ML20081B637
| ML20081B637 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/29/1984 |
| From: | Boggs R, Cheney C, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20081B631 | List: |
| References | |
| TAC-54449, TAC-54894, WCAP-10474, NUDOCS 8403090259 | |
| Download: ML20081B637 (90) | |
Text
_ - - _______ ___ _______________________________
WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION WCAP-10474 l
f ANALYSIS OF CAPSULE U FROM THE ALABAMA POWER COMPANY l
JOSEPH M. FARLEY UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM R.S.Boggs S. E. Yanichko p;
C. A. Cheney l
W. T. Kaiser l
February 1984 l
l Work performed under Shop Order No. AOCP 106 APPROVED:
T. R. Mager, Manager Metallurgical and NDE Analysis 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 without the customer's approval.
r WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 8403090259 840301 PDR ADOCK 05000348 P
s r
PREFACE This report has been technically reviewed and verified.
Reviewer Sections 1 thru 5 and 7 M. K. Kc;ika Section 6 S. L. Anderson Appendix A F. J. Witt i-s r
r_ __
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 6-2.
Discrete Ordinates Analysis 6-1 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
).
1 i
4 LIST OF TABLES Table Title Page 4-1 Chemical Composition and Heat Treatment of The 4-3 Farley Unit 1 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for The Farley Unit 1 5-5 Lower Shell Plate B 6919-1 at 550* F, Fluence 1.65 x 10'(Transverse) irradiated n/cm (E > 1 MeV) 5-2 Charpy V-Notch Impact Data for The Farley Unit 1 5-6 Lower Shell Plate B 6919-1 (Longitudinal)
Irradiated at 550*F, Fluence 1.65 x 10"(E > 1 MeV) 5-3 Charpy V-Notch Impact Data for The Farley Unit 1 5-7 Pressure Vessel Weld Heat Affected Zone Metal Irradiated at 550*F, Fluence 1.65 x 10"(E > 1 MeV)
/
5-4 Charpy V-Notch Im aact Data for The Farley Unit 1 5-8 Pressure Vessel We d Metal irradiated at 550*i:,
Fluence 1.65 x 10" n/cm (E > 1 MeV) 2 5-5 Instrumented Charpy impact Test Results for 5-9 Farley Unit 1 Lower Shell Plate B 6919-1 (Transverse) 5 Instrumented Charpy impact Test Results for The 5-10 Farley Unit 1 Lower Shell Plate B 6919-1 (Longitudinal) 5-7 Instrumented Charpy impact Test Results for 5-11 Farley Unit 1 Heat Affected Zone Metal 5-8 Instrumented Charpy impact Test Results for 5-12 Farley Unit 1 Weld Metal 5-9 The Effect of 550*F Irradiation at 1.65 x 10" 5-13 (E > 1 MeV) on the Notch Toughness Properties of The Farley Unit 1 Reactor Vessel Materials 10 Tensile Properties for Farley Unit 1 Reactor Vessel 5-14 Material Irradiated to 1.65 x 10" n/cm 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 1 Surveillance Capsules 6-4 Calculated Neutron Energy Spectra at the Center 6-13 of Farley Unit 1 Surveillance Capsules 6-14 il
LIST OF TABLES (cont.)
d Table Title Page 6-5 Spectrum-Averaged Reaction Cross Sections at the 6-15 Center of Farley Unit 1 Surveillance Capsules 6-6 Irradiation History of Faridy Unit 1 Reactor Vessel 6-16 Surveillance Capsule U 6-7 Comparison of Measured and Calculated Fast-Neutron 6-19 Flux Monitor Saturated Activities for Capsule U 6-8 Results of Fast-Neutron Dosimetry fcr Capsule U 6-20 6-9 Results of Thermal-Neutron Dosimetry for Capsule U 6-21 6-10 Summary of Neutron Dosimetry Results for Capsule U 6-22 A-1 Reactor Vessel Toughness Data A-7 s
-g e
t lii
f LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in Farley 4-4 Unit 1 Reactor Vessel (Updated 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 1 Reactor Vessel Lower Shell Plate B 6919-1 (Transverse Orientation) 5-2 Irradiated Charpy V-Notch Impact Properties for 5-16 Farley Unit 1 Reactor Vessel Lower Shell Plate B 6919-1 (Longitudinal Orientation) 5-3 Irradiated Charpy V-Notch Impact Properties for 5-17 Farley Unit 1 Reactor Pressure Vessel Weld Heat Affected Zone Metal y
5-4 Irradiated Charpy V-Notch Impact Properties for 5-18 Farley Unit 1 Reactor Pressure Vessel W6fd Metal 5-5 Charpy impact Specimen Fracture Surfaces for 5-19 Farley Unit 1 Pressure Vessel Lower Shell Plate B 6919-1 (Transverse Orientation) 5-6 Charpy impact Specimen Fracture Surfaces for 5-20 Farley Unit 1 Pressure Vessel Lower Shell Plate B 6919-1 (Longitudinal Orientation) 5-7 Charpy impact Specimen Fracture Surfaces for 5-21 Farley Unit 1 Weld-Heat-Affected-Zone Metal 5-8 Charpy impact Specimen Fracture Surfaces for 5-22 Farley Unit 1 Weld Metal 5-9 Comparison of Actual versus Predicted 30 ft Ib 5-23 Transition Temperature increases for the Farley Unit 1 Reactor Vessel Material based on the Prediction Methoda of Regulatory Guide 1.99 Revision 1 5-10 Tensite Properties for Farley Unit 1 Reactor Vessel 5-24 Lower Shell Plate B 6919-1 (Transverse Orientation)
'I' 5-11 Tensile Properties for Farley Unit 1 Reactor Vessel 5-25 Lower Shell Plate B 6919-1 (Longitudinal Orientation) 7 iv
LIST OF ILLUSTRATIONS (cont.)
i Figure Title Page 7
5-12 Tensile Properties for Farley Unit 1 Reactor 5-26 Vessel Weld Metal 5-13 Fractured Tensile Specimens of Farley Unit 1 5-27 Reactor Vessel Lower Shell Piate B 6919-1 (Transverse Orientation) 5-14 Fractured Tensile Specimens of Farley Unit 1 5-28 Reactor Vessel Lower Shell Plate B 6919-1 (Longitudinal Orientation) 5-15 Fracture Tensile Specimens of Farley Unit 1 5-29 Reactor Vessel Weld Metal 5-16 Typical Stress-Strain Curve for Tension Specimens 5-30 6-1 Farley Unit 1 Reactor Geornetry 6-23 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-24 6-3 Calculated Azimuthal Distribution of Maximum 6-25 Fast-Neutron Fiux (E > 1.0 Mev) within the Pressure Vessel Surveillance Capsule Geometry 6-4 Calculated Radial Distribution of Maximum 6-26 Fast-Neutron Flux (E > 1.0 Mev) within the y
Pressure Vessel 6-5 Relative Axial Variation of Fast-Neutron Flux 6-27 (E > 1.0 Mev) within the Pressure Vessel 6-6 Calculated Radial Distribution of Fast-Neutron Flux 6-28 (E > 1.0 Mev) within the Reactor Vessel Surveillance Capsules 6-7 Calculated Variation of Fast-Neutron Flux Monitor 6-29 Saturated Activity within Capsules U, X, and Y 6-8 Calculated Variation of Fast-Neutron-Flux Monitor 6-30 Saturated Activity within Capsules W, V, and Z A-1 Effect of Fluence and Copper Content on A-8 ARTNDT for Reactor Vessel Steels Exposed to Irradiation at 550 F A-2 Fast Neutron Fluence (E > 1.0 Mev) as a Function A-9 of Full Power Service Life A-3 Farley Unit 1 (ALA) Reactor Coolant System A-10 Heatup Limitations Applicable for the First 7 EFPY A-4 Farley Unit 1 (ALA) Reactor Coolant System A-11 l
Cooldown Limitations Applicable for the First 7 EFPY V
I l
SECTION 1
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule U, the second capsule to be removed from the Farley Unit 1 reactor pressure vessel,ied to the following conclusions:
a The capsule received an average fast neutron fluence (E>1.0 Mev) of 1.65 x 10'8 n/cm2 I
e irradiation of the reactor vessel lower shell plate B 6919-1 to 1.65 x 10' n/cm2 resulted in 30 and 50 ft Ib transition temperature increases of 90 and 95' F, b
respectively, for specimens oriented normal to the principal rolling direction of the plate and 105 and 120* F, respectively, for specimens oriented parallel to the plate principal rolling direction.
m Weld metal irradiated to 1.65 x 10 n/cm2 resulted in both 30 and 50 ft Ib tran-sition temperature increases of 80 and 85* F respectively.
m Comparison of the 30 ft Ib transition temperature increases for the Joseph M.
Farley Unit 1 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99 Revision 1 shows that the plate material did not embrittle as much as predicted. The weld metal shift is also less than predicted by the methods of NRC Regulatory Guide 1.99 Rev.1.
1 1-1
f SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule U, the second capsule to be removed from the reactor in the continuing surveillance program which moni-tors the effects of neutron irradiation on the Joseph M. Farley Unit 1 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Joseph M. Farley Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corpora-tion. A description of the surveillance program and the preirradiation mechanical I
properties of the reactor vessel materials are presented by Davidson and Yanichko. I'l The surveillance program was pianned to coverthe 40-yeardesign life of the reactor i;
pressure vessel and was based on ASTM E-185-73," Recommended Practice forSur-veillance Tests for Nuclear Reactor Vessels".lal Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance Capsule U removed from the Joseph M. Farley Unit 1 reactor vessel and discusses the analysis of these data.
I 2-1
r SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist f racture 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 SJ 533 Grade B Class 1 (base material of the Joseph M. Farley Unit 1 reactor pressure vesse! beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irraation.
{
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 til of the ASME Boiler and Pressure Vesse: Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RT N DT)-
RTNOT 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 R curve) which appears in Appendix G of the ASME Code.
I The K IR curve is a lower bound of dynamic, crack arrest, and static fracture tough-ness results obtained from several heats of pressure vessel steel. When a given material is 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 factors.
