ML20154S181
| ML20154S181 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 01/31/1988 |
| From: | Albertin L, Colburn D, Lamantia L WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20154S180 | List: |
| References | |
| WCAP-11726, NUDOCS 8810050002 | |
| Download: ML20154S181 (94) | |
Text
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WESTING 50USE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION WCAP-11726 I
l ANALYSIS OF CAPSULE Y FROM TB SOUTE CAROLINA ELECTRIC AND CAS COMPAhT VIRGIL C. St%(RR UNIT 1 REACTOR YESSEL RADIATION SURYRILLANCE PROGRAM i
i D. J. Colburn l t
L. A. Lamantia L. Albertin t
January 1988 APPROYED: N- ~
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Y. A. Meyer, Manager I
Structural Materials and Reliability technology Work performed under Shop Order No. CFZJ-106 Prepared by Westinghouse Electric Corporation for the 1 South Carolina Electric and Cas Company Although information contained in this report is nonproprietary no distribution shall be ande outside Westinghouse or its licensees without the customer's approval WESTINCHOUSE ELECTRIC CORPORATION Nuclear Energy Systems e
P. O. Box 2728 Pittsburgh, Pennsylvania 15230
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1 PREFACE
. This report has been technically reviewed and verified. !
Reviewer l Sections 1 thro gh 5 and 7 S. 5. Yanichko [ .
b s.ction 6 5. L. Anderson M dmA sL A ,
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t CONTENTS s i
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- 1. SUWHARY OF RESULTS ........................................... 1-1
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- 2. INTRODUCTION ................................................. 2-1 l
- 3. EACKGR0UND.................................................. . 3-1
- 4. DESCRIPTION OF PR0GRAW........................................ 4-1
- 5. TESTING 0F SPECIMENS FROW CAPSULE V........................... 5-1 5.1 Overview ............................................... 5-1 5.2 Charpy V-Notch Impact Test Results ..................... 5-3 :
i 5.3 Tension Test Results ................................... 5-4 ,
5.4 Compact Tension Tests ...................~............... 5-5
- 6. RADIATION ANALYSIS AND NEUTRON DOSIMITRY ..................... 6-1 6.1 Introduction ........................................... 6-1 -
6.2 Discrete Ordinates Analysis ............................ 6-2 [
t 6.3 Radiometric Monitors ................................... 6-6
- i 6.4 Neutron Transport Analysis Results ..................... 6-11
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6.5 Influence of an Energy Dependent Damage Model........... 6-13 i i
6.6 Neutron Dosimetry Results .............................. 6-14
- 7. SURVEILLANCE CAPSULE REMOYAL SCIEDULE ........................ 7-1
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- 8. REFERENCES ................................................... 8-1 j 1
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LIST OF ILLUSTRATIONS l UAEt f. alt 4-1 Arrangement of surveillance capsules in the V. C. Summer Unit I reactor vesse1................................... 4-5 4-2 Capsule V diagram showing location of specimens, thermal monitors and dosimeters.......................... 4-6 !
5-1 Charpy Y-notch impact properties for V. C. Summer Unit 1 ,
reactor vessel shell plate A9154-1 (longitudinal '
orientation)............................................. 5-13 5-2 Charpy Y-notch impact properties for V. C. Summer Unit 1 reactor vessel shell plate A9154-1 t orientation) . . . . . . . . . . . . . ...................... . . . . . . . . . .(transverse 5-14 L L
5-3 Charpy V-notch impact properties for V. C. Summer Unit 1 !
reactor vessel weld meta 1................................ 5-15 l 5-4 Charpy Y-notch impact properties for V. C. Summer Unit i reactor vessel weld heat af f ected some metal. . . . . . . . . . . . . 5-16 5-5 Charpy impact specimen fracture surfaces for V. C.
l Summer Unit i reactor vessel shell plate A9154-1 (longitudinalorientation)............................... 5-17 l 5-6 Charpy impact specimen fracture surfaces for V. C.
Summer Unit i reactor vessel shell plate A9154-1 L l (transvers e orientation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-18 F
5-7 Charpy impact specimen fracture surfaces for V. C. ;
Summer Unit I reactor vessel weld metal.................. 5-19 5-8 Charpy ispact specimen fracture surf aces for V. C.
Summer Ur.it 1 reactor vessel weld RAZ metal. . . . . . . . . . . . . . 5-20 l i
. 5-9 Tensile properties for V. C. Summer Unit I reactor vessel L shell plate A9154-1 (longitudinal orientation) . . . . . . . . . . . 5-21 5-10 Tensile properties for V. C. Summer Unit 1 reactor vessel shell plate A9154-1 (transverse orientation) . . . . . . . . . . . . . 5-22 ;
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l Finure ,P,,gg 6-11 Tensile properties for V. C. Summer Unit 1 reactor vessel ,
i weld metal............................................... 5-23 5-12 Frmetured tensile specimens from V. C. Summer Unit i reactor vessel shell plate A9154-1 (longitudinal orientation)............................................. 5-24 5-13 Fractured tensile specimens from V. C. Summer Unit i reactor vessel shell piste A9154-1 (transverse orientation)............................................. 5-25 l<
5-14 Fractured tensile specimens from V. C. Sammer Unit 1 ;
reactor vessel weld meta 1................................ 5-26 .
5-15 Typical stress-strain curve for tension specimens........ 5-27 6-1 V. C. Summer Unit I re actor geometry. . . . . . . . . . . . . . . . . . . . . 6-42 6-2 F1mn view of a dual reactox vessel surveillance capsule.. 6-43
- 6-3 Relative animi varinace of f ast (E > 1.0 WeV) neutron [
flux sad fluence within the reactor vessel wall. . . . . . . . . . 6-44 +
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LIST OF TABLES Table P_a.ge, 3-1 Y. C. Summer Unit 1 Reactor Vessel Toughness Data (Unirradiated).......................................... 3-3 4-1 Chemical Composition of the V. C. Summer Unit 1 Reactor Vessel Surveillance Wateria1s............................ 4-3 4-2 Beat Treatment of the V. C. Summer Unit 1 Reactor Yessel Surveillance Wateris1s............................ 4-4 5-1 Charpy V-Notch Impact Data for the V. C. Summer Unit 1 Shg1PlageA9154-1Irradiatedat550*F, Fluence 147x 10 n/ca (E > 1. 0 WeV) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-6 5-2 Charpy Y-Notch Impact Data for the V. C. Sunser Unit 1 ReactorVesselWeldMeta}gandEg3MetalIrradiatedat 550'F, Fluence 1.47 x 10 n/cm (E > 1.0 WeV) . . . . . . . . . . . 5-7
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5-3 Instrumented Charpy Impact Test Results for V. C.
Unit 1ShellPlateA9154-1Irradiatedat1.47x10{ peer n/cm (E > 1 . 0 M e V) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-8 5-4 Instrumented Charpy Impact Test Results for V. C. Summe i Unit 1 Weld Metal and RAZ Metal Irradiated at 1.47 x 10{g n / cm (E > 1. 0 Me V) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5-5 Effect of 550*F Irradiation at 1.47 x 10 18 n/ca2 (R > 1.0 WeV) on Notch Toughness Properties of V. C.
Sumeer Unit 1 Remetor Yessel Waterials................... 5-10 5-6 Comparison of V. C. Summer Unit 1, 30 f t-lb Transition Temperature Results with Regulatory Guide 1.99 Revision 2 Predictions.............................................. 5-11 5-7 Tensile Properties for V. C. Summer Unit WaterialIrradiatedat550*Fto1,47x10{gReactgrYessel n/cm
( E > 1 . 0 W e V) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-12 1
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Table Pagg 6-1 SAILOR 47 Neutron Energy Group Structure................. 6-16 _
6-2 Nuclear Constants for Radiometric Monitors Contained in the V. C. Summer Unit 1 Surveillance Capsules............ 6-17 6-3 Fast (E > 1.0 MeV) Neutron Exposure at the Reactor Vessel Inner Radius--Asinuthal Angle of 0'............... 6-18 6-4 Fast (E > 1.0 MeV) Neutron Exposure at the Reacter Yessel Ir ir Radius--Asisuthal Angle of 12*.............. 6-19 6-5 . 3 . 0 MeV) Neutron Exposure at the Reactor Ve Inner Radius--Asinuthal Angle of 20'.............. 6-20 6-6 Fast (E > 1.0 'JeV) Neutron Exposure at the Reactor Vessel Inner Radius--Asinuthal Angle of 30*.............. 6-21 6-7 Fast (E > 1.0 MeV) Neutron Exposure at the Reactor Vessel Inner Radius--Asinuthal Angle of 45*.............. 6-22 6-8 Fast (E > 1.0 MeV) Neutron Exposure at the 16.94 Degree Surveillance Capsule Center.............................. 6-23
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6-9 Fast (E > 1.0 MeV) Neutron Exposure at the 19.72 Degree Surveillance Capsule Center. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-24 6-10 Calculated Relative Fast Neutron Exposure Parameters for V. C. Summer Unit 1............ .................... 6-25 6-11 Ratios of Fast Neutron Exposure Parameters to Fast (E > 1.0 MeV) Neutron Flux for the Reactor Vessel and Surveillance Capsules.................................... 6-28 6-12 Calculated Neutron Energy Spectra at the Center of the V. C. Summer Unit i Surveillance Capsule 'V' . . . . . . . . . . . . . 8-27 6-13 Spectrum Averaged Reaction Cross Sections at the Center of the V. C. Summer Unit 1 Sury iillance Capsules. . . . . . . . . 6-28 6-14 V. C. Summer Unit 1 Fower History, Capsule U from NUREG-0020............................................... 6-29 6-15 Y. C. Summer Unit 1 Power History, Capsule Y from NUREC-0020............................................... 6-30 ,
6-16 Comparison of Weasured and Cs'culated Radiometric Monitor Saturated Activities for V. C. Summer Unit 1 .
Surveillance Capsule U................................... 6-33 6-17 Results of Fast Neutron Dosimetry for V. C. Summer Unit 1 Surveillance Capsule U................................... 6-35 i viii
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. 6-18 Product Nuclide Burnout Assessment for V. C. Summer Unit i Surveillance Capsule U................................... 8-38 6-19 Comparison of Measured and Calculated Radiometric Monitor Saturated Activities for V. C. Suraer Unit 1 Survelliance Capsule Y................................... 6-37 6-20 Results of Fast Neutron Dosimetry for V. C. Summer Unit 1 Surveillance Capsule V................................... 6-39 6-21 Product Nuclide Burnout Assessment for V. C. Summer Unit i Surveillance Capsule V.................................. 8-4J 6-22 Suasary of V. C. Summer Unit 1 Fast Neutron Fluence Results Based on Surveillance Capsule V.......................... 0-41 9
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- 1.
SUMMARY
OF RESULTS ,
The analysis of the reactor vessel material contained in surveillance Capsule V, the second capsule to be removed from the V. C.
Summer Unit 1 pressure vessel, led to the following conclusions:
- The capsule received an ggerage 2 f"'t "*"t# " II"*""'
(E > 1 MeV) of 1.47 x 10 n/cm
- Irradiation of the r9getor 5"'*1 1"t'r"*di"t* ' hell Pl ate A9154-1 to 1.47 x 10 n/cm resulted in 30 and 50 ft-lb trannition temperature increases of 60*F and 65'F, respectively, for specimens oriented parallel to the major working direction (longitudinal orientation) and increases of 40*F and 55'F, respectively for specimens oriented normal to the major working direction (transverse orientation).
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- Weld metal irradiated to 1.47 x 10 19 n/cm resulted in both a 30 and 50 f t-lb transition temperature increase of 45'F.
- The average upper shelf energy (transverse orientation) of the plate A9154-1 increased from 75 to 78 ft-lb, and the limiting weld decrggsed fgom 91 to 85 ft-lb after irradiation to 1,47 x 10 n/cm . Both materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life of the vessel as required by 10CFR50, Appendix 0.
- The surveillance capsule test results do not indicate any significant changes in the RT values projected for the reactor vessel, and, therefor!!Ta low risk of vessel failure from pressurized thermal shock (PTS) events is postulated.
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- 2. INTRODUCTION This report presents the results of the examination of Capsule V, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the V. C. Summer Unit i reactor pressure vessel materials under actual operating conditions.
The surveillance program for the V. C. Summer Unit i reactor pressure vessel asterials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel asterials are presented by Davidson and Yanichko.(1) The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E-185-79, ' Standard .
Practice for Conducting Surveillance Tests f.r Light Water Nuclear Power Reactor Yessels'. 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 Center where the postirradiation mechanical testing of the Charpy Y-notch impact and tensile surveillance specimens were performed.
This report summarized the testing of and the postirradiation data obtained from surveillance Capsule V removed from the V. C. Summer Unit 1 reactor vessel and discusses the analysis of these data.
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- 3. BACKGROUND l
l The ability of the large steel pressure vessel containing the -
reactor core and its primary coolant to resist fracture constitutes an l 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 ber.ause it is subjected to significant fast neutron k 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 V. C. Summer Unit I reactor pressure vessel beltline) are well documented in the literature.
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Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels have been presented in ' Protection Against Nonductile Failure', Appendix G to Section III of the ASME Boiler and Pressure Yessel Code. The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTg7) .
