ML20082C546
| ML20082C546 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/31/1983 |
| From: | Cheney C, Kunda M, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20082C540 | List: |
| References | |
| TAC-53266, TAC-54234, WCAP-10425, NUDOCS 8311220115 | |
| Download: ML20082C546 (86) | |
Text
..
WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION N
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1 ANALYSIS OF CAPSULE U FROM THE ALABAMA POWER COMPANY JOSEPH M. FARLEY UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM l
l l
M. K. Kunka
(
S. E. Yanichko C. A.' Cheney
(
W. T. Kaiser October 1983 l
Work performed under Shop Order No. WKA 6820 APPROVED:
T. R. Mager, Manager Metallurgical and NDE Analysis j
Prepared by Westinghouse for the Alabama Power Company.
Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees 6
without the customer's approval.
g WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems 0311220115 831110 P.O. Box 355 DR ADOCK 05000 DR Pittsburgh, Pennsylvania 15230 p
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TABLE OF CONTENTS Section Title Page 1
SUMMARY
OF RESULTS 1-1 2
INTRODUCTION 2-1 3
BACKGROUND 3-1 4
DESCRIPTION OF PROGRAM 4-1 5
TESTING OF SPECIMENS FROM CAPSULE U 5-1 5-1.
Overview 5-1 5-2.
Charpy V-Notch Impact Test Results.
5-3 5-3.
Tension Test Results 5-4 5-4.
Compact Tension Test Results 5-4 6
RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1.
Introduction 6-1 l
6-2.
Discrete Ordinates Analysis 6-1 l
6-3.
Neutron Dosimetry 6-4 6-4.
Transport Analysis Results 6-8 6-5.
Dosimetry Results 6-8 7
SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8
REFERENCES 8-1 Appendix HEATUP AND COOLDOWN LIMIT CURVES FOR A-1 A
NORMAL OPERATION
= a i
t
LIST OF TABLES Table Title Page 4-1 Chemical Composition and Heat Treatment of The 4-3 Farley Unit 2 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for The Farley Unit 2 5-5 Intermediate Shell Plate B7212-1 (Transverse) Irradiated at 550 F, Fluence 5.61 x 10 a n/cm* (E > 1 MeV) i 5-2 Charpy V-Notch Impact Data for The Farley Unit 2 5-6 Intermediate Shell Plate B7212-1 (Longitudinal)
Irradiated at 550 F, Fluence 5.61 x 10 (E > 1 MeV) 5-3 Charpy V-Notch Impact Data for The Farley Unit 2 5-7 Pressure Vessel Weld Heat Affected Zgne Metal Irradiated at 550 F, Fluence 5.61 x 10' (E > 1 MeV) 5-4 Charpy V-Notch Impact Data for The Farley Unit 2 5-8 Pressure Vessel Weld Metal Irradiated at 550 F, Fluence 5.61 x 10's n/cm (E > 1 MeV) 5-5 Instrumented Charpy impact Test Results for 5-9 Farley Unit 2 Intermediate Shell Plate B7212-1 (Transverse) 5-6 Instrumented Charpy Impact Test Results for The 5-10 Farley Unit 2 Intermediate Shell Plate B7212-1 (Longitudinal) 5-7 Instrumented Charpy impact Test Results for 5-11 Farley Unit 2 Heat Affected Zone Metal 5-8 Instrumented Charpy impact Test Results for 5-12 Farley Unit 2 Weld Metal 5-9 The Effect of 550 F Irradiation at 5.61 x 10's 5-13 (E > 1 MeV) on the Notch Toughness Properties of The Farley Unit 2 Reactor Vessel Materials 5-10 Tensile Properties for Farley Unit 2 Reactor Vessel 5-14 Material Irradiated to 5.61 x 10'8 n/cm2 6-1 47 Group Energy Structure 6-10 6-2 Nuclear Parameters for Neutron Flux Monitors 6-11 6-3 Calculated Fast-Neutron Flux (E > 1.0 Mev) and 6-12 Lead Factors for Farley Unit 2 Surveillance Capsules 6-4 Calculated Neutron Energy Spectra at the Center 6-13 of Farley Unit 2 Surveillance Capsules 6-14 ii
LIST OF TABLES (cont.)
Table Title Page
~
6-5 Spectrum-Averaged Reaction Cross Sections at the 6-16 Center of Farley Unit 2 Surveillance Capsules 6-6 Irradiation History of Farley Unit 2 Reactor Vessel 6-16 Surveillance Capsule U 6-7 Comparison of Measured and Calculated Fast-Neutron 6-17 Flux Monitor Saturated Activities for Capsule U 6-8 Results of Fast-Neutron Dosimetry for Capsule U 6-18 6-9 Results of Thermal-Neutron Dosimetry for Capsule U 6-19 i
6-10 Summary of Neutron Dosimetry Results for Capsule U 6-20 4
et 6
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+ e iii
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LIST OF ILLUSTRATIONS Title Page Figure
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4-4 4-1 Arrangement of Surveillance Capsules in Farley Unit 2 Reactor Vessel (U adated Lead Factors for The Capsules Shown in Parentheses.)
4-2 Capsule U Diagram Showing Location of Specimens, 4-5 Thermal Monitors, and Dosimeters 5-1 Irradiated Charpy V-Notch Impact Properties for 5-15 Farley Unit 2 Reactor Vessel Intermediate Shell Plate B7212-1 (Transverse Orientation) 5-2 Irradiated Charpy V-Notch Impact Properties for 5-16 Farley Unit 2 Reactor Vessel Intermediate Shell Plate 87212-1 (Longitudinal Orientation) 5-3 Irradiated Charpy V-Notch Impact Properties for 5-17 Farley Unit 2 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5-4 Irradiated Charpy V-Notch Impact Properties for 5-18 Farley Unit 2 Reactor Pressure Vessel Weld Metal 5-5 Charpy impact Specimen Fracture Surfaces for 5-19 Farley Unit 2 Pressure Vessel Intermediate Shell Plate B7212-1 (Transverse Orientation) 5-6 Charpy impact Specimen Fracture Surfaces for 5-20 Farley Unit 2 Pressure Vessel Intermediate Shell Plate B7212-1 (Longitudinal Orientation) 5-7 Charpy impact Specimen Fracture Surfaces for 5-21 Farley Unit 2 Weld Heat Affected Zone Metal 5-8 Charpy impact Specimen Fracture Surfaces for 5-22 Farley Unit 2 Weld Metal 5-9 Comparison of Actual versus Predicted 30 ft Ib 5-23 Transition Temperature Increases for the Farley Unit 2 Reactor Vessel Material based on the Prediction Methods of Regulatory Guide 1.99 Revision 1 10 Tensile Properties for Farley Unit 2 Reactor Vessel 5-24 Intermediate Shell Plate B7212-1 (Transverse Orientation) 5-11 Tensile Properties for Farley Unit 2 Reactor Vessel 5-25 Intermediate Shell Plate B7212-1 (Longitudinal Orientation) iv
LIST OF ILLUSTRATIONS (cont.)
Figure Title Page 5-12 Tensile Properties for Farley Unit 2 Reactor 5-26 Vessel Weld Metal Fractured Tensile Specimens of Farley Unit 2 5-27 5-13 Reactor Vessel Intermediate Shell Plate B7212-1 (Transverse Orientation) 5-14 Fractured Tensile Specimens of Farley Unit 2 5-28 Reactor Vessel Intermediate Shell Plate B7212-1 (Longitudinal Orientation) 5-15 Fracture Tensile Specimens of Farley Unit 2 5-29 Reactor Vessel Weld Metal 5-16 Typical Stress-Strain Curve for Tension Specimens 5-30 6-1 Farley Unit 2 Reactor Geometry 6-21 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-22 l
6-3 Calculated Azimuthal Distribution of Maximum 6-23 l
Fast-Neutron Flux (E > 1.0 Mev) within the Pressure Vessel Surveillance Capsule Geometry 6-4 Calculated Radial Distribution of Maximum 6-24 l
Fast-Neutron Flux (E > 1.0 Mev) within the Pressure Vessel 6-5 Relative Axial Variation of Fast-Neutron Flux 6-25 (E > 1.0 Mev) within the Pressure Vessel l
6-6 Calculated Radial Distribution of Fast-Neutron Flux 6-26 l
(E > 1.0 Mev) within the Reactor Vessel l
Surveillance Capsules l
6-7 Calculated Variation of Fast-Neutron Flux Monitor 6-27 l
Saturated Activity within Capsules U, X, and V 6-8 Calculated Variation of Fast-Neutron-Flux Monitor 6-28 Saturated Activity within Capsules W, Y, and Z A-1 Predicted Adjustment of Reference Temperature, A-6 t
"A", as a Function of Fluence and Copper Content A-2 Fast Neutron Fluence (E > 1.0 Mev) as a Function A-7 of Full Power Service Life A-3 Farley Unit 2 (APR) Reactor Coolant System A-8 Heatup Limitations Applicable for the First 4.3EFPY A-4 Farley Unit 2 (APR) Reactor Coolant System A-9 Cooldown Limitations Applicable for the First 4.3EFPY O e V
SECTION 1
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule U,the First capsule to be removed from the Farley Unit 2 reactor pressure vessel, led to the following conclusions:
m The capsule received an average fast neutron fluence (E > 1.0 Mev) of 5.61 x 10'8 n/cm.
Irradiation of the reactor vessel intermediate shell plate B7212-1 to 5.61 x 10'8 m
n/cm resulted in 30 and 50 ft Ib transition temperature increases of 133 and 141 F, respectively, for specimens oriented normal to the principal rolling direction of the plate and 103 and 135 F, respectively, for specimens oriented parallel to the plate principal rolling direction.
m Weld metal irradiated to 5.61 x 10'8 n/cm resulted in both 30 and 50 f t Ib tran-sition temperature increases of 10 F.
m Comparison of the 30 ft Ib transition temperature increases for the Joseph M.
Farley Unit 2 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99 Revision 1 shows that the plate materialdid not embrittle as much as predicted. The weld metal shift is minimal and thus can not validly be compared to a predicted shift.
O 9
1-1
SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule U, the first capsule to be removed from the reactor in the continuing surveillance program which moni-tors the effects of neutron irradiation on the Joseph M. Far:ey Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Joseph M. Farley Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corpora-tion. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by Davidson and Yanichko.I IThe surveillance program was planned to cover the 40-yeardesign life of the reactor pressure vessel and was based on ASTM E-185-73," Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels".r21 Westinghouse
^
Nuclear Energy Systems oersonnel were contracted for the preparation of procedures for removing tne capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance speci-mens was performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance Capsule U removed from the Joseph M. Farley Unit 2 reactor vessel and discusses the analysis of these data.