3-1 l
l
RT NDT and, in turn, the operating limits of nuclear power plants can be adjusted to I
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 1 Reactor Vessel Radiation Surveillance Program, D1 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 RT NDT) due to irradiation is added to the original RT NDT to adjust the RT NDT for radiation embrittlement. This adjusted RT NDT (RTNDTinitial
+ A RTNOT) 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 tie effects of irradiation on the reactor vessel materials.
l r
3-2 1-D
r SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Joseph M. Farley Unit 1 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 vessei 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 from the reactor after 3.02 effective full power years of plant operation. This capsule contained Charpy V-notch impact specimens from the limiting core region plate (lower shell plate B 6919-1), core region weldment and weld heat affected zone (HAZ) material. All heat affected zone specimens were obtained from within the HAZ of plate B 6919-1. The capsule also contained compact tensile (CT) and tensile specimens from the lower shell plate B 6919-1 and weldment, and a bend bar specimen from the lower shell plate B 6919-1. The chemistry and heat treatment of the program surveillance materials are presented in Table 4-1.
Al! test specimens were machined from the 1/4 thickness location of the plate. Test specimens represent material taken at least one plate thickness from the quenched end of the plate. 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 E399-70T. The precracked bend bar was machined in the transverse orientation. Charpy V-notch specimens from the weld metal were I
oriented with the longitudinal axis of the specimens normal to (transverse orientation) the weld direction. Tensile specimens were oriented with the 4-1
I longitudinal axis of the specimen normal to (transverse orientation) the weld direction.
s Capsule U contained dosimeter wires of pure copper, iron, nickel, a'nd aluminum -
0.15% cobalt (cadmium-shielded and unshielded). In addition cadmium shieided dosimeters of Np*87 and U*** 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 Melti,ng 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 Capsule U are shown in Figure 4-2.
I t'
r 4-2
4 TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE FARLEY !) NIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Chemical Composition (WT-%)
Plate B 6919-1 Weld Metal I""
Combustion Engineering Westinghouso Analysis Analysis C
0.20 0.13 S
0.015 0.009 Na C.003(bl 0.005 Co 0.008 0.018 Cu 0.14 0.14 Si 0.18 0.27 Mo 0.56 0.50 Ni 0.55 0.19 Mn 1.39 1.06
{
Ibl 0.063 Cr 0.13 II V
<0.001 0.003 P
0.015 0.016 Sn 0.008 *3 0.005 l
At 0.025 0.009
[a] All elements not listed are less than 0.010 weight %.
[b] Westinghouse Analysis.
Heat Treatment Heat Treatment Material
- 7 Time (br)
Coolant p
Intermediate 1550 /1650* F 4
Water quenched Shell Plate B 6919-1 1225'F 25'F 4
Air cooled 1150 F 25' F 40 Furnace cooled to f
600 F
}
Weidment 1150* F 25 F 16 Furnace cooled 4-3
+
0*
REACTOR VESSEL (3.12) Y (2.70) Z CORE BARREL
-NEUTRON PAD (2.70) V (3.12) X
<r 90' 270*
U (3.12) hW (2.70)
C[
l y
180*
NOTE:
Capsule identifications Have Changed From Those Identified in WCAP-8810 9
Figure 4-1.
Arrsngement of Surveillanco Capsules in
(
Farley Unit 1 Reactor Vessel (Updated Lead Factors for the Capsules Shown in Parentheses.)
4-4
9
)
i e
SPECIMEN NUMBERING CODE:
AT - PLATE B 6919-1 (TRANSVERSE ORIENTATION)
I AL - PLATE B 6919-1 (LONGITUDINAL ORIENTATION)
AW - CORE REGION WELD METAL AH - HEAT-AFFECTED-ZONE METAL COMPACT COasPACT COMPACT COMMCT SENO S AR TENSli2 TENSION TE NSION CHARPY CHARPV CHARPV T E NSION TE NS ON CHARPY Awef Awa astS 4 Met set; AM12 Awt A*et AMs U
Att Am2 A$4 AW)
AW7 AW1 4W14 4M14 AW t '
AL2 All AWS 4M4 set AW13 AM13 4 Wit AMt0 Aw?
AMF And AM4
.mmmmm gl' Cu M-tSN Ce Ce ie i l
ummll T.
SF9 F OOO 590 F MONITOR a ll ll 5 m-IS% Co ICet you, yon 8 ll ll 8 1is It t
,A l 8. Il l TO TOP OF VESSft.
1
\\
.l
1 g NEUTRON SHIELD PAD l
hk :
d'-
.* Y 5/v l
i
.e**
d CAPSULE U d
CORE CORE BARREL T.
{~~
VESSEL WALL Aleo Available On CAR) wrii>re c-a CAPsuta u N~
Pv OOtawETER COMPACT COMPACT CHARPV SLOCK TEN 5 ELE CHARPV CHARPV CHARPV CHARPT CHARPV YEN 410N TENSION TENStLE AM3 b
Af ts At tl AL 2 ATg att ALS AT3 AL3 AT3 aW2 AM2 tA3 AL3 AT14 AL14 471*
Allt A?O AL4 A9 ALS AT3 AL3 AT4 AT3 AT3 Att A T3 LW1 LMt At t Aft)
AL 13 A tto Atle A TF AL 7 Af4 AL4 ATI At t AT_'
II I
3 II g
asl 5
y yV W m-iss C.
l; C,
gg g3 p__ u_,,s e,
]
I l
ll Ili e a a i el l T+1 es, r.
- r,, g
lw a a e-A y w w, "ll l N M
9lll -
m-m CA Cai 1
1 m-m C. <C.,
.O sre r l
70.
}jO e0 l i lI Il i i
18 1 a_!! 'I '
8..!'8
-CENTf3 KEG 3CN Or VESSEL TO SOTTOW Or VERSEL Figure 4-2.
Capsule U Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters 4-5 1
n409090259-OI
i 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 personnel. Testing was performed in accordance with 10CFR50, Apoendices G and H, ASTM Specification E185-82 and Westinghouse Procedure MHL 7601, Revision 3 as modified by RMF Procedures 8102 and 8103.
2 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-8810.D3 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 l
Procedure 8103 on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy machine is instrumented with an EffectsTechnology model 500instrumenta-tion system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (E D ). From the load-time curve, the load of general yielding (P GY), the time to general yielding (t GY), the maximum load (P M ), 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 et which fast fracture was initiated is identified as the fast f racture load (P F ),
and the load at which fast fracture terminated is identified as the arrest load (P A)-
1 5-1
(
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 (E c )
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 usirg the three point bend formula.
Percent shear was determined from postfracture photographs using the rat lo-of-areas methods in compliance with ASTM Specification A370-77. The lateral exoan-sion 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 R c 45. The upper e
pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constar.t crosshead speed of 0.05 inch per minute
[
throughout the test.
Deflection measarements were made with a linearvariable 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 procedure was used to monitor specimen temperature. Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In test configuration, with a slight load on the spacimen, a plot of specimen temperature versus upper and lower grip and contretter temperatures w:3 developed over the range room temperature to 550* F (288'C). The upper grip was used to control the furnace temperature. During t
5-2
the actual testing the grip temperatures were used to obtain desired specimen temperatures. Experiments indicated that this method is accurate to plus or minus 2* F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used to calculate the 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 con-tained in Capsule U irradiated at 1.65 x 10 n/cm are presented in Tables 5-1 through 5-8 and Figures 5-1 through 5-4. A summary of the transition temperature in-creases and upper shelf energy decreases for the Capsule U material is shown in
/
Table 5-9.
l Irradiation of plate B 6919-1 material (transverse orientation) to 1.65 x 10 n/cm (Figure 5-1) resulted in 30 and 50 ft Ib transition temperature increases of 90 and 95'F, respectively, and an upper shelf energy decrease of 8 ft Ib. Irradiation of plate 2
B 6919-1 materiel (longitudinal orientation) to 1.65 x 10 n/cm (Figure 5-2) resulted in 30 and 50 f t Ib transition temperature increases of 105 and 120' F, r6spectively, and an upper shelf energy decrease of 30 ft Ib.
l Weld metal irradiated to 1.65 x 10 n/cm (Figure 5-4) resulted in 30 and 50 ft Ib l
transition temperature increases of 80 and 85* F respectively and an upper shelf energy decrease of 41 ft Ib.
Weld HAZ metalirradiated to 1.65 x 10 n/cm'(Figure 5-3) resulted in 30 and 50 ft Ib transition temperature increases of 120 and 135"F respectively and an upper shelf energy decrease of 40 ft Ib.
i The fracture appearance of each irradiated Charpy specimen from the various ma-terials is shown in Figures 5-5 through 5-8 and show an increasing ductile or tougher appearance with increasist test temperature.
5-3
l f
Figure 5-9 shows a comparison of the 30 ft Ib transition temperature increases for the various Farley Unit 1 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99 Revision 1.l31 This comparison shows that 2
the transition temperature increases resulting fromirradiation to 1.65 x 10n/cm are less than predicted by the Guide for plate B 6919-1, transverse and longitudinal orientations. The weld metal transition temperature increase resulting from 1.65 x 10 n/cm2 is also less than prendicted by the Guide.
5-3.
TENSION TEST RESULTS The results of tension tests performed on plate B 6919-1 and weld metalirradiated to 1.65 x 10 n/cm* are shown in Table 5-10 and Figures 5-10, 5-11 and 512, respectively. These results show that irradiation produced an increase in 0.2 percent yield strength of 10 to 19 ksi for plate B 6919-1 and approximately 8 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 y
The 1/2T compact tension fracture mechanics specimens that were contained in
[
Capsule U will be reported at a later time.
t 5-4 j
I TABLE 5 CHARPY V-NOTCH IMPACT DATA FOR THE FARLEY UNIT 1 r
LOWER SHELL PLATE B 6919-1 (TRANSVERSE) lRRADIATED AT 550*F, FLUENCE 1.65 x 10 n/cm* (E > 1 MeV)
Temperature impact Energy Lateral Expansion SW No.
C (F)
Joules (ft-lb)
MM (mils)
(%)
-4
( 25) 8.0
( 6.0) 0.05 ( 2.0) 5 AT9 10 ( 50) 32.5 (24.0) 0.28 (11.0) 11 AT11 24 ( 75) 36.5 (27.0) 0.30 (12.0) 14 AT13 24 ( 75) 43.5 (32.0) 0.46 (18.0) 17 AT10 38 (100) 34.0 (25.0) 0.38 (15.0) 14 AT14 38 (100) 34.0 (25.0) 0.38 (15.0) 21 AT2 52 (125) 44.5 (33.0) 0.51 (20.0) 26 AT6 52 (125) 42.0 (31.0) 0.46 (18.0) 25 AT3 66 (150) 54.0 (40.0) -
0.64 (25.0) 29
)
AT15 93 (200) 74.5 (55.0) 0.97 (38.0) 55 l
AT8 107 (225) 111.0 (82.0).
