ML20151U127
| ML20151U127 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 11/30/1987 |
| From: | Albertin L, Lippincott E, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20151U117 | List: |
| References | |
| WCAP-11651, NUDOCS 8804290266 | |
| Download: ML20151U127 (84) | |
Text
l WCAP-11651 WESTINGHOUSE CLASS 3 CUST0ER DESIGNATED DISTRIBUTION ANALYSIS OF CAPSULE U FROM THE COMMONWEALTH EDISON CD.
BYRON UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko E. P. Lippincott L. Albertin November 1987 APPROVED:
7A' b 24 T. A. Meyer, Manager Structural Materials and Reliability Technology Work performed under Shop Order No. BE0J-106 Prepared by Westinghouse Electric Corporation for the Consonwealth Edison Company i
i Although information contained in this report is nonproprietary no distribution shall be made outside Westinghouse or its licensees without the customer's approval WESTINGHOUSE ELECTRIC CORPORATION Power Systems Division P. O. Box 2728 Pittsburgh, Pennsylvania 15230 4
i 8804290266 880426 A
PDR ADOCK 05000454 P
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FREFACE This report has been technically reviewed and verified.
Reviewer Sections 1 through 5, 7 and 8 C. C. Reinecke Section 6
- 5. L. Anderson g j Mon %
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CONTENTS 1.
SUWWARY OF RESULTS...........................................
1-1 2.
INTRODUCTION.................................................
2-1 3.
BACKGROUND..................................................
3-1 4.
DESCRIPTION OF PR0 GRAM....................................'....
4-1 5.
TESTING OF SPECIMENS FROM CAPSULE U...........................
5-1 5.1 Overview...............................................
5-1 5.2 Charpy V-Notch Impact Test Results.....................
5-3 5.3 Tension Test Results...................................
5-4 5.4 Hardness Test Resu1tn...................................
5-4 5.5 Compact Tension Tests..................................
5-5 6.
RADIATION ANALYSIS AND net / TRON DOSIMETRY.....................
6-1 6.1 Introduction...........................................
6-1
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6.2 Discrete Ordinates Ann'ysis............................
6-2
{
6.3 Radiometric Monitors...................................
6-4 6.4 Neutron Transport Analysis Results.....................
6-12 i
1 6.5 Dosimetry Results......................................
6-13 7.
SURVEIILANCE CAPSULE REMOVAL SCHEDULE........................
7-1 8.
REFERENCES...................................................
8-1 i
iv
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LIST OF ILLUSTRATIONS Figure hge 4-1 Arrangement of surygillance espsules in the reactor vessel...................................................
4-6 4-2 Capsule U diagram showing location of specimens, thermal monitors and dosimeters..........................
4-7 5-1 Charpy Y-notch impact properties for Byron Unit i reactor vessel shell forging 5P-5933 (tangential orientation).... 5-14 t
l 5-2 Charpy Y-notch impact properties for Byron Unit i reactor vessel shell forging 5P-5933 (axial orientation).........
5-15 5-3 Charpy Y-notch impact properties for Byron Unit I reactor vessel weld meta1........................................
5-16 5-4 Charpy Y-notch impact properties for Byron Unit I reactor vessel weld heat affected sone metal............
5-17 5-5 Charpy impact specimen fracture surfaces for Byron Unit 1 reactor veswel shell forging 5P-5933 (tangential j
o r i e n t at i o n)............................................. 5 - 18 i
5-6 Charpy impact specimen fracture surfaces for Byron Unit i reactor vessel shell forging 5P-5933 (axial orientation).............................................
5-19 1
5-7 Charpy impact specimen fracture surfaces for Byron Unit I reactor vessel weld metal................................
5-20 5-8 Charpy impact specimen frseture surfaces for Byron Unit i reactor vessel weld heat affected sono metal.............
5-21 5-9 Tensile properties for Byron Unit i reactor vessel shell f orging 5P-5933 (tangential orientation).................
5-22
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5-10 Tensile properties for Byron Unit I reactor vessel shell i
forging 5P-5933 (axial orientation)......................
5-23 4
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f.M1 5-11 Tensile properties for Byron Unit i reactor vessel weld meta1...................................................