8 e
2-1
SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA 533 Grade B Class 1 (base material of the Joseph M. Farley Unit 2 reactor pressure vessel beltline)are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and tough-ness under certain conditions of irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure," Appendix G to Section 111 of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temper-ature (RT NDT)'
RT NDT s defined as the greater of either the drop weight nil-ductility transition i
temperature (NDTT per ASTM E-208) or the temperature 60 F less than the 50 f t Ib (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (K IR curve) which appearsin Appendix G of the ASME Code.
The K IR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given materialis indexed to the K IR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity facto.s.
3-1
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to
" ~
account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as
~'
the Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program,Ul in which a surveillance capsule is periodically removed from the operating nuclear reactor and the' encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft Ib temperature ( A RTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted NDT nitial+ A RTNDT ) is used 'to indeIx the dia~terial to the K IRcurve i
RT NDT (RT 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.
4 e
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3-2
SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Joseph M. Farley Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. 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 core.
Capsule U was removed f rom the reactor af ter 1.1 effective full power years of plant operation. This capsule contained Cnarpy V-notch, tensile and compact tension (CT) specimens from weld metal representative of the reactor vessel core region weld metal, and Charpy V-notch, tensile, compact tensile (CT) and bend bar specimens from the intermediate shell plate 87212-1. The capsule also contained Charpy V-notch specimens from weld heat affected zone (HAZ) metal. All neat affected zone specimens were obtained from the weld heat affected zo'ne of plate 87212-1. The chemistry and heat treatment of the program surveillance materials is presented in Table 4-1.
All test specimens were machined from the 1/4 thickness location of the plate. Test specimens represent matcrial taken at least one plate thickness from the quenched n
end of the plate. Some base metal Charpy V-notch and tensile specimens were oriented with the longitudinal axis of the specimens normal to (transverse orientation) and some parallel to (longitudinal orientation) the major working direction of the plate. The CT test specimens were machined such that the crack of the specimen would propagate normal to (longitudinal specimens) and parallel to (transverse specimens) the major working direction of the plate. All specimens were fatigue precracked per ASTM E399-72. The precracked bend bar was
}~
machined in the transverse orientation. Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens normal to
+
(transverse orieni
- tion) the weld direction. Tensile specimens were oriented with 4-1 m
,i
the longitudinal axis of the specimen normal to (transverse orientation) the weld direction.
Capsule U contained dosimeter wires of pure copper, iron, nickel, and aluminum -
0.15% cobalt (cadmium-shielded and unshielded). In addition cadmium shielded 27 2
dosimeters of Np and U a were contained in the capsule.
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. 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 arrangement of the various mechanical test specimens, dosimeters and
-thermal monitors contained in Cepsule U are shown in Figure 4-2.
O n
4 s
9 8
9 4-2
TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE FARLEY UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS Chemical Composition (WT-%)
Plate B7212-1 Weld Metal Combustion Engineering Combustion Engineering Analysis Analysis C
0.21
<0.086 S
0.016 0.014 N2 0.006lbl 0.007 Co 0.027 0.010 Cu 0.20 0.03 Si 0.24 0.34 Mo 0.49 0.23 Ni 0.60 0.90 Mn 1.30 0.95 Cr 0.15lbl
< 0.01 V
0.003lbl O.006 P
0.018 0.004 Sn 0.011 Ibl 0.002 AI 0.040 0.003 (al All elements not listed are less than 0.010 weight %.
[b] Westinghouse Analysis.
Heat Treatment Heat Treatment Material Time (hr)
Coolant p
intermediate 1550 /1650'.
4 Water quenched Shell Plate B7212-1 1225 F t 25 F 4
Air cooled 1150 F i 25 F 18 Furnace cooled to 600 F
~'
Weldment 1150 F t 25 F 13 Furnace cooled 4-3
REACTOR VESSEL (3.12) U (2.70) Z CORE BARREL
- NEUTRON PAD (2.70) Y (3.12) X L
o 270 90*
V (3.12)
W (2.70) 180' l
l I
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Figure 4-1.
Arrangement of Surveillance Capsules in Farley Unit 2 Reactor Vessel (Updated Lead Factors for the Capsules Shown in Parentheses.)
l l
4-4 l
1 f
SPECIMEN NUMBERING CODE:
CT - PLATE B7212-1,(TRANSVERSE ORIENTATION)
CL - PLATE B7212,1 (LONGITUDINAL ORIENTATION)
CH - HEAT-AFFECTED-ZONE METAL 4
e CouPACT COMPACT COMPACT COuPACT BEND B AR TENSILES TENSION TENSION CHARPY CHARPY CHARPV TEN $10N TENSION CHARPY CHA
-i CW3 CW15 CH15 CW12 CM12 CW9 CH9 CW6 CHS CW3 CT1 CW2 CW4 CW3 CW2 CW1 CW14 CHte CW11 CH11 CWS CH3 CL4 CL3 CL2 CL1 CWS CHS CW2 1
CW1 CW1]
CM13 CWIC CH10 CW7 CH7 CW4 CH4 CW1 a6 3
ll l C.
Axiss Co Cu l
18 I i 1 3 is i i !i 1
u 81 i LJ 6J be w g.,
l 579 F MMM
$90' F MI ll ll 8 Al 15 Co(Cd) ll MONITOR 3
MCNITOR g
3 13 ll 1 l li Il lig at l g
! h 11 1 d
C TO TOP OF VESSEL i
d
(
b i
i NEUTRON SHIELD PAD N
/
/
A' s
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l
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v p
am
/
CAPSULE U d
CORE a
/
CORE BARREL
/
s'
'-, VESSEL WALL
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CC UPA CT COMPACT 2PY DOSCETER TENSILE CHARPV CHARPY CHARPY CHARPV CHARPy TENSION TENS 6CN TEN 5tLES CH3 CL3 CT15 CL15 CT12 CL12 cts CLs CT4 CLE CT3 CL3 CT3
=
CH2
{
CL2 Cita CL14 CT11 CL 11 CT8 CLS CT5 CLS CT2 CL2 cts CT3 CT2 CT1 CT2 y
CM1 CL1 CT13 CL13 CT10 CLIO CT7 CL7 CT4 CL4 CT1 CL1 CT1 trT
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is ;-
As-is Co C.
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A,_ iS, Co i l i Ii 11 I il 1I el i i. Il l le a iW u u u:
IO 5F9 F ll -
Al.15 Co (Cd)
MONITOR j l llil, A'-15 c Co (Cd)
II i 3 Is l
11 s 3 8 13 1
...i d. i ir.
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u INTE3 KEctON OF vf1SEL TO BOTTOM OF VESSEL Figure 4-2.
Capsule U Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters 4-5
[
SECTION 5 TESTING OF SPECIMENS FROM CAPSULE U 5.1.
OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems per-sonnel. Testing was performed in accordance with 10CFR50, Appendices G and H, ASTM Specification E185-79 and Westinghouse Procedure MHL 7601, Revision 3 as modified by RMF Procedures 8102 and 8103.
Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8956.N No discrepancies were found.
Examination of the two low-melting 304 C (579 F) and 310 C (590 F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304 C (579 F).
The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 instru-mentation system. With this system, load-time and energy-time signals can be recorded in additinn to the standard measurement of Charpy energy (E o). From the load-time curve, the load of general yielding (PGY), the time to general yielding (t GY ), the maximum load (PM ), and the time to maximum load (t M ) 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 (P F), and the load at which 'ast fracture terminated is identified as the arrest load (PA)-
5-1 L
e 4,
The energy at maximum load (EM ) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is roughly equivalent to the energy required to initiate a crack in the specimen.Therefore,the propagation energy for the crack (E p ) is the difference between the total energy to fracture (ED) and the energy at maximum load.
The yield stress (ay) is calculated from the three point bend formula. The flow stress is calculated from the average of the yield and maximum loads, also using 4
the three point bend formula.
Percent shear was determined from postfracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
Tension tests were performed on a 20,000-pound Instron, split-console test machine-(Model 1115) per ASTM Specifications E8-81 and E21-79, and RMF
- Procedure 8102. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test.
Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67.
. Elevated test temperatures were obtained with a three-zone electric resistance
~
split-tube furnace with a 9-inch hot zone. All tests were conducted in air.
Because of the difficulty in remotely attaching a thermocouple directly to the speci-men, the following proceduce was used to monitor specimen temperature.
- Chromel-alumel thermocouples were inserted in shallow holes in the center and
]
each end of the gage section of a dummy specimen and in each grip. In test con-figuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550 F (288 C). The upper grip was used to controi the furnace temperature. During the actual testing the grip temperatures were used to 4
5-2
obtain desired specimen temperatures. Experiments indicated that this method is accurate to plus or minus 2 F.
The yield load, ultimate load, fracture load, total elongation, and uniform elonga-tion were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-
. fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5-2.
CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials contained in Capsule U Irradiated at 5.61 x 10'8 n/cm are presented in Tables 5-1 through 5-8 and Figures 5-1 through 5-4. A summary of the transition temperature increases and upper shelf energy decreases for the Capsule U material is shown in Table 5-9.
Irradiation of plate B7212-1 material (transverse orientation) to 5.61 x 10'8 n/cm (Figure 5-1) resulted in 30 and 50 ft Ib transition temperature increases of 133 and 141* F, respectively, and an upper shelf energy decrease of 26 ft Ib. Irradiation of plate B7212-1 material (longitudinal orientation) to 5.61 x 10'8 n/cm (Figure 5-2) resulted in 30 and 50 ft Ib transition temperature increases of 103 and 135 F, respectively, and an upper shelf energy decrease of 36 ft Ib.
28 Weld metal irradiated to 5.61 x 10 n/cm (Figure 5-4) resulted in 30 and 50 ft Ib transition temperature increases of 10 F and an upper shelf energy decrease of 12 ftIb.
Weld H AZ metal irradiated to 5.61 x 10'8n/cm (Figure 5-3) resulted in 30 and 50 f t b transition temperature increases of 58 and 50 F respectively and an upper shelf energy decrease of 47 ft Ib.
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasing ductile or tougher appearance with increasing test temperature.
5-3 L
Figure 5-9 shows a comparison of the 30 ft Ib transition temperature increases for the various Farley Unit 2 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99 Revision 1.l 1 This comparison shows that the transition temperature increases resulting fromirradiation to 5.61 x 10'8 n/cm2 are less than predicted by the Guide for plate B7212-1, transverse and longitudinal
. orientations. Since the 10 ft Ib shift the weld metal experienced is less than the 50 ft Ib shif t level, below which the Guide has not been proven valid, no valid comparison can be made. The weld metal shift is minimal due to the low copper and phosphorus levels of the weld wire.