RT NDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the ,'
temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specinens oriented normal ,
(transverse) to the major working direction of the material. The RTET '
of a given material is used to index that material to a reference stress -
intensity factor curve (K IR curve) which appears in Appendix 0 of the -
ASME Code. The K IR curve is a lower bound of dynamics, crack arrest,
. and static fracture toughness results obtained from several heats of precsure vessel steel. When a given material is indexed to the K IR curve, allowable stress intensity factore can be obtained for this 1
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material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors. ,
RT NDT 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 changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the V. C. Summer Unit 1 Reactor Yessel Radiation Surveillance Program,(1) in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 f t-lb temperature (ARTNDT) d"* D I##"di"Di "
is added to the original RT NDT t adjust the RT NDT f r radiation embrittlement. This adjusted RTNDT (RTNDT initial + ARTNDT) is used to index the material to the K IR curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials. The unirradiated vessel material data for V. C. Summer Unit 1 is shown in Table 3-1.
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Table 3-1. V. C. Summer Unit 1 Reactor Vessal Toughnaza D:,ta (Usirradicted) 50 ft-lb MWD (*) NMWD(b) 35 mil Shelf Shelf RT Cu P Ni NDT Temp NDT Energy Energy Component Heat No. Crade % % % *F *F *F ft-lbs it-lbs Closure Dome A9231-1 A533B C1. 1 -
.009 .46 -20 40 -20 -
106.0 Head Flange 5297-Y1 A508 C1. 2 -
.009 .70 10 <60 10 -
129.0 Vessel Flange 5301-VI ' ' -
.007 .70 0 (60 0 -
172.0 Inlet Nozzle 436B-1 * * -
.005 .76 -20 (40 -20 -
130.0 436B-2 * * -
.005 .81 0 <60 0 -
114.5 436B-3 " " -
.005 .81 -20 (40 -20 -
135.0 Outlet Nozzle 437B-1 * * -
.007 .81 -10 <50 -10 -
146.0 437B-2 * * -
.006 .80 -10 <50 -10 -
165.0
- 437B-3 ' " -
.006 .78 0 <50 0 -
150.0 y Nozzle Shell C9955-2 A533B C1. 1 .13 .010 .57 -20 78 18 -
100.5 00123-2 * *
.12 .009 .58 -30 86 26 -
91.0 Inter. Shell A9154-1 * *
.10 .009 .51 -20 90 30 136 80.5
"
- A9153-2 ' "
.09 .006 .45 -20 40 -20 141 106.5 Lower Shell C9923-2 ' '
.08 .005 .41 -10 70 10 161 91.5 C9923-1 ' '
.08 .005 .41 -30 70 10 148 106.0 Botton Hd. Ring A9249-1 * * -
.010 .53 -40 23 -37 -
107.0 Botton Dome A9231-2 * * -
.010 .45 -10 42 -10 -
134.0 Inter. to Lower Shell Cirth Weld .06 013 .89 -50 16 -44 -
84.0 Inter. & Lower Shell Long. Welds .06 .013 .89 -50 16 -44 -
84.0 Weld HAZ - - -
-70 -37 -70 -
130.0
(*) Major Working Direction
) Normal to Major Working Direction
- 4. DESCRIPTION OF PROGRAM Six surveillance capsuler, for monitoring the effects of neutron exposure on the V. C. Summer Unit 1 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup., The six capsules were pocitioned in the reactor vessel between the neutron shielding pads and the vessel wall as shown in Figure 4-1.
i The vertical center of the capsuiss is opposite the vertical center of l the core.
Capsule Y was removed after 2.93 effective full power years of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2 T compact tension (CT) specimens (Figure 4-2) from the intermediate shell plate A9154-1 and submerged arc weld metal representative of the intermediate to lower shell beltline weld seaa of the reactor vessel and l Charpy V-notch specimens from weld heat-af fected zone (HAZ) material.
l All heat-affected zone specime.is were obtained from within the HAZ of plate A9154-1 of the representative wald.
The chemistry and heat treatment data from the V. C. Summer Oult 1 surveillance material are presented in Tables 4-1 and 4-2, respectively.
All 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. Base metal Charpy Y-notch impact and tension specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation) and also normal to the n.ajor working direction (transverse orientation). Charpy Y-notch and tensile j -
specimens from the weld metal were oriented with the longitudinal axis l of the specimens transverse to the welding direction. The CT specimens in Capule V were sachined such that the simulated crack in the specimen would propagate normal and parallel to the major working direction for the plate specimen and parallel to the weld direction.
4-1
Capsule V contained dosisater wires of pure copper, iron, nickel, and aluminum-0.15% cobalt (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were contained in the capsule.
Thermal monitors ande from the two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. The composition of the two alloys and their molting points are as follows.
2.5% Ag, 97.5% Pb Welting Point: 579'F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Welting Point: 590'F (310'C)
The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in capsule V are shown in Figure 4-2.
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Table 4-1
- Chemical Composition of the Y. C. Summer Unit 1 Reactor Vessel Surveillance Materials Plate A9154-1 Weld Metal (*)
Lukens Steel Co. Lukens Steel Co.
Element Analysis Analysis 0 0.22 0.085 S 0.015 0.012/0.007(b)
N 2
0.0076 0.015 r
Co 0.010 0.060/0.01(b)
Cu 0.10 0.05/0.04(b)
Si 0.24 0.48/0.42b)
Mo 0.49 0.49/0.46b)
Ni 0.51 0.91/0 95 (b)
Mn 1.30 1.32/1.50(b)
Cr 0.08 0.14/0.12(b)
Y 0.001(*) 0.005 P 0.009 0.013/0.009(b)
Sn 0.007 0.0047 Al 0.024 0.007/0.03(b)
B 0.0004 0.0005 Ti 0.0002 0.001 Pb <0.005 0.0206 Zr 0.001 0.001 As 0.006 0.006 W (0.01 0.01 r l
(") Westinghouse Analysis, b) Analysis performed on hrradiated weld specimen CW14.
(*) Surveillance weld was ande of the same RACO INWM wire Heat #4P4784 and Linde 124 Flux Lot No. 3930 as the
. beltline welds of the reactor vessel.
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Table 4-2 Heat Treatment of the V. C. Summer Unit 1 .
Reactor Vessel Surveillance Materials Waterial Tencera' cure (*F) Time (Hr) Coolant Shell Plate 1550*/1650' 1/2 hr/in., sin. Water quenched A9154-1 1225' *25* 1/2 hr/in., sin. Air cooled 1150' *25' 43 Furnace cooled to 600*F Weldsent 1150'
- 25* 12 Furnace cooled e
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Iso-Figure 4-1. Arrangement of surveillance capsules in the V. C. Summer Unit i reactor vessel 4-5
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- 5. TESTING OF SPECIMENS FROM CAFSULE V 5.1 OVB1 VIEW The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Center with consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H,(2) ASTM Specification E185-82, and Westinghouse )
Procedure RMF-8402, Revision 0 as modified by RMF Procedures 8102 and l 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-9234.(1) No discrepancies were found.
Examination of the two low-melting point 304*C (579'F) and 310'C 1- (590'F) eutectic alloys indicated no melting of either type of thermal 1
monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304'C (579'F) . l The Charpy impact tests were performed per ASTM Specification E23-82 and RMP Procedure 8103 on a Tinius-Olsen Model 74,358J machine.
l The tup (striker) of the Charpy machine is instrum.ented with an Effects l Technology Model 500 instrumentation system. With this system, load-time and v.nergy-time signals can be recorded in addition to the standard measurement of Charpy energy D(B ) . From the load-time curve, the load of general yielding (Pgy), the time to general yielding (tgy),the j
maximum load (P g ), and the time to maximum load (ty ) can be determined.
Under some test conditions, a sharp drop in load indicative of fast
)- fractur* was observed. The load at which fast fracture was initiated is identified as the fast fracture load (P ),
p and the load at which fast fracture terminated is identified as the arrest load (/g).
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The energy at maximum load (Eg ) was determined by comparing the energy-time record and the load-time record. The energy at maximum load '
is roughly equivalent to the 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 2 and maxinua loads, also using the three-point bend formula.
Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTW 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-l l
console test machine (Model 1115) per ASTW Specification E8-83 aid
! E21-79, and RWF Procedure 8102. All pull rods, grips, and pins were made of Inconel 718 hardened to Re 45. The upper pull rod was connected ,
through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches 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 ASTW E83-67.
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests wars conducted in air.
Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature. Chrosel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy a-2
specimen and in each grip. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range of room temperature to 550'F (288'C) . The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures.
Experiments indicated that this method is accurate to *2*F.
The yield load, ultimate load, fracture load, total elongation, and unifora elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional ares. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress i (true stress at fracture) and percent reduction in area was compuced using the final diameter seasurement.
5.2 CHARPY V-NOTCH IMPACT TBST RBSUI,TS The results of Charpy Y-notch impact tests performed on the
, various materials contained in Capsule V irradiated at 1.47 x 10 19 n/cm are presented in Tables 5-1 through 5-4 and Figures 5-1 through 5-4. !
The transition temperature increases and upper shelf energy decreaseu for the Capsule V materials are summarized in Table 5-5.
l Irradiation of vessel intermediate shell plate A9154-1 material (longitudinal orientation) speclaens to 1.47 x 10 1 n/cm2 (Figure 5-1) resulted in a 60*F and 65'F increase in 30 and 50 f t-lb transition temperatures respectively, and an upper shelf energy cecrease of 10 I ft-lb.
1
( Irradiation of vessel intermediate sbw.. r: a 9154-1 material I
(transverse orientation) specimens to 1.47 x 10 19 n/m 9Jigure 5-2) l resulted in 30 and 50 ft-lb transition temperature increases of 40'F and l 55'F, respectively, The irradiated upper shelf energy experienced an l 1
increase of 1 ft-lb when compared to the unirradiated data. l s-a 1
l _ __
l 1
Weld metal irradiated to 1.47 x 1019 ,7 2 (Figure 5-3) resulted in both 30 and 50 ft-lb transalon temperature increases of 45'F and an .
upper shelf energy decrease of 6 ft-lb.
Weld RAZ metal irradiated to 1.47 x 1019 ,7,2 (Figure 5-4) resulted in 30 and 50 ft-lb transit!on temperature increases of 45'F and 55'F, respectively, and an upper sholf energy decrease of 19 ft-lb.
The fracture appearance of each irradiated Charpy specimen f om the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile or tougher appearance with increasing test temperature.
A comparison of the 30 ft-lb transition temperature increases for the various V. C. Summer Unit i surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2(3) is presented in Table 5-6. This comparison indicates that the transition temperature increases resulting from irradiation to 2
1.47 x 10 18 n/cm are conservative compared to the Guide predictions.
5.3 TBNSION TBST RESULTS The results of tension tests performed on plate A9154-1 (longitudinal and transverse orbatation) and the weld mecal irradiated 2
to 1.47 x 10 19 n/a are shown in Table 5-7 and Figures 5-9, 5-10 and 5-11. Plate A9154-1 test results are shown in Figures 5-9 and 5-10 and indicate that irradiation to 1.47 x 10 19 n/cm2 caused a less than 10 ksi increase in the 0.2 percent offset yield strength and ultimate tensile strength. Weld metal tension test results shown in Figure 5-11, show that the ultimate tensile strength and the 0.2 percent offset yield strength increased by arproximately 5 ksi with irradiation. The fractured tension specimens for the plate material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14. A typical stress-strain curve for the tention tests is shown in Figure 5-15. ,
8-4
5.4 COMPACT TBNSION TBSTS Per the surveillance capsule testing contract with South Carolina Electric and Gas, 1/2 T-compact tensinn (CT) specimens will not be tested. CT specimen will be stored at the Hot Cell at the Westinghouse R&D Center.