5-24 5-12 Fractured tensile specimens from Byron Unit I reactor vessel shell forging 5P-5933 (tangential orientation)........................................
5-25
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I 5-13 Fractured tensile specimens from Byron Unit i reactor vessel shell forging 5P-5933 (axial orientation)........
5-26 4
5-14 Fractured teasile specimens from Byron Unit i reactor vessel weld meta1.......................................
5-27 5-15 Stress-strain curve for tension specimen AL3............
5-28 6-1 Plan view of a dual reactor vessel surveillance Capsule.................................................
6-29 6-2 Calculated asinuthal distribution of maximum fast (E > 1.0 WeV) neutron flux within the reactor vessel-surveillance capsule geometry...........................
6-30
.i 6-3 Calculated radial distribution of maximum f ast (E > 1.0 MeV) neutron flux within the reactor vessel....
6-31 6-4 Reintive axial variation of f ast (E > 1.0 MeV) neutron flux within the reactor vesse1..........................
6-32 1
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i LIST OF TABLES 4
Table Pagg 4-1 Chemical Composition and Heat Treatment of the Byron Unit 1 Reactor Vessel Surveillance Materials............
4-3 4-2 Chemical Composition of Byron Unit 1 Capsule U Irradiated Charpy Impact Specimens......................
4-4 4-3 Byron Unit i Reactor Yessel Toughness Data..............
4-5 5-1 Charpy V-Notch Impact Data for the Byron Unit 1 Shell Fo{ging5g-5933.Trradiatedat550'FFluence3.50x 10 n/cm (E > 1. 0 M eV)................................
5-6
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5-2 Charpy V-Notch Impact Data for the Byron Unit 1 Reactor
.i VessalWeldMetalgndHAjMetalIrradiatedat550*F g
Fluence 3.50 x 10 n/cm (E > 1. 0 MeV).................
5-7 5-3 Instrumented Charpy Impact Test Results for Byron Unit 1 ShellForgingg5P-59g3Irradiatedat550*FtoaFluence of 3.50 x 10 n/cm (E > 1.0 MeV).......................
5-8 1
5-4 Instrumented Charpy Impact Test Results for Byron Unit 1 WeldMetalandHAZWQalIrisdiatedat550*Ftoa i
Fluence of 3.50 x 10 n/cm (E > 1.0 MeV)...............
5-9 18 2
5-5 Effect of 550'F Irradiation at 3.50 x 10 n/cm i)
(E > 1.0 MeV) on Notch Toughness Properties of Byron Unit 1 Reactor Vessel Materials..........................
5-10 1
3-6 Comparison of Byron Unit 1 30 ft-lb Transition Temperature Results with Proposed Regul6 tory Guide 1.99 Revision 2 Predictions..............................................
5-11 i
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.i P gg Table 3
5-7 Tensile Properties for Byron Unit 1 Reactg Vess51 i
Mrd.orial Irradiated at 550*F to 3.50 x 10 n/ca i
(E > 2.0 MeV)...........................................
5-12 I
5-8 Eardness (IRB) of the Byron Unit 1 Reactor Vessel
^
Sugeillatee Materials Irradiat.d at ss F to 3.s0 x 10 n/cm (E > 1. 0 Me V)...............................
5-13 l
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6-1 SAILOR 47 Neutron Energy Group Structure................
6-15 i
i 6-2 Nuclear Constants for Radiometric Monitors Contained in i
the Byron Unit 1 Surveillance Capsules...................
6-16 j
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6-3 Calculated Fast Neutron Exposure Parameters for the Peak j
Location of the Byron Unit 1 Reactor Vessel..............
6-17 1
6-4 Cale.ulated Fast Neutron Exposure Parameters and the Lead Factors for the Byron Unit 1 Surveillance Capsules.......
6-18 6-5 Calculated Neutron Energy Spectrum at the Center of.
j Byron Unit 1 Surveillance Capsule U......................
6-19 i
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6-6 Irradiation History of Byron Unit 1 Surveillance j!
Capsule U...............................................
6-20 6-7 Measured Radionotric Monitor Activities and Reaction i
1 Rates for Byron Unit 1 Surveillance Capsule U............
6-21 i
i 6-8
' "ults of Neutron Dosimetry for Byron Unit 1 Surveillance I
wapsule U...............................................