1.55 (61.0) 100 AT12 121 (250) 116.5 (86.0) 1.45 (57.0) 100 AT1-149 (300) 107.0 (79.0) 1.40 (55.0) 100 AT7 177 (350) 119.5 (88.0) 1.42 (56 9) 100 AT4 204 (400) 103.0 (76.0) 1.41 (55.5) 100 i
l 5-5
TABLE 5-2
(
CHARPY V-NOTCH IMPACT DATA FOR THE FARLEY UNIT 1 LOWER SHELL PLATE B 6919-1 (LONGITUDINAL) lRRADIATED AT 550'F, FLUENCE 1.65 x 10" n/cm* (E > 1 MeV)
Temperature Impact Energy Lateral Expansion SW No.
C (F)
Joules (ft-lb)
MM (mils)
(%)
AL3 10 ( 50) 15.0 ( 11.0) 0.05 ( 2.0) 5 AL8 24 ( 75) 53.0 ( 39.0) 0.56 (22.0) 27 AL6 24 ( 75) 47.5 ( 35.0) 0.46 (18.0) 13 AL11 38 (100) 46.0 ( 34.0) 0.51 (20.0) 15 AL12 38 (100) 44.5 ( 33.0) 0.43 (17.0) 18 AL15 52 (125) 54.0 ( 40.0) 0.66 (26.0) 20 AL9 52 (125) 68.0 ( 50.0) 0.74 (29.0) 29 AL13 66 (150) 72.0 ( 53.0) 0.81 (32.0) 28 ALS 66 (150) 80.0 ( 59.0) 0.79 (31.0) 40 g
AL4 93 (200) 92.0 ( 68.0) 1.09 (43.0) 55 AL7 107 (225) 127.5 ( 94.0) 1.75 (69.0) 100 t
AL14 121 (250) 146.5 (108.0) 1.80 (71.0) 100 AL1 149 (300) 149.0 (110.0) 1 88 (74.0) 100 AL10 177 (350) 162.5 (120.0) 1.93 (76.0) 100 AL2 204 (400) 143.5 (106.0) 1.70 (67.0) 100 9
5-6
I TABLE 5-3 i
CHARPY V-NOTCH IMPACT DATA FOR THE FARLEY UNIT 1 PRESSURE VESSEL WELD HEAT AFFECTED ZONE METAL 2
IRRADIATED AT 550* F, FLUENCE 1.65 x 10" n/cm (E > 1 MeV)
Temperature impact Energy Lateral Expansion SW No.
C (F)
Joules (ft-lb)
MM (mils)
(%)-
AH10
-73 (-100) 4.0 ( 3.0) 0.00 ( 0.0) 5 AH14
-46 ( -50) 27.0 ( 20.0) 0.18 ( 7.0) 11 AH7
-46 ( -50) 40.5 ( 30.0) 0.28 (11.0) 21 AH9
-32 (-25) i 20.5 I 15.0) 0.15 ( 6.0) 26 AH1
-18 (
0) 26.0 ( 19.0) 0.38 (15.0) 48 AH6
-18 (
0) 40.5 ( 30.0) 0.46 (18.0) 51 AH15
-4
( 25) 69.0 ( 51.0) 0.74 (29.0) 49 AH3 10 ( 50) 114.0 ( 84.0) 1.19 (47.0) 95
?
AH12 24 ( 75) 124.5 ( 92.0) 1.19 (47.0) 75 AH13 38 ( 100) 114.0 ( 84.0) 1.24 (49.0) 67
}
AH2 38 ( 100) 171.0 (126.0) 1.83 (72.0) 100 AHS 66 ( 150) 130.0 ( 96.0) 1.54 (60.5) 86 AH4 93 ( 200) 165.5 (122.0) 1.75 (69.0) 100 AH8 149 ( 300) 141.0 (104.0) 1.75 (69.0) 100 AH11 177 ( 350) 162.5 (120.0) 1.73 (68.0) 100 5-7 I
(
TABLE 5-4 CHARPY V-NOTCH IMPACT DATA FOR THE FARLEY UNIT 1 PRESSURE VESSEL WELD METAL IRRADIATED AT 550* F, FLUENCE 1.65 x 10" n/cm (E > 1 MeV) 2 I
Temperature impact Energy Lateral Expansion SW No.
C (F)
Joules (ft-lb)
MM (mils)
(%)
AW6
-46 (-50) 7.0 ( 5.0) 0.03 ( 1.0) 13 AW1
-32 (-25) 35.5 ( 26.0) 0.33 (13.0) 14 AW4
-18 ( 0) 43.5 ( 32.0) 0.46 (18.0) 28 AW13
-18 ( 0) 42.0 ( 31.0) 0.43 (17.0) 37 AW14
-4
( 25) 51.5 ( 38.0) 0.58 (23.0) 37 AW5
-4
( 25) 65.0 ( 48.0) 0.74 (29.0) 50 AW8 10 ( 50) 81.5 ( 60.0) 0.89 (35.0) 58 AW10 24 ( 75) 97.5 ( 72.0) 1.17 (46.0) 63 AW3 24 ( 75) 110.0 ( 81.0) 1.42 (56.0) 76 AW12 38 (100) 111.0 ( 82.0) 1.40 (55.0) 76 AW7 66 (150) 130.0 ( 96.0) 1.73 (68.0) 100 AW15 93 (200) 146.5 (108.0) 1.88 (74.0) 100 AW11 121 (250) 138.5 (102.0) 2.01 (79.0) 100 AW2 149 (300) 143.5 (106.0) 1.93 (76.0) 100 AW9 177 (350) 154.5 (114.0) 2.11 (83.0) 100
\\
5-8
~
TABLE 5-5 l
INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR FARLEY UNIT 1 LOWER SHELL PLATE B 6919-1 (TRANSVERSE)
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 No.
(" C) (Joules) (kJ/m*)
(kJ/m*)
(kJ/m )
(N)
( Sec)
(N)
(pSec)
(N)
(N)
(MPa) (MPa) j ATS
-4 8.0 102 68 34 17000 95 16800 105 16800 0
874 870 AT9 10 32.5 407 386 21 16900 110 20400 380 20400 0
867 958 AT11 24 36.5 458 366 91 15900 100 19600 380 19400 2600 819 914 AT13 24 43.5 542 479 63 15400 105 20200 480 20200 0
792 916 AT10 38 34.0 424 342 82 16200 110 19500 360 19500 2800 833 919
)
AT14 38 34.0 424 283 141 13400 95 16700 355 16400 1700 688 773 AT2 52 44.5 559 425 135 15700 115 19900 445 19800 3700 808 917 AT6 52 42.0 525 429 96 15300 110 19500 450 19500 3600 785 895 AT3 66 54.0 678 468 210 13900 85 18900 505 18400 5700 714 842 AT15 93 74.5 932 429 503 11900 80 17300 505 16900 10400 613 751 AT8 107 111.0 1390 484 906 14100 105 19400 510 723 861 AT12 121 116.5 1458 581 877 14300 105 19500 595 737 871 AT1 149 107.0 1339 508 831 13600 80 19400 520 700 849 AT7 177 119.5 149i 546 945 14100 85 19600 555 728 869 AT4 204 103.0 1288 426 862 9400 75 18000 500 486 684
TABLE 5-6 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR FARLEY UNIT 1 LOWER SHELL PLATE B 6919-1 (LONGITUDINAL) l Normalized Energies Test Charpy Charpy Maximum Prop
'rseld Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress No.
( C) (Joules) (kJ/m*)
(kJ/m*)
(kJ/m*)
(N)
( Sec)
(N)
( Sec)
(N)
(N)
(MPa) (MPa) l AL3 10 15.0 186 140 47 16800 90 18600 170 18100 0
863 910 AL8 24 53.0 661 573 88 15900 125 21500 560 21500 1100 817 963 o,
AL6 24 47.5 593 502 92 16200 100 21000 485 21000 0
833 958 AL11 38 46.0 576 527 49 15800 110 20400 530 20400 700 815 933 AL12 38 44.5 559 517 42 16700 115 21200 500 21200 1200 858 975 AL15 52 54.0 678-587 91 15600 105 20600 580 20600 3800 801 929 AL9 52 68.0 847 625 223 16000 105 20800 605 20400 3600 824 946 AL13 66 72.0 898 598 300 -#5300 115 20400 600 19600 5200 790 919 ALS 66 80.0 1000 656 344 16100 105 21200 620 20900 4800 831 961 AL4 93 92.0 1152 679 473 16000 115 21500 645 21100 12100 822 964 AL7 107 127.5 1593' 554 1039 14500 110 19300 585 744 869 AL14 121 146.5 1830 628 1202 14800 85 21200 600 762 927 AL1 149 149.0 1864 589 1276 14800 85 20700 570 760 912 AL10 177 162.5 2034 648 1385 13300 80 20400 640 684 867 AL2 204 143.5 1796 519 1277 12800 85 18500 570 659 807 c
1 TABLE 5-7 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR FARLEY UNIT 1 WELD 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 Yleid Load Maximum Load Load Stress Stress No.
( C) (Joules) (kJ/m*)
(kJ/m*)
(kJ/m*)
(N)
(ySec)
(N)
( Sec)
(N)
(N)
(MPa) (MPa)
AH10
-73 4.0 51 34 17 11900 75 11900 75 11900 0
611 611 AH14
-46 27.0 339 293 46 20d00 135 22100 310 22100 0 1048 1093 (ny AH7
-46 40.5 508 455 53 20100 110 22900 405 22900 0
1035 1106 AH9
-32 20.5 254 96 158 17900 90 19400 125 18700 5100 922 960 AH1
-18 26.0 322 71 251 19100 120 19100 120 19100 6800 982 982 AH6
-18 40.5 508 103 406 18700 95 19700 130 18700 9400 064 989 AH15
-4 69.0 864 449 416 17000 85 20900 430 20500 13000 874 974 AH3 10 114.0 1424 723 701 17700 90 22000 655 11000 7300 913 1022 AH12 24 124.5 1559 608 952 15200 100 18900 630 15100 6000 783 879 AH13 38 114.0 1424 684 739 16200 95 21000 665 17900 11100 833 957 AH2 38 171.0 2135 113 2023 17200 95 16200 145 883 859 AHS 66 130.0 1627 584 1043 16000 95 20100 585 823 930 AH4 93 165.5 2068 735 1332 15300 115 20200 740 790 913 AH8 149 141.0 1763 623 1139 13100 75 19200 650 673 831 AH11 177 162.5 2034 686 1348 13500.
