr-46 21 185.000 991.14 34 250.000 1939.81 - 9 125.000 649-4a -(palt 22 190.000 1037.24 35 255.000 2054.00 ' 20 130.000 669-90 23 195.000 1O86.87 36 260 000 2475 54 11 135.000 699-+4 J 24 200.000 t140.24 37 265 000 2305 52 12 140.000 703.71 25 205.000 1197.59 38 270 000 2444.05 - 13 145.000 726.04 26 210.000 1259.09 4F%e < ephe d 3 tal Cl + o E
S AP GO & 'OO DEG-F/HR HEATUPS REG.GU10E 1.99.REV.2 iWITHOUT MARGIN 03/19/92 COMPOSITE CURVE PLOTTED fOR HEATUP PROFILE 3 HEATUP RATEt5) (DEG F /HR ) 100.0 = IRRADIATION PERIOD 15.000 E F P YEARS = FLAW OEPTH * (t-AOWIN)T IPOICATED 1FOICATED INDICATED I PO IC A T E D Ito!CATED IPOICATED TEMPERATURE PRES 5URE 1EMPERATURE PRESSURE TEMPERaIURE PRESSURE (DEG. F ) (PSI) (OEG.F ) (PSI) (DEG.F ) (PSI) 1 85.000 GW 15 155.003 650.14 29 225.000 1136.51 2 90.000 599-9e 16 160.000 668.41 30 230 000 t195.95 3 95.000 59iP-e9 874.D 17 165.000-688.77 34 235.000 1259.71 4 100.000 Se+-e4 18 170.000 711.18 32 240 000 1328 17 5 105.000 Ste-esJ 19 175 000 735,98 33 245.000 1409.71 6 910.000 574.93 20 180.0 % 763.08 34 250.000 1480.59 7 115.000 575.32 21 185.000 792.49 35 255.000 1564.98 8 120.000 577.75 22 190.000 824.78 36 260.000 1655.37 9 125.000 582.29 23 195.000 859 65 37 265.000 8752 26 10 130.000 588.76 24 200.000 897.39 38 270.000 1855.96 11 135.000 597 09 25 205.000 938.23 39 275.000 1966 60 12 140.000 607.45 26 210.00u 982.34 40 280.000 2084.85 13 145.000 619.75 27 '95.000 1029.94 41 285. 00 2210.97 14 150.000 634.03 28 220 000 1081.24 42 290 000 2345.58 tD e tab l SP e
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SAP COOLDowN CURVES REG. GUIDE t 99.REV.2 WfitOUT MARGIN 05/19/92 THE fOLLOWING DA?A WERC PtOfIED FOR COOLDOWN PROF ILE 2 4 20 DEG-F / HR COct Dowt4 ) IRRADIAllON PERIOD = 18 000 EFP VEARS e FL AW CEPitt
- ADWIN i ItotCATED INDICATED INDICATED IPOICATIO ItOICATED ItotCATED T E MPE R A T UR E PRESSURE TEMPER *iURE PRES 5URE TEMPERATURE PRESSURE (DE G.F )
(PSI) (DEG.F) (PSI) (DEG.F) (PSI) 8 85.000 GiHH4a 1 7 115.000 799-t9'l 13 145.000 916.99 2 90,000 999-69 8 12O 000 743 4 ' NI Y 14 150.000 955.06 esa-tf 9 125.000 6' 3 95.000 r 15 155 000 ,o3,5 99 99 4 100.000 m , f,.2 i + io i3o.ooo o,, oc is iso.ooo 97 5 105.000 ++s-e9 11 135.000 e*e-*+ 17 165.000 1087.29 6 110.000 ttS-t5-12 140.000 881,51 18 170.000 1838.14 h y reg e +4.& W + to 4 (Al 4.0
SAP COOLDOWN CURVES REG. L 'OE t.99,REV.,2 WITHOUT MARGIN 05/19/92 THE FOLLOWING DATA WERE Pt011ED FOR COOLDOWN PROFILE 3 ( 40 DE G-F / HR
- 18. EK)O E F P vEARS IRRADIATION PERIOD
. FLAW DEPTH = ADWIN T INDICATED I PO I C A T E D IN01CATED INDICATED I PO !C A T E D INDICAIED TEMPERAIURE PRESSURE IEMPERATURE PRESSURE T E MPE R A f t*RE PRESSURE (DEG. F ) (PSI) (DEG.F) (PSI) (DEG.F) (PSI) 1 85.000 588.40 I i15 000 +++ 13 $45.000 903 31 2 S0.000 605 22. 8 120.000 N ' W + 14 ?" ' '" 150.000 944.22 3 95.000 6f9-5+' 9 125.000 15 155.000 988.26 4 100 000 6+9-04 to I30.000 Tse-99 16 160.000 1035 62 5 105 000 M
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SAP COOLDOWN CURVES REG. GUIDE 1.99.REV.2 WiifOUT MARGIN 05/19/92 THE FOLELwlNG l'AT A WERE PLOllED FOR COOLDOWN FROFILE 5 ( 100 DEG-F/HR COOLDOWN ) 16.000 EfP YEARS IRRADIATION PERIOD = FLAW DEPTH
- A0 WIN i INDICATED lt4)l C A I E D