6-23 6-9 FERRET-SAND II Results for Byron Unit 1 Surveillance Capsule U...............................................
6-24 6-10 Comparison of Measured and Calculated Reaction Rates j
Used in the Analysis of Byron Unit 1 Capsule U...........
6-26
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i 6-11 Integral Neutron Flux Results Derived from Adjusted Spectrum................................................
6-27 i
6-12 Summary of Byron Unit 1 Fast (E > 1.0 MeV) Neutron i
Fluence Results Based Upon Surveillance Capsule U.......
6-28 f
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SUMMARY
OF RESULTS i
'The analysis of the reactor vessel material contained in surveillance Capsule U, the first capsule to be removed from the Commonwealth Edison Company Byron Unit i reactor pressure vessel, led to the following conclusions:
- The capsule received an avggage f p t neutron fluence (E > 1.0 MeV) of 3.50 x 10 n/cm after 1.15 EFPY of plant operation.
- Irradiationofthereactorvesselintggaedingeshellforging SP-5933 Charpy specimens to 3.50 x 10 n/cm (E > 1.0 MeV) resulted in no 30 and 50 ft-lb transition temperature increases for specimens oriented parallel to the major working direction (tangential orientation) or normal to the major working direction (axial orientation).
ld HAZ metal Charpy specimens irradiated to
- Weldmeta}8""n/cmd "f (E > 1.0 MeV) also resulted in no 30 a 3.50 x 10 ft-lb transition temperature increases.
^
- The average upper shelf enargy of the shell forging 5P-5933 showednogecreageinenergyafterirradiationto g
3.50 x 10 n/cm (E > 1.0 WeV). Weld metal showed that the initial 74 ft-lb upper shelf energy decreased by 4 ft-lb.
Both materials exhibit a more th u adequate upp:r 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 G.
- The surveilluce capsule test results do not indicate any significut chuges in the RT values projected for the reactorvessel,and,thereforkTa low risk of vessel 15ilure from pressurized thermal shock (PTS) events is postulwed.
- The calculated end-of-life maximum neutron fluence (E > 1.0 MeV) for the Byron Unit i reactor vessel is as follows:
18 2
Yessel inner radius - 3.14 x 10 n/ca 18 2
Vessel 1/4 thickness - 1.72 a 10 n/cm Vessel 3/4 thickness - 3.24 x 1018,7e,2 1-1
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2.
INTRODUCTION 3
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This report presents the results of the examination of Capsule l
I U, the first capsule to be removed from the reactor in the continuing l
surveillance program which monitors the effects of neutron irradiation i
on the Byron Unit i reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Byron Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation auchanical properties of the reactor vessel materials are l
a presented by J. A. Davideon.II) The surveillmace program was p anned to cover the 40-year design life of the reactor nressure vessel and was
[
I hased on AS1)I E-185-73, ' Standard Practice for Surveillmace Tests for Nuclear 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 Cnarpy Y-notch impact and tensile surveillance specimens were performed.
This report summarised the testing of and the postirradiation j
data obtained from surveillance Capsule U removed from the Byron Unit 1 reaktor vessel and discusses the saalysis of these data, i
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8.
BACKGROUND The abil!.ty of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vess=4 is the most critical
)
region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 508 Class 2 (base material of the Byron Unit I reactor pressure vessel beltline) are v. ell documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ducti.lity and toughness under certain conditions of irradiation.
A method for performing analyses to guard against fast fracture j
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 (RTET)
- RT is defined as the greater of either the drop weight nil-ET ductility transition temperature (NDTT per ASTM E-208) or the temperature 60*F less than the 50 f t-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material. The RTET of a given material is used to inder that material to a reference stress intensity factor curve (K c2rve) which appears in Appendix G of the yg ASME Code. The KIR curve is a lower bound of dynnaics, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given matarial is indexed to the K IR curve, allowable stress intensity factors can be obtained for this 3-1
j anterial as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
RT and, in turn, the operating limits of nuclear power plants NDT 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 i
monitored by a reactor surveillance program such as the Byron Unit 1 Reactor Vessel Radiation Surveillance Program,(1) in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 f t-lb temperature (ARTNDT) d"* D* I#~#"dI"DI "
is added to the original RT t adjust the RT f r radiation NDT NDT embrittlement. This adjusted RTNDT (RTNDT initial + ARTNDT) I' "**d D
)
indsx the material to the Kyg curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects i
of irradiation on the reactor vessel materials.