90 18500 740 693 823
TABLE 5-8 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR FARLEY UNIT 1 WELD METAL l
Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yleid Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yleid Load Maximum Load Load Stress Stress No.
( C) (Joules) (kJ/m*)
(kJ/m )
(kJ/m )
(N)
( Sec)
(N)
( Sec)
(N)
(N)
(MPa) (MPa)
AW6
-46 7.0 85 41 44 13800 80 13800 85 13800 0
712 711 AW1
-32 35.5 441 407 34 17100 120 19700 425 19700 0
881 948 m
AW4
-18 43.5 542 442 100 17400 150 19300 485 19200 4400 897 944 b
AW13
-18 42.0 525 462 64 18400 95 2i000 435 21000 3500 948 1015 AW14
-4 51.5 644 517 127 16800 110 20300 510 20300 3200 863 953 AWS
-4 65.0 813 563 250 16800 110 19800 560 19400 4400 863 940 ASN8 10 81.5 1017 650 367 17700 110 21600 595 21300 6600 909 1009 AW10 24 97.5 1220 584 636 15200 150 19300 630 18900 8700 783 888 AW3 24 110.0 1373 631 742 15600 105 19500 630 17900 11000 803 904 AW12 38 111.0 1390 612 777 16200 105 19800 605 16200 10200 833 926 AW7 66 130.0 1627 568 1059 13800 110 17300 645 709 801 AW15 93 146.5 1830 581 1250 15400 90 19200 595 792 890 AW11 121 138.5 1729 570 1159 14300 85 18600 600 735 846 AW2 149 143.5 1796 551 1246 11500 55 18000 605 590 759 AW9 177 154.5 1932 584 1348 11200 80 16400 700 577-709 l
l w
TABLE 5-9 THE EFFECT OF 550 F IRRADIATIOli AT 1,65 x 10(E > 1 MeV)
ON THE NOTCH TOUGHNESS PROPERTIES OF THE FARLEY UNIT 1 REACTOR VESSEL MATERIALS Average Average 35 mil Averagc Average Energy Absorption 30 f t Ib Temp (* F)
Lateral Expansion Temp (* F) 50 ft Ib Temp (*F) at Full Shear (ft Ib)
(ftkb)
Material Unitradiated irradiated AT Unirradiated Irradiated AT Uni < radiated irradiated AT Unitradiated irradiated Plate 15 105 90 45 175 130 70 165 95 90 82 8
B 6919-1 (transverse)
?
Plate
-30 85 105
-10 135 145 0
120 120 140 110 30 B 6919-1 (longitudinal)
HAZ Metal
-150
-30 120
-105 45 150
-125 10 135 155 115 40 Weld Metal
-80 0
80
-50 45 95
-55 30 85 149 108 41 D
TABLE 5-10 TENSILE PROPERTIES FOR FARLEY UNIT 1 REACTOR VESSEL MATERIAL IRRADIATED TO 1.65 x 10 n/cm*
Test
.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp Strength Strength Load Stress Strength Elongation Elongation in Area No.
Material
(* F)
(ksi)
(ksi)
(kip)
(ksi)
(ksi)
(%)
(%)
(%)
AT2 Plate 150 79.0 97.2 3.58 161.5 72.9 11.3 19.3 55 B 6919-1 (transverse)
AT3 Plate 300 75.4 93.1 3.55 154.6 72.3 10.1 18.6 53 l
8 6919-1 (transverse)
AT1 Plate 550 70.7 93.7 3.60 133.9 73.3 9.2 17.3 45 g
.L B 6919-1 (transverse)
All Plate 150 80.5 98.8 3.25 172.2 66.2 11.1 23.0 62 8 6919-1 (songitudinal)
AL2 Plate 300 74.6 94.7 3.20 161.2 65.2 9.8 20.4 60 l
B 6919-1 l
(longitudinal)
AL3 Plate 550 70.3 93.1 3.37 183.3 68.7 9.3 13.5 63 8 6919-1 (longitudinal)
AW-1 Weld 71 83.5 96.8 3.07 191.2 62.5 11.2 24.3 67 Metal AW-3 Weld 200 81.5 91.7 2.90 196.7 59.1 10.2 22.8 70 Metal AW-2 Weld 550 74.8 90.7 2.95 181.1 60.1 9.5 20.7 67 Metal
~
l TEMPERATURE ('C) 100
-50 0
50 100 150 200 250 100 80 i
7 O
60 Ea 40 e
O 3
2 0
2.5 100 2.0
_ 80 2
'E
- 1.3 _
7 60 e
E a
E U
1.0
~
130*F 40 o
o,5
" 20 I
O 2
0 0
100 y
UNIRRADIATED 120 90 80 O
100 O
IRRADI ATED (550' F) 70 1.65 x 10 n/cm2 80 E
0 60 2
2 N
Q 95' F w
0 50 60 y
0 e
w 40 0
90*F e
g 4c 30 o
o,
- 0 2
20 20 o
3, t
i i
I t
I e
0
-100 0
100 200 300 400 500 TEMPER ATURE (* F)
Figure 5-1.
Irradiated Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Lower Shell Plate B 6919-1 (transverse orientation) 5-15 t
1
TEMPERATURE (* C)
-100
-50 0
50 100 150 200 5
5 5
5 5
\\
\\
^
r r
100 3
_ 80 6
$ 60
.I*
M e
O e
e y\\
N 2
0 100 2.5 g
s 80 D
2.0 2
1
\\
y 60 1.5 -
2 O
e j,g -
40 145*F g
4
\\
29 0.5 2g 2
3 0
0 160 0
140 UNIRRADIATED 160 120
=
r e
e 100 g
IRRADI ATED (550*F) 120
.6 x O' M q
> 80
=
E
~
w O
E
'U 8
80
~
120'F O
303.,
3 2
20 0
0
-100 0
100 200 300 400 TEMPERATUR E ('F)
Figure 5-2.
Irradiated Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Lower Shell Plate B 6919-1 (longitudinal orientation) 5-16 1
1
TEMPER ATURE (' C) 100
-50 0
50 100 150 200 i
e i
i i
e 100 g
7 O
O k
~
o E
8
- 40,,-
O
a.,,
[,.
~
,9 e
e ME t h'p'bgh 9
AT 14 AT 6 AT 2 AT 15 AT 3 i
l l
r m.
AT 8 AT 12 AT 1 AT 7 AT 4 Figure 5-5.
Charpy impact Specimen Fracture Surfaces for Farley Unit 1 Pressure Vessel Lower Shell Plate f
B 6919-1 (Transverse Orientation) 5-19
I i
h
- 1 s
i y
N..,,
I i
7d:S 4 ;.. -i pj + ^]U ;
- jf %?
. f ii-l ' '
{;.f* :.Qe.,
+x : u p
- J lllg,J,N,
j l
-l, lO' kjS r,
AL 3 AL6 AL 8 AL 12 AL 11 n ~-~,sr7 l
,4n i
~
. e s.
I l
\\
5.
9k$$f
^%?
A?8 4gg QQ
-Q AL S AL 4 AL 15 AL 9 AL 13 p
'U"'
Md'i s' t
. yrg
'.' J,'
e
>L AL 7 AL 10 AL 2 AL 14 AL1 Figure 5-6.
Charpy impact Specimen Fracture Surfaces for Farley Unit 1 Pressure Vessel Lower Shell Plate B 6919-1 (Longitudinal Orientation) y 5-20
4 4 g..
S
\\
. t
-4
'_r
'(
A.4
~~
~; : ~. r. ;,,1 A>-"'
.. a, AH 10 AH 14 AH7 AH9 AH 1
, y- - -
immac L,,..
w.
jg
+
s 9
y-I 4
t..
AH 6 AH 15 AH3 AH 12 AH 13
- s<
- w' W
=
l i
- i A 1}
^ '
a M%
AH 2 AH 5 AH 4 AH 8 AH 11 Figure 5-7.
Charpy impact Specimen Fracture Surfaces for
/
Farley Unit 1 Weld Heat Affected Zone Metal 5-21
~
h[fikid.
. yte..
ggj)lLc
'. J M.
- I
- . -(.gg l
. ),/ f gh
-Q[ j a
.. :,+,,:
AW 6 AW 1 AW 13 AW 4 AW 14
": m g
. f.j 7 --,
7 M, ' ]
k '
/
cif'k/
. K.
j
?
>)
~
l 0
.. w s_
AW 5 AW 8 AW 10 AW 3 q
AW 12 g
j
=
AW 7 AW 15 AW 11 AW 2 AW 9 l
l Figure 5-8.
Charpy impact Specimen Fracture Surfaces for Farley Unit 1 Weld Metal l
5-22 l
=i 500 400 C 300
.s N
S g 200 56 U
5 o#*
l E
2 100 916s h
a:
y 90 o
l 80 y
~
O m
p 60
.14% Cu Q PLATE B 6919-1 (transverse)
U E
50 z
_oiSo,p PLATE B 6919-1 (longitudinal) 40 ai 30 O we'o ueT^'
H s
N 2
o 1
l t
I e
I i iiii e
i i
i f I ii 10
2 3
4 5 6 78910" 2
3 4 5 6789 l
FLUENCE (n/cm2)
Figure 5-9.
Comparison of Actual versus Predicted 30 ft Ib Transition T5mperature Increases for the Farley Unit 1 Reactor Vessel Material Band on the Prediction Methods of Regulatory Guide 1.99 Revision 1 r
TEMPERATURE ('C) 1 0
50 100 150 200 250 300 350 120 800 700 i
100 C
1 e
y O
ULTIMATE
/
STR GTH N w
ac 500 0.2% YlELD
/
o STRENGTH s0 n
v v
400
~
40 100 200 300 400 500 600 LEGEND o unirradiated e irradiated at 1.65 x 10 n/cm8 80 O
O 8
80 e --
REDUCTION IN 2
AREA
$ 40 P
O3 O
[
TOTAL m
m 20 y ELONGAYlON v
b C
h UNIFORM 2
- ELONGATION 0
O 100 200 300 400 500 600 TEMPER ATURE (* F)
Figure 5-10.
Tensile Properties for Farley Unit 1 Reactor Vessel I
Lower Shell Plate B 6919-1 (transverse orientation) 5-24
TEMPERATURE (* C) 0 50 100 150 200 350 300 350 120 I
i i
I 3
800 700 100 A
2 ULTf MATE TENSILE 2
600 T
/
STRENGTH W
2 g 80 x
g g
500 g
0.2% YlELD m
STRENGTH 2
k
/
n 60 400 v
y 300 I
40 100 200 300 400 500 600 LEGEND O unirradiated e irradiated at 1.65 x 10" n/cm2 80 C
O 1
n i
m n
60 REDUCTION IN AREA E
$ 40 P
N 2
o 8-
/
6 TOTAL x
m
$ ELONGATION 20 2
7 W 0 UNIFORM 0
Y ELONGATION O
O 100 200 300 400 500 600 TEMPERATURE ( F)
Figure 5-11.