5-3.
TENSION TEST RESULTS The results of tension tests performed on plate B7212-1 and weld metalirradiated i8 to 5.61 x 10 n/cm are shown in Table 5-10 and Figures 5-10, 5-11 and 5-12, respectively. These results show that irradiation produced an increase in 0.2 percent yield strength of 12 to 18 ksi for plate B7212-1 and approximately 5 ksi for the weld metal. Fractured tension specimens for each of the materials are shown in Figures 5-13,5-14 and 5-15. A typical stress-strain curve for the tension specimens is shown in Figure 5-16.
5-4.
COMPACT TENSION TEST RESULTS The 1/2T compact tension fracture mechanics specimens that were contained in Capsule U will be reported at a later time.
G 5-4
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE FARLEY UNIT 2 INTERMEDIATE SHELL PLATE B7212-1 (TRANSVERSE)
~~.
IRRADIATED AT 550 F, FLUENCE 5.61 x 10'8 n/cm" (E > 1 MeV)
Temperature impact Energy Lateral Expansion Sqle No.
( F)
(ft Ib)
.(mils)
(%)
CT4 50 15.0 10.5 16 CT8 75 22.5 15.5 23 CT10 100 21.5 20.0 27 CT1 110 32.5 26.5 31 CTS 125 33.5 30.0 37 CT12 150 37.0 30.0 37 CT14 175 46.0 40.0 46 CT9 175 42.5 39.5 48 CT11 200 49.5 50.0 57 CT6 225 69.0 69.5 100 CT13 250 67.5 62.5 100 CT2 275 68.0 67.0 100 CT7 300 67.0 63.5 100 CT3 350 69.0 60.0 100 CT15 400 73.0 68.5 100 l
5-5 i
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE FARLEY UNIT 2 INTERMEDIATE SHELL PLATE B7212-1 (LONGITUDINAL)
IRRADIATED AT 550 F, FLUENCE 5.61 x 10 n/cm* (E > 1 MeV)
Temperature Impact Energy Lateral Expansion No.
( F)
(ft Ib)
(mils)
(%)
CL8 0
13.5 8.0 2
CL3 50 23.5 21.0 17 CL4 75 33.0 27.0 25 CL5 100 40.5 33.5 31 CL6 125 42.0 30.5 38 CL14 150 46.0 37.5 44 CL15 150 54.0 40.0 49 CL12 175 53.0 40.0 56 CL13 200 55.0 63.0 58 CL9 200 73.5 59.0 73 CL11 225 88.5 81.0 96 CL2 250 90.0 71.5 100 CL7 300 91.5 76.0 100 CL1 350 98.5 80.0 100 CL10 400 97.5 80.0 100
- e e
5-6
TABLE 5-3 CHARPY V-NOTCH IMPACT DATA FOR THE FARLEY UNIT 2 PRESSURE VESSEL WELD HEAT AFFECTED ZONE METAL lRRADIATED AT 550 F, FLUENCE 5.61 x 10'8 n/cm* (E > 1 MeV)
Temperature Impact Energy Expansion S e ple No.
( F)
(ft Ib)
(mile)
(%)
CH1
-100 21.0 16.5 5
CH12
-50 29.5 28.5 35 CH2
-25 75.5 63.0 52 CH4 0
68.0 57.5 62 CH3 0
102.0 75.0 87 CH11 50 98.0 70.0 84 CH8 50 103.5 82.5 82 CH10 75 57.5 50.0 73 CH7 75 119.0 89.0 100 CH14 100 105.5 89.0 100 CH6 125 124.5 92.5 100
~
CH9 175 107.5 89.0 100 CH15 200 113.5 90.0 100 CH13 250 96.5 87.5 100 CH5 300 108.0 92.0 100 G
5-7
TABLE 5-4 CHARPY V-NOTCH IMPACT DATA FOR THE FARLEY UNIT 2 PRESSURE VESSEL WELD METAL lRRADIATED AT 550 F, FLUENCE 5.61 x 10 n/cm" (E > 1 MeV)
Temperature impact Energy Expansioni No.
(* F)
(ft Ib)
(mils)
(%)
CW3
-50 13.0 16.0 24 CW12
-25 79.0 68.0 53 CW4 0
95.0 77.0 64 CW7 0
36.5 31.5 42 CW9 25 89.5 74.0 67 CW10 50 85.0 70.0 73 CW8 50 96.0 78.0 91 CW14 75 112.5 84.5 99 CW11 100 108.0 80.5 98 CW1 125 140.0 91.5 100 CW5 150 132.5 92.0 100
~
CW2 175 122.0 91.0 100 CW6 200 114.0 90.5 100 CW13 250 154.5 90.0 100 CW15 300 127.5 96.0 100 l
l i
5-8
TABLE 5-5 l
lNSTRUMENTED CHARPY IMPACT TEST RESULTS FOR FARLEY UNIT 2 INTERMEDIATE SHELL PLATE B7212-1 (TRANSVERSE) j I
Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp. Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2
2 2
No.
( C) (Joules) (kJ/m )
(kJ/m )
(kJ/m )
(N)
( Sec)
(N)
( Sec)
(N)
(N)
(MPa) (MPa)
CT4 10 20.5 254 172 82 14300 145 16900 265 16600 0
734 802 y
CT8 24 30.5 381 281 101 14900 130 17900 360 17700 200 766 844 CT10 38 29.0 364 224 141 14600 100 17300 280 17200 1800 751 821 CT1 43 44.0 551 378 173 14400 140 18900 460 18300 2400 742 857 CTS 52 45.5 568 375 193 14100 135 18800 455 18300 3000 726 846 l
CT12 66 50.0 627 396 231 13600 90 17900 450 17900 3200 701 812 CT9 79 57.5 720 385 335 14200 120 18600 450 18000 6600 732 844 CT14 79 62.5 780 460 319 14100 125 18400 535 18300 5900 723 835 l
CT11 93 67.0 839 444 395 13400 115 18000 525 17300 8800 691 808 CT6 1 37' 93.5 1169 440 730 13100 115 18000 515 675 801 CT13 121 91.5 1144 351 793 12200 110 17300 440 628 759 CT2 135 92.0 1152 354 798 12500 120 17300 450 644 766 CT7 149 91.0 1136 414 721 12200 130 16800 535 629 747 CT3 177 93.5 1169 414 756 12400 100 16600 510 640 748 CT15 204 99.0 1237 399 838 11100 120 16100 530 568 698 r
t TABLE 5-6 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR FARLEY UNIT 2 INTERMEDIATE SHELL PLATE B7212-1 (LONGITUDINAL)
Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2
No.
( C)
(Joules) (kJ/m )
(kJ/m )
(kJ/m )
(N)
( Sec)
(N)
( Sec)
(N)
(N)
(MPa) (MPa)
CL8
-18 18.5 229 155 74 - 15900 110 18400 205 17800 0
816 881
?
CL3 10 32.0 398 324 74 15400 100 19200 360 19100 0
790 888 5
CL4 24 44.5 559 465 95 14300 115 19400 515 19400 0
736 868 CL5 38 55.0 686 477 210 15000 95 19400 505 19100 0
771 884 CL6 52 57.0 712 552 160 15000 130 19300 610 19200 1600 771 883 CL14 66 62.5 780 456 323 14100 95 18700 505 18300 4900 723 844 CL15 66 73.0 915 547 368 14300 130 19700 605 19000 3500 734 874 CL12 79 72.0 898 544 354 14400 125 19100 605 18900 7200 740 861 CL13 93 74.5 932 450 482 13500 120 18100 530 17800 8100 697 815 CL9 93 99.5 1246 531 715 13100 140 19100 615 16300 8800 672 828 CL11 107 120.0 1500 519 981 13500 105 18400 585 693 819 CL2 121 122.0 1525 521 1004 13100 120 18300 605 675 808 CL7 149 124.0 1551 490 1061 12700 120 17500 590 655 777 CL1 177 133.0 1661 494 1167 12300 125 17400 605 633 765 CL10 204 132.0 1652 472 1180 11700 120 16700 600 601 731 l
l
J 4
TABLE 5-7 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR FARLEY UNIT 2 HEAT AFFECTED ZONE METAL Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2
2 No.
(* C)
(Joules) (kJ/m )
(kJ/m )
(kJ/m )
(N)
(pSec)
(N)
( Sec)
(N)
(N)
(MPa) (MPa)
CH1
-73 28.5 356 253 103 17100 115 18900 290 19300 0
881 927 CH12
-46 40.0 500 381 119 16000 125 18700 440 18700 400 825 893
?
CH2
-32 102.5 1280 474 806 15900 115 19000 515 16000 3200 816 897 O
CH4
-18 92.0 1152 454 699 14400 115 17900 525 15600 6100 741 830 CH3
-18 138.5 1729 548 1180 14900 115 18700 600 12000 7100 768 865 CH11 10 133.0 1661 525 1136 14400 95 18300 575 13700 8700 741 841 CH8 10 140.5 1754 538 1216 15200 120 18300 600 15800 0
784 862 CH10 24 78.0 974 342 632 13'400 120 16500 440 688 768 CH7 24 161.5 2017 611 1406 13000 155 18700 725 670 817 CH14 38 143.0 1788 580 1208 13700 110 17200 675 705 794 CH6 52 169.0 2110 595 1515 13900 110 18100 675 717 825 CH9 79 146.0 1822 556 1266 13400 130 16800 685 691 778 f17100 CH15 93 154.0 1924 559 1365 12400 115 675 638 759 CH13 121 131.0 1635 501 1135 11200 90 15800 635 576 694 CH5
.49 146.5 1830 522 1308 11500 120 15800 680 590 702 l
l l
l
TABLE 5-8 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR FARLEY UNIT 2 WELD METAL Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2
No.