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DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Byron Unit i reactor pressure vessel core region asterial n re inserted in the reactor vessel prior to initial plant startup. The six capsules nre positioned in the reactor vessel between the neutron shield pads and the vessel wall as shown in Figure 4-1.
The-vertical center of the capsules is opposite the vertical center of the core.
Capsule U was removed after 1.15 effective full pour years of plant operation. This capsule conta!.ned Charpy V-notch, tensile, and 1/2 T compact tension (CT) specimens (Figure 4-2) from the intermediats i
shell forging 5P-5933 ud submerged are nld metal representative of the intermediate to lone shell beltline nld sena of the reactor vessel and Charpy Y-notch specimens from n ld heat-affected none (EAZ) material.
All hest-affected sono specimens n re obtained from within the HAZ of forging 5P-5933 of the representative weld.
The chemistry and heat treatment of the Byron Unit i surveilluce material is presented in Table 4-1.
Additional chemical analyses performed on irradiated Charpy specimens is presented in Table 4-2.
All test specimens were machined from the 1/4 thickness location of the plate. Test specimens represent asterial taken at least one forging thickness from the quenched end of the forging.
Bace netal Charpy V-notch impact ud tension specimens nre oriented with the longitudinal uis of the specimen parallel to the major working direction of the forging (tugential orientation) and also normal to the l.
major working direction (uial orientation). Charpy Y-notch and tensile specimens from the n ld metal w re oriented with the longitudinal axis of the specimens transverse to the welding direction. The CT specimens in Capsule Y nre machined such that the simulated crack in the specimen
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would propagate normal and parallel to the major working direction for the plate specimen and parallel to the weld direction.
l 4-1 1
Capsule U contained dosimeter wires of pure copper, iron, In nickel, and aluminus-0.15% cobalt (cadmium-shielded and unshielded).
addition, cadmium shielded dosimeters of neptunium (NpD7) and uranium (8) were contained in the capsule.
Thornal monitors ende 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 melting points are as follows, i
2.5% As, 97.5% Pb Melting Point: 579'F (304'C) 1.75% As, 0.'75% Sn, 97.5% Pb Melting Point:
590'F (310'C)
The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in capsule U are shown in Figure 4-2.
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Table 4-1 CEMICAL COMPOSITION AND EAT TREATWENT OF THE BYRON UNIT 1 REACTOR YESSEL SURVEILLANCE MATERIALS Chestemt Composition (wt%)
F,lement Inter. Shel: Ferrinz 5P-5933 WsidMetal(*)
0 0.21 0.04 1
Mn 0.69 1.45 l
P 0.010 0.011 1
S 0.009 0.010 l Si 0.27 0.57
)(b)
Ni 0.73 0.71 l
Mo 0.56 0.44 I
Cr 0.36 0.11 l
Cu 0.05 0.026 )
Al 0.010 0.013 Co 0.01 0.011 (b)
Sb
<0.002
<0.002 W
(0.002 (0.002 Ti
<0.001 0.002 Zr (0.002 (0.002 V
0.003 0.006 (b)
Sn
<0.002 0.008 (b)
As 0.005 0.004 Cb
<0.002
<0.002 N
0.005 0.011 2
B
<0.005 (0.003 Heat Treatment Histon Waterial Temperature ('F)
Time (Hr)
Coolant Inter. Shell Austenitising 1630-1650 3.6 Water quenched (Forging 5P-5933) Austenitising 1575-1600 3.5 Water quenched Tempered 1230-1280 5.5 Air cooled Stress Relief 1100-1150 12.25 Furnace cooled WeldMetal(*)
Stress Relief 1100-1150 12.25 Furnace cooled (a)This weldsent was f abricated by The Babcock and Wilcox Co., using 5/32 inch MnWoNi weld filler wire, heat number 442002 and Linde 80 flux, lot number 8873 and is identical to that used in the actual fabrication of the reactor vessel intermediate to lower shell girth weld.