O I
o e
8-8
Table 5-1 Charpy V-Notch Impact Data for the V. C. Summer Unit 1 '
ShellPlateA9154jgIrr:4[atedat550'F, Fluence 1.47 x 10 n/cm (B > 1.0 MeV)
Temperature Impact Energy Lateral Expansion Shear Sample No. I'F1 ('0) (ft-lb) IJ1 (mils) 1331 (%)
Longitudinal Orientation CL19 0 (-18) 6.0 ( 8.0) 6.0 (0.15) 2 0030 25 (- 4) 24.0 32.5) 24.0 0.61) 10 CL28 25 (- 4) 32.0 43.5) 25.5 0.65) 15 CL20 50 10) 38.0 51.5) 34.5 0.88) 15 CL22 50 10) 44.0 59.5) 34.5 0.88) 20 CL29 77 25) 73.0 99.0) 52.5 1.33) 45 CL26 78 26) 79.0 (107.0 53.0 1.35) 50 CL21 100 38) 23.0 (31.0 22.5 0.57 20 CL16 100 38) 70.0 (95.0 51.5 1.31 50 CL18 125 (52) 65.0 88.0 49.0 1.24 50 CL24 150 (66) 100.0 135.5 79.5 2.02) 90 CL23 200 (93) 125.0 169.5) 83.5 2.11) 95 CL17 250 121) 121.0 (164.0 82.5 (2.10) 100 CL25 350 177) 119.0 (161.5 85.5 (2.17) 100 CL27 400 204) 125.0 (169.5 80.5 (2.04) 100 ,
Transverse Orientation CT21 -50 -46) 5.0 7.0 5.0 (0.13) 2 ,
CT19 0 -18) 14.0 19.0 13.0 (0.33 5 CT24 50 10) 25.0 34.0 26.0 (0.66 15 CT18 50 10) 44.0 59.5 25.0 (0.64 20 CT16 77 25) 35.0 47.5 32.0 0.81) 20 CT30 78 26) 33.0 44.5 28.0 0.71 20 CT20 100 40.0 54.0 41.0 1.04 35 CT28 100 44.0 60.0 34.5 0.88 35 CT22 125 34.0 46.0 36.5 0.93 30 CT27 125 46.0 62.5 42.5 1.08) 35 CT17 150 52.0 70.5 49.5 1.26) 45 CT26 250 1 77.0 104.5 62.5 1.59) 100 CT29 300 14 73.0 99.0 71.0 1.80 100 CT23 375 19 77.0 104.5 64.5 1.64 100 8-e
Table 5-2
. Charpy Y-Notch Impact Data for the V. C. Summer Unit 1 Reactor VesselWeldMetalandH{gMeta{Irradiatedat550*F, Fluence 1.47 x 10 n/cm (E > 1.0 MeV)
Temperature Impact Energy Lateral Expansion Shear Sample No. (*F) (*C) (ft-lb) 2), (mils) (mm) (%)
Weld Metal CW17 -90 (-68) 22.0 ( 30.0) 22.5 0.56) 10 CW25 -60 -51) 19.0 (26.0 17.5 0.44) 15 CW23 -25 -32) 10.0 13.5 16.0 0.41) 15 CW20 -25 -32) 10.0 13.5 14.5 0.37 20 CW19 -10 -23) 37.0 50.0 32.0 0.81 30 CW26 0 -18) 33.0 44.5) 31.5 0.80 45 CW24 0 -18) 42.0 57.0) 36.5 0.93) 45 CMO 25 (- 4) 29.0 39.5) 28.5 0.72) 35 CW18 25 - 4) 46.0 62.5) 42.0 1.07) 70 CW27 50 10) 62.0 84.0) 54.5 1.38) 70 CW28 50 10) 78.0 (106.0) 65.0 1.65) 95 CW29 100 38) 73.0 99.0) 63.0 1.60) 95 CW21 150 66) 84.0 114.0) 72.5 1.84) 100 CW'22 200 93) 79.0 107.0) 68.5 1.74) 100 CW16 300 149) 92.0 (124.5) 85.5 2.17) 100 HAZ Wetal
. CH28 -90 -68) 22.0 30.0) 19.0 (0.48) 10 CH26 -60 -51) 26.0 35.5 20.0 (0.51) 15 CH27 -50 -48) 27.0 36.5 24.0 0.61) 20 CH29 -25 -32) 26.0 35.5 26.0 0.66) 20 CH25 -25 -32) 24.0 32.5) 21.5 0.55) 20 CH30 -10 -23) 77.0 104.5) 52.0 1.32) 50 CH23 -10 -23) 41.0 55.5) 31.0 0.79) 30 CH18 0 -18) 68.0 (92.0) 49.5 1.26 60 CH22 0 -18) 63.0 (85.5) 47.5 1.21 50 CH17 50 10) 112.0 152.0) 77.0 1.06 100 '-
CB19 50 10) 72.0 97.5 52.5 1.33) 65 i CH21 100 38 95.0 129.0 73.5 1.87) 95 CH24 150 66 112.0 152.0 76.0 1.93) 100 CH16 200 ( 93 103.0 139.5 74.0 1.88) 100 CH2O 300 (149) 117.0 159.0) 84.0 (2.13) 100 e
8 'r
Tabla 5-3 V Su Instrumented Shell Charpy Plate A9154-1 Impact Test Irradiated at 1.47 Results x 10 fopg n/cm. C.2 (mmer E> Unit 1 1.0 MeV)
Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Saaple Temp Energy Ed/A Ee/A Ep/A Load to Yield Load Maximum Load Load Stress Stress Number f*FJ (ft-lb) (ft-lo/in ) (kips) (ssec) (kips) (ssec) (kips) (kips) (ksil (ksi)
. Longitudinal Orientation CL19 0 6.0 48 30 18 3.2 85 3.80 110 3.80 -
106 116 CL30 25 24.0 193 172 21 3.25 85 4.25 395 4.25 -
108 124 CL28 25 32.0 258 232 26 3.50 95 4.60 500 4.60 -
116 134 CL2O 50 38.0 306 231 75 3.40 95 4.55 505 4.55 .15 113 132 CL22 50 44.0 354 318 36 3.45 85 4.65 650 4.65 -
114 134 1
CL29 77 73.0 588 303 285 3.15 90 4.35 665 4.10 1.20 105 124
, CL26 78 79.0 636 308 328 3.25 85 4.50 655 3.95 .55 107 128 CL21 100 23.0 185 88 97 3.15 85 3.65 235 3.65 1.10 104 112 CL16 100 70.0 564 306 258 3.10 80 4.45 650 4.15 1.00 103 126
- CL18 125 65.0 523 294 229 3.00 85 4.30 655 4.05 2.00 99 120 CL24 150 100.0 805 304 501 3.10 90 4.40 670 2.75 1.65 102 124
[ CL23 200 125.0 1007 269 737 2.70 75 4.00 635 - -
89 111 CL17 250 121.0 974 295 679 2.9 35 4.20 670 - -
95 117 CL25 350 119.0 958 289 670 2.75 90 4.00 675 - -
91 112 CL27 400 125.0 1007 297 709 2.85 95 4.10 685 - -
94 115 Transverse Orientation l CT21 -50 5.0 40 29 12 3.45 80 4.00 95 3.95 -
115 124
- CT19 0 14.0 113 107 5 3.65 85 4.10 250 4.10 -
121 128 CT24 50 25.0 201 161 40 3.30 85 4.15 370 4.10 -
110 123 CT18 50 44.0 354 115 239 3.45 90 4.00 280 4.00 .40 113 123 CT16 77 35.0 282 155 127 3.15 85 3.95 375 3.95 1.35 105 118 CT30 78 33.0 266 176 90 3.20 90 4.30 405 4.20 .95 106 124 i CT20 100 40.0 322 217 105 3.05 80 4.15 495 4.15 1.40 101 120 Cr28 100 44.0 355 - - - - - - - - - -
CT22 125 34.0 274 137 137 3.15 110 3.90 360 3.75 1.25 103 117 CT27 125 46.0 370 218 152 3.05 85 4.15 505 4.00 1.20 100 119 CT17 150 52.0 419 214 205 3.05 85 4.10 500 4.05 2.15 100 118 CT25 200 67.0 540 - - - - - - - - - -
CT26 250 77.0 620 223 397 3.10 90 4.30 500 - -
103 122 l CT29 300 73.0 588 197 391 2.70 80 3.90 475 - -
89 110 CT23 375 77.0 620 200 420 2.75 100 3.80 505 - -
91 108 i
l l
Table 5-4 Instrumented Charpy Impact Test Results for V ggC . Sunser Unit 1 Weld Metal and HAZ Metal Irradiated AT 1.47 x 10 n/cm (E > 1.0 MeV)
Normalized Energies Test Charpy Charpy Maximus 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 Number f*F1 (ft-lb) (ft-lb/in") (kips) (psec) (kips) (psec) (kips) (kipsT (ksi) (ksi)
Weld Metal CW17 - 90 22.0 177 71 106 3.90 90 4.35 175 4.30 -
128 136 CW25 - 60 19.0 153 124 29 3.80 90 4.50 275 4.45 -
126 137 CW23 - 25 10.0 81 44 36 3.50 85 3.80 130 3.80 .65 115 121 CW20 - 25 10.0 81 44 37 3.60 90 3.80 130 3.75 .30 119 123 CW19 - 10 37.0 298 249 49 3.70 90 4.60 510 4.50 .35 123 137 CW26 0 33.0 266 198 68 3.60 90 4.35 430 4.35 1.05 118 131 CW24 0 42.0 338 244 95 3.65 90 4.60 505 4.55 .75 121 137 CW30 25 29.0 234 112 121 3.45 90 4.00 275 3.95 2.35 114 124 CW18 25 46.0 370 234 136 3.35 85 4.30 510 4.20 2.4 110 126 CW27 50 62.0 499 227 272 3.30 85 4.35 500 4.05 2.20 109 126 CW28 50 78.0 628 230 348 3.50 85 4.60 575 - -
116 134
{ CW29 CW21 100 150 73.0 84.0 588 676 262 278 326 398 3.35 3.2 95 85 4.35 4.40 575 605 110 105 127 125 CW22 200 79.0 636 265 371 3.05 85 4.15 600 - -
101 119 Cir16 300 92.0 741 256 485 2.80 85 4.00 605 - -
92 112 HAZ Metal CH28 - 90 22.0 177 159 18 4.30 95 4.75 320 4.75 -
143 150 CH26 - f4 26.0 209 181 28 4.25 105 4.80 370 4.75 -
141 150 CH27 - 50 27.0 217 179 38 3.85 90 4.65 375 4.60 .25 127 140 CH25 - 25 24.0 193 152 41 3.80 90 4.40 330 4.40 .30 126 136 CH29 - 25 26.0 209 160 49 3.85 95 4.45 350 4.40 .55 128 138 CH23 - 10 41.0 330 294 36 3.90 110 4.80 595 4.65 1.45 129 144 CH30 - 10 77.0 620 339 281 3.85 85 4.85 655 4.45 2.00 127 144 CH22 0 63.0 507 291 216 3.90 95 4.80 580 4.65 1.45 129 144 CHIP O 68.0 548 333 214 3.75 90 4.85 655 4.75 2.80 125 143 CH19 50 72.0 580 325 255 3.55 135 4.75 700 4.65 2.60 117 137 CHt7 50 112.0 902 317 585 3.45 90 4.60 665 - -
114 133 CH21 100 95.0 765 263 502 3.30 90 4.35 575 - -
109 127 CH24 150 112.0 902 312 590 3.40 90 4.55 660 - -
112 132 CH16 200 103.0 829 297 533 3.15 85 4.30 660 - -
104 123 CH2O 300 117.0 944 - - - - - - - - - -
Table 5-5 I8 Effect of 550*F Irradiation at 1.47 x 10 m/cm2 (E > 1.0 MeV) on htch T-J-- - Properties of Y. C. Summer Unit 1 Reactor hemel Materiale Average 35 all Average Energy Average 30 ft-lb Lateral Espanelos Average 50 ft-lb Aboerytion at Temperature (*F) Temperature (*F) Temperatere (*F) Full Shear (ft-Ib)
Waterial Unirradiatd Irradiated E Usirradiated Trradiated E Unirradiatd Irradiated E Usirradioted Irrediated Afft-lM Flate A9614-1 -25 36 80 0 60 50 0 as SE 132 122 10 (Longitediaal)
Flate A9614-1 26 46 40 55 96 40 75 133 56 75 78 1(*}
(Treaeverse) hld Metal -40 -15 45 -35 0 35 -15 30 45 91 85 S EAZ h tal -90 -46 45 -80 -15 45 -70 -16 EE 130 111 19 IO) Increase is shelf energy 1
i t
i i
I I
I i
l
, ._y,, 7- . _ , , . , . , . ._,..m.. _. , - - . _ . .
1 Table 5-6 Comparison of f. C. Summer Unit 1, 30 ft-lb Transition Temperature Results with Regulatory Guide 1.99 Revision 2 Prediction l
R.C. 1.99 Rev. 2 Predicted 30 ft-lb Transition Actual Shift Based
- Material Caosule Fluence Temperature Shift On Surveillance Data Plate A9514-1 U 6.39 x 10 57*F 40*F 4
(Lougitudinal) Y 1.47 x 10 72*F 60*F 8
Plate A9514-1 U 6.39x10fg 57'F 30*F (Transverse) Y 1.47 x 10 72*F 40*F I Weld Metal U 6.39 x 10 59'F 30*F
, Y 1.47 x 10 75'F 45*F b
i i
Table 5-7 TensilePropertiesforV.C.SummerUng1ReptorYesselMaterial Irradiated at 550*F to 1.47 x 10 n/cm (E > 1.0 MeV)
Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp. Strength Strength Load Stress Strength Elongation Elongation in Area (ksi) (ksi)
Number Material f*F1 (ksi) (ksi) (kip) (5) (5) (5) l CL5 Long. 100 69.0 91.7 3.05 212.1 62.1 12.0 25.4 71 CL4 Long. 250 68.2 86.6 2.90 201.7 59.1 11.3 23.9 71 CL6 Long. 550 61.6 88.6 3.05 194.8 62.1 10.5 23.0 63 CT6 Trans. 74 69.3 92.7 3.33 168.1 67.8 12.8 26.0 60 CT4 Trans. 200 65.5 87.6 3.10 162.4 63.2 10.5 21.9 61 CTS Trans. 550 61.1 88.2 3.45 211.5 70.3 10.5 19.2 67
, CW6 Weld O 81.0 98.8 3.30 202.3 67.2 13.5 25.8 67
,L CW5 Weld 100 76.9 91.7 3.00 176.5 61.1 12.0 23.4 65
" CW4 Weld 550 70.3 88.6 3.25 152.5 66.2 9.8 19.4 57 l
Curva 7550hc <A
(
- C)
-150 -100 - 50 0 50 100 150 200 250 120 '
I I I I i I ' '
2 100 - _ e c : : -
- e 2 f 80 n
gg -
e m 0 - 2 I -
\
e -
0 lM i i i i i i i i i 15 5 80
^
e . 2.0 o -
e o
-M - -
L5 e 8O -
go -2
- 1. 0 e
&w 3
15 i i n i i i i i 0 0
200 , , , , , , , , ,
180 -
240 160
. - o 200 1
_ 140 e ~
160 d -
Unirradiated o -
@1M Irradiated at -
120 0 1
g 80 -
- . 1. 47 x 1019 n/cm2
" 60 -
,/
65 F 0
40 - / I MoF s
- _ m ,
20 -
7o 0 i i i i i i i 0
- 200 -100 0 100 200 300 00 500 Temperature ( F)
Figure 5-1. Charpy Y-notch impact properties for V. C. Summer Unit i reactor vessel shell plate A9154-1 (longitudinal orientation) 8-13
_-- . _ . . - . . - . - - . _ . , - - - _ - -- . .~ . - . .