( C)
(Joules) (kJ/m )
(kJ/m )
(kJ/m )
(N)
( Sec)
(N)
( Sec)
(N)
(N)
(MPa) (MPa)
CW3
-46 17.5 220 66 154 16300 130 16200 135 16000 2000 838 837 CW12
-32 107.0 1339 584 755 14000 110 17000 680 15200 3600 721 797 cp CW7
-18 49.5 619 437 182 13800 110 17000 525 17000 5200 712 794 iG CW4
-18 129.0 1610 660 950 13400 110 16600 780 13900 6700 688 772 CW9
-4 121.5 1517 570 947 13000 100 16700 680 13700 6500 668 764 CW10 10 115.0 1441 553 887 13400 105 16400 675 13400 7500 689 765 CW8 10 130.0 1627 563 1064 11000 95 16200 705 14300 8500 566 699 CW14 24 152.5 1907 608 1299 12200 95 16000 750 9900 5900 629 727 CW11 38 146.5 1830 541 1289 12600 100 16100 675 646 737 CW1 52 190.0 2373 586 1787 11600 100 15800 750 597 704 CW5 66 179.5 2246 585 1661 11100 100 15600 755 568 685 CW2 79 165.5 2068 582 1485 10700 90 15400 765 551 672 CW6 93 154.5 1932 577 1355 11000 135 15000 795 564 668 CW13 121 209.5 2618 553 2065 10000 110 14700 785 512 635 CW15 149 173.0 2161 554 1607 9300 70 14200 775 477 604
{
i TABLE 5-9 l
THE EFFECT OF 550 F IRRALIATION AT 5.61 x 10 a (E > 1 MeV) i ON THE NOTCH TOUGHNESS PROPERTIES OF THE FARLEY UNIT 2 REACTOR VESSEL MATERIALS l
Average Average 35 mit Average Average Energy Absorbtion 30 ft Ib Temp (* F)
Lateral Expansion Temp (*F) 50 ft Ib Temp (*F) at Full Shear (ft Ib)
Material Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradicted irradiated AT Unirradiated irradiated (ft Ib)
Plate
-13 120 133 32 160 128 50 191 141 95 69 26 B7212-1 ui (transverse) a W
Plate
-23 80 103
-10 118 128 8
143 135 130 94 36 B7212-1 (longitudinal)
HAZ Metal
-131
-73 58
-91
-38 53
-83
-33 50 158 111 47 Weld Metal
-30
-20 10
-20
.-17 3
-1 9
10 144 132 12
TABLE 5-10
)
TENSILE PROPERTIES FOR FARLEY UNIT 2 l
REACTOR VESSEL MATERIAL IRRADIATED TO 5.61 x 10 e n/cm i
. Test Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp Strength Strength Load Stress Strength Elongation Elongation in Area i
No.
Material
(*F)
(ksi)
(ksi)
(kip)
(ksi)
(ksi)
(%)
(%)
(%)
CT2 Plate 135 86.6 105.3 3.77 213.3 76.8 11.6 23.5 64 B7212-1 l
(transverse) 1 CT3 Plate 225 79.9 101.9 3.80 182.1 77.4 10.6 21.0 57 87212-1 (transverse) l CT1 Plate 550 75.0 98.2 4.00 180.4 81.5 9.4 16.9 55 c.n B7212-1 k
(transverse)
CL1 Plate 135 86.2 104.9 3.50 195.4 71.3 11.3 24.2 64 87212-1 (longitudinal)
CL2 Plate 250 81.5 101.9 3.48 207.9 70.9 9.8 21.0 66 B7212-1 (longitudinal)
CL3 Plate 550 76.4 98.8 3.68 205.5 75.0 9.7 19.6 64 B7212-1 (longitudinal)
CW2 Weld 75 70.3 85.0 2.45 235.9 49.9 11.6 25.7 79 Metal CW3 Weld 300 65.2 77.8 2.40 219.5 48.9 9.6 22.5 78 Metal CW1 Weld 550 62.1 79.9 2.45 209.6 49.9 10.2 23.6 76 Metal t
]
/
100
_ 80 1
2
$ 60 W
I e
O e'
20 2
O a
3 l
100
~
E O
E a -
7 60 e
5 g 40 128*
g 20 3
I I
I I
I I
O e
120 UNIRRADIATED o
100 9
h 80 a
= 60
/...\\
z e-141 IRRADIATED (550' F)
W 5.61 x 10'" n/cm2 40 133*
20 0
I i
i i
I i
-100 0
100 200 300 400 TEMPERATURE ( F)
Figure 5-1.
Irradiated Charpy V-Notch Impact Properties for Farley Unit 2 Reactor Vessel Intermediate Shell Plate B7212-1 (transverse orientation) 5 I
3
/ -
- : e 100 80 e
7 e; %
O
~
e 40 e',
I I
I I
I O
100 o
2 80 ee O
60 n
o/
- j l
0 UNIRRADIATED 0
120 7
o 100 e
80 d
IRRADI ATED (550* F)
O 5.61 x 10'8 n/cm
{ 60 2
o g
e#
.135' w
o e
6 40 fo /
,o i
i i
i i
g_
-100 0
100 200 300 400 TEMPER ATURE (* F)
Figure 5-2.
Irradiated Charpy V-Notch Impact Properties for Farley Unit 2 Reactor Vessel Intermediate Shell Plate B7212-1 (longitudinal orientation) 5-16
s o
^?
? O rd:
100 o
e 80
?
$ 60 y
40 20 0
l 0
O
^O O
e
^
^
m
- 80 O
j o.
o
{%
O
~
o e
.j 40 53' 20 0
O o
180 UNIRRADIATED 160 O
o 140 d120 O
b e
w
$100 o
o e
0 z 80 IRRADIATED (550 F) 5.61 x 10 n/cm 60 50
/
58 20 0
-200
-100 0
100 200 300 TEMPERATURE (" F)
Figure 5-3.
Irradiated Charpy V-Notch Impact Properties for Farley Unit 2 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5-17
3 100 t^
- qr %
2 80 o
e
^
e "O 60
~'
5 40
/*
20 O
100 g.
2
- 80 j
5
{ 60 U
.j 40
~3' N
20 e
0
~
0 0
UNIRRADIATED o
140 e
g 120 e
O e
IRRADI ATED (550* F) 100 5.61 x 10'" n/cm' 4
e
- s 80 e
$ 60 10'
- 40 0
10 20 0
0
-100 0
100 200 300 400 TEMPERATURE ( F)
Figure 5-4.
Irradiated Charpy V-Notch Impact Properties for Farley Unit 2
~
- Reactor Pressure Vessel Weld Metal I
5-18 1
0"'
.i
~
ggg CT4 CT8 CT10 CT1 CTS CT12 CT14 CT9 CT11 CT6
[
' s f ".
f;. 4;lfK spyg;ng;;;l li-4t. ;};p.((
. gi -
. ; =. 7..
p 7l cwa -
.:e, -
^ ' " '
(,.'.. ll-[
ll?
];hI gl..
- .l
-)
.p.'h
^1 frl;ll;.4hby.-~
- p. ;p. ;;y
.l y.
8 -
.:. g.
hekk.5Q f{.l*h f0gl hf... ? _ ~.{
CT13 CT2 CT7 CT3 CT15 Figure 5-5.
Charpy impact Specimen Fracture Surfaces for Farley Unit 2 Pressure Vessel Intermediate Sheli Plate B7212-1 (Transverse Orientation) 5-19
4t a
~
f
~
.&%f ?.'.!r j *h$.sq
'p%
'm
~t 1* e = >;--
- n N
i CL8 CL3 CL4 CL5 CL6
- s e
X
?
i
- =
3 ti m
CL14 CL15 CL12 CL13 CL9 i
i t
I?f#$p.
SQf.,.,,
i y
o w,
..;e.
.l ;
- j. 3.g 1
^
,a.
n.
?
y:y%
gi - :i!.: /,;. ?;...
.r.- y..;.
m:.c.
.3 - '.. -
1 cc :: e 4 6
..m e.
CL11 CL2 CL7 CL1 CL10 g
Figure 5-6.
Charpy impact Specimen Fracture Surfaces for Farley Unit 2 Pressure Vessel Intermediate Shell Plate B7212-1 (Longitudinal Orientation)
{
1 5-20 1 1
w fg$d-Y,,g'-.'f} -
ca
.ve....
i
...v,.. _,
4 Y ~%f:.
1 V
- l CH1 CH12 CH2 CH4 CH3 CH11 CH8 CH10 CH7 CH14 l
CH6 CH9 CH15 CH13 CHS Figure 5-7.
Charpy impact Specimen Fracture Surfaces for Farley Unit 2 Weld Heat Affected Zone Metal 5-21
1,$: }
....-,.. ?
.. 7_.
! ::... n.g:
.,;.j
.,,_:~f,
g,.
-- f. A
.y
, y pv y
.* 's f
yx. ;w....,.1" p : 8.,. ; :[.;
CW3 CW12 CW4 CW7 CW9 CW10 CW8 CW14 CW11 CW1 5.
.m.-
yggt 3
T;-
CWS CW2 CW6 CW13 CW15 Figure 5-8.
Charpy impact Specimen Fracture Surfaces for Farley Unit 2 Weld Metal 4
5-22
.;. g r!. ~.e....: v. m t,.s e.
- .s v.n.:.
.>.,;.. > c,c.; :n 1.w
, r ; :. a _. x.m.:y_....; : < gen
a 4
- 7 6%
500 s
400 C 300 Oq?,
g&st L
w E 200 w
Eo E
O gib 100 W
/g
\\
i z
70 9
t-g 50 40 m
n h
y 30 g-20 a
l O PLATE B7212-1 (transverse)
~
PLATE B7212-1 (longitudinal) 10 d
A WELD METAL 9
8 l
7 6
10
2 4
6 8 10
2 4 i 6 FLUENCE (n/cm')
Figure 5-9.
Comparison of Actual versus Predicted 30 ft Ib Trancition Temperature increases for the Farley Unit 2 Reactor Vessel l Material Based on the Prediction Methods of Regulatory Guide 1.99 Revision 1 l
l
=
120 i
100 l"
ULTIMATE 7
STRENGTH
-g E
-O g 80 E
0.2% YlELD 7
G 60 I
I I
I I
I 40 LEGEND 0 unirradiated e Irradiated at 5.61 x 10 n/cm' 80 60 0-o RE CTION
.Y N; 40 503 C
~ g TOTAL 20
- ELONGATION O
O-0 UNIFORM r
' ELONG ATION i
1 0
0 100 200 300 400 500 600 TEMPERATURE ( F) l l-Figure 5-10.
Tensile Properties for Farley Unit 2 Reactor Vessel Intermediate Shell Plate B7212-1 (transverse orientation) 5-24
120 100 t
ULTIMATE TENSILE STRENGTH g
f v
$M 0.2% YlELD
("
,0 40 LEGEND O unirradiated e irradiated at 5.61 x 10" n/cm' l
80 l
o REDUCTION IN
_ _ AREA 60
^
c N
5 40 ED O
2F
^
O
- TOTAL 20
'2-7 ELONGATION O'
y 6 --
y UNIFORM
- ELdNGATION O
O 100 200 300 400 500 600 TEMPERATURE ( F) 4 Figure 5-11.