(b) Analyses performed by the Babcock and Wilcox Co.
4-s
Table s a CHEMICAL COMPOSITION FOR BYRON UNIT 1 CAPSULE U IRRADIATED CHARPY IMPACT SPECIMENS Chemical Composition (wt.%)(*)
Weld Metal:
Specimen No.
Cu Ni C
Mn P
S Si Cr Mo V-AW-1 0.023 0.670 0.084 1.40 0.0085 0.016' O.523 0.091 0.419--
0.025 AW-9 0.022 0.665 0.075
_1.45 0.0091 0.015 0.513 0.096 0.393 0.025 AW-1 0.021 0.714 AW-2 0.021 0.741 AW-3 0.022 0.713 AW-4 0.021 0.714 AW-5 0.020 0.704 AW-6 0.020 0.694 AW-7 0.020 0.706 L
AW-8 0.021 0.677 AW-9 0.023 0.677 AW-10 0.021 0.680 AW-11 0.021 0.680 AW-12 0.021 0.667 AW-13 0.024 0.677 AW-14 0.022 0.697 AW-15 0.021 0.634 Forging SP-5933 Specimen No.
Cu Fi C
Mn P
S Si Cr Mo V
AL-13 0.034 0.730 0.185 0.679 0.0054 0.012 0.241 0.34 0.564 0.059 AL-13 0.032 0.791
(") Method of analysis--Inductively coupled Plasma Spectrometry (ICPS) for all elements except C, S and Si.
t Table 4-3 BYRON UNIT 1 REACTOR YESSEL TOUGHNESS DATA Upper Shelf Enerum Normal to Principal Principal W rking Working T
RT Material Cu Ni P
NDT NDT Direction Direction Component Code No.
Spec. No.
(%)
(%)
(%)
f*F1 (*F)
(ft-lb)
(ft-lb)
Closure Head Dome C3486-1 A533B CL1
.10
.65
.016 -10
-10 151 Closure Head Ring IV4566 A508 CL2
.11
.75
.007 20 20 125 Closure Head Flange 124K358VA1
.74
.011 60 60 145 Vessel Flange 123J219VA1
.73
.012 10 10 152 Inlet Nossle IV4684/3V1320
.12
.82
.008 -10
-10 117 O'
.12
.82
.008 -20
-20 116 1Y4684/3V1320 1Y4695
.13
.79
.007 -20
-20 116 IV4695
.12
.78
.006 -20
,-20 119 Outlet Nossle 1Y4656
.11
.84
.007 0
0 131 IV4656
.11
.84
.007 -20
-20 131 l
2V2557
.11
.85
.007 -20
-20 112 2V2557
.11
.84
.008 -10
-10 94 l
N ssle Shell 123J218
.05
.72
.010 20 20 138 184 o
l Inter. Shell 5P-5933
.05
.73
.010 40 40 139 156 Lower Shell 5P-5951
.04
.64
.014 10 10 150 160 Botton Head Ring IV4672
.80
.012 0
0 115 Bottom Head Dome C2815-1 A533B CL1
.19
.64
.009 -30
-20 118 Inter to Lower WF336*
SAW
.024
.70
.010 -30
-30 77 Shell Cirth Weld
' Welded using 5/32 inch MnMoNi weld wire Heat No. 442002 and Linde 80 Flux Lot. No. 8873.
O.
REACTOR VESSEL CCRE BARREL 330 (3.85) Z NEUTRCN PAD CAPSULE U (3.85)
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3,5 3,,,.
270*
go.
3*
2.5*
I
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5t Aa. 7,-
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(3.65) Y (3.85) X i
W (3.85) 180*
PLAN VIEW Figure 4-1.