Curvo 755039- A
(
- C)
-150 -100 - 50 0 50 100 150 200 250 120 i i ' ' '
i i i 13 100 -
3 80 -
%w E 2
- mm - y -
20 -
~2 0
l@ i i i i 15 i i i i i 5 80 -
10 e o "
? -
L5 E-
'2 I40 - 40* F *. -
1.0 5 ym i
8 i
'2 i i i i i
- 0.5 0
t 0 _
100 i , , , , , , , ,
90 -
120 -
o 80 -
n . . -
100
- 70 -
[* Irradiated at E@
~
80
~
Unirradlated 1 p 50 55*F .47 x 1019 n/cm2 _ g g 40 -
m .
30 -
8 40 40* F 20 -
~
10
' ' ' ' ' ' ' ' 0 0
- 200 -100 0 100 200 300 400 500 Temperature (* F)
Figure 5-2. Charpy f-notch impact properties for V. C. Summer Unit i reactor vessel shell plate A9154-1 (transverse -
orientation) 8-14
curva 755037-A
(
- C)
-150 -100 - 50 0 50 100 150 200 250 120 8 ' '
'2 '2' 13' i i 100 -
s' -; = = _
f 80 - _
$M - _
di e -
- 2, 20 - _
' ' ' ' ' ' I 0
IN i i i i i i i i i 15
- E 80 -
. n -
10 6 0 v
7 60
- 1. 5 3 3 40 - -
1.0 3
- 35 F 4m 0.5 "0 ' -
' ' ' ' ' ' 0 200 i , , , , , , , ,
~
180 -
240 160 T
o 140 y120 -
_ ig
~ 100 -
Unirradiated o -
120 C c 80 -
o a \,
" 60 - Irradiated at -
80
- 45 F 1. A7 '.019 n/cm2 40 -
45 F * -
40 20 -
~
0 i t *-8 ' ' ' ' ' 0
- 200 -100 0 100 200 300 400 500 ,
Temperature ( F)
Figure 5-3. Charpy Y-notch impact properties for V. O. Summer Unit 1 reactor vessel weld metal i
a-18
curvo 755038-A
( 80)
-150 -100 - 50 0 50 100 150 200 250 120 100 -
i i i o i{ 2s' '3 f
3 80 - -
Bs
.c
- o .
m 40 - --
20 -
.2 '
0' ' ' ' ' ' '
100 , ,. , , , , , , 2.5 3Y -
,k 9 * -
10 3 J - 2 -
- 1. 5 's ol w 40 -
45* F 1.05
-m - -
E5 3 0 ' ' ' I i i_m_
i i 0 2% , , , , , , , , ,
180 -
240 160 2%
3 140 -
o
- S2 Unirradiated g120 -
160 o
O
~ 1@ - * -
@ * (Irradiated at 2 -
120 C g80 o * . 1,47 x 1019 n/cm o -
M
- 55* F 40 -
45*F -
40 20 - t ,
0__ __ i i i i i ' i 0 100 0 100 200 300 400 500 Temperature (* F) .
Figuts -
. apact prcpertie= for V. C. Sumner Unit 1 sold heat af fected zone metal -
a-se
i I
, ?. --
- . ; ,.. I CL19 CL30 CL23 CL20 CL22 i
l 2
1
~
i
-- 1 _ ,
Y 5' i l
. :. g i . . 9 .2,'.L . .g ,ivs . 1
%+$Ti; j.. V, -
3' .
_,g ;
i-
):__ _ '( ,
l CL29 CL26 CL21 CL16 CL18 l
'r;,
f ,s; nv-sJ ?,[O j .t. t ..,
y ,
- j g
{ t
~
i ..
S. L d'
- j' ;%%., .
- T w - *
'Q .
I CL24' CL23 CL17 CL25 CL27 l l
4 Figure 5-5. Charpy impact specimen fracture surf aces for V. C. Su=er Unit i rea,ctor vessel shell Plate A9154-1 (longitudinal !
! orientation) 4 e
- h i
i
i f
I CT21 CT19 CT24 CT18 CT16 l
-
- NI Q'S 4,, .
,f
.g.....,.
s v ., ,
- y. .s a .. ->
l
\ , .
?,5 - ,. !.- h-$t CT30 CT20 CT28 CT22 CT27 r
W M tgy)t :
Nr' ,, , - ...?..,.
] 3 * ? ,. g . . .
g ,
~
, .k. =. , . ,f ], f, 5
' t eroemorr ;
M"' f 1t,'l ' QD% b *: x o
..-i 6 g ,. ' VI ' , h I'), '[$
i . _ , I . ,_ , <+.)-
0:i s %:,1 j CT 25 CT26 CT29 CT23 s'T17 i
l Figure 5-6. Charpy impact specimen fracture surfaces for V. C. Susser !
! du.t I reactor vessel shell plate A9154-1 (transverse orientation) j l
l .
s-is m-nene
- 1 i
l
&. .J k,
,4
.,.E. I.
L ,
C#17 CW25 C#23 CW20 C#19
&l, .; g ,i t gr ;w~
- q;p. ,.
w + .
Q yii\ f [ :~. -l $. [ 3 'M i ;d?'i j_',tyi 'i'j);[,
[ f,j;g t,J
,wg ..
T ,
CW26 CW24 CW30 CW18 C#27 l
" ' I
<,k *
,+ .
k.N(,i -
. :. _ q 23 ~ ;,, ?2,
. . Vg. 5.t-4 , ,, ,; YW-;1 b-r. (_ ., ..
y b { -
- p. ,, 1%
sacr --
w m- ,w_ \
C#28 C#29 C#321 C#22 C#16 Figure 5-7. Charpy impact specimen fracture surf aces for V. C. Su=mer Unit I reactor vessel weld metal 5-19 P3-15 W l
T' . ;*. : '
pf. ., p E . 1:pd
%#,j)$
h )J.16 3, - 4s..,je 5 -7 -
U CH28 CH26 CH27 CH29 CH25 L, -
.,.._m.
}
?g h J~ b
. c, ((::6 l
. g. ,
.P' ,
t- ;_ ?&.V 4 )*Pn .l.2.'*>
CH30 CH23 CHIS CH22 CM17 ,
,W l}5[- ' f. , $ 2
.Q'*Eb i r ,
gj _
i ygq Yw
.- ;,4 % y
!$ t I
,c g , ' 6 -r . _ <
CH19 CH21 CH?4 CH16 CH2O Figure 5-8. Charpy impact specimen fracture surf aces f or V. C. Sur.=er Unit i reactor vessel weld heat affected zone metal 5-20 FM-15ee!
Curve 755035-A Temperature ( 'C) 0 % 100 1% M 2% E 110 i i i i i i i 100 -
= 90 -
% [UltimateTensile Strength -
~
S _ 2 r-0,2 % Yleid Strength
@80 g 70 -
0: : j> e 1 500 O O 60 o Unirradiated 400 19 2 g _
- Irradiated at 1.47 x 10 n/cm 300 40 S l 1 i i i i i 2
70 - 0; '
- -f 60 - Reduction in Area -
3g -
E 40 -
5 T g 30 -
g otal Elongation ,
20 -
N' -
0-_ e n 10 -
~
o #-
g ,
b:Unif9rm Elongation ,
0 100 200 300 400 500 57 5 Temperature ( 'F)
Figure 5-9. Tensile properties for V. C. Surser Unit i reactor vessel shell plate A9154-1 (longitudinal orientation) 5-31
curve 755036-A 1 Temperature ( *C) 0 50 100 150 200 250 300 ,
110 i i i i i i i
~
700 '
. 100 -
= 90 - N (Ultimate Tensile Strength
$m 80 -
/ m 1~ 5c m
- s 500 -
h 70 m p 0. 2 % Yield Strength 60
' / o i- 400 Unirradiated 19 2 50 -
. Irradiated at 1.47 x 10 n/cm - 300
' ' ' ' I 40 I I i i i i i 70 - Reduction in Area '
n b e o
60 - : ,_
350 E 40 r-Total Elongation
]a 30 -
_n li 20 -
=
9-t, -
e ;
10 -
Uniform Elongation i i t i 0 !
0 100 200 300 400 500 57 5 l
Temperature ( *F) ,
Figure 5-10. Tensile properties for V. C. Summer Unit i reactor vessel
- shell plate A9154-1 (transverse orientation)
I J
8-22
curve 755034-A Temperature ( "C) 100 150 200 250 300 0 50 110 i i i i i i i Ultimate Tensile Strength 700 100<
= 90 -
I SA ;
1 600 g
- 80<q_ 0,2 % Yield Strength - 3 E
g 500 g 70 -
- 0 i
60 400 o Unirradiated 19 2 50 -
- Irradiated at 1.47 x 10 n/cm 300 i
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Reduction in Area 350-x _
E 40
- Total Elongation f 30, -
~
20 - "
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1 niformU Elongation i i i i i 0
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l l
5-23 f
- -------_______________________3
/.
\ g i
- ,s
. ,1.,.
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wd W
Specimen CL5 100*F i
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... .a
. i Specimen CL4 250*F .
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- m ., ., , g; ;._ j l
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l
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f .
N' b
'Q, Specimen CL6 55a F Figure 5-12. Fractured tensile specimens from V. C. Surmer Unit I reactor vessel shell plate A9154-1 (longitudinal -
orientation) !
i I
l 8-24 sy. l h H I
I f
i r
I r&T. .
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t .
, ,I e; is '
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- i.. --
l, Specimen CT4 200*F i
i i
W I
i ..
i I
i i
1 l e 4
)
ll Specimen CT5 550*F 1
l .
l Figure 5-13. Fractured tensile specimens fres V. C. Summer Unit 1
{
reactor vessel shell plate A91541 (transverse orientation) 1 8-25 p,y. ; g ; )
j
r
^
l Specimen 06 0'F
- 4
= .i
. s _.
,e , , _ .- .. :.
- 55. l _
Specimen Os 100*F ,
Specimen 04 550'F Figure 5-14. Frsetured tensile specimens from V. C. Summer Unit I reactor vessel veld sets 1 3-26 M' ' 5. a ? 1
P i
i 120 i 1 i ( ;
l@ -
80 -
I
$ , \ _
wi me -
i 20 -
Soecimen CL6 -
(550'F) !
0 i i i i 0 0,05 0.10 0.15 0.20 0.25 Strain, in/in l
t Figure 5-15. Typical stress-strain curve for tension specimens .
t i
1 i
i
. I 1
- 6. RADIATION ANALYSIS AND NEUTRON DOSIMETRY
6.1 INTRODUCTION
Knowlelse of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LFR reactor prerJure vessel surveillance programs for two reasons.
First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to thich the test specimens were 1
exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship aust be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally set by employing a combination of rigorous analytical techniquas and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter inforzation is derived solely from analysis.
This section describes a discrete ordinates S, transport analysis performed for the V. C. Sunser Unit 1 reactor to determine the
- l f ast (B > 1.0 MeV) neutron flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules. The analysis data were then used to develop lead factors for use in relating -
neutron exposure of the reactor vessel to that of the surveillance capsules. Based on the use of spectrun-averaged reaction cross sections derived from this calculation and the V. C. Summer Unit 1 power history, the analyeis of the neutron dosimetry contained in Capa le V and an ,
- updated evaluation of dosimetry from Capsule U is presented.
r l
l e-1
6.2 DISCRBTE ORDINATES ANALYSIS A plan view of the V. C. Summer Unit 1 geometry at the core ,
midplane is shown in Figure 6-1. Since the reactor exhibits 1/8 the core symmetry, only zero to 45-degree sector is depicted. Six irradiation capsules attached to the neutron pads are included in the design to constitute the reactor vessel surveillance program. The capsules are located at 16.94 and 19,72 from the major axis at O' shown in Figure 6-1.
A plan view of a single curveillance capsule attached to the neutron pads is shown in Figure 6-2. The stainless steel specimen container is 1.182 by 1-inch and approximately 56 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot-high reactor core.
From a neutron transport standpoint, the surveillance capsule struttures are significant. They have a marked effect on both the distribution of neutron flux and the neutron energy spectrum in the .
water annulus between the thermal shield and the reactor vessel. In order to properly determine the neutron environment at the test specimen -
locations, the capsules themselves must be included in the analytical model. This requires at least a two-dimensional calcula' .on.