Tensile Properties for Farley Unit 2 Reactor Vessel Intermediate Shell Plate B7212-1 (longitudinal orientation) 5-25
120 100 T
.s ULTIMATE TENSILE 5
TRENGm 6
g' _
'2 0.2% YlELD STRENp2TH ~
Ne e
i i
i a
a g
LEGEND l
0 unirradiated e Irradiated at 5.61 x 10 n/cm' l
80 7
l l
O REDUCTION IN o
AREA I
60 O
-6 0
C 2 40 o
o TOTAL ELONGATION W
O B
20 g
S UNIFORM ELONGATION O
a a
a a
a 0
100 200 300 400 500 600 TEMPER ATURE (* F) a Figure 5-12.
Tensile Properties for Farley Unit 2 Reactor Veesel Weld Metal 5-26
PER TENSILE SPECIMEN CT2 we,%%...d-sik Tested at 13S'F 4e 1234 6789 0
10THS 1
INCHES TENSILE SPECIMEN CT3 1
I Tested at 225 F tJ -
+ w w. e
% %,,s/g %
1234 6789 0
10THS 1
INCHES
~
s N,. M~ ~~
TENSILE SPECIMEN CT1 Tested at 550* F 1234 6789 0
10THS 1
INCHES Figure 5-13.
Fractured Tenslie Specimens of
~
Farley Unit 2 Reactor Vessel Intermediate Shell Plate B7212-1 (Transverse Orientation) 5-27
s
. m..
TENSILE SPECIMEN CL1 Tested at 135'F 6" N;i,-
e 1 234 6789 0
10THS 1
INCHES i
TENSILE SPECIMEN CL2 Tested at 250 F i
p. L.,
..,m v-
,q.,4 I
1 234 6789 O
10THS 1
INCHES l
l Tested at 550 1234 6789 0
10THS 1
INCHES t
~ '
Figure 5-14.
Fractured Tensile Specimens of Farley Unit 2 Reactor Vessel Intermediate Shell Plate B7212-1 (Longitudinal Orientation) j 5-28
TENSILE SPECIMEN CW2
.f%
Tested at 75'F
- A.,3 p h.%.
3,..,
1234 6789 0
10THS 1
INCHES l'
h TENSILE SPECIMEN CW3 r,- _ m 7 --.- _.
,)I, Tested at 300*F 1 234 6789 I
0 10THS 1
INCHES e!ted t 0*
i 1234 6789 0
10THS 1
INCHES j
l l
l l
~
Figure 5-15.
Fracture Tenslie Specimens of Farley Unit 2 Reactor Vessel Weld Metal i
l' 5-29
i l
120 1
1 i
108 96 84 7 72 I
.x i
in l
E 60 "
l vs 48 '
36 24 SPECIMEN CL1 12 f
a e
0 0
.03
.06
.09
.12
.15
.18
.21
.24
.27
.30 STRAIN (in/in) l t
t Figure 5-16.
Typical Stress-Strain Curve for Tension Specimens 5-30
SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1.
INTRODUCTION Knowledge of the neutron environment within the pressure vessel surveillance capsu% geometry is required as an integrcl part of LWR pressure vessel surveillance programs for two reasons. First,in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second,in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship between the environment at various positions within the reactor vessel and that experienced by the test specimens must be established. The former 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, on the other hand, is derived solely from analysis.
This section describes a discrete ordinates Sn transport analysis performed for the Farley Unit 2 reactor to determine the fast-neutron (E > 1.0 Mev) flux and fluence as
.all as the neutron energy spectra within the reactor vessel and surveillance aapsules, and, in turn, to develop lead factors for use in relating the neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averaged reaction cross sections derived from this calculation, the analysis of the neutron dosimetry contained in Capsule U is discussed and comparisons with analytical predictions are presented.
~
6-2.
DISCRETE ORDINATES ANALYSIS A plan view of the Farley Unit 2 reactor geometry at the core midplane is shown in Figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a 0 to 45 sector
~ ^
is depicted. Six irradiation capsules attached to the neutron pad are included in the design for use in the reactor vessel surveillance progi im. Three capsules 6-1
(
are located symmetrically at 16.94 and 19.74 from the cardinal axes, as shown in Figure 6-1.
A plan view of surveillance capsules attached to a neutron pad is shown in Figure 6-2. The stainless steel specimen container has a 1.25 inch by 1.07 inch cross section and is approximately 56 inches high. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 4-2/3 feet of the 12-foot-high reactor core.
From a nedtronid standpoint, the surveillance capsule structures are significant. In fact, as will be shown later, they have a marked impact on the distributions of neutron flux and energy spectra in the water annulus between'the neutron pad and the reactor vessel. Thus, in order to properly ascertain the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model. Use of at least a two-dimensional computation is, therefore, mandatory.
In the analysis of the neutron environment within the Farley Unit 2 reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the DOTl4 two-dimensional discrete ordinates code. The radial and aximuthal distributions were obtained from an R, O computation wherein the geometry shown in Figures 6-1 and G-2 was described in the analytical model. In addition to the R, O computation, a second calculation in R, Z geometry was also made to obtain relative axial variations of neutron flux throughout the geometry of interest.
In the R, Z analysis the reactor core was treated as an equivalent volume cylinder and, of course, the surveillance capsules were not included in the model.
Both the R, O and R, Z analyses employed 47 neutron energy groups and a P3 expansion of the scattering cross-sections. The cross-sections used in the
~
~~
analyses were obtained from the SAILOR cross-section libraryl51 which was developed specifically for light water reactor applications. The neutron energy group structure used in the analysis is listed in Table 6-1.
i A key input parameter in the analysis of the integrated fast neutron exposure of the reactor vessel is the core power distribution. For this analysis, power distributions l
representative of time averaged conditions derived from statistical studies of long
'~
term operation of Westinghouse 3-loop plants were employed. These input dis-tributions include rod-by-rod spatial variations for all peripheral fuel assemblies.
6-2
l It should be noted that this generic design basis power distribution is intended to
(
provide a vehicle for long term (end-of-life) projection of vessel exposure. Since plant specific power distributions reflect only past operation, their use for projection into the future may not be justified; and the use of generic data which reflects long term operation of similar reactor cores may provide a more suitable l
approach.
5eric~hmaik iesting of these generic power distributions and the S Il OR cross-
~
~
~
~
l sections against surveillance capsule data obtained from 2-loop and 4-loop l
Westinghouse plants indicates that this analytical approach yields conservative results with calculations exceeding measurements by from 10-25%.1 One further point of interest regarding these analyses is that the design basis i
assumes an out-in fuel loading pattern (fresh fuel on the periphery). Future commitment to low leakage loading patterns could significantly reduce the calculated neutron flux levels presented in Section 6-4. In addition capsule lead l
factors could be changed, thus, impacting the withdrawal schedule of the remaining surveillance capsules.
Having the results of the R, O and R, Zealculations.three-dimensional variations of l
neutron flux may be approximated by assuming that the following relation holds for the applicable regions of the reactor.
l
$(R,Z,0,Eg)
$(R,0,Eg) F(Z,Eg)
=
where:
$(R,Z,0,Eg)
= neutron flux at point R,Z,0 within energy group g (R,0,Eg)
= neutron flux at point R,9 within energy group g obtained from the R,0 calculation F(Z,Eg)
= relative axial distribution of neutron flux within energy group g obtained from the R,Z calculation 6-3
6-3.
NEUTRON DOSIMETRY The passive neutron flux monitors included in Capsule U of Farley Unit 2 are listed in Table 6-2. The first five reactions in Table 6-2 are used as fast-neutron monitors to relate neutron fluence (E > 1.0 Mev) to measured material property changes.
Bare and cadmium-covered cobalt-aluminum monitors were also included in the table to determine the magnitude of the thermal neutron flux at the monitor location, which is necessary to account for burnout of the product isotope generated by fast-neutron reactions.
The relative locations of the various monitors within the surveillance capsules are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in Table 6-2 does not yield a direct measure of the energy-dependent flux level at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average 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:
a The operating historv of the reactor a
The energy response of the monitor a
The neutron energy spectrum at the monitor location a
The physical characteristics of the monitor The analysis of the passive monitors and subsequent derivation of the average neutron flux requires completion of two procedures. First, the desintegration rate of product isotope per unit mass of monitor must be determined.Second,in order to define a suitable spectrum averaged reaction cross section, the neutron energy s-spectrum at the monitor location must be calculated.
6-4
I k
The specific activity of each of the monitors is determined using established ASTM procedures.tr.8mni Following sample preparation, the activity of each monitor is l
determined by means of a lithium drifted germanium (Ge(Li)) gamma I
spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting,and the acceptable error in detector calibration. For the samples removed from Farley Unit 2, the overall 2 o deviation in the measured data is 10 percent. The neutron energy spectra are determined analytically using the method described in Subsection 6-1.
I Having the measured activity of the monitors and the neutron energy spectra at the locations of interest, the calculations of the neutron flux proceeds as follows:
The reaction product activity in the monitor is expressed as f
N t
a(E)$(E){P
{1~
j)
- d (6-1) fY R=
i
- E max pi where:
R = induced product activity t
No = Avogadro's number i
A = atomic weight of the target isotope f = weight fraction of the target isotope in the target material i
Y = number of product atoms produced per reaction a(E) = energy-dependent reaction cross section
$(E) = energy-dependent neutron flux at the monitor location with the reactor at full power Pj = average core power level during irradiation period j 6-5 L
Pmax = - maximum or reference core power level A = decay constant of the product isotope tj = length of irradiation period j td = decay time following irradiation period j Since neutron flux distributions are calculated using rnu'!tigroup transport methods and, further, since the prime interest is in the fast-neutron flux above 1.0 Mev, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-1) is replaced by the following relation:
a(E) $(E)dE F$(E > 1.0 Mev)
=
-E where:
s N
- 9#9 a(E) & (E)dE
- o G-1 0
- em N
J t 0 Me, g
G G g p yg, Thus, equation 6-1 is rewritten N
R.
I i y 5 & (E 1.0 Mev)
(1-e '^ i) e-t d (6-2) p i 1 8
4 6-6
or, solving for the neutron flux, R
$ (E > 1.0 Mev) =
N N
g Pj
-Atj)
-At d fIY I
(6-2)
A-P max p1 The total fluence above 1.0 Mev is then given by l
N p,
(E > 1.0 Mev){max I
4 (E > 1.0 Mev) =
tj (6-3) l pi where:
N max j = total effective full power seconds of reactor operation t
P up to the time of capsule removal l
l An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered Co '(n,y) Co* data by means of 5
cadmium ratios and the use of 37-barn,2200 m/sec cross section. Thus, hD-1 J
@Th =
bare h
D R
(6-4)
N No
-Pj
-A t ) e-A t j
d f yo (1-0 i
A P max j=1 where:
are D is definea as R Cd covered 6-7 i,..... _...
f-6-4.