Arrangement of surveillance capsules in the reactor vessel 4-6
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s LEGE!D:
AL - IhTERMEDIATE SHELL FORGIf;G SP-5933 (TANGEhTIAL ORIEF.TATION)
AT - IffrERMEDIATE SHELL FORGI!.'G 5F-5933 (AXIAL ORIEf(TATION)
AW - WEIE METAL AH - liEAT-AFFECTED-ZONE (HAZ)
CCMPACT CO WACT COMPACT CCWACT S=ACER TENSILE TENSION TENSICH CHARPY __
CHARFY 04ARPY TENSICH TENSION CHARPY CHAFtl AW3 AW15 AH15 AW12 AH12 AW9 AH9 AW6 AH6 AV3 A
AW2 Ad4 AW3 AW2 AM AW14 AH14 AW11 A}i11 AW8 AH8 AL4 AL3 AL2 ALI AWS AHS AW2 Ad AW1 l
AW13 AH13 AW10 AH1C Aiff Alf7 AW AH4 Av1 Cu
- Al
.15% Ce Cw -
UUU 0 01 RRQ 579 *r
=90 'r f
- Al
.15% Co (Cd)
MCNI TOR "
l biONITOR
=
l
s
%D r
I l
i e
l CVM U 237 Np u238 OCSI C ER CC@ACT Co* ACT
]
EE.OCM TD6iLE CHARPY CHARPY CHAPpY CHARPY CHARPY TD6 ION TENSION TENS:LE AL3 AT15 AL15 AT12 AL12 AT9 AL9 Alf A16 AT3 l AL3 AT) 2 366 M2 AT14 Allt AT11 AL11 AT8 ALA AT5 ALS AT?
A12 AT4 AT3 AT2 ATI AT?
1 AL3 AT13 AL13 ATIC AL10 AT7 ALT AT4 Att AT1 Ali Att Al
.15% Ce C-Al
.35% C.
J O LJ Ll
]
Al
.l5% Co (Cd A l
. 4 5% Co (Cd) l l l l l, lr j MONITCH Ni Ni s
i PC310N OF VESSEL TO SOTTCM CF VESSEL l
,i Figure 4-2.
Caprule U diagram showing location of specimens, thermal monitors and dosimeters p
e i i 4-7 TI APERTUllE
[
CAISO ppgg g/
b Aho Available On Aperture Card i
5.
TESTING OF SPECIMENS FROM CAPSULE U 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch and tensile specimen was performed at the Westinghouse Research and Development Center with consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 100FR50, Appendices G and H,(2) ASTM Specification E185, and Westinghouse Procedure RMF-8402, Revision 0 as modified by RMF Procedures 8102 and 8103.
Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-9517.(1) No discrepancies were found.
Examination of the two low-melting point 304*C (579'F) and 310*C (590*F) eutectic alloys indicated no melting of either type of thermal mon.i tor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304*C (579'F).
The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74,358J machine.
The tup (striker) of the Charpy machine is instrumented with an Effects Technology Model 500 instrumentation system. With this system, load-time and energy-time sigt.als can be recorded in addition to the standard measurement of Charpy energy (E ).
From the load-time curve, the load D
of general yielding (Pgy), the time to general yielding (tGY), the maximum load (P ), and the time to maximum Ivad (t ) can be determined.
y g
Under some test conditions, a sharp drop in load indicative of fast
~
fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (P ), and the load at which f ast p
fracture terminated is identified as the arrest load (P )
- A 5-1 I
The energy at maximum load (E ) was determined by comparing the g
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 ) is the p
difference between the total energy to fracture (E ) and the energy at j
p maximum load.
The yield stress (a ) is calculated from the three-point bend y
formula. The flow stress is calculated from the average at the yield and maximum loads, also using the three-point bend formula.
i Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specifier. tion A370-77. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
l Tension tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-83 and E21-79, and RMF 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 aiuute throughout the test.
Deflection measurements were made with a linear variable displacement transducer (LVDT) extensoneter. The extensoneter knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 Leh. The extensometer is rated as Class B-2 per ASTM E83-67.
Elevated test temperatures were obtained with a three-sone electric resistance split-tube furnace with a 9-inch hot sone. All tests were conducted in air.
Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen tesperature.
Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy l
a-a
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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 uniform elongation were determined directly froa the load-extension The yield strength, ultimate strength, and fracture strength curve.
were calculated using the original cross-sectional area. The final diameter and final sage length were determined from post-!racture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed j
using the final diameter measurement.
5.2 CHARPY Y-NOTCH IMPACT TBST RESULTS I
The results of Charpy V-notch impact tests performed on the 18 2
various materials contained in Capsule U irradiated at 3.50 x 10 n/cm (E > 1.0 MeV) are presented in Tables 5-1 through 5-4 and are compared l
with unirradiated results(I) as shown in Figures 5-1 through 5-4.