!POICATED INDICATED INDICATED INDICAIED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG. F ) (PSI) (DEG.F ) (PSI) IDEG.F ) (PSI) i 1 85.000 485.49 7 115.000 N 12 140.000 829.92 2 90.000 506.17 8-120.000 630-68 13 145.000 877,82 3 95.000 528.66 9 125.000 4GG-GG ' h3.1 t 14 150.000 929.37 4 100 000 552.81 10 130 000 9++-1+ 15 s55.000 984.79 5 105.000 579.04 tt 135.003 444-63 16 160,000 1044 66 6 110.000 607.20 k Ft
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l a E APPEllDIX C-Calibration Reports 5 l I' k -C-0 i-l' d m
WESTINGHOUSE SCIENCE & TECHNOLOGY CEhTER CALIBRATION RECORD DATE: 11/16/91 INSTRUMENT BEING CALIBRATED:Tinius-Olsen Mod.,74, Ser. # 123159 Impact machine LOCATION: Bldg. 302A low level cell OPERATOR: L. M. Thomas CALIBRATION METHOD: NIST standards per ASTM E23-91 RESbLTS: Series SRM No. Average Value Expected Range Acceptable Not Acceptable it-lb it-lb LL23 2002 12.4 11-15 X HH24 2096 71.8 66-78 X CALIBRATION ACCEPTABLE X NOT ACCEPTABLE COMMENTS: Machine dial values oniv. '?- ti bibl I APPROVAL: )
- k. P. Shogan Date Vanager, Nuclear Services & Materials Testing l
l C-1 L l
c Ecport & Gertificate of Berification-C&M COLLINS CALIBRATION SERVICE THIS IS TO CERTIFY that the following described testing machine has been Calibrated by C&M Call. BRATION and the loading range (s) shown-below found to be within a maximum tolerance of ). MACHINE [N.S I/ O A/ 2 'f E / TYPE / M fr//- b dee LOCATION MP3 Nac kowr 8/O r Iw. CAPACITY 20 CW /N /f 9 D (en 4 m Chunk // lA SERIAL NO. It(A DATE OF VERIFICAllON Jum" * /M/ Method of verification and below recorded data is in accordance with A.S.T.M. E 4 #9 & Mll. STO. 45662 A. The testing device (s) used for this callbration have Certifications traceable to the NationalInstitute of Standards Technology. Machine Readings G Pounds O Newtons O Kilograms Readings temperature corrected for _t 3
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u came causmanon otvictl uacamsmaaoa c.o uaesma causmanon ocvice unenmicaaoa eo l aeAomo ne tomo jgf No atAomo atAoms /p wo. CO C*o ? ? -?? ."+ I 200 2 0 / !! ~ 1.1/ .TL i 10 D toa 39 .39 .39 l 't0 D 40? 1s ~ ? !1 .CE 2 200 ee! ?) 1.23 .G? I E00 90+(? .uL? .rt 2 200 3 0) 3c. -l39 xS L /200 1707 ? ~ 2. ? .bo ? %O uo 3.1 t ~ 1 I? 7% i JbOO /G OI. 3 -I. 3 .ct E 500 yo y so -7.so .so t ? coo ? oc s.s - s.s .31 3 /00 Joo.?1 - 21 23 i S~o o so?st ~ t. 31 .4E 2 200 Po l.ss -) a .IJ l /000 soor.? 7z 7? ? 400 vo? 71 - ?., s 6E E 2000 20s1.2 - 13.8 .st J l f00 Ec3 t? ~ 3 ?! .s3 E 1000 lo ?!. 7 - t r.1 .n ) 70 0 to s..ti ~ g 11 ,sb ? 4000 uo ? t. I ~ ?V. i .70 ) l000 lo c r. - 6 - c. l .?G 5000 5004.s ~ + S~ ,o9 CAllBRATION APPARATUS - Morehouse Proving Rings. Stainsense load cells & Troernner Dead Weights. V traceable to the NationalInstitute of Standards Technoloog in accordance with A S.T Milatest speelf6 cations C.D. SERI AL NO. LO ADING RANGli CALIB. NATION A.L INST;TUTE oF STANDARDS TECHNOLOGY LAB N0., CODE CLASS A VALUd DATE TEMP. oF C.D. 1
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'C M ACHlhE RANGE LDADikG RANGE Piienary Leadindicating Device 1L ' t 1 a ci,,,,. + O 5"oo Fo-Soo C & M COLLINS CALIBRATION SERVICE O ,yg fgo, ,ey n 230 Haymont Dr., Box 149, Gibsonia, PA 15044 O to oo en o. ceno -(412)443-7631 O ' ann o soo sooo B y'- Ae fN t < service suomeeen - Witnessed Dy: C-2