TRANSPORT ANALYSIS RESULTS Results of the Sn transport calculations for the Farley Unit 2 reactor are sum-marized in Figures 6-3 through 6-8 and in Tables 6-3 through 6-5. In Figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius,1/4 thickness location, and 3/4 thickness location are presented as a function of azimuthal angle. The influence of the surveillance capsules on the fast-neutron flux distribution is evident. In Figure 6-4, the radial distribution of maximum fast-neutron flux (E > 1.0 Mev) through the thickness of the reactor pressure vessel is shown. The relative axial variation of neutron flux within the vessel is given in Figure 6-5. Absolute axial variations of fast-neutron flux may be obtained by multiplying the levels given in Figtires 6-3 or 6-4 by the appropriate values from Figure 6-5.
In Figure 6-6, the radial variations of fast-neutron flux within surveillance capsules at 16.94 and 19.72* are presented. This data, in conjunction with the maximum vessel flux, is used to develop lead factors for each of the capsules. Here, the lead factor is defined as the ratio of the fast-neutron flux (E > 1.0 Mev) at the dosimeter block location (capsule center) to the maximum fast-neutron flux at the pressure vessel inner radius. Updated lead factors for all of the Farley Unit 2 surveillance capsules are listed in Table 6-3.
Radial v triations of analytically determined reaction rate gradients foreach of the fast-neu tron monitors are shown in Figures 6-7 and 6-8 for capsules at 16.94 and 19.72, espectively.
In order to derive neutron flux and fluence levels from the measured disintegration rates, suitau'e spectrum-averaged reaction cross sections are required.
Calculations of the neutron energy spectrum existing at the center of each of the Farley Unit 2 surveillance capsules are listed in Table 6-4. The associated
. spectrum-averaged cross sections for each of the fast-neutron reactions are given in Table 6-5.
6-5.
DOSIMETRY RESULTS The irradiation history of the Farley Unit 2 reactor up to the time of removal of Capsule.U is listed in Table 6-6. Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsule U, based on the irradiation history shown in Table 6-6, are given in Table 6-7.
l 6-8
r l
The fast-neutron (E > 1.0 Mev) flux and fluence levels derived for Capsule U are presented in Table 6-8. The thermal-neutron flux obtained from the cobalt-aluminum monitors is summarized in Table 6-9. Due to the relativelylow thermal-neutron flux at the capsule location, no burnup correction was made to any of the
. measured activities. The maximum error introduced by this assumption is esti-
- mated to be less than 1 percent for the Ni"(np) Co" reaction and even less for all of the other fast-neutron reactions.
An eF3mination of Table 6-8 shows that the fast-neutron flux (E > 1.0 Mev) derived from the five threshold reactions ranges from 1.56 x 10".to 1.92 x 10" n/cm'-sec,a total span of less than 24%. It may also be noted that the calculated flux value of 1.98 x 10" n/cm'-sec exceeds all of the measured values with calculation to
)
experimental ratios ranging from 1.03 to 1.27. Tnis behavior is consistent with prior benchmarking studies.
Comparisons of measured and calculated current fast neutron exposures for Capsule U as well as for the inner radius of the pressure vessel are presented in_
Table 6-10. Measured values are given based on the Fe"(np) Mn" reaction alone.
Based on the data given in Table 6-8, the best estimate exposure of Capsule U is:
9T = 5.61 x 10 e n/cm' (E > 1.0 Mev)
Since the calculated fluence levels were based on conservative representations of core power distributions derived for long term operation while the Capsule U data l
are representative only of cycle 1 operation,it is recommended that projections of vessel toughness into the future be based on design basis calculated fluence levels.
l Withdrawals of future surveillance capsules should further substantiate the l
adequacy of this approach.
6-9 p
~
l TABLE 6-1 47 GROUP ENERGY STRUCTURE Lower Energy Lower Energy Group (Mev)
Group (Mev) 1 14.19*
25 0.183 2
12.21 26 0.111 3
10.00 27 0.0674 4
8.61 28 0.0409 5
7.41 29 0.0318 6
6.07 30 0.0261 7
4.97 31 0.0242 8
3.68 32 0.0219 9
3.01 33 0.0150 10 2.73 34 7.10 x 10-3 11 2.47 35 3.36 x 10-12 2.37 36 1.59 x 10-13 2.35 37 4.54 x 10-*
14 2.23 38 2.14 x 10-4 15 1.92 39 1.01 x 10-*
16 1.65 40 3.73 x 10-5 17 1.35 41 1.07 x 10-5 18 1.00 42 5.04 x 10-8 19 0.821 43 1.86 x 10-8 20 0.743 44 8.76 x 10-7 21 0.608 45 4.14 x 10-7 22 0.498 46 1.00 x 10-7 23 0.369 47 0.00 x 24 0.298
- The upper energy of group 1 is 17.33 Mev e
6-10
c 1
TABLE 6-2 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS i
l W* %
i l
of Target Product Fission i
m Monitor Material Reaction of interest g monitor Response Range Half-Ufe Yield (%)
i Copper Cu' (n.a) Co" 0.6917 E > 4.7 Mev 5.27 years l
Iron Fe ' (n.p) Mn '
O.0585 E > 1.0 Mev 314 days 5
5 Nickel Ni58 (n.p) Co 0.6777 E > 1.0 Mev 71.4 days l
sa Uranium *I'l U a (n,f) Cs' 7 1.0 E > 0.4 Mev 30.2 years 6.3 2
237tal Neptunium Np2 7 (n,f) Cs' 7 1.0 E > 0.08 Mev 30.2 years 6.5 I
Cobalt-aluminum *I Co*' (n,y) Co" 0.0015 0.4 ev < E < 0.015 Mev 5.27 years Cobalt-aluminum Co ' (n,y) Co" 0.0015 E < 0.015 Mev 5.27 years 5
[a] Denotes that monitor is cadmium shielded.
O
TABLE 6-3 CALCULATED FAST-NEUTRON FLUX (E > 1.0 Mev)
AND LEAD FACTORS FOR FARLEY UNIT 2 SURVEILLANCE CAPSULES Capsule Azimuthai
& (E > 1.0 Mev)
Lead identification Angle (n/cm*-sec)
Factor U
16.94 1.98 x 10" 3.12 X
16.94 1.98 x 10" 3.12 V
16.94 1.98 x 10" 3.12 W
19.72 1.73 x 10" 2.70 Y
19.72 1.73 x 10" 2.70 Z
19.72 1.73 x 10" 2.70 A
V 4
S 6-12
TABLE 6-4 a
CALCULATED NEUTRON ENERGY SPECTRA AT ThE CENTER OF FARLEY UNIT 2 SURVEILLANCE CAPSUi.ES Neutron Flux (n/cm*-sec)
Group No.
Capsules U, X, V Capsules W, Y, Z 7
7 1
2.98 x 10 2.82 x 10 8
8 2
1.05 x 10 1.03 x 10 8
8 3
3.68 x 10 3.60 x 10 4
6.79 x 10' 6.60 x 10' 5
1.14 x 10' 1.10 x 10' 6
2.56 x 10' 2.47 x 10' 7
3.57 x 10' 3.42 x 10' 8
7.45 x 10' 7.01 x 10' 9
6.99 x 10' 6.45 x 108 10 5.88 x 10' 5.40 x 10' i
11 7.12 x 10' 6.51 x 10' 12 3.57 x 10' 3.25 x 10' 13 1.09 x 10' 9.89 x 10' 14 5.54 x 10'
-5.03 x 10' 15 1.53 x 10' 1.07 x 10' 16 2.12 x 10' 1.89 x 10' 17 3.38 x 10' 2.98 x 10' 18 8.15 x 10' 7.07 x 10' 19 6.54 x 10' 5.42 x 10' 20 3.03 x 10' 2.52 x 10' 21 1.20 x 10" 9.87 x 10' 22 9.47 x 10' 7.71 x 10' 23 1.26 x 10" 1.03 x 10" 24 1.28 x 10" 1.04 x 10" 25 1.29 x 10" 1.04 x 10" 26 1.41 x 10" 1.14 x 10" 27 1.17 x 10" 9.40 x 10' 28 7.36 x 10' 5.88 x 10' 29 1.96 x 10' 1.56 x 10' 30 1.05 x 10' 8.31 x 10'
~
6-13
TABLE 6-4 CONTINUED CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF FARLEY UNIT 2 SURVElLLANCE CAPSULES Neutron Flux (n/cm -sec)
Group No.
Capsules U, X, V Capsules W, Y, Z 31 3.18 x 10' 2.56 x 10'*
32 2.28 x 10' 1.84 x 10' 33 3.02 x 10' 2.43 x 10'*
34 3.72 x 10' 2.98 x 10' 35 6.72 x 10' 5.38 x 10' 36 6.74 x 10' 5.36 x 10'*
37 9.02 x 10' 7.15 x 10' 38 4.56 x 10' 3.61 x 10' 39 5.01 x 10' 3.98 x 10' 40 6.88 x 10' 5.45 x 10' 41 7.97 x 10' 6.29 x 10' 42 4.21 x 10' 3.34 x 10' 43 4.36 x 10' 3.47 x 10' 44 2.43 x 10' 1.94 x 10' 45 1.66 x 10' 1.32 x 10' 46 1.84 x 10' 1.46 x 10' 47 1.90 x 10' 1.89 x 10' 9
S-14
TABLE 6-5 A~
SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF FARLEY UNIT 2 SURVEILLANCE CAPSULES o (bams)
Reaction Capsules U, X, V Capsules W, Y, Z Fe" (np) Mn '
O.0517 0.0548 5
Nise (np) Co 0.0741 0.0767 se Cu" (n,a) Co" 0.000429 0.000469 Npas7 (nf) FP 3.42 3.33 U238 (nf) FP 0.301 0.306 x
~
a(E) (E)dE g:
O
&(E)dE 1 O Mev e
6-15
TABLE 6-6 1RRADIATION HISTORY OF FARLEY UNIT 2 REACTOR'~
~
~
VESSEL SURVEILLANCE CAPSULE U Irradiation Decay PJ P max Pj/
Time Time Month - Year (MW)
(MW)
Pmax (day)
(day) 5 1981 52.