The transition temperature increases and upper shelf energy decreases for the Capsule U materials are summarized in Table 5-5.
Irradiation of vessel intermediate shell forging 5P-5933 18 material (tangential and axial orientation) specimens to 3.50 x 10 n/cm (Figure 5-1 and 5-2) resulted in no increase in the 30 and 50 ft-lb transition temperatures and no upper shelf energy decrease.
Weld metal irradiated to 3.50 x 1018,j,2 (Figure 5-3) resulted in no 30 and 50 ft-lb transition temperature increase and an upper shelf energy decrease of 4 ft-lb.
Weld HAZ setal irradiated to 3.50 x 10 37,,2 (Figure 5-4) 18 resulted in no 30 and 50 ft-lb transition temperature increase and no upper shelf energy decrease. A large scatter in data was observed which
.is typical of many HAZ Charpy tests for other surveillance programs.
5-s
.l The fracture appearance of each irradiated Charpy specimen from j
N
}
the v'arious 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 Byron Unit 1 surveillance asterials 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 3.50 x 1018,je,2 are less than the Guide predictions.
5.8 TBNSION TEST RBSULTS The results of tension tests performed on shell forging 5P-5933 (tangential and axial orientation) and the weld metal irradiated to 3.50 x 1018,7,,2 are shown in Table 5-7 and are compared with unirradiated results(I) as shown in Figures 5-9, 5-10 and 5-11.
Forging 5P-5933 test results are shown in Figures 5-9 and 5-10 and indicate that irradiation to 3.50 x 1018,j,,2 caused a less.than 3 kai 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 l
less than 5 kai with irradiation. The small increases in 0.2% yield strength and tensile strength exhibited by the forging material and weld metal indicate that these asterials are not highly sensitive to radiation at 3.50 x 1018,j,,2 as is also indicated by the Charpy impact test results. The fractured tension specimens for the forging material are shown in Figures 5-12 and 5-13, while the fractured specimens for 4
l the weld metal are shown in Figure 5-14.
A typical stress-strain curve for the tension tests is shown in Figure 5-15.
1 5.4 EAIDNESS TBST RBSULTS
.l Rockwell B (HRB) hardness tests performed on irradiated Charpy impact test specimens are presented in Table 5-8.
a-4
~
h 5.5 COMPACT TENSION TESTS Per the surveillance capsule testing program with the Commonwealth Edison Company, 1/2 T-compact tension fracture mechanics specimens will not be tested and will be stored at the hot cells at the Westinghouse RAD Center.
e a
5-8
Table 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE BYRON UNIT 1 SHELL AT 550'F, FORGING 5P-5933 p/cmIp(E > 1.0 MeV)
FLUENCE 3.50 x 10 n
Temperature Impact Energy Lateral Expansion Shear Sample No. fill
(*0)
(ft-lb) IJ1 (sils)
(as)
(%)
Tangentini Orientation
~~
AL2
-100
-46 5.0 7.0 5.0 0.13 0
AL4
- 75
-59 20.0 27.0 17.5 0.44 5
- 75