In the analysis of the neutron environment within the V. C. i Summer Unit i reactor geometry, two sets of transport calculations were carried out. The first, a single calculation in the conventional forward mode, was used to obtain spectrum-averaged reaction cross sections and gradient correctie: for dosimetry reactions. The second set of calculations consisted of a series of adjoint mode neutron transport calculations relating the f ast (E > 1.0 MeV) neutron flux at the surveillance capsulo locations and at selected azimuthal locations on the reactor vessel to the power distributions in the rea eor core.
These adjoint importance functions, when combined with cycle-specific .
core power distributions, yield plant-specific f ast neutron exposure at the surveillance capsule and pressur e vesJel locatioPs for daCh .
operating fuel cycle.
e-2
The forward transport calculation was carried out in R, 0 geometry with an S6 angular quadrature using the DOT two dimensional discrete ordinates code (4) and the SAILOR cross-section library.(5) The SAILOR library is a 47 group, ENDF-B/V based data set which was developed specifi.cally for light water reactor applications.
An.8.sotropic scattering is treated with a P3 expansion of the scattering cross-sections. The energy gr4Jp structure used in the analysis is listed in Table 6-1.
The design basis core power distribution used in the transport calculation was derived from statistical studies of long-tera operation of Westinghouse 3-loop plants. Inherent in the development of this design basis core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery.
Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely
. that a single reactor would have a power distribution at the nominal +2a level for a large nuber of fuel cycles, the use of this design basis distribution is expected to yield conservative results. This is especially true in cues where low leakage fuel management has been employed.
The adjoint analyses were also carried out using the 47 neutron energy group, P3 cross sections from the SAILOR library and an S 6 angular quadrature. Adjoint source locations were chosen at the center of each of the surveillance capsules *J well as at positions along the inner diameter of the presrure vessel. These calculations were run in R, e geometry to provide power distribution importance functions for the neutron exposure parmasters of interest. Havirig the adjoint importance functions and appropriate core power distributions, the response of interest is calcalated as P P P
, (,,) ( , , ) dEd6RdR RR,' 0' "
- r F E l
e-a l I
E where.:
Rg .,,, = Response of interest (e.g., p (E > 1.0 WeV), dpa, ,
etc.) at radius R' and asinuthal angle d'.
I(R,#,E) = Adjoint importance function at radius R and asinuthal angle # for neutron energy group E.
F(R,#,E) = Full power fission neutron density at radius R and asinuthal angle # for neutron energy group E.
The fission neutron density distributions used include the enrichment and burnup dependent effects of the fissions of other actinides in addition to U-235.
Core power distributions for use in the V. C. Summer plant specific evaluations are derived from nessured assembly and cycle burnups for each operating fael cycle to date. The adjoint results are in terms of fuel cycle averaged neutron flux which when multiplied by the fuel cycle length yields the incremental fast neutron fluence.
The core power distributions used in the plant specific fast .
neutron exposure analysis of the V. C. Summer surveillance capsule and reactor vessel were derived from the following fuel cycle nuclear design -
reports:
Fuel Crele Nuclear Desira Report i WCAP-9685 2 VCAP-10663, Rev. 1 3 VCAP-10874 Three regions of the core consisting of subsets of fuel assemblies are defined. In performing the adjoint evaluations, the relative power in the fuel assemblies comprising region 3 has been adjusted to account for known biases in the prediction of power in the peripheral fuel assemblies while the relative power in the fuel .
assemblies comprising region 2 has been maintained at the cycle average value. Due to the extreme self-shielding of the reactor core, neutrons .
born in the fuel assemblies comprising region 1 do not contribute e-4
r significantly to the neutron exposure of either the surveillance capsules or the reactor vessel.
In each of the adjoint evaluatioas, within-assembly spatial gradients have been superimposed on the average assembly power levels.
For the peripheral asoemblies (Region 3), these spatial gradients also (
include adjustments to account for analytical deficiencies that tend to occur near the boundaries of the core region.
Reactor vessel and surveillance capsule neutron fluence projections are made to 32 effective full power years (EFPY). Current neutron fluences, based on past core loadings, are defined as of the end b of Cycle 3.
Several key assumptions are required to make the neutron fluence projections. In particular the time weighted average neutron flux for ,
Cycles 2 and 3 and an 80 percent capacity f actor are assumed to be representative of all future operation. Thus, the neutron fluence projections reflect a continued commitment to the low leakage fuel management strategy shown by the second and third core loading.
Finally, it is assume that the V. C. Sum =er core will continue to be operated at the current power level of 2775 Wit.
The Westinghouse neutron transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oak Ridge National Laboratory (ORNL) Foolside Critical l Assembly (PCA) f acility,(6) the Venus PWR engineering Wockup,(7) and the Westinghouse power reactor surveillance capsule data base.(0) The benchmarking studies show that the use of SAILOR cross sections and design basis core power distributions produces neutron fluxes t'ant tend to be conservative with calculation exceeding measurements by 10 to 25 percent. When plant specific core power distributions are used with the adjoint itportance functions, the benchmarking studies show that neutron fluence predictions are distributed within plus or minus 15 percent of
, measured values at surveillance capsule locations. This analysis is consistent with established ASTV standards. (9'I3) n e-5
6.3 1ADI0 METRIC MONITORS The passive radiometric monitors included in the V. C. Summer .
Unit 1 surveillance program are listed in Table 6-2. The first five reaction in Table 6-2 are used as f ast neutron monitors to relate f ast (R > 1.0 WeV) neutron fluence to measured natorial property changes. In order to assess the potential for burnout of the product nuclides generated by f ast neutron reactions, it is necessary to also determine the angnitude of the thermal and resonance region neutron flux at the monitor location. Therefore, bare and cadmium-shielded cobalt-aluminua monitors are miso included.
The relative locations of the various radiometric monitors within the surveillance capsule are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminua 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 axial center of the capsule. .
The use of passive monitors such as those listed in TaFlo 6-2 does not yield a direct seasure of the energy-dependent neutros flux -
level at the point of interest. Rather, the activation or fiss!.a process is a measure of the integrated effect that the time arJ enert;y dependent neutron flux has on the target satorial 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 sensurements only if the irradiation parameters are well known. In particular, the following variables are important:
- The operating history of the reactor
- The energy response of the monitor
- The neutron energy spectrum at the monitor location
- The physical characteristics of the monitor .
e-e
The analysis of the passive monitors and the subsequent derivation of the average neutron flux require two operations. First, the disintegration rate of product nuclido 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.
The spe ific activity of each monitors is determined using established ASTM procedures. (14-22) Following sample preparation, the ,
activity of each monitor is determined by means of a lithium-drifted k germanium,Ge(Li)e Camma ray spectrometer. The overall standard [
deviation of the measured data is a function of the precision of sample , l weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed from V. C. Summer Unit 1, the overall 20 deviation in the measured data is determined to be plus -
or minus 10 percent. The neutron energy spectra at the monicor p l locations are determined analytically using the method described in A8*J l Section 6-2.
Having the measured activity of the monitors and the neutron energy spectrum at the monitor locations of interest, the calculation of the neutron flux proceeds as follows.
The reaction product activity in the monitor is expressed as
. n P5 f -At ' -At d 3
A = N,FY a(E) p(E) dB 2 p C I (1-e '
e (6-1) d
'E j=l max where A = induced product activity (dps per gram)
N, = number of target element atoms per gram F = weight fraction of the target nuclide in the target
. material Y = number of product atoms produced per reaction
, a(E) = energy dependent reaction cross section s .
e-7
p(E) = time averaged energy dependent neutron flux at the monitor location with the reactor at full (reference) ,
power
= average core pocer level during irradiation period j P)
P,, = maximum or reference core power level X = decay constant of the product nuclide
= length of irradiation period .i t) t d
u decay time following irradiation period j n = total number of irradiation periods
= flux (E > 1.0 WeV) during irradiation period j divided C) by the average flux (E > 1.0 MeV) over the total irradiation. C) is calculated with the adjoint neutron transport method and accounts for the change in neutron monitor response caused by core power distribution variations fran fuel cycle to fuel cycle. P)/P,,,,
rhich accounts for the month-by-month variation of power level within a fuel cycle, is applied to the full power .
based flux ratio, C).
Because the neutron flux distributions are calculated using multigroup transport methods and, further, because the main interest is in the f ast (E > 1.0 WeV) neutron flur, spectrum averaged ramction cross sections are defined such that the integral term in Equation 6-1 is 6 replaced by the following relation.
f where:
e-s
~
'" a(E) ((E) a a,p r
- = 3 ,,y p(E) d =
1
'O fg t gul -
i pg = ' " WeV p(E) 4 = p 8
1 g=1 s = group number from Table 6-1.
3 Thus, Equation 6-1 is rewritten, i
. P -At'
- 3. e-Atd i,
A=N,TYpffpI j=1 x C) ,1 - e or, solvird for the f ast (E > 1.0 WeV) neutron flux, A
J f* f P
-At' (6-2)
N,FY i yp*I** ,1 - e 3. e Atd C) '
j=1 The total f ast (B > i.0 MeV) neutron fluence is then given by 4
f 6 = pf I t) , (6-3) 3=1 x ;
where l r
l I
- t i
4-9 t 4 i
, .-.-._-,,_.-,------..,n_ . , , _ . - , . . - _ . - - ..--n _ - . . - _ . - . - . . . . - , , , , , . . , . , , , , _ , - - .
r I
t) = total effective full power seconds of reactor operation up to the time of capsule j=1 ass .
removal ,
An assessment of the potential for product nuclide burnout may ,
be made using the bare and cadalus shielded cobalt sensured activities .
and published data or the 2200 m/s absorption cross-section and the resonance integral. This is done by rewriting Equation 6-1 in terms of ;
a monitor 2200 m/s neutron flux and a monitoi resonance flux as follows i I
P -At ' l Aba o " oFY[a2200 2200*IIfes)j=1 #
r P U
j 'I ~ ' J '-Atd
' (6-4) L sax i
P 3
-Ati -At d i A cd=N,FY(rip,,,)g, 2, p
,1 - e . e (6-5) j=1 max C) where I
Abue = bare induced product activity (dps per gram) ,
A cd = cadalua shielded induced product activity (dps per gram) j a 2200
- published 2200 m/s absorption cross-section for nuclide ,
l of interest f
RI = published epicadalua dilute resonance integral for
{
nuclide of interest :
! p2200 = monitor 2200 m/s neutron flux to be determined from j
! measured activities L l p,,, = monitor resonance neutron flux to be determined from !
! measured activities i r
l Equations 6-4 and 6-5 are solved for p2200 ""d #res "'I"8 th' l average seasured bare and cadmius shieldad cobalt activities at the f j nonitor location. >
l h
i l
l I
t
! .-io 1
._. _ _ . _ - _ _ - . - . _ . . - - - , _ . - - - - , _ . . . - - _ _ _ , . - _ - _ _ _ _ _ . - . _ - . . _ , - , , . _ _ . . . . _ - , _ . _ . . _ . _ _ . - - -- _ _ _.L
l l
The total loss rate of a product nuclide any then be expressed as the sua of its radioactive decay rate si.d the neutron absorption rate in that nuclide while the reactor is at power. The product nuclide neutron absorption rate any be estimated from the published data for a
2200 and II and the smaller fluxes determined above. If the neutron absorption rate is small when compared to the decay rate then there is no concern regarding burnout.
6.4 NEUTION TRANSPORT ANALYSIS 118 ULT 5 j Results of the discrete ordinates transport calculations for the )
V. C. Summer Unit I reactor are sunnarised in this section. Calculated 1 i f ast (E > 1.0 WeV) neutron exposure results are presented in Tables 6-3 through 6-9. Date are presented at several asinuthal locations at the
, inner radius of the reactor vessel as well as at the center of each surveillance capsule.
Tables 6-3 through 6-7 list plant specific naximum fast neutron
, flux levels at O', 15*, 20', 30', and 45' at the reactor vessel inner radius for the first three fuel .i.les, and projected to the expiration
, . date (March 21,2013) of the current operating license and a reference
, date of 32 EFPY, Plant specific beltline cumulative fluence levels for the three completed fuel cycles and design basis cumulative fluence s
levels based on a design basis 3 loop core power distribution (at the l
nominal plus 2 signa level) are presented for each completed fuel cycle.
Similar data for the center of the surveillance capsules loented at 16.94 and 19.72 are given in Table 6-8 and 6-9, respectively. The measured f ast neutron fluence for surveillance capsule U (withdrawn at the end of cyclt: 1) and surveillance capsule V (withdrawn at the end of cycle 3) is also presented in Table 6-9 for comparison with analytical results (see Section 6-5). The maxiana neutron flux levels reflect a core axial power distribution peak to average *stio of 1.20.
It should be noted that implementation of a more severe low leakage pattern would act to reduce the projectione of fluence at key
, locations. On the other hand, relaxation of the current low leakage e-11
1 l
l t
patterns or a return to out-in fuel management would increase those projections. ,
Table 6-10 provides the relative radial variation of f ast neutron flax or fluence through the reactor vessel at sero and forty l five degrees.
l Figure 6-3 provides the relative axial variation of f ast neutron l flux and fluence over the beltline region of the reactor vessel. A 1
three dimensional description of the fast neutron exposure of the reactor vessel wall can be constructed using the data given in Tables 6-4 through 6-10 and Figure 6-3 along with the relation.