IReport & (!Iertificate of Berification C&M COLLINS CAllBRATION SERVICE THIS IS TO CERTIFY that the following described testing meChine has been Calibrated by C&M Call. BRATION and the loading range (s) shown below found to be within a maximum tolerance of /,
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MACHINE INblSU N 3 Y El TYPE Lmd C= // ~ 're. rsiv LOCATION OFJ Nap b n e 8/ec h!r Ice' A CAPACITY Pooc>u/6e kf & le,, J r / ( %,e t, 1 i FA SERIAL NO. /td DATE OF VERIFICATION Juoe u- /99 / Method of verifiestion and below recorded data :s in accordance with A.S.T.M. E.4 f9. & Mll. STO. 45662.A. The testing devicets) used for this calibratien have cert!!! cations traceable to the Nationalinstitute of Standaros Technology, Machine Peadings Cil Pounds O Nevitons O Kilograms Readings temperature corrected for 7? 'C. M A C man E C AutRAYiON DEvK.E MACM4NE ERROR C.Q M ACMIN E CAL 18AAftON DEVICE MACMINE ERROR CD R E ADINo NE A0sNd jg, No READINQ REALIN:)
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/ r_ 0 0 /0 0 3 t, ~36 .36 ? O L) O 2 0/ +. ] -N $ 7? E i L+000 L.* c t 7 ? 17 ? N3 E (COO ?M66 ~%l 61 E V 000 50GT o Lco .tt c '000o /00 R - 18 3t 2000 2005.s ~ T. S .?8 ? 4000 19 &S' fr 4%? .</ ? @dCo 7999.6 4/0 .o / 2 12000 tp-i/ -1/ .04 E /6000 /& 0 ?3 ~ 2.1 .sv ? i 20000 20otV - 18' . o '/ CALtBRATION 4.PPARATUS Morehouse Proving Rings Stainsense load cells & Troemner Dead Weights. Verifications traceable to the Natfor'al Institute of Standards ".Jehnotony, in accordance with A.S.T.M lateet specifications. C.D. S E RI Al. N O. LOADING RANGil C ALIB. N ATION AL INSTITUTE OF STANDARDS TECHNOLOGY LAB No. TEMP. CODE CLAS$ AVALUM DATE OF C.D. 1 47.S'7 2 vr. Ph /6s CVT. to 717 2 ? Cp 'r.3 'r 2] 'C 2 Le 7 5y I WW-?v 4 r ? V. o o 717 t ? 9 '! T 'r 912O?l7+T547 t3' 'C 3
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/o oce i&i / re o o - eooo C & M COLLINS C AllBRATION SERVICE d-Jo mu /6f Poo s m voo 230 Haymont Dr., Box 149, Gibsonia, PA 15044 (412)443 7631 By: lM~ f T't!~ ' ' (SERVICE ENGINEER) Witnessed By: C-3
Attacitment 2 to ET 92-0216 Page 1 of 2 Current and Projected Values of RTPTS l l-l l r,
...... _ ~. - to ET 92-0216 Page 2 of 2 In accordance with the revised Pressurized Thermal Shock Rule. - 10 CFR 50.61, the current and projected values of RTPTS were calculated using methods from Revision 2 of Regulatory Guide ~ 1.99, Chemistry Information from previous submittals, and current surve ' lance capsule results. These values do not exceed the 10 CFR 50,61 screening criteria of 270'F for plates, forgings, and axial weld materials or-300'F for circumferential veld materials. RT,,3 = 1+M+MTr13 1 l I = Initial RT, un M = Uncertainty Margin ART,3= UP*'*" p f = Neutron Fluence CF = Chemistry Factor 4.79 EFPY 32.0 EFPY l' Material Cu (1) Ni(1) CF (2) I(1) M (2) f(3) RT m f(3) RT m R2005-1 0.04 0.66 26 -20 34 0.362 32.70 2.5 .40 R2005-2 0.04 0.64 26 -20 34 0.362 32.70 2.5 46.40 R2005-3 0,05 0.63 31 -20 34 0.362 36.30 2.5 52.63 i R25081 0.09 0.67 58 0 34 0.362 75.72 2.5 '106.28 R2508-2 0.06 0.64 37 10 34 0.362 70.62 2.5 90J11 R2508-3 0.07 0.62 44 40 34 0.362 105.65 2.5 128.83 1.ong Welds 0.04 0.04 27.8 -50 56 0.362 26,00 2.5 40.64 Cire Weids 0.05 0.05 31.75 50 56 0.362 28.84 2.5 45.57
References:
(1) Previous WCNOC 10 CFR50 61 submittat KMLNRC 86-015 p21 (2) New 10CFR50.61 rule . (3) WC AP.13365, surveillance capsulo Y onolysis report, p6-27 l I l L l l. 1: .}}