Tensile Properties for Fariey Unit 1 Reactor Vessel I
Lower Shell Plate B 6919-1 (longitudinal orientation) 5-25
TEMPERATURE (* C) 0 50 100 150 200 250 300 350 120 800 J
700 ULTIMATE TENSILE 100 STRENGTH k
(
3 600 m
a 2
m c.
W 80 lE a
h LD STRENGTH N
500 60 400 300 40 100 200 300 400 500 600 j
TEMPER ATURE (* F)
LEGEND o unirradiated l
e Irradiated at 1.65 x 10" n/cm2 80
/
s 60 REDUCTION IN AREA 7_
N l
i l
y 40 b
"3 l
2 f
TOTAL ELONGATION f
h 20 l
l 5
Q l
UNIFORM ELONGATION 0
O 100 200 300 400 500 600 TEMPER ATUR E (* F)
Figure 5-12.
Tensile Properties for Farley Unit 1 Reactor Vessel Weld Metal t
5-26
o e ted at 1 0*
6 7 8 9 1 2 1 2 3 4 5
0 1
10THS INCHES 5,I" a n.,',,,_,' u 4 TENSILE SPECIMEN AT 3 w :a m:me
, :]
7 f-Tested at 300*F 6 7 8 9 1 2 1 2 3 4 5
0 1
10THS INCHES
(
TENSILE SPECIMEN AT 1 i
6 7 8 9 1 2 1 2 3 4 5
0 1
10THS INCHES Figure 5-13.
Fractured Tensile Specimens of Iarley Unit 1 Reactor Vessel Lower Shell Plate B 6919-1 (Transverse Orientation) 5-27
TENSILE SPECIMEN AL 1 Tested at 150* F i 55dj5i55 i 5 0
1 10THS INCHES e ted at 3 0* F i 55dj5i55 i 2 0
1 10THS INCHES
'""*Ie!tedit sYON ^'
i 554j5i55 i 2 0
1 10THS INCHES Figure 5-14.
Fractured Tensile Specimens of Farley Unit 1 Reactor Vessel Lower Shell PI' ate B 6919-1 (Longitudinal Orientation) k 5-28 1
l Tested a 71 ggbM*MAE57~4jiAW 6 7 8 9 1 2 1 2 3 4 5
0 1
10THS INCHES TENSILE SPECIMEN AW 3 94#,, %.g %,,,.
..,w, a3mmeA*
Tested at 200*F 6 7 8 9 1 2 1 2 3 4 5
0 1
10THS INCHES h.
TENSILE SPECIMEN AW 2 Tested at 550 F T
i i
i i
i g
6 7 8 9 1 2 1 2 3 4 5
0 1
10THS INCHES Figure 5-15.
Fracture Tensile Specimens of Farley Unit 1 Reactor Vessel Weld Metal 5-29
\\
^
5 2
^
5 2
2 3-s T
n A
e N
m E
i 2
c M
t e
C p
E S
P e
S l
5 i
7 s
1 neT ro f
5 e
1 vruC p
N n
5 i
nI a
2 iA r
1
/nR t
iT S-S sse r
a 1
tS lac ip 5
y 70 T
6 1
5 0
5 e
ru g
iF 5
2 0
I o
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
5 0
5 0
5 0
5 2
0 9
7 6
4 3
1 1
1
^_b m*Nwm x
mbO
'Il llllll l
l l
l 1
l l
lI
SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1.
INTRODUCTION Knowledge of the neutron environment within the pressure vessel surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second,in relating the changes observed in the test specimens to the present and future 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 surveiilance 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 1 reactor to determine the last-neutron (E.> 1.0 Mev) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules, and,in turn, to develop lead factors for usein relating the neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averaged reaction cross secticr5 derived from this calculation, the analysis of the neutron dosimetry contA>ed.in Capsule U is discussed and comparisons with analytical predictions are presented.
6-2.
DISCRETE ORDINATES ANALYSIS A plan view of the Farley Unit 1 reactor ge metry at the core midplane is shown in o
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 p_rogram. Three capsules are located
{
symmetrically at 16.94* and 19.72 from the cardinal axes, as shown in Figure 6-1.
6-1
i 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 neutronic standpoint, the surveiilance capsule structures are significant. In fact, as will be shown later, they have a marked impact on the distributienc of neutron flux and energy spectra in the waterannulus between the neutron pad and the reactor i
vessel. Thus, in order to properiy 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 1 reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the DOTI'l two-dimensional discrete ordinates code. The radial and azimuthal distributions were obtained from an R. O computation wherein the geometry shown in Figures 6-1 and 6-2 was described in the analytical model. In adoition to the R, e computation, a second calculation in R, Z geometry was also made to obtain relative axial variations
~cf neutron flux throughout the geometry of interest. In the R, Z analysis the reactor a
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 library!51 which was developed specifically for light water reactor applications. The neutron energy group structure used in the analysis is listed in Table 6-1.
J
- A key input parameter in the analysis of the integrated fast neutron exposure of the reactor vessel is the core power distribution. For this analysis, power distributions representative of time averaged conditions derived from statistical studies of long term operation of Westinghouse 3-loop plants were employed. These input distribu-tions include rod-by-rod spatial variations for all peripheral fuel assemblies.
It sh'ould 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 6-2
l specific power distributions reflect only past ooeration, 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 approach.
Benchmark testing of these gencric power dis'tributions and the SAILOR cross-sections against surveillance capsule data obtained from 2-loop and 4-loop Westing-house plants indicates that this analytical approach yields conservative results with calculations exceeding measurements by from 10-25%.W One further point of interest regarding these analyses is that the design basis 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 factors could be changed, thus, impacting the withdrawal schedule of the remaining surveillance capsules.
Having the results of the R, O and R, Z calculations, three-dimensional variations of neutron flux may be approximated by assuming that the following relation holds for the applicable regions of the reactor.
/.
(R, Z, O, E ) = (R, O,E ) F(2. E )
g g
g where:
/
m
((R, Z, O, E ) = neutron flux at point R, Z, O within erLigy group g g
(4R, O, E ) = neutron flux at point R, O within energy group g obtained from the g
R, O calculation F(Z, E ) = relative axial distribution of neutron flux within energy group g
g obtained from the R, Z calculation 6-3 L
1-3.
NEUTRON DOSIMETRY The passive neutron flux monitors included in Capsule U of Farley Unit 1 are listed in Table 6-2. The first five reactions in Table 6-2 are used as fast-neutron monitors to elate neutron fluence (E > 1.0 Mev) to measured material property changes. Base and cadmium covered cobalt-aluminum monitors were also included in the table to determine the magnitude of the thermal neutron flux at the monitorlocation,which is
-)ecessary to account for burnout of the product isotope generated by fast-neutron
'eactions.
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 wir 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 cf the energy-dependent flux level at the point of interest. Rather, the activa-tion or fission process is a measure of the bitegrated offect 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 iie activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:
s The operating history of the reactor a
Tha 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 spectrum at the monitor location must be calculated.
6-4
l f
The specific activity of each of the monitors is determinod using established ASTM procedures.i"""I Following sample preparation, the actisity of each moriitor is determined by means of alithium drifted germanium (Ge(Li)) gamma spectrometer.
The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed from Farley Unit 1, the overall 2 a deviation in the measured data is 10 parcsnt. The neutron energy spectra are determined analytically using the method described in Subsection 6-1.
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 expressad as N
t a(E)&(E){ P d
6-1 fY I1, '
I' R=
i
- E max p,
where:
R = induced product activity No = Avogadro's number A = atomic weight of ?% target lsotope 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 powe Pj = average core power level during irradiation period j 6-5
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 multigroup transport mett. ids 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:
F&(E> 1.0 Mev) a(E) d(E)dE
=
E where:
N 9 #9 a(E) $ (E)dE O
Gat
{.
t= =
N 9
d 10 Mew G-G 10 Mew Thus, eauation 6-1 is rewritten N
PI d
(6-2)
Rr f j y & c (E > 1.0 Mev)
(1-e-Atj)e-At A
P max js1 6-6
f or, solving for the neutron flux, R
4 (E > 1.0 Mev) =
N N
P
-At
-At d f; y &
(1-e
)e (6-2) je t The total fluence above 1.0 Mev is then given by N
p, I
4 (E > 1.0 Mev) = c (E > 1.0 Mev){
max tj (6-3)
- 3.,
where:
N tj = total effective full power seconds of reactor operation P max up to the time of capsule removal An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered Co" (n,y)Co" data by means of cadmium ratios and the use of a 37-barn 2200 m/sec cross section. Thus, fD-1 i
Rbare hD 4Th =
(6-4)
N No Pj
-A t ) e-A t j
d f ycr (1.e i
P A
max j=1 where:
D is defined as l
rcd covered 6-7
1 6
64.
TRANSPORT ANALYSIS RESULTS
' Results of the S n transport calculations for the Farley Unit 1 reactor are summarized 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 surveiilance 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-neutror, flux (E > 1.0 Mev) through the thickness of the reactor pressure vesselis 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 Figures 6-3 or 6-4 by the appropriate values from Figure S-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 ir.ner radius. Updated lead factors for all of the Farley Unit 1 surveillance capsules are listed in Table 6-3.
Radial variations of analytically determined reaction rate grsdients for each of the fast-neutron monitors are shown in Figures 6-7 and 6-8 for capsules at 16.94* and 19.72*, respectively.
In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. Calculations of the neutron energy spectrum existing at the center of each of the Farley Unit 1 surveillance capsules are listed in Tablo 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 1 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 Cepsule U. based on the irradiation history shown in Table 6-6, are given in Table 6-7.
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 1
6-8
l monitors is summarized in Table 6-9. Due to the relatively low 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 estimated to be less than 1 percent for the Ni"(n.p) Co" reaction and even less for all of the other fast-neutron reactions.