1 p(R,8,Z) = p(8) F(R) G(Z) where:
p(1,8,Z) = f ast neutron fluence at location R,8,3 within the reactor vessel wall '
p(8) = f ast neutron fluence at asinuthal location 8 on the reactor vessel inner radius from Table 6-4 through ,
6-7 F(R) = relative f ast neutron flux at depth R into the reactor vessel from Table 6-10 ,
G(Z) = relative fast neutron flux at axial position Z from ,
Figure 6-3 Analysis has shown thst the radial and axial variations within the reactor vessel wall are relatively insensitive to the implementation of low leakage fuel management schemes. Thus, the above relationship provides a vehicle for a reasonable evaluation of fluence gradients ,
within the reactor vessel wall. !
In order to derive neutron flux and fluence levels from the ,
measured disintegration rates, suitable spectrum-averaged reaction cross sections are reqaired. The calculated neutron energy spectrum at the ,
e-is
centar of the V. C. Summer surveillance capsules, listed in Table 6-12
- , was taken from the forward calculation. The resulting calculated l spectrum-averaged cross sections for each of the five fast neutron reactions are given in Table 6-13.
J 6.5 INFLUENCE OF AN ENB1GY DEPENDENT DAMAGE MODEL The use of f ast (E > l 1 MeV) neutron fluence to correlate measured material property changes to the neutron exposure of the materials for light water reactor applications has traditionally been '
accepted for development of damage trend curves as well as for implementation of trend curve data to assess reactor vessel condition.
In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule location and positions within the reactor vessel wall could lead to a reduction in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage
- . gradients through the reactor vessel wall. ;
Because of this potential shift away from a threshold fluence towards an energy dependent damage function for data correlation, ASTM E-853, ' Standard Practice Analysis and Interpretation of Light Water Reactor Surveillance Results', recommends reporting calculated displacements per iron atos (dpa) in addition to f ast neutron fluence to provide a dhts base for future reference. The energy dependent dpa I
function to be used for this evaluation is specified in ASTM E-693, l
' Standard Practice for Characterizing Neutron Exposures in Ferritic l Steels in Terms of Displacement per Atom (dpa)'. [
.i For the V. C. Summer Unit i reactor vessel, iron aton displacement rates at each surveillance capsule location and at 1 positions within the reactor vessel wall have been calculated. The analysis has shown that for a given location the ratio of dpa/ flux
- , (E > 1.0 VeV) is inrensitive to changing core power distribution. That i is, while implementation of low leakage loading patterns significantly j . inpact the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum at a given location are i 6-13
of second order. The dpa/ flux (E > 1.0 MeV) ratios calculated for key locations in the V. C. Susser Unit i reactor geometry are given in Table .
6-11. The data in Table 6-11 any be used on conjunction with the fast l neutron fluence data provided in Section 6-4 to develop distributions of j dpa within the surveillance capsules and the reactor vessel. Also provided ja Table 6-11 we ratios of flux (E > 0.11 MeV) to flux (E >
1.0 MeV) calculated for key locations in the V. C. Summer reactor e
geometry.
6.6 NEUT10N D0SIMETRY BESULTS The irradiation history of the V. C. Summer Unit I reactor is given in Tables 6-14 and 6-15 for surveillance capsules U and Y respectively. The data were obtained from NUREG-0020.(23) All of the radiometric moniters are located at the radial center of the surveillance capsule.
Comparisons of measured and calculated saturated activity of the radiometric monitore contained in Capsule U and Capsule V are given in .
I Tables 6-16 and 6-10 respectively. Note that the results presented for
! capsule U are an update of earlier work (24) based upon the current plant .
[
l specific analysis, f l
, The f ast (E > 1.0 MeV) neutron flux and fluence levels derivad l for capsule U and capsule V using the spectrum averaged c.ross sections l listed in Table 6-13 are given in Tables 6-17 and 6-20 respectively.
j Table 6-18 and 6-21 summarise the key nuclear data and results of the ,
i <
l product nuclide burnout assessments that were performed for capsule U ;
and capsula V. Due to the relatively low thermal and resonance neutron l !
fluxes at the surveillance capsule location, the neutron absorption rate j is negligibly small when compared to the radioactive decay rate. !
l Therefore, no correction has been made for product nuclide burnout.
An examination of Table 6-17 shows that the average fast (E >
1.0 MeV) neutron flux derived from the five threshold reactions for , [
2 capsule U is 1.795E+11 n/cm -see with a la standard deviation of *6.6 !
2 percent. The calculated flux value of 1.692E+11 n/ca -see based on the ,
plant specific core power distribution is within six percent of the <
average value derived from the measurements.
i 6-14
... - ~_ _ _. _ _, , _ . -_ _ -___ _ ,_ _ ._ _ _ _ _ _ .
An examination of Table 6-20 shows that the average f ast (E > .
~
1.0 MeV) neutron flux derived from the five threshold reactions for 2
capsule V is 1.600E+11 n/cm -see with a la standard deviation of *8.9 percent. The calculated flux value 1.5598+11 n/cm 2-see based on the plant specific core power distribution is within three percent of the l average value derived from the measurements. .
A summary of measured and calculated current f ast neutron exposures for capsule V and for key reactor vessel locations is presented in Table 6-22. The measured value is given based on the ;
average of all five threshold reactions listed in Table 6-20. End of life (BOL) reactor vessel f ast neutron fluence projections are also included in Table 6-22. l Based on the data given in Table 6-20, the best estimate f ast neutron exposure of Capsule V is l 4 = 1.47E+19 n/cm (E > 1.0 MeV) at 2.93 EFPY.
[
l 4
e-15
l l
l l
9 Table 6-1 SAILOR 47 Neutron Energy Group Structure .
Group Group Energy Lower Energy Energy Lower Energy i Group (VeV) Group (VeV) 1 14.19(*) 25 0.183 .
1 2 12.21 26 0.111 4 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 7.108-3 11 2.47 35 3.368-3 12 2.37 36 1.59B-3
- 13 2.35 37 4.54E-4 14 2.23 38 2.14E-4
. 15 1.92 39 1.015-4 ,
t 16 1.65 40 3.735-5 ;
17 1.35 41 1.075-5 ;
i, 18 1.00 42 5.04E-6 :
t j
19 0.821 43 1.86E-6 ,
20 0.743 44 8.768-7 l
- 21 0.608 45 4.148-7
) 22 0.498 46 1.008-7 -
23 0.369 47 0.00 24 0.298 .
(*)The upper energy of group 1 is 17.33 WeV.
I s-te ,
i
Table 6-2 Nuclear Constants for Radiometric Monitors Contained in the V. C. Summer Unit 1 Surveillance Capsules 2 Reaction Target Fission of Weight Product Yield Monitor Waterials Interest Fraction Half-life (4)
Iron wire fen (n,p) Mn54 0.058 312.2 dy Nickel wire NiS8 (n,p) CoS8 0.6827 70.91 dy Copper wire Cu63 (n,a) Co60 0.6917 5.272 yr
' 137 Uranium-238 in U 380 ( ) U 38 (n,f) Cs 1.0 30.17 ,vr 6.0 Neptunium-237 in Np0 (*) Np237(n3f) Cs I37 1.0 30.17 yr 6.5 Cobalt-aluminum wire *) Co 59 (n,7) Co60 0.0015 5.272 yr 60 Cobalt-aluminum wire Co S9 (n,7) Co 0.0015 5.272 yr
(*) Denotes that the monitor is cadmium-shielded.
t d
l l
i l ,
6-17
Table 6-3 Fast (E > 1.0 MeV) Neutron Exposure at the Reactor ;
Vessel Inner Radius - Asinuthal Angle of O' i
Beltline Region '
Cumulative 2 t Irradiation Cycle Average Fluence (n/cm ) Cumulative Cyale Time Fly Plant Desisy Basis ,)
Time !
No. (EFPS) (n/cm -sec) Specific (EFPY) ,
1 3.567E+07 5.55E+10 1.985+18 2.35E+18 1.13
. 2 2.115E+07 4.28E+10 2.88E+18 3.75E+18 1.80 i 3(b) 3.550E+07 4.22E+10 4.38E+18 6.10E+18 2.93 EOL2013(*) 7.799E+08 4.25E+10(d) 3.32E+19 5.76E+19 26.51 EOL 2022 1.052E+09 4 25E+10(d) 4.47E+19 1.27E+20 35.14 I l
) .
(*) Design Basis Flux: 6.60E+10 2 n/cm ,,,
s (b)End of Cycle 3 (Shutdown): March 6, 1987 i
(*)Present Operating Licenen Expires: March 21, 2013 .
l (d) Time Weighted Average Neutron Flux for Projections i
l-t
, l I
i i
f i j
l I
i i
I
' 6-18 ,
I i
p h
g ,
i o
Table 6-4 i Fast (E > 1.0 WeV) Neutron Exposure at the Reactor i j Yessel Inner Radius - Asinuthal Angle of 15' l I
Beltline Re81on l Cumulative 2 Irradiation Cycle Average Fluence (n/ce ) Cumulative Cycle Time Figx Plant Time Deatsy,)
No. (EFPS) (n/cm -see) [ggsiQ9 Basis (EFPY) [
1 3.567E+07 3.62E+10 1.29E+18 1.58E+18 1.13 2 2.115E+07 3.04E+10 1.93E+18 2.53E+18 1.80 l 3(b) 3.550E+07 2.99E+10 3.00E+18 4.11E+18 2.93 1
EOL2012(*) 7.799E+08 3.01E+10(d) 2.35E+19 3.88E+19 26.51 EOL 2022 1.052E+09 3.01E+10(d) 3.16E+19 8.57E+19 35.41 I
(*) Design Basis Flux: 4.45E+10 n/cm 2 ,,,,
j (b)End of Cycle 3 (Shutdown): Warch 6, 1987 '
(*)Present Operating License Expires: March 21, 2013 ;
(d) Time Weighted Average Neutron Flux for Projections l
i i
. r i
{
Y j e j i
l.
! l l~
t
- .-2 l
Table 6-5 Fast (E > 1.0 MeV) Neutron Exposure at the Reactor Yessel Inner Radius - Asinuthal Angle of 20' i
Beltline Region :
Cumulative 2 Irradiation Cycle Average Fluence (n/ca ) Cumulative :
Cycle Time Fly Plant Desisy Time No. (EFPS) (n/cm -sec) Soecific Basis ,) (EFPY) ,
! 1 3.567E+07 3.0d+10 1.08E+18 1.35E+18 1.13 2 2.115E+07 2.87E+10 1.69E+18 2.15E+18 1.80 i 3(b) 3.550E+07 2.66E+10 2.64E+18 3.49E+18 2.93 r
EOL 2013(*) 7.799E+08 2.74E+10(d) 2.14E+19 3.30E+19 26.51 EOL 2022 1.052E+09 2.74E+10(d) 2.88E+19 7.28E+19 35.14 !
(*) Design Basis Flux: 3.78E+10 n/cm 2 ,,,
(b)End of Cycle 3 (Shutdown): March 6, 1987 3
(*)Present Operating License Expires: March 21, 2013 .
.; (d) Time Weighted Average Neutron Flux for Projections l r
i I
j i
I J
I i I j
l l I !
t i !
- e-.o .
i
Table 6-6 Fast (E > 1.0 MeV) Neutron Exposure at the Reactor Vessel Inner Radius--Asinuthal Angle of 30' Beltline Region Cumulative Irradiation Cycle Average Fluence (n/cm2 ) Cumulative Cycle No.
Time (EFPS)
Fly (n/cm -sec)
Plant Desigg Basis ,)
Time 12151111 (EFPY) ,,
1 3.567E+07 2.39E+10 8.55B+17 1.04E+18 1.13 2 2.115E+07 2.185+10 1.31E+18 1.65E+18 1.80 3(U 3.550E+07 2.00E+10 2.02E+18 2.69E+18 2.93 EOL 2013(*) 7.7995+08 2.075+10(d) 1.61E*19 2.54E+19 26.51 EOI 2022 1.052E+09 2.07E+10(d) 2.17E+19 5.61E+19 35.14
(*) Design Basis Flux: 2.912E+10 n/es 2 ,,,,
(b)End of Cycle 3 (Shutdown): March 6, 1987
(*)Present Operating License Expires: March 21, 2013 (d) Time Weighted Average Neutron Flux for Projections l
l s-n
Table 6-7 l Fast (E > 1.0 MeV) Neutron Exposure at the Reactor Yessel -
Inner Radius--Asinuthal Angle of 45' ,
Beltline Region ,
Cumulative 2 Irradiation Cycle Average Fluence (n/en ) Cunulative Cycle Time Figx Plant Desigg Time ;
No. (EFPS) (n/cm -sec) Specific Basis ,) (EFPY) r 1 3.567E+07 1.71E+10 6.10E+17 7.40E+17 1.13 ;
I 2 2.115E+07 1.35E+10 8.96E+17 1.17E+18 1.80 1 3(b) 3.550E+07 1.35E+10 1.37E+18 1.91E+18 2.93 i EOL 2C13(*) 7.799E+08 1.35E+10(d) 1.05E+18 1.81E+19 26.51 EOL 2C22 1.052E+09 1.35E+10(d) 1.42E+19 3.99E+19 35.14 2
(*) Design Basis Flux: 2.078E+10 n/ca ,,,
j (b)End of Cycle 3 (Shutdown): March 6, 1987 ' !
) (*)Present Operating License Expires: March 21, 2013 '
i (d) Time Weighted Average Neutron Flux for Projections f i,
I, t i
1 1
h I, I 3
I l
f i . !
J :
l I
i !