2652
.020
'24 785 6
1981 366.
2652
.138 30 755 7
1981 158.
2652
.060 31 724 8
1981 2436.
2652
.918 31 693 9
1981 2537.
2652
.957 30 663 10 1981 2624.
2652
.989 31 632 11 1981 2528.
2652
.953 30 602 12 1981 2595.
2652
.978 31 571 1
1982 2289.
2652
.863 31 540 2
1982 8.
2652
.003 28 512 3
1982 2023.
2652
.763 31 481 4
1982 2568.
2652
.968 30 451 5
1982
- 2543, 2652
.959 31 420 6
1982 2503.
2652
.944 30 390 7
1982 2539.
2652
.957 31 359 8
1982 2590.
2652
.977 31 328 9
1982 2477.
2652
.934 30 298 10 1982 2574.
2652
.971 22 276 7
EFPS = 3.46 x 10 Sec EFPY = 1.1
[a] Decay time is referenced to 7-25-1983.
e 6-16
TABLE 6-7 COMPARISON OF MEASURED AND CALCULATED FAST-NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE U
^
'I Reaction Radial Measured (dps/gm) and Location Activity Axlai Location (cm)
(dps/gm)
Capsule Calculated Fe * (n p) Mn'd 5
Top 186.21 1.68 x 10' 5.50 x 108 Middle 186.21 1.62 x 10' 5.31 x 108 Bottom 186.21 1.71 x 10' 5.60 x 108 8
Average 5.47 x 10 6.60 x 10e NiS8 (n p) Coss 8
7 Top 186.21 5.17 x 10 8.24 x 10 8
7 Middle 186.21 4.84 x 10 7.71 x 10 Bottom 186.21 5.29 x 10" 8.43 x 107 7
8 Average 8.13 x 10 1.01 x 10 Cu" (N,a) Co*
Top 186.21 6.42 x 10' 5.33 x 105 Middle 186.21 6.47 x 10' 5.37 x 105 Bottom 186.21 6.79 x 10*
5.64 x 10 5 5
Average 5.45 x 10 5.64 x 10' Np" ' (n,f) Cs' 7 8
7 8
Middle 186.21 2.15 x 10 8.83 x 10 1.10 x 10 U238 (n/) Cs' '
5 8
Middle 186.21 2.29 x 10 9.40 x 10' 9.12 x 10 e
6-17
i TABLE 6-8 RESULTS OF FAST-NEUTRON DOSIMETRY FOR CAPSULE U Saturated Activity
$ (E > 1.0 Mev)
& (E > 1.0 Mev) 2 (dps/gm)
(n/cm -sec)
(n/cm )
lT Reaction Measured Calculated Measured Calculahd Measured Calculated as Fe ' (n.p) Mn '
5.47 x 10 6.60 x 10 1.62 x 10" 1.98 x 10" 5.61 x 10'8 6.85 x 10's 5
5 8
8 Nisa (n.p) Co 8.13 x 10 1.01 x 10 1.56 x 10" 1.98 x 10" 5.39 x 10's 6.85 x 10'8 S8 7
8 Cu (n,a) Co 5.45 x 10 5.64 x 10 1.92 x 10" 1.98 x 10" 6.64 x 10'8 6.85 x 10'8 S
8 5
8 Np237 (n,f) Cs' 7 8.83 x 10 1.08 x 10 1.56 x 10" 1.98 x 10" 5.40 x 10 a S.85 x 10's 7
8 i
U238 (n,f) Cs' 7 8.27 x 10' 9.12 x 10 1.72 x 10" 1.98 x 10" 5.96 x 10'8 6.85 x 10'8 8
- U" adjusted saturated activity has been multiplied by 0.88 to correct for 350 ppm U'
- impurity.
I I
L TABLE 6-9 i
RESULTS OF THERMAL-NEUTRON DOSIMETRY FOR CAPSULE U
" ~
Saturated Activity (dps/gm)
A.x!al
$th Location Bare Cd - Covered (n/cm'-sec) 8 7
Top 1.13 x 10 6.45 x 10 8.50 x 10' Middle 1.18 x 10s 6.72 x 10 8.90 x 10' I
7 8
7 Bottom 1.16 x 10 6.50 x 10 9.00'x 10' e
e 6-19
i TABLE 6-10
SUMMARY
OF NEUTRON DOSIMETRY RESULTS FOR CAPSULE U Current & (E > 1.0 Mev)
EOL $ (E > 1.0 Mev) 7 (n/cm )
(n/ctn)
~
Location Measured Calculated Measured Calculated Capsule U 5.61 x 10'8 6.85 x 10'8 VesselIR 1.81 x 10'8 2.21 x 10'8 5.30 x 10'8 6.47 x 10'8 l
Vessel 1/4 T 9.64 x 10'7 1.18 x 10 2.99 x 10
3.64 x 10'8 18 i
Vessel 3/4 T 2.11 x 10'7 2.58 x 10'7 6.15 x 10'8 7.52 x 10'8 NOTE: EOL fluences are based on operation at 2652 MWt for 32 effective-full-power years.
%e I
l
?
9 0
N 6-20
1 T.94 CAPSULES U, X, V 16 O'
I/
19.72 CAPSULES W, Y, Z l
I 45' r
1
(
r}
r1 1 1 j
REACTOR VESSEL l
l su r 1
/
1 2
/
1 j
NEUTRON PAD i
1 i
REACTOR I
/
CORE BARREL CORE 1
/
/
1
/
1
//l Figure 6-1.
Farley Unit 2 Reactor Geometry 6-21 a
gCHARPY SPECIMEN r
,2 r
/ !
//
%% % % % % % % % % % % %% A N N K
NEUTRON PAD N
\\ \\ \\ \\ \\ \\ \\ \\ \\\\ \\ \\\\ \\ \\\\ \\
N.
l Figure 6-2.
Plan View of a Reactor Vessel Surveillance Capsule 6-22
10
9 8
7 6
5 4
l 3
2 u
h10" 9
"y 8
SURVEILLANCE 5
CAPSULES 6
l 5
5 aa 4
2 3
g I
w PRESSURE z
2
- VESSELIR 1/4T LOCATION m
10"'
9 8
7 6
5 4
3
- 3/4T LOCATION 2
f I
I t
I I
i 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEGREES)
Figure 6-3.
Calculated Azimuthal Distribution of Maximum FasuWeutron Flux (E > 1.0 Mev) within the Pressure Vessel Surveillance Capsule Geometry 6-23
2 l
l 10" 9
8 199.39 7
l 6
IR 5
204.39 4
l o
3
'h 3
l
{
1/4T u
3 2
d z
O cr
>=3 Lu z 10v, L
214.39 9
l 3
l l
7 6
3/4T 5
219.39 l
l 3
OR 2
10
194 196 198 200 202 204 206 208 210 212 214 216 218 220 222 R ADIUS (cm)
Figure 6-4.
Calculated Radial Distribution of Maximum Fast-Neutron Flux (E > 1.0 Mev) within the Pressure Vessel 6-24
10" 9
6 4
2 x
3 J
- u. 10-'. -
z o
8 E
'f 6
Z w
4
?_
d E
2 10'#
8 6
5 4
5 w
E O
2 TO VESSEL
- CLOSURE HEAD 10 '
I I
I I
t
-200
-200
-100 0
100 200 300 DISTANCE FROM CORE MIDPLANE (CM)
Figure 6-5.
Re!ative Axial Variation of Fast-Neutron Flux (E > 1.0 Mev) within the Pressure Vessel
+
G-25
ZW EO Z.u 3 X E # i* 2 7 g
1 3
g 0
gs 2
m 2
3 4
5 6 7 69 "
2 3
4 5 6 789 a 1
8 2
F gi u
1 re 6
.W C.U C A
A
.Z P.x P 6
S S
1 8
I 3
YU vU L
L E
E S
S wC t
ial thc inul 1
t a
8 i
h te 4
e d
R R
ea a i
c d t
ia ol r
1 6
D R8 I
Vi e s A5 26 s
D tr s i I
eb U
u S
i l
Sti
(
u o c
r n m
v
)
1 eo 8
i if 6
OOw 2W C l
F l
. ooQ O "Eq a
n a J
s ct i
e -N Ce a u 1
p tr 8
i s
o 7
ul n es Fl u
l x
(
E 1
8 I
8 1
0 M
I ev
)
1 8
9
)1 Il i
3 i
~
Np" (n,f) Cs'
5 Ni'8 (np) Co'*
h 10' 9
8 7
E wz 6
Ho 5
~
Ek 4
MO Oo Oa 3
2 I
U (n.f) Cs"'
- osf, Fe (n.f) Mn
o 10 9
>8 b
7
[-
6 0
5 at 4
0w Q3 m
3
~~
$2 m
10 Cu (n.a) Co*
9 8
7 6
5 4
g 3
l 2
10' 182 183 184 185 186 187 188 189 R ADIUS (cm)
Figure 6-7.
Calculated Variation of Fast-Neutron-Flux Monitor Saturated Activity within Capsules U, X, and V
- ~ '
)
=
M4&3C$WO OP>![tE?$
~
1 1
1 1
0 0
0 0
2 3
4 5 6789" 2
3 4
2 3
4 5 6789*
2 3
4 56789 7 1
8 2
- - - ~ - - - -
- - - ~ - - - - - - -
~
NN F
C FU p i" gi e
e '
a"
'(
u
.n
(
r
(
(
(
n
.n.n
.np e
.n p )f C
f) 1 6
8
)
i 3
)
)
Co 8
C MC s' "
s o
n 5 '
n AC c a t
l 1
i c v
8 e
i u 4
t l
y a t
we id t
t hiV na 6
Cai R
r A1 8
2 a i D
t 8
p o I
5 U
s n u
S o
l
(
e f
c I
s m
F Wa
)
s t
1 8
,Y N 6
Q06s$wx a e 8N-
"8U n u J00g9z t
dr e
o Zn -
F ul 1
8 a
x 7
M o
n i
t o
r 1
S 8
I a
8 tu ra t
e e
d 1
8 9
l; i
l' 1j
)
'l I
i l
SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE l
l l
l 1
The following removai schedule is recommended for future capsules to be removea from the Farley Unit 2 reactor vessel:
j Lead Removal Estimated Fluence Capsule Factor TimeQ n/cm x 10" 2
l U
3.12 Removed (1.1)
.56 l
W 2.70 4
2.18 X
3.12 6
3.781ni o
Z 2.70 12 6.54rci V
3.12 18 11.34 Y
2.70 Standby r
l
[a] Effective full power years from plant startup
[b] Approximates vessel end of life 1/4 thickr.,sss walllocation fluence
[c] Approximates vessel end of life inner wall location fluence 7-1
l 1
~ '
SECTION 8 REFERENCES
- 1. Davidson, J. A., Yanichko, S.E., " Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program,"
WCAP 8956, August,1977.