An examination of Table 6-8 shows that the fast-neutron flux (E > 1.0 Mev) derived from the five threshold reactions ranges from 1.71 x 10" to 2.01 x 10" n/cm'-sec, a total span of less than 18%. It may also be noted that the calculated flux value of 1.98 x 10" n/cm'-sec exceeds all but one of the measured values with calculation to experimental ratios ranging from 0.985 to 1.16. This behavior is consistent with prior bencnmarking 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 estimatc exposure of Capsule U is:
4T = 1.65 x 10 n/cm (E > 1.0 Mev) h 6-9
I 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-'
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-*
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-*
19 0.821 43 1.86 x 10-*
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 6-10
TABLE 6-2 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Wt %
of Target 8 * '*
- Product Fission Monitor Material Reaction of Interest gmmonitor Response Range Half-Ufe Yield (%)
Copper Cu (n,a) Co" 0.6917 E > 4.7 Mev 5.27 years Iron Fe" (np) Mn" 0.0585 E > 1.0 Mev 314 days S8 Nickel Ni'8 (n.p) Co 0.6777 E > 1.0 Mev 71.4 days 2 alal Uranium U (n.f) Cs
1.0 E > 0.4 Mev 20.2 years 6.3 Neptunium' '
Np # (n,f) Cs
1.0 E > 0.08 Mev 30.2 years 6.5 Cobalt-aluminum *3 Co" (n,y) Co" 0.0015 0.4 ev < E < 0.015 Mev 5.27 years l
Cobalt-aluminum Co** (n,y) Co" 0.0015 E < 0.015 Mev 5.27 years
[a] Denotes that monitor is cadmium shie!ded.
- d TABLE 6-3 CALCULATED FAST-NEUU!ON FLUX (E > 1.0 Mov)
AND LEAD FACTORS FOR FARLEY UNIT 1 SURVEILLANCE CAPSULES Capsule Azimuthal (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 Y
16.94*
1.98 x 10" 3.12 W
19.72 1.73 x 10" 2.70 V
19.72 1.73 x 10" 2.70 l
Z 19.72 1.73 x 10" 2.70 1
P 6-12
TABLE 6-4 l
k CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF FARLEY UNIT 1 SURVEILLANCE CAPSULES Neutron Flux (n/cm*-sec)
Group No.
Capsules U, X, Y Capsules W, V, 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 10e 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 10' 10 5.88 x 10' 5.40 x 10' 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 e 5.88 x 10' i
29 1.96 x 10' 1.56 x 10' 30 1.05 x 10' 8.31 x 10' N
6-13
TABLE 6-4 CONTINUED CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF FARLEY UNIT 1 SURVEILLANCE CAPSULES Neutron Flux (n/cm -sec)
Group No.
Capsules U, X, Y Capsules W, V, 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' 1.94 x 10' 44 2.43 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' t
6-14
TABLE 6-5 SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF FARLEY UNIT 1 SURVEILLANCE CAPSULES a (bams)
Reaction Capsules U, X, Y Capsules W, V, Z Fe'd (n.p) Mn
O.0517 0.0548 se NiS8 (n p) Co 0.0741 0.0767 Cu' (n,a) Co' O.000429 0.000469 Np* (n,f) FP 3.42 3.33 U '8 (n,f) FP 0.301 0.306 2
=
a(E)&(E)dE
-g=
o "c(E)dE 1.0 Mev i
l t
l 6-15
TABLE 6-6 IRRADIATION HISTORY OF FARLEY UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE U Irradiation Decay PJ P max Pj/
Time Time Month - Year (MY!)
(MW)
P mar.
(day)
(day) 8 1977 7
2652'
.027 23 2263 9
1977 1075.
2652
.405 30 2233 10 1977 416.
2652
.157 31 2202 11 1977 1556.
2652
.587 30 2172 12 1977 1745.
2652
.658 31 2141 1
1978 1805.
2652
.681 31 2110 2
1978 2094.
2652
.789 28 2082 3
1978 2528.
2652
.953 31 2051 4
1978 2288.
2652
.863 30 2021 5
1978
-2203.
2652
.831 31 1990 6
1978 2442.
2652
.921 30 1960 7
1978 2388.
2652
.900 31 1929 8
1978 2356.
2652
.888 31 1898 9
1978 1275.
2652
.481 30 1868 10 1978 2002.
2652
.755 31 1837 11 1978 2613.
2652
.985 30 1807 12 1978 2374.
2652
.895 31 1776 1
1979 2066.
2652
.779 31 1745 2
'1979 2369.
2652
.893 2R 1717 3
1979 504.
2652
.190 31 1666 4
1979 0.
2652 0.000 30 1656 5
1979 0.
2652.
0.000 31 1625 6
1979 0.
2652 0.000 30 1595 7
1979 0.
2652 0.000 31 1564 8
1979 0.
2652 0.000 31 1533 9
1979 0.
2652 0.000 30 1503 10 1979 0.
2652 0.000 31 1472 11 1979 597.
2652
.225 30 1442 12 1979 2370.
2652
.894 31 1411 i
1 6-16
TABLE 6-6 CONTINUED IRRADIATION HISTORY OF FARLEY UNIT 1 REACTCR VESSEL SURVEILLANCE CAPSULE U Irradiation Decay PJ P max Pj/
Time Time Month - Year (MW)
(MW)
P max (day)
(day) 1 1980 2308.
2652
.870 31 1380 2
1980 910.
2652
.343 29 1351 3
1980 2411.
2652
.909 31 1320 4
1980 2576.
2652
.971 30 1290 5
1980 2545.
2652
.960 31 1259 6
1980 1115.
2652
.421 30 1229 7
1980 1548.
2652
.584 31 1198 8
1980 2178.
2652
.821 31 1167 9
1980 1842.
2652
.695 30 1137 10 1980 2563.
2652
.966 31 1106 11 1980 525.
2652
.198 30 1076 12 1980 0.
2652 0.000 31 1045 1
1981 0.
2652 0.000 31 1014 2
1981 0.
2652 0.000 28 986 3
1981 0.
2652 0.000 31 955 4
1981 1544.
2652
.582 30 925 5
1981 2162.
2652
.815 31 894 6
1981 2478.
2652
.934 30 864 7
1981 2444.
2652
.922 31 833 8
1981 2461.
2652
.928 31 802 9
1981 727.
2652
.274 30 772 10 1981 0.
2652 0.000 31 741 11 1981 0.
2652 0.000 30 711 12 1981 0.
2652 0.000 31 680 1
1982 0.
2652 0.000 31 649 2
1982 0.
2652 0.000 28 621 3
1982 1634.
2652
.616 31 590 4
1982 2051.
2652
,773 30 560 5
1982 2558.
2652
.964 31 529 6-17
TABLE 6-6 CONTINUED IRRADIATION HISTORY OF FARLEY UNIT 1 REACTOR VESSEL SURVC!LLANCE CAPSULE U Irradiation Decay PJ P max Pj/
Time Time Month - Year (MW)
(MW)
P max (day)
(day) 6 1982 2509.
2652
.946 30 499 7
1982 2561.
2652
.966 31 468 8
1982 2174.
2652
.820 31 437 9
1982 2448.
2652
.923 30 407 10 1982 2234.
2652
.842 31 376 11 1982 2619.
2652
.988 30 346 12 1982 2443.
2652
.921 31 315 1
1983 2319.
2652
.874 14 301 EFPS = 9.51E + 07 SEC l
EFPY = 3.02 l Decay time is referenced to 11-11-83.
l 6-18
TABLE 6-7 l
COMPARISON OF MEASURED AND CALCULATED FAST-NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE U
^
"Y Reactior.
Radial Measured E' U*}
and Location Activity Axial Location (cm)
(dps/gm)
Capsule Calculated Fe * (n p) Mn '
5 5
Top 186.21 1.88E + 06 5.78E + 06 Middle 186.21 1.86E + 06 5.72E + 06 Bottom 186.21 1.94E + 06 5.97E + 06 8
8 5.82 x 10 6.60 x 10 Average ss Nisa (n.p) Co Top 186.21 4.16E + 06 8.87E + 07 Middle 186.21 4.03E + 06 8.59E + 07 Bottom 186.21 4.33E + 06 9.23E + 07 7
8 8.90 x 10 1.01 x 10 Average Cu' (N,a) Co*
Top 186.21 1.35E + 05 5.25E + 05 Middle 186.21 1.34E + 05 5.22E + 05 Bottom 186.21 1.42E + 05 5.53E + 05 5
5.33E + 05 5.64 x 10 Average
^
Np 7 (n,f) Cs' 7 8
Middle 186.21 6.05E + 06 9.47E + 07 1.10 x 10 U a (n,f) Cs' '
8 Middle 186.21 7.00E + 05 1.10E + 07 9.12 x 10 6-19
'~
~
1 l
1ABLE 6-8 RESULTS OF FAST-NEUTRON DOSIMETRY FOR CAPSULE U Saturated Activity
$ (E > 1.0 Mev)
$ (E > 1.0 Mev)
(dps/gm)
(n/cm*-sec)
(n/cm')
Reaction Measured Calculated Measured Calculated Measured Calculated Fe64 (n.p) Mn 5.82 x 10 6.60 x 10 1.73 x 10" 1.98 x 10" 1.65 x 10
1.88 x 10
54 8
8 Nisa (n.p) Co 8.90 x 10 1.01 x 10s 1.71 x 10" 1.98 x 10" 1.63 x 10
1.88 x 10
sa 7
Cu (n,a) Co*
5.33 x 10 5.64 x 10 1.88 x 10" 1.98 x 10" 1.79 x 10
1.88 x 10
5 5
Np 7 (n,f) Cs' '
9.47 x 10 i.08 x 10
1.68 x 10" 1.98 x 10" 1.60 x 10
1.88 x 10
7 U238 (n,f) Cs' '
9.68 x 10' 9.12 x 10' 2.01 x 10" 1.98 x 10" 1.91 x 10
1.88 x 10
- U
- adjusted saturated activity has been multiplied by 0.88 to correct for 350 ppm U'" impurity.
TABLE 6-9 RESULTS OF THERMAL-NEUTRON DOSIMETRY FOR CAPSULE U Saturated Activity (dps/gm)
Axial th Location Bare Cd - Covered (n/cm'-sec)
Top 1.39E + 08 Middle 1.39E + 08 Bottom 1.40E + 08 7.47E + 07 1.15 x 10"
- Wires were not found in Cd cover.
6-21
.1 TABLE 6-10
SUMMARY
OF NEUTRON DOSIMETRY RESULTS FOR CAPSULE U 46 Current & (E > 1.0 Mev)
EOL & (E > 1.0 Mev)
(n/cm )
(n/cm')
Location.