I e-na
~
. _ . _ . - _ _ _ = _ _ _ _ - - . . - -- >
O Table 6-8 9
Fast (E > 1.0 MeV) Neutron Exposure at the 16.94 Degree Surveillance Capsule Center Beltline Region Cumulative Irradiation Cycle Average heyce (n/ca2) Capsule Cycle Time Fly Plant M Basis Desigg
,)
Data 2
No. (EFPS) (n/cm -see') (n/cm )
1 3.567E+07 1.69E+11 6.03E+18 7.45E+18 6.39E+18:U 2 2.115E+07 1.52E+11 9.26E+18 1.18E+19 3(b) 3.550E+07 1.46E+11 1.44E+19 1.92E+19 1.47E+19:V EOL 2013(c) 7.799E+08 1.48E+11(d) 1.16E+20 1.82E+20 EOL 2022 1.052E+09 1.48E+11(d) 1.56E+20 4.02E+20 l
l
. (*) Design Basis Flux: 2.09E+11 n/ca 2 ,,,,
(b)End of Cycle 3 (Shutdown): March 6, 1257 l
(*)PresentOperatingLicenseExpires: March 21, 2013 l (d) Time Weighted Average Neutron Flux for Projections l
l l
l 9
5 e=33
~
Table 6-9 Fast (E > 1.0 MeV) Neutron Exposure at the 19.72 Degree Surveillance Capsule Center Beltline Begion l Cumulative Irradiation Cycle Average Fluence (n/ca2) Cumulative Cycle Time F1p Plant Time No. MFPS) (n/cm -sec) Soecific BasisDesist .) (EFPY) 1 3.567E+07 1.46E+11 5.21E+18 6.44E+18 1.13 2 2.115E+07 1.38E+11 8.14E+18 1.02E+19 1.80 3(b) 3.550E+07 1.28E+11 1.26E+19 1.66E+19 2.93 EOL 2013(*) 7.799E+08 1.32E+11(d) 1.02E+20 1.57E+20 26.51 EOL 2022 152E+09 1.32E+11 1.385+20 4.02E+20 35.14 l
(*) Design Basis Flux: 1.81E+11 n/cm 2 (b)End of Cycle 3 (Shutdown): March 6, 1987
(*)Present Operating License Expires: March 21, 2013 '
(d) Time Weighted Average Neutron Flux for Projections
[
l i
f i
i l
9 4
1 i
I e-se
Table 6-10 Calculated Relative Fast Neutron Exposure Parameters for V. C. Summer Unit 1 Reintive Radial Variation of Fast (R > 1.0 MeV)
Neutron Flux or Fluence Through the Reactor Yessel Radial Radius Asisuthal Anale Location (ca) 0 Dearees 45 Dezrees IR 199.390 1.0 1.0 1/4 T 204.390 0.592 0.601 1/2 T 209.391 0.294 0.305 3/4 T 214.392 0.138 0.147 01 219.392 0.0567 0.0669 G
m O
e-s.
l
Table 6-11 ,
I i
1 Ratios of Fast Neutron Exposure Parameters to Fast r (E > 1.0 MeV) Neutron Flux for the Reactor i 1 Vessel and Surveillance Capsules i Fast (E > 1.0 MeV) Iron Displacement t Neutron Flux Ratio Rate (dos /see' Ratio >
M 0 Dem. 45 Dem. O Dem. 44 Dez.
RV IR 2.42 1.94 1.598-21 1.55E-21 i
RV 1/4 T 3.61 2.96 1.825-21 1.70E-21 L i
RV 1/2 7 5.30 4.51 2.275-21 2.13E-21 i
RV 3/4 T 7.38 6.02 2.85E-21 2.44E-21 ,
RV OR 9.10 8.23 3.37E-21 3.20E-21 (
i 17' Capsule 5.00 2.1365-21
. 20' Capsule 4.86 2.0998-21 l
r 2
y,gg,, n/en -see (E > 0.1 MeV) dos /see 2 2 -
- n/ca -see (E > 1.0 WeV) n/ca -see (E > 1.0 MeV) f -
1 c
? (
I i t i !
- t
! i
- t 1
i I i l
. t
- i l
I e-..
Table 6-12 Calculated Neutron Energy Spectra at the Center of the V. O. Summer Unit 1 Surveillance Capsule Y Energy Neutrgn Flux Energy Neuttgn Flux Croup (n/cm -see) Croup (n/es -see) 1 3.00E+07 25 1.29E+11 2 1.11E+08 26 1.42E+11 3 3.89E+08 27 1.18E+11 4 7.17E+08 28 7.38E+10 5 1.21E+09 29 1.96E+10 6 2.71E+09 30 1.04E+10 7 3.80E+09 31 3.16E+10 l 8 7.95E+09 32 2.25E+10 9 7.46E+09 33 2.96E+10 10 8.27E+09 34 3.63E+10
. 11 7.61E+09 35 6.65E+10 12 3.81E+09 36 6.74E+10 13 1.16E+09 37 9.04E+10 14 5.90E+09 38 4.54E+10 15 1.64E+10 39 5.04E+10 16 2.25E+10 40 6.95E+10 17 3.60E+10 41 8.04E+10 18 8.62E+10 42 4.26E+10 19 6.58B+10 43 4.37E+10 20 3.06E+10 44 2.41E+10 21 1.21E+11 45 1.61E+10 22 9.51E+10 46 1.68E+10 23 1.28E+11 47 1.68E+10 24 1.29E+11
/
e l
e-ar
L l
Table 6-13 Spectrum Averased Reaction Cross Section At the :
Center of the V. C. Summer Unit 1 Surveillance Capsules Spectrun-Aver i l
Cross Section ;
l Reaction of Interest (barns) l Fe" (n,p) Mn" 0.0515 Ni&S (n,p) Co58 0.0726 i Cu63 (n,a) Co60 0.000428 U238 (n,f) Cs137 I 0.301 l Np237 (n,f) Cs137 3.43
(") "
a(E) p(E) dB
& = ,0" ,
! a(E) dB 4 '1 MeV -
1
[
i'
[
\
l l i
l ;
i l
t I
i I
-l :
I -
l \
\ j i r 6
i e-aa
)
i l
u l
Table 6-14 V. C. Su=mer Unit 1 Po,er History, Capsule U from NUREG-0020 Reactor (MWT) 2775 NSSS Power (MWT) 2785 -
Cycle 1 Date Cross Therent Energy (VWR) m_ g, Wonthly Lifeties -
11 82 143789 143789 l 12 82 651071 794860 -
l 1 83 975351 1770211 2 83 844552 '
2614763 3 83 566778 3181541 4 83 0 3181541 5 83 265111 3446652 6 83 1673672 5120324 7 83 1863508 6983832 8 83 1758090 8741922 9 83 1749405 10491327
. 10 83 1810240 12301567 11 83 1518120 13819687 12 83 800616 14620303 1 84 1996935 16617238 2 84 1666309 18283547 3 84 1459743 19743290 4 84 119028 19862318 5 84 1704230 21566548 6 84 1847339 23413887 7 84 833316 24247203 8 84 1844750 26091953 9 84 1405054 27497007 Cycle 1 part=eters Startup Nov. 16, 1982 Shutdown: Sept. 28, 1984 EFPY = 1.13 EFPS = 3.567E 07 Capsule U Removed 6-29
~
Table 6-15 V. C. Summer Unit 1 Power History, Capsule V from NUREC-0020 Reactor (MWT) 2775 i NSSS Power (Wirt) 2785 Cycle 1 Date Cross Thersal Enerrr (VfB)
Monthly L<,fetime
- 5. Z.
11 82 143789 143789 12 82 651071 794860 1 83 975351 1770241 2 83 844552 2614763 3 83 566778 3181541 4 83 0 3181541 5 83 265111 3446652 6 83 1673672 5120324 7 83 1863508 6983832 8 83 1758090 8741922 9 83 1749405 10491327 10 83 1810240 12301567 . ,
11 83 1518120 13819687 I 12 83 800616 14620303 1 84 1996935 16617238 .
2 84 1666309 18243547 3 84 1459743 19743290 4 84 119028 19862318 5 84 1704230 21566548 i 6 84 1847339 23413847 7 84 833316 24247203 J
8 84 1544750 26092953 9 84 1405064 27497007 :
- Cycle 1 Parameters i Startup
- Nov. 16, 1982 r Shutdown: Sept. 28, 1984 EPPY = 1.13 EFPS = 3.567E+07 i
I G-30
Table 6-15 (cont'd.)
Cycle 2 Date Gross Thermal Ener m WWE)
Monthly .r,fetime L L 12 84 455882 27952449 1 85 1515633 29464522 2 85 1578246 31046808 3 85 1896308 32943116 4 85 1682110 34625226 5 85 1256696 35481922 6 85 1989561 37871483 7 85 2053141 39924624 8 85 1742390 41667014 9 85 1836663 43503677 10 85 292854 43796531 Cycle 2 Parameters
, Startupt Dec. 19, 1984 l Shutdown: Oct. 5, 1985 1
. EFPY = 1.80 EFPS = 2.11E+07 l .
1 l
l 1
l 9
8-31
Table 6 15 (cont'd.)
Cycle 3 .
h Cross Thermal Ener r (WWE) g, g, Monthly L;,fetime 12 85 661338 44487869 1 86 2039156 46497025 2 86 1638382 48135407 3 86 2061270 50196677 4 86 1898100 52094777 5 86 2021310 54116047 6 86 1461204 55877291 7 86 1898412 57476103 8 86 2037220 59513323 9 86 1957929 61471252 10 86 1835749 63307001 11 86 1862802 65169803 12 86 1844820 67014623 1 87 1977210 68991833 2 87 1854160 70829993 3 47 333666 71163659 Cycle 3 Parameters .
Startup: Dec. 16, 1985 Shutdown: March 6, 1987 ETPY = 2.93 ETPS = 3.55E+07 l Capsule Y removed 4-33
Table 6-16 Comparison of Measured and Calculated Radiosotric Monitor Saturated Activities for V. C. Summer Unit 1 Surveillance Capsule U Radiometric Monitor !
Saturated Activity Monitor and (Disinterrations/Second-Graa Axial Location (*) Weasured Ragig Calculated Q ,
Fe54(n.o1Wn54 (,,og ,g,,) f Top 5.514E+6 Middle 5.460E+6 Botton 5.929E+6 l
Average (b) 5.634E+6(*4.6%) 5.652E+6 1.003 f 58 gn,,3c,58 N1 (,, ,g ,g,,)
Top 9.173E+7 Widdle 8.674E+7 Botton 9.363E+7 !
Average (b) 8.726E+7 0.962 9.070E+7 (*3.9%)
Cu63(n.a)Co60 [,, ,g ,g,,)
Top 5.568E+5
- Widdle 5.491E+5 Bottes 5.728E+5 Ayarase(b) 5.596E+5 (*2.2%) 4.798E+5 0,857 I
U238,n,f)CsI3I(*) (gs of wire) !
Midd;,e 1.010E+7 i Corrected (d) 8.059E.6 7.750E+6 0.962 ;
I N 237g ,, g)c,137(c) (,, ,g ,g,,) j le 9.489E+7 9.616E+7 1.013 l 1 c I
I e
e=33 l
l
Table 6-16 (cont'd)
Radiometric Monitor Saturated Activity Monitor acd (Disintemrations/Second-Gran Asial Location (*) Measured M Calculated M Co89(n.1) Co60(c) (,, ,g ,g,,)
Top 6.800E+7 Middle 6.922E+7 Botton 7.028E+7 Average b) 6.913+7 (*1.6%) 5.911E+7 0.855
) Co89(n.11Co60 (,,ng ,g,,)
! Top 1.235E+8 i Middle 1.269E+8
- Botton 1.2685+8 Average b) 1.25 3+8 (*1.6%) 6.936E+7 0.552 1
(*) Refer to Figure 4-2 for the location of the varioun radiometric .
nonitors, of the mean saturated activity is .
b)The expressed standard deviation as a percentage o (la)f the mean.
(*)This radiometric monitor was cadalus shielded.
(d)The measured value gg been multi lied by 0.7gggte correct for the effect of 323 ppa U andthebuIldupofru 4
e-34
[
l l
i i
Table 6-17 ;
- I Results of Fast Neutron Dosimetry for V. C. Summer Unit 1 (
Surveillance Capsule U !
Current b)
Radiometric Monito{,) Fast (E > 1.0 MeV) Fast E : 1.0 MeV) l Saturated Activity NeuttgaFlux !
Reaction of (dos /ma'- (n/ca -soci Neutron (n/en {)1uence ,
Interest Measured Ca;,culated Measured Calculated Measured Calculated !
l fen (n,p) man 5.634E+6 5.652E+6 1.749E+11 6.240E+16 !
I Ni5t(n,p)Co64 9.070E+7 8.726E+7 1 783E+11 6.362E+18 f Cu63(n a)Co60 5.596E+5 4.798E+5 1.995E+11 7.115E+18 f U238(n,f)Cs137 8.059E+6 7.750E+6 1.7768+11 6.301E+18 !
Np237(n,f)Cs137 9.489E+7 9.6165+7 1.675E+11 5.9768+18 l
Average 1.795E+11 1.692E+11 6.3988+18 6.034E+18
(*6.65]
l i i I") Refer to Table 6-16.
b) Total irradiation time for surveillance Capsule U is 3.567E+07 offective l full power second (EFPS).
i l
l !
t l
l
' \
i i
l !
l' l l .
l l l l
l e-u j 1 \
(- i I'
1 [
I i ,
L Table A-18 :
l Product Nuclide Burnout Assessment for Y. C. Summer t Unit i Surveillance Capsule U !