- 2. ASTM Stardard E185-73, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" in ASTM Standards, Part 10 (1973), American Society for Testing and Materials, Philadelphia, Pa.1973.
- 3. Regulatory Guide 1.99, Revision 1," Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commiss;on, April 1977.
- 4. Soltesz, R. G., Disney, R. K., Jearuch, J., and Zeigler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation. Vol. 5
- Two-Dimensional Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.
- 5. SAILOR RSIC Data Library Collecuen DLC-76, " Coupled, Self-shielded,47 Neutron,20 Gamma-ray, P3, Cross Section Library for Light Water Reactors."
- 6. Berchmark Testing of Westinghouse Ne tron Transport Analysis Method-ology - to be published.
- 7. ASTM Designation E261-77, Standard Practice for Measuring Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 915-926, American Society for Testing and Materials, Philadelphia. Pa.1981.
8-1
Y
.8. ASTM Designation E262-77, " Standard. Method for Measuring Thermal Neutron Flux by Radioactivation Techniques,"in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 927-935, American. Society for Testing and Materials, Philadelphia, Pa.,'1981.
9.- ASTM Designation E263-77, " Standard Method for Measuring Fast-Neutron Flux by Radioactivation of iron," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 936-941, American Society for Testing and Materials, Philadelphia, Pa.,1981.
l
-10. ASTM Designation E481-78, " Standard Method of Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver,"in ASTM Standards (1981),
Part 45,' Nuclear Standards, pp.1063-1070, American Society for Testing and Materials, Philadelphia, Pa.,1981.
I i
- 11. ASTM Designation E264-77, " Standard Method for Measuring Fast-Neutron I
Flux by Radioactivation of Nickel,"in ASTM Standards (1981), Part45, Nuclear f
Standards, pp. 942-945, American Society for Testing and Materials.
Philadelphia, Pa.,1981.
e 8-2
i APPENDIX A FARLEY UNIT 2 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1.
INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of i
RTNDT (reference nil-ducti!ity temperature). The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ARTNDT.RT i
I NDT s designated as the higher of either the drop weight nil-ductility transition temperature (TNDT) or the temperature at which the material exhibits at least 50 it Ib of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60 F.
RTN DT increases as the material is exposed to fast-neutron radiation. Thus, to find the most limiting RT NDT at any time period in the reactor life, AR r NDTdue to the radiation exposure associated with that time period must be added to the original unirradiated RT NDT. The extent of the shift in RT i
NDT s enhanced by certain chemical elements (such as copper and phosphorus) present in reactor vessel steels. The Regulatory Guide 1.99 trend curves which show the effect of fluence and copper and phosphorus contents on ART NDT for reactor vessel steels are shown in Figure 1.
Given the copper and phosphorus contents of the most limiting material, the radiation-induced ARTNDT can be estimated from Figure A-1. Fast-neutron f!uence (E > 1 Mev) at the 1/4T (wall thickness) and 3/4T (wall thickness) vessel locations are given as a function of full-powerservice life in Figure A-2.The data for all other ferritic materials in the reactorcoolant pressure boundary are examined to insure that no othar component will be limiting with respect to RT NDT. The fracture toughness properties of the ferritic materialin the reactorcoolant pressure A _-
boundary are determined in accordance with the NRC Regulatory Standard 1
Review Plan.111 The postirradiation fracture toughness properties of the reactor vessel beltline material were obtained directly from the Farley Unit 2 Vessel Material Surveillance Program.
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A-3.
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE l
RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the I
combined thermal and pressure stresses at any time during heatup and cooldown cannot be greater than the reference stress intensity factor, K~lR, for the metal temperature at that time. KIR'is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code.tal The KIR curve is given by the equation:
K IR = 26.78 + 1.223 exp (0.0145 (T-RT NDT + 160))
(A-1) where K R s.the reference stress intensity factor as a function of the metal I
i temperature T and the metal reference nil-ductility temperature RT NDT.Thus, the governing equation of the heatup-cooldown analysis is defii.ed in Appendix G to the ASME Code 1 as follows:
r2 C K IM + K lt 5 KIR (A-2) where:
i IM s the stress intensity factor caused by membrane (pressure) stress i
K Klt is the stress intensity factor caused by the thermal gradients KIR is a function of temperature to t; e HT NDT of the material C= 2.0 for Level A and Level B service limits C=
1.5 for hydrostatic ar.d leak test conditions during which the reactor core is not critical.
A-2
At any time during the heatup or cooldown transient, K IR s determined by the i
metal temperature at the tip of the postulated flaw, the appropriate value of RT NDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed. From equation (A-2), the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure-versus-coolan.t temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal g radients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.
From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooidown analysis is necessary because control of the cooidown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This con-dition, 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 K IR at the 1/4T location for finite cooldown rates than for steady-state operation.
Furthermore, if conditions exist such that the increase in K IR exceeds K lt, 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 tempera-ture at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures con-servative operation of the system for the entire cooldown period.
A-3
Three separate calculations are required to deterrnirIe the limit curves for finite s
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 1/4T defect at the inside of the vessel wall.
The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K f
IR orthe 1/4T crack during heatup is lower than the K IR or the 1/4T crack during steady-f state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower K IR'r do not offset each other, and the pressure-temperature curve based on stead /-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered.
Therefore, both cases have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of pressure-l temperature limitations for the case in which a 1/4T deep outside surface flaw is 1
1 assumed. Unlike the situation at the vessel inside surface, the thermal gradients l
established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ran i Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it i:, possible for conditions to exist wherein, over the course of the heatup ramp, the control'ing 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. Then, composite curves for the heatup rate data and the A-4
\\
cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.
A-4.
HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant
- - System have been calculated using the methods discussed previously. The deriva-tion of the limit curvesis presented in the NRC Regulatory Standard Review Plan.131 Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program.
Charpy test specimens from Capsule U irradiated to 5.61 x 10'8 n/cm' indicate that the representative core region weld metal and the limiting core region shell plate B7212-1 exhibited maximum shifts in RTNDT of 10 F and 133* F, respectively. The shifts are well within the appropriate design curve (Figure 1) prediction. Heatup and cooldown limit curves for normal operation of Farley Unit 2 for up to 5*
~
effective-full power years (EFPY) have been generat'ed, and are shown in Figures A-3 and A-4.
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line, shown in Figure A-3. This is in addition to other criteria which must be met before the reactor is made critical.
The leak-test limit curve shown in Figure A-3 represents minimum temperature re-t2 3 quirements at the leak test pressure specified by applicable codes Up to 4.3 EFPY service period assuming margins of 10 F. and 60 psig for possible instrument errors.
e A-5
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A [40 + 1000 (% Cu - 0.08) + 5000 (ao P - 0.008)] (f /10"l 1/2 g
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gppER 300 w
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b200 E
b o#
o? e
=
?
8 z 100 w
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ch
@o c 50
~0.35 0.30 0.25 0.20 o Cu 0.15% Cu 0.10 o Cu w
c P = 0.012 LOWER LIMIT o
o 5
% Cu = 0.08 W
% P = 0.008
~
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$ Plate B7212-1
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a t iI e I
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e i a B R R
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f f I I
2 x 10
4 6
8 10
2 4
6 8
10
2 4
6 2
FLUENCE. n/cm (E > 1 Mev)
Figure A-1.
Predicted Adjustment of Reference Temperature, "A", as a Function of Fluence and Copper Content.
For Copper and Phosphorus Contents other than those Plotted, use the Expression for "A" given on the Figure.
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sT#
T#
2 g
3 4
4 E
/
/
1 3
C A
0 F
3 p
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3 no it 62 c
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^~EuE~ w02WSu. bEawz 5
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?~
-4 Curve Applicable for Heatup Rates up to LEAK TEST LIMIT 60* F/HR, for the Service Period up to 5 l
EFPY, and Contains Margins of 10' F and 60 PSIG for.Possible Instrument Errors for the Service Period up to 4.3 EFPY.
Materials Basis:
l Controlling Materials - RV Inter Shell l
Copper Content 0.20%
l
~
RT Original, -10*F RT Af terd.3 EFPY, 1/4T, 137' F hM
~
3/4T. 57' F n.
un 1800 E
HEAT UP RATES
$1600 TO 60"F/HR g
8 1400 B
y h1200
=
5z 1000
~ CRITICALITY LIMIT 800 Based on Inservice Hydrostatic Test Temperature (277'F) for the Service Period up to 5 EFPY.
600 400 200 a
Z f
f I
I a
a
~
0 50 100 150 200 250 300 350 400 INDICATED TEMPCR ATURE (* F)
Figure A-3.
Farley Unit 2 (APR) Reactor Coolant System Heatup Limitations Applicable for the First 4.3EFPY
7
~
4 r
2400 2200 i
Curve Applicehle for the Serv.t.c4 Period up to 5 EFPY and Contains 2000 Margins of 10' F and 60 PSIG for Possible Instrument Errors for the 6gg Service Period up to 4.3 EFPY.
is a.
~
w 1000
=
ac3
$ 1400 u.ini e. se i.:
=
Controlling Material-RV Inter Shell
$ 1000 Af ter41,3EFPY,1/4T 137' F j
R 0
l T
3/4T. 57 F l,,
l COOLDOWN RATE
- F/HR.
W 0
20 60 400 100 200 a
a a
a a
e 0
50 100 150 200 250 300 350
,' 400 a
a INDICATED TEMPERATURE (* F)
Farley Unit 2 (APR) Reactor Coolant System Cooldome Linaltations
. Figure A-4.
Applicable for the First 5 'EFPY
1 l
i D
)
APPENDIX A REFERENCES 1
i k
(1) " Fracture Toughness Requirements", Branch Technical Position - MTEB S-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Re-l ports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981.
f l
(2) ASME Boiler and Pressure Vessel Code, Section lil, Division 1 - Apprendices,
" Rules for Construction of Nuclear Vessels", Appendix G," Protection Against Nonductile Failure", pp.461-469,1980 Edition, American Society of Mechanical Engineers, New York,1980, p
(3) " Pressure-Temperature Limits", Chapter 5.3.2 in Standard Revisw Plan for the Review of Safety Analyses Reports for Nuclear Power Plants, LWR Edition.
ll
- NUREG-0800,1981.
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