Measured Calculated Measured Calculated Capsule U 1.65 x 10
1.88 x 10
VesselIR 5.29 x 10
6.03 x 10
5.62 x 10
6.40 x 10
Vesse! 1/4 T 2.81 x 10
3.20 x 10
2.98 x 10
3.40 x 10
Vessel 3/4 T 3.27 x 10'7 7.02 x 10" 3.47 x 10
7.46 x 10_
l; NOTE: EOL fluences are based on operation at 2652 MWt for 32 effective-full-power years.
l r
e a
6-22
l T
.9A* CAPSULES U, X, Y 16
/
C' j
19.72* CAPSULES W. V Z l
45' lll//
A\\\\\\\\)
1, r1 1 1 REACTOR VESSEL f
/
1,,.,
//
NEUTRON PAD
/
/
REACTOR d
bCORE BARREL I
f CORE
/
l/
d l Figure 6-1.
Fariey Unit 1 Reactor Geometry 6-23
l l
gCHARPY SPECIMEN r e ' ' ' 'y r
r
(
/ l
///M
//
NN N N N % N N N N N N N N N N N K NEUTRON PAD g
N N s s N N N N NN N \\N N \\\\ \\ \\
Figure 6-2.
Plan View of a Reactor Vessel Surveillance Capsule 6-24
l 10
8 7
6 S
4 3
2 7j10" j
8 SURVEILLANCE 5
CAPSULES 6
5 5
d 4
z 3
5 PRESSURE w
2
-VESSELIR z
m 1/4T LOCATlON 8
7 6
5 4
3
-3'4T LOCATION 2
I I
I I
I i
10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEGREES)
Figure 6-3.
Calculated Azimuthal Distribution of Maximum Fast-Neutron Flex (E> 1.0 Mev) within the Pressure Vessel Surveillance Capsule Geometry 6-25
199 39 I
IR 204 39 I
I 1/4T 214 39 l
I 3 /4T 219 39 s
\\
OR I
I i
t t
I t
i I
f i
1 I
4 196 198 200 202 204 206 208 210 212 214 216 218 220 222 R ADIUS (cm)
Figure 6-4.
Calculated Radial Distribution of Maximum Fast-Neutron Flux (E > 1.0 Mev) within the Pressure Vessel 6-26
i 10" a
9 6
4 2
5 d 10" z
O 8
m
$6
~
w 4
E 4
~
de 2
10
8 6
5 4
5 E
8 2
TO VESSEL CLOSURE HEAD 10
-300
-200
-100 0
100 200 300 DISTANCE FROf.i CORE MIDPLANE (CM)
Figure 6-5.
Relative Axial Variation of Fast-Neutron Flux (E > 1.0 Mev) within the Pressure Vessel 6 _. _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _
2 10'8 9
l 8
l 7
l l
6 5
7 4
'y CAPSULES U,X,Y E
g 3
CAPSULES 8
s5 w z,v M
3 2
a z
O
=
E c
H wz 3"
b9 z 10 gQ 9
mo 8
@g 7
6 5
4 3
2 I
10
182 183 184 185 186 187 188 189 R ADIUS (cm)
Figure 6-6.
Calculated Radial Distribution of Fast-Neutron Flux (E > 1.0 Mev) within the Reactor Vessel Surveillance Capsules 6-28
3 2
Np* (n.h Cs'8' E
Ni" (n p) Co
I N5 S
5 yp Gb 4
00 C"
3 2
T U (nf) Cs'
8
- onk, Fe"(nf) Mn" a 10 E: :
N 4
O N
~
3 E
@ 2 en l
10' Cu (n.a) Co*
1 :
7 6
2
-m 3
2 10' 182 183 194 185 186 187 188 189 R ADIUS (cm)
Figure 6-7.
Calculated Variation of Fast-Neutron-Flux Monitor Saturated
. Activity within Capsules U, X, and Y
,_6-29 s
4 3
E 2
Ni" (n.p) Co"
~
Np" (n.f) Cs'8-10' :
9 8
7 6
5 E2 4
t9 7
{3 EQ eo
.u OO CJ
{2 b
~
h U"' (n,f) Cs
Q' Fe'* (n.p) Mn'*
O c 9 w 8 E7 g6 Q5 :
m 4
=
3 2
10'
=
l Cu (n.a) Co" 7
6 4
3 g
~
2 10' s
I e,
e t
I e
e i
e I
i 182 183 184 185 186 187 188 189 R ADIUS (cm)
Figure 6-8.
Calculated Variation of Fast-Neutron-Flux Monitor Saturated Activity within Capsules W, V, and Z 6-30
)
SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule is recommended for future capsules to be removed from the Farley Unit 1 reactor vessel:
Lead Removal Estimated Fluence Capsule Factor
.. Time '1 n/cm'. x 109 l
Y 3.12 Removed (1.13)
.583' U
3.12 Removeo (3.02) 1.65 X
3.12 6
3.05 tb1 W
2.70 12 5.28 Icl V'
2.70 J 21 9.28 Z
2.70 Standby
. [a] Effective full power years from plent startup
[b] Approximates vessel end of life 14 ihickness walllocahn fluence
[c] Approximates vessel end of life itiner wall Ic0ation fluence
~
,e T
[
e
/
r
< y-j
~
5
/
a.t
/
1 4
i8 '
g
SECTION 8 REFERENCES
- 1. Davidson, J. A., Yanichko, S.E., " Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program,"WCAP 8810, December,1976.
- 2. ASTM Standard 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 Commission, April 1977.
j
- 4. Soltesz, R. G., Disney, R. K., Jedruch, 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 Collectiun "DLC-76," Coupled, Self-shielded, 47 Neutron,20 Gamma-ray, P3, Cross Section Library for Light Water Reactors."
- 6. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology
- to be published.
'7. ASTM Designation E261-77, Standard Practic.e for Measuring Neutron Flux, Fluence, and Spectra o' y Radioactivation Techniques," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 916 926, American Society for Testing and Matenais, Philadelphia, Pa.,1981.
8-1
- 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 Star.dards (1981), Part 45, Nuclear Standards, pp. 936-941 American Society for Testing and Materials, Phila-delphia, Pa.,1981.
- 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.
- 11. ASTM Designation E264-77," Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel," in AGTM Standards (1981), Part 45, Nuclear Standards, pp. 942-945, American Society for Testing and Materi91s, Phila-delphia, Pa.,1981.
1
)
8-2
~
APPENDIX A FARLEY UNIT 1 R
HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1.
INTRODUCTION Heatup and cooldown limit curves are calculated usirg the most limiting value of RTNDT (reference nil-ductility 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-NDT s designated as the higher of either the drop weight nil-ductility transition i
RT temperature (NDTT) or the temperature at which the material exhibits at least 50 f t Ib k
of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60* F.
NDT ncreases as the material is exposed to fast-neutron radiation. Thus, to find i
RT the most limiting RT NDT at any time period in the reactor's life, ART NDT due to the radiation exposure associated with that time period must be added to the original un-irradiated RT NDT. The extentof the shif t in RT NOT is enhanced by certain chemical elements (such as copper) present in reactor vessel steels. Design curves which show the effect of fluence and copper content on ARTNDT for reactor vessel steels exposed to 550'F are shown in Figure A-1.
Given the copper content of the most limiting material, the radiation-induced ART NDT can be estimated from Figure A-1. Fast-neutron fluence (E >1 Mev) at the 1/4T (wall thickness) and 3/4T (wall thickness) vessel locations are given as a func-tion of full-power service life in Figure A-2. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other com-ponent will be limiting with respect to RTNDT.
A-1 l
l
I A-2.
FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture-toughness properties of the Farley Unit 1 reactor vessel materials are presented in Table A-1. 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.M The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the Farley Unit 1 Vessel Material Surveillance Program.
A-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, K i, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K IR, for the meal temperature at that time. KIR s obtained from the reference fracture toughness curve, defined in i
Appendix G to the ASME Code.I21 The K R curve is given by the equation:
I K IR = 26.78 + 1.223 exp [0.0145 (T-RT NDT + 160)]
(A-1)
[
where KIR is the reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G to the ASME Coder 21 as follows:
CK IM+K it 5K IR (A-2)
- 1.
- Fracture Toughness Requirements." Branca Technical Position iATEB No. 5-2. Section 5 3.2-14 in Standard Rewew Plan NUREG-75/087.1975
- 2. ASME Boder and Pressure Vessef Code. S6stion lit. Division 1 - Appendices. " Rules for Construction of Nuclear Vessels." Appendix G. " Protection Against Nonductile Failure." pp. 559-569.1983 Edition, Amencan Society of Mechanical Engineers. New York,1983.
A-2
+
L_i______________._
where:
K IM s the stress intensity factor caused by membrane (pressure) stress i
K lt is the stress intensity factor caused by the thermal gradients IR s a function of temperature to the RTNDT of the material i
K C = 2.0 for Level A and Leve! B service limits
~
C=
1.5 for hydrostatic and leak test conditions during which the reactor core is not critical l
At any time during the heatup or cooldown transient, K IR s determined by the metal i
temperature at the tip of the postulated flaw, the appropriate value for RT NDT, and the reference fracture toughness curve. The thermal stresses resulting from j
temperature gradients through the vessel wall are calculated and then the
(
corresponding (thermal) stress intensity factors, K lt, 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-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall.
During cooldown,the controlling location of theflawis alwaysattheinsideof thewall because the thermal gradients produce tensile stresses at the inside, which increase l
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 cooldown analysis is necessary because control of the cooldown 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 1/4Tvessellocation 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 highervalue of K IR at the 1/4T location for finite cooldown rates than for steady-state operation.
A-3
Furthermore, if conditions exist such that the increase in K IR exceeds K it, the calculated allowable pressure during cooldown will be greater than the steady-state a
value.
The above procedures are needed because there is no direct centrol on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased atvarious intervals along a cooldown ramp.The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldcwn 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 I
relationships are developed for steady-state conditions as well as finite heatup rate conditions assurning the presence of a 1/4T defectattheinsideof thevesselwall.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 IR for the 1/4T crack during heatup is lower than the K IR orthe 1/4T crack during steady-state f
conditions at the same coolant temperature. During hehtup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower KIR'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 1/4T 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-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel's inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any oressure stresses present. These thermal stresses are dependent on both the 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.
A-4
)
l
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 en 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 composito curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition awitches 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 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 snd cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in paragraph A-3.The derivation of the limit curves is presented in the NRC Regulatory Standard Review Plan.DI Transition temperature shifts occurring in the pressure vessel materials due to
(
radiation exposure have been obtained directly from the reactor pressure vessei surveillance program.