Nuclear Data i 2200 RI J Mulide IglL-Idi.E 1hACAE1 M l Ma N 312.2 dy 10.0 -
88
- Co 70.91 dy 1880 8490 f
I i Co ' Stable 37.2 75.5 !
l Co" 5.272 yr 2.0 4.3 f III
! Cs 30.17 yr 0.11 0.50 I Surveillance Caosule U Avernaed Saturated Co60 Activity f a 1
- Bare Co-Al Vire: A = 1.265+08 dps/ga l l [
Cd Shielded Co-Al Vire: A = 6.92E+07 dps/ga
- l f
'a Monitor Fluxes Derived Free Co 60 Saturated Activity 2
i p(2200) = 9.925 10 n/ca ,,,, ;
i 2
(
p(RES) = 5.982+10 n/ca ,,,, [
Froduct Nuclide Loss late Connarison '
r
? Decay Absorption I constaat Rate l 3g3,183 (1/see) (1/see) [
54 l Mn 2.57E-04 9.925-13 64 1.13E-07 5.982-10 Co
(
Co" 4.17E-09 4.55E-13 [
! 137 I 1
Cs 7.28E-10 2.99E-14 .
l i -
1 t
e-se .
l
- . . . _ - _ _ ~ , - . , .
Table 6-19
- Compariser of Weasured and Calculated Radiometric Monitor Saturated Activities for V. C. Su=ser Unit 1 Surveillance Capsule V Radiometric Monitor Saturated Activity Wonitor and (Di sir.t erration s /S e cond-Cram Arial Location (*) Weasured Basis Calculated M F 54cn , ,) yn 54 (ga of wire)
Top 5.177E+6
' Widdle 5.351E+6 Bottos 3.056E+6 Average (b) 5.195E+6(*2.9%) 5.235E+6 1.008 58 gn,,)c,58 N1 (ga of wire)
Top 7.939E+7 Widdle 8.084E+7 Botton 7.615E+7 Average b) 7.879E+7 (*3.0%) 8.082E+7 1.026 63 gn,,)e,60
. Cu (Em of wire)
Top 5.098E+5 Widdle 5.237E+5
+ Botton 5.098E+5 Average (b) 5.144E+5 (*1.6%) 4.444E+5 0.864 U238(n.f)CsI37 (*) (ga of wire)
Widdle 9.540E+6 Corrected (d) 7.090E+6 7.178E+6 1.012 N 237gn,f)c,137(c) (gn of wire) i le 8.206E 7 8.907E+7 1.083
+
6-37
Table 6-19 (cont'd)
Radiometric Monitor Saturated Activity Monitor and (Disinterrations/Second-Graal Arial Location (*) Measured M Calculatte M Co 49g ,,,)e,M(c) (,, ,g ,gg,)
Top 7.040E.7 Middle 6.381E+7 Bottos 6.341E+7 Average O) 6.601B+7 (*5.8%) 5.475E+7 0.829 Co69 (n .11 Co60 (g, og ,g,,)
Top 1.120E+8 Middle 1.169E+8 Botton 1.155E+8 Average O) 1.144E+8 (*2.2%) 6.424E+7 0.560
(*) Refer to Figure 4-2 for the location of the various radionotric monitors.
(b)The studard deviation (la) of the mean satursted activity is expressed as a percentage of the sema.
(*)This radiometric monitor was cadstua shielded.
(d)The nessured value gp been multi lied by 0.jg to correct for the effect of 232 pa U2 andthebui$dupofFu e=38
Table 6-20
. I Results of Fast Neutron Dosisotry for V. C. Summer Unit 1 Surveillance Capsule Y Current (b)
RadiosotricWonito5 Fats E > 1.0 WeV)
Saturated Activity ") Fast Neutran (E > 1.0Flux WeV) Neutron Eluence Reaction of (dp:/1st (n/es*-see) (n/es*)
Interest Weasured Ca culated Weasured Calculated Weasured Calculatej Fest(,,p)g,54 5.195E+6 5.235E.6 1.613E+11 1.489E+19 Ni38(n.p)C058 7.879E+7 8.082E+7 1.549E+11 1.430E+19 Cuo3(,,4g,60 5.144E+5 4.444E+5 1.834E+11 1.693E+19 U238(n,f)Cs137 7.088E+6 7.178E+8 1.554E+11 1.435E+19 Sp 37(n,f)Cs137 8.226E+7 8.907E+7 1.452E+11 1.341E+19 Average 1.600E+11 1.559E+11 1.477E-19 1.44E+19
(*8.9%)
(*) Refer to Table 6-19.
I ) Total irradiation time for surveillance Capsule V is 9.232E+07 ef fective full power second (EFFS).
1 6-39
5 i
1 !
i
- Table 6-21
1 Product Nuclide Burnout Assessment for V. C. Sunser Unit 1 i i
Surveillance Capsule Y [
Nuclear Data i i
1 2200 RI i
- Muslidt tall-MLA ikAEEAl IAEAll [
Ma N 312.2 dy 10.0 -
Co 88 70.91 dy 1880 8490 f
l !
. Co&9 Stable 37.2 75.5 I Co" 5.272 yr 2.0 4.3 137 j Cs 30.17 yr 0.11 0.50 f i
Surveillance Cassule Y Avermae Saturated Co" Activity i Bare Co-Al V!.re: A = 1.15E+08 dps/ga
! Cd Shielded Co-Al Vire: A=6.60E+07 dFs/gu *i l !
l Monitor Fluzes Derived from Co " Saturated Activity l
((2200) = 8.565+10 n/ce2 7,,, l 2
((RES) = 5.70E+10 n/ca j,,,
A J 5 Product Nuclide Loss Rate Cossarison ;
l i j
Decay Absorption !
j Constaat Rate b i
M (1/see) (1/see) 54 i Mn 2.575-04 8.545-13 ;
) 68 Co 1.135-07 5.542-10 60 Co 4.17E-09 4.16E-13 i Cs I3I 7.28E-10 2.855-14 . .
l 1 . l 1 i i r
! I
+
4
- e-4o ,
3
j ..
Table 6-22 Summary of V. C. Summer Unit 1 Fast Neutron Fluence Results Based on Surveillance Capsule V End of Life Current Fast (E > 1.0 MeV) Fast (E > 1.0 MeV)
Neutron {luence NeutronFlyence) n/cm n/cm Location Weasured Calculated Weasured Calculated Capsule V 1.47E+19 1.44E+19 Vessel IR 4.72E+18 4.63B+18 6.73E+19 6.60E+19 Vessel 1/4 T 2.80E+18 2.75E+18 3.99E+19 3.92E+19 Vessel 3/4 T 6.50E+17 6.37E+17 9.27E+18 9.08E+18
(*) Current fluences are based on operation at 2775 MWt for 2.93 ".FPY.
(b)EOL fluences are based on operation at 2775 MWt for 32 EFPY.
(c)The measured results of surveillance Capsule V were extrapolated to the reactor vessel locations using the following-calculated lead factors:
Inner Radius - 3.11
, 1/4 Thickness - 5.24 3/4 Thickness - 22.6 9
O 6-41
1 16.94 DEG. (CAPSULES U, V, X) .
00 19.72 DEG. (C PSULES W. Y, Z) g I
i REACTOR VESSEL 450
\\\\\x /
s-,
/
/
/
I :
I :
l s /
l '
I : -
/
I
/
l / .
I
-l/
/
l / <
e l
/
l /
f /
l
/
t /
/
I e /
l /
I /
b k___________________________
Figure 6-1. V. C. Summer Unit 1 reactor geometry 6-42
- 16.94 DEG. - 19.72 D EG.
T 4 T
7 -
, - 73.31 IN.
s I '
NEUTRON PAD l l
I l I i Figure 6-2. Plan view of a dual reactor vessel surveillance capsulo 6-43
10 0 8 -
6 -
4 -
2 -
5 10-1 -
~
u'. 8 -
Z 6 -
8 -
g 4 -
y .
$ 2 -
P 5
E 10-2 _, ,
8 -
6 - ~
l 4 -
CORE MIDPLANE 2 - TO VESSEL l
CLOSURE HEAD I ! ! ! !
10-3
-300 -200 -100 0 100 200 300 400 l
l DISTANCE FROM CORE MIDPLANE (cm) l I
l Figure 6-3. Relative axial variation of fast (E > 1.0 MeV) neutron flux within the reactor vessel wall .
6-44 l \
l .
i 7. SURVEILLANCE CAPSULE REMOVAL SCHEDULE l
l The following removal schedule is based on ASTM E185-82 and is recommended for future capsules to be removed from the V. C. Summer Unit i reactor vessel:
Vessel Estinated Location Lead Fluen e Capsule (der.) Factor Resoval Time (") (n/cm )
18 U 343 3.11 1.13 (Removed) 8.39 x 10 19 Y 107 3.11 2.93 (Removed) 1.47 x 10 X 287 3.11 6 3.85 x 10 19(b)
W 110 2.69 12 6.65 x 10 19 (")
, Y 290 2.69 20 11.10 x 10 19 Z 340 2.69 Standby -
(*) Effective full power years from plant startup.
(b) Approximate fluence at 1/4 thickness reactor vessel wall at end of life.
(*) Approximate fluence at reactor vessel inner wall at end of life.
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- 8. REFERENCES
- 1. J. A. Davidson and S. E. Yanichko, "South Carolina Electric and Gas Company, V. C. Summer Nuclear Plant Unit No.1 Reactor Yessel Radiation Surveillance Program', WCAP-9234, January 1978.
- 2. Code of Federal Regulations, 10CFR50, Appendix G, ' Fracture Toughness Requirements', and Appendix H, 'Rev.ctor Vessel Material Surveillance Program Requirements', U.S. Nuclear Regulatory Commission, Washington, D.C.
- 3. Regulatory Guide 1.99, Proposed Revt.sion 2, ' Radiation Damage to Reactor Vessel Materials', U.S. Nuclear Regulatory Commission, February, 1986.
- 4. R. G. Soltess, R. K. Disney, J. Jedruch, and S. L. Ziegler,
' Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique', WANL-PR(LL)-034, Vol 5, August
- 1970.
- 5. '0RNL RSIC Data Library Collection DLC-76, SAILOR Coupled Self- ,
Shielded, 47 Neutron, 20 Camma-Ray, P3, Cross Section Library for Light Water Reactors'.
- 6. S. L. Anderson and K. C. Tran, WCAP-11428, ' Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology , POA Evaluations', April 1987.
- 7. A. H. Pero, WCAP-11173 (NUREG/CR-4827), ' Neutron and Gamma-Ray Flux Calculations for the Venus PWR Engineering Vockup', January 1987.
- 8. Benchmark Testing of Westinghouse Neutron Transport Analysis j Wethodology - Surveillance Capsule Data Base (to be published) .
- 9. ASTM Designation E482-82, ' Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance', in ASTM Standards, Section 12, American Society for Testing and l Waterials, Philadelphia, PA, 1986. ,
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- 10. ASTM Designation E560-84, ' Standard Practice for Extrapolating Reactor Yessel Surveillance Dosimetry Results', in ASTM Standards,
- Section 12, American Society for Testing and Materials, Philadelphia, PA, 1986.
- 11. ASTM Designation E693-79 (Reapproved 1985), ' Standard Practice for Characterising Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)', in ASTM Standards, Section 12, American Society for Testing Materials, Philadelphia, PA, 1986.
- 12. ASTM Designation E706-84, ' Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards', in ASTM Standards, Section 12, American Society for Testing and Matarials, Philadelphia, PA, 1986.
- 13. ASTM Designation E853-84, ' Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1986.
- 14. ASTM Designation E261-77, ' Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1986, e
- 15. ASTM Designation E262-85, ' Standard Method for Determining Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques',
e in ASTM Standards, Section 12, American Society for Testing and ,
Materials, Philadelphia, PA, 1986.
- 16. ASTM Designation E263-82, ' Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1986.
- 17. ASTM Designation E264-82, ' Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1986.
- 18. ASTM Designation E481-78, ' Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver', in ASTM ;
Standards, Section 12, American Society for Testing and Materials, '
Philadelphia, PA, 1986.
. 19. ASTM Designation E523-82, ' Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper', in ASTM Standards, Section 12, American Society for Testing and Materials, i
. Philadelphia, PA, 1986. l l
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- 20. ASTM Designation E704-84, ' Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238', in ASTM Standards, Section 12, American Society for Testing and Materials, ,
Philadelphia, PA, 1986.
- 21. ASTM Designation E705-84, ' Standard Method for Measuring Reaction !
Rates by Radioactivation of Neptunium-2378, in ASTM Standards, 1 Section 12, American Society for Testing and Materials, Philadelphia, PA, 1986. J
- 22. ASTM Designation E1005-84, ' Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance',
in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1986.
- 23. NUREG-0020, ' Licensed Operating Reactors Status Summary Report',
June itJ1 through October 1988.
- 24. A. H. Fero, R. S. Bogg, W. T. Kaiser, WCAP-10814, ' Analysis of Capsule U from the South Carolina Electric and Gas Company V. C.
Summer Unic 1 Reactor Yessel Radiation Surveillan.e Program', June 1985, e
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