Charpy test specimens from Capsule U indicate that the representative core region we!d metal and the limiting core region plate (B 6919-1) exhibited shifts in RTNDT0f 80 and 90*F, respectively. These shifts at a fluence of 1.65 x 10" n/cm2 are well within the appropriato design curve (Figure A-1) predction. Heatup and cooldown calculations were based on the ARTNDT predicted for the core region weld material located in the intermediate to lower shell weld seam. This materialis limiting because l
it occurs at the peak fluence. The intermediate shell and lower shell longitudinal weld seams are located at an azimuthal angle of 45* from the peak fluence. Figure 6-3
(
shows that the fluence at these longitudinal welds is significantly lower than the peak fluence. Heatup and cooldown limit curves for normal op ? ration of the reactor vessel are presented in Figures A-3 and A-4 and represent an operational time period of 7
. effective-full-power years.
- 1.
- Pressure-Temperature Limits " Section 5.3 2 in Standard Review Plan, NUREG-75/087.1975.
A-5
}
l 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 a requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of references.[1.21 Figures A-3 and A-4 define limits for insuring prevention of nonductile failure.
l
- 1. " Pressure-Temperature Limits." Section 5.3.2 in Standard Revrew Plan, NUREG 75M7.1975.
- 2. ASME So#ler and Pressure Vesset Code, Section Ill. Division 1 - Appendices, S for Construction of Nuclear Vessels," Appendix G. " Protection Against Nonductile Failure,"pp.559-569,1983 Edition, American Society of Meenanical Engineers, New York,1983.
A-6
I TABLE A-1 i
REACTOR VESSEL TOUGHNESS DATA l
Material Cu P
TNDT TMWD TNMWD RTNDT Upper Shelf Energy Comanent Code No.
Type
(%)
(%)
( F)
(* F)
(* F)
(* F)
MWD NMWD l
Closure head dome B6901 A5338, Cl.1 0.16 0.009
-30 20 40!*1
-20 140 Closure nead segment B6902-1 A5338, Cl.1 0.17 0.007
-20
-10 10l*1
-20 138 O *l 60 75'*3 l
Closure head fla, ige B6915-1 A508, Cl. 2 0.10 0.012 60 *l
-20 l
tal l
Vessel flanae B6913-1 A508 Cl. 2 0.17 0.011 60 *l
-30
-10l*1 60 106 inlet nozzle B6917-1 A508, Cl. 2 0.010 60 *I 45 60 110 Inlet nozzle B6917-2 A508, Cl. 2 0.008 60 *l 115 60 80 0.308 60 *l 35 60 98 l
Inlet nozz!e B6917-3 A508, Cl. 2 l
96.5 i
Outlet nozzle B6916-1 A508, Cl. 2 0.007 60 *l 60 60 l
I
[
Outlet nozzle B6916-2 A508, Cl. 2 0.011 60 *3 30 60 97.5 l
100 Outlet nozzle B6916-3 A508, Cl. 2 0.009 60 *1 50 60 Nozzle shell B6914-1 A508, Cl. 2 0.010 30 70 90 *l 30 148 l
inter. shell B6903-2 A533B, Cl.1 0.13 0.011 0
-25 40 0
151.5 97 I
inter. shell B6903-3 A533B, Cl.1 0.12 0.014 10 5
52 10 134.5 100 Lower shell B6919-1 A533B, Cl.1 0.14 0.015
-20
-5 75 15 133 90.5 Lower shell B6919-2 A533B, Cl.1 0.14 0.015
-10 0
65 5
134 97 Bottom head ring B6912-1 A508, Cl. 2 0.010 10
-25
-S *3 10 163.5 l
Bottom head segment B6906-1 A533B, Cl.1 0.15 0.011
-30
-50
-30"3
-30 147 Bottom head dome B6907-1 A533B, Cl.1 0.17 0.014
-30
-10 10 *3
-30 143.5 l
O*3
<60 0
l 0.27 0.015 Inter. shell long.
l weld seam Oal
<60 0
l 0.24 0.011 Inter, to lower shell weld seams O*l
<60 0
l l
0.17 0.022 Lower shell long.
l weld seams MWD - Major Working Direction
[a] Estimated per NRC Regulatory Standard Review Plan, Section 5.3.2, NMWD - Normal to Major Working Direction TMWD - 50 f t Ib energy temperature in Major Working Direction TNMWD - 50 f t tb energy temperature Normal to Major Working Direction
NOT = [40 + 1000 (% cu -0.08) + 5000 (% P -0.008)] [f/10"] '/'
}400 UPPER g 300 l
r 2M 3
0.35 t
s*0 E
0.30 l
I '" /
e 0.25 z
0.20% Cu l-3
[
@ 50 0.15% Cu 0.10% Cu LOWER LIMIT
%P = 0.012
% Cu = 0.08
)
o
% P = 0.008 N9 8
E I
I I
I I
t 1
i 10
2 4
6 8
10" 2
4 6
FLUENCE (N/CM8 E > 1 MeV)
AWELD METAL GSHELL PLATE B E919-1 l
l l
l 4
NDT or Reactor f
l Figure A-1.
Effect of Fluence and Copper Content on ART Vessel Steels Exposed to Irradiation at 550 F 1
A-8
m I;
S 1
)YPF 2
E T
T i
3
(
4 4
/
e
/
3 1
f iL 0
3 ec iv r
82 e
S re 6
w 2
o P
l 4
lu 2
F fo 22 n
o itc 0
n 2
u F
a 8
s 1
a
)
Y V
P e
6 F 1 E M
1 4
1 E
(
e 2
c i
1 neu l
0 F
e 1
no r
t 8
u i
eN ts 6
a e
F 4
i 2
-A e
2 e
rug I : _ _ -
iF 0
0 0
0' o,
1 1
1 eE$;ozmad z@>an a
r
>6 I
- . T * & Ilk):: ' f. '.fV ? ' ' ' ? J -:&. %x 9 _g >L; W *
. V
$%.. ).; ; \\ '.'..;
_ ;, p.
,. ; W.;.l _.... f l' L.6
%. b " :: ' ' _ L '. r..WQ,;.
- :$.N. < N '*9 c
' n;! C
- :: '.l:.Q.i.s $.I g
=
.a
?.%,.,.
y
- %?:.
- y'.;.
G:p :
hf'.}'
- pL IA
(.,.p
.J.;-}S'
- i. :r 3000 g
I l
'W.
. $y i
~.
"$)Y
.f MATERIAL PROPERTY BASIS:
CONTROLLING MATERIAL: WELD METAL i
$l
.~
"~
J.N.i d.'
COPPER CONTENT: 0.24 WT%
g.f
~
PHOSPHORUS CONTENT: 0.011 WT%
U: -?
.r RT NDT INITIAL: 0* F LEAK TEST LIMIT
^
Mais
- M
\\ /
f f
.?-
3/4 T, 87* F
\\ [
j j
h}j G ' -l 1/4 T 185*F
" ~
((ap l
f f
- h. :k CURVE APPLICABLE FOR HEATUP
- A
?. b "
RATES UP TO 60*F/HR FOR THE
.. !. S'-
SERVICE PERIOD UP TO
[
}
}
M/
7 EFPY AND CONTAINS I
f W
4 7. '-
MARGINS OF 10*F AND 60 l
M h,ig5' 3
2000 PSIG FOR POSSIBLE
~~
f j
1:
INSTRUMENT ERRORS.
7 f
_j; 9
% f(sn 1-i f
y l
f
}
g.
?.G ~'. )
F'- 9
~
. (.h:.:
E A'
(
e v
- )
9 'p m
f
.f
-f
- .,- l.
gn l
)
w f.: ~_i; f
[?
HEATUP RATES
[
/
J.',,
- k.ji#h c
UP TO 60* F/HR' y r
h...
i,.j.
!l2 i
i
- v.
4a-
&g&
- e :c u o
)
.l,Wl:- !.h 5
7 z
i :?.. n j
t.
- ?-.,y 1000
/
/
CRITICALITY LIMIT i
E " <. :9 (p'.lrg
/
BASED ON INSERVICE
- HYDROSTATIC TEST p
/
., [*....
+
V TEMPERATURE (325' F) i.$fl.
/
FOR THE SERVICE
...,,, p '
/
i I' 4l. l
'Yf PEjRIOD UP TO 7 EFPY
. ?,
kM l Wm
. r.:.? -
.=.
..:,5 %-
.T q.,
. ; ' v.;:
s 4
.c
- >,/,.p y
,1
..p
..' b
_j...y.~ t
~. ',%
.!'i
....p
- f. e.: 3 0
100 200 300 400 500 S.*
- .W 0
v.s S
INDICATED TEMPERATURE (*F)
- .v
' j.4 -
.. :: ?
.x
..{ ..(<f
- a. c,g
- f.yn
< 4,
((
r ?;N
- b Figure A-3.
Fariey Unit No.1 Reactor Coolant System I[S -
Heatup Limitations Applicable for the
[&
M Y.
First 7 EFPY
- f..f
.b s
..,r,
< :y 7;;
..Q,.g,y1 A-10
- 7..
9.
?:$.3.i ?l
..q sc
- j,>.g
,u
.c r
,./.m
...*'y
, as
_i. ' -
.r' d.,..,
..g.-
e.
..m..,, p,....Q, ;...,,2.,. ;. '..
_,.;.. c,,
_3 ;
4
,S
~
3000 l
1 e
)
- MATECIAL PROPERTY BASIS:
CONiROLLING MATERIAL:
WELD METAL l
COPPER CONTENT: 0.24 WT%
PHOSPHORUS CONTENT: 0.011 WT%
f
~~~
RTNOT INITI AL: 0* F j
RTNDT AMER 7 EFPY:
1/4 T,185* F 3/4 T,87* F j
g 2000 CURVE APPLICABLE FOR
/
CCOLDOWN RATES UP TO r
100*F/HR FOR THE SERVICE J
'~
~~
PERIOD UP TO 7 EFPY AND f
y
[
CONTAINS MARGINS OF a:
10*F AND 60 PSIG FOR J
~~
POSSIBLE INSTRUMENT
/
ERRORS.
y
/
g
?a
=
/
9
)
o iE J
r 1000 A
V
.\\
A P)
COOLDOWN RATES *F/HR M
O scG5 V/
L 20 -
E5d@35#1/
~~
40 e
,,,e-
"""~
60 "
~
gg.
0 0
100 200 300 400 500 INDICATED TEMPERATURE (* F)
Figure A-4.
Farley Unit No.1 Reactor Coolant System Cooldown Limitations Applicable for the First 7 EFPY A-11 J