-59 32.0 43.5 30.0 0.76 20 AL7
- 60
-51 10.0 13.5 9.5 0.24 5
AL13
- 60
-51 34.0 46.0 24.0 0.61 15 AL8
- 50
-46 37.0 50.0 28.0 0.71 15 AL3
- 50
-46 43.0 58.5 31.0 0.79 25 AL15
- 25
-32 78.0 106.0 60.5 1.54 55 ALS
- 25
-32 79.0 107.0 61.0 1.55 55 AL10 0
-18 127.0 172.0 81.5 2.07 65 AL9 50 10 144.0 '195.0 81.0 2.06 70 AL11 75 24 170.0 230.5 90.5 2.30 90 AL12 125 52 174.0 236.0 83.0 2.11 100 AL5 250 121 169.0 229.0 85.0 2.16 100 AL14 350 177 158.0 214.5 88.0 2.24 100 Axial Orientation ATil
-100
-73 8.0 11.5 7.5 0.19 5
AT5
- 75
-59 5.0 7.0 5.5 0.14 2
AT2
- 75
-59 8.0 11.0 7.0 0.18 5
AT15
- 60
-51 14.0 19.0 11.0 0.28 10 AT4
- 60
-51 20.0 27.0 15.0 0.38 10 AT13
- 50
-46 24.0 32.5 16.0 0.41 15 AT3
- 50
-46 47.0 63.5 33.0 0.84 25 AT12
- 25
-32 67.0 91.0 47.0 1.19 30 AT14
- 25
-32 80.0 108.5 60.0 1.52 60 ATS 0
-18 93.0 126.0 64.5 1.64 65 AT7 50 10 124.0 168.0 75.0 1.91 80 AT1 75 24 163.0 221.0 84.0 2.13 95 AT8 125 52 155.0 210.0 70.0 1.78 100 i
l AT10 250 121 147.0 199.5 84.5 2.15 100 AT9 350 177 141.0 191.0 82.0 2.08 100 1
s-a i
f Table 5-2 CHARPY Y-NOTCH IMPACT DATA FOR THE BYRON UNIT 1 REACTOR VESSELWELDMETALANDHjjMETAgIRRADIATEDAT 1
550*F, FLUENCE 3.50 x 10 n/cm (E > 1.0 MeV)
Temperature Impact Energy Lateral Expansion Shear Sample No. (*F)
I'01 (ft-lb) fJ1 (mils)
(mm)
(%)
Weld Metal AW7
-100
-73 15.0 20.5 10.5 0.27 5
AW2
-50
-46 23.0 31.0 20.0 0.51 10 AW13
-25
-32 28.0 38.0 25.0 0.64 15 AW9
-25
-32 33.0 44.5 27.0 0.69 25 AW6
-0
-18 32.0 43.5 26.0 0.66 30 AW1
-0
-18 39.0 53.0 34.0 0.88 40 AW10 25
-4 46.0 62.5 38.5 0.98 55 AW11 25
-4 50.0 68.0 40.5 1.03 65 AW12 50 10 44.0 59.5 41.0 1.04 50 AW14 50 10 54.0 73.0 46.0 1.17 60 AW15 75 24 60.0 81.5 40.5 1.23 90 AW5 125 52 70.0 95.0 81.0 1.55 100 AW3 200 93 67.0 91.0 58.5 1.49 100 l
AW8 250 121 70.0 95.0 63.0 1.60 100 AW4 300 149 73.0 99.0 66 5 1.69 100 i
HAZ Metal AH15
-150 -101 25.0 34.0 16.5 0.42 10 AH10
-150 -101 15.0 20.5 8.5 0.22 5
i AH12
-100 - 73 111.0 150.5 54.5 1.38 60 AH2
-100 - 73 29.0 39.5 15.5 0.39 10 AH13 59 55.0 74.5 29.0 0.74 35 AH1 59 32.0 43.5 17.0 0.43 15 AH11 46 94.0 127.5 44.0 1.12 60 AH4 46 45.0 61.0 25.0 0.64 30 AH6 0 - 18 69.0 93.5 43.5 1.10 55 AH7 50 10 58.0 78.5 37.0 0.94 50 AH5 75 24 135.0 183.0 71.0 1.80 95 AH9 125 52 77.0 104.5 56.0 1.42 100 AH3 200 93 89.0 120.5 56.0 1.42 100 AH14 300 149 173.0 234.5 79.0 2.01 100 l
5-7
Table 5-3 1
l SHELLFORCING5P-5933IRRADIATEDAT550*FTOAFLUENCEOF3.50x10@n/cm (E >
INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE BYRO T
i j
j Normalised Energies i
Test Charpy Charpy Maximus Prop Yield Time Maximum Time to Fracture Arrest Yield Flow J
Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2
Number 1*F1 (ft-1b)
(ft-lb/in )
(kips)
(psec)
(kips)
(ssec)
(kips)
(kips)
(ksi)
(ksi)
Tanzential Orientation AL2
-100 5.0 40 29 11 3.60 80 4.00 95 3.90 0.15 118 125 AL4
- 75 20.0 161 166
- 5 3.85 90 4.70 345 4.70 128 142 j
- 75 32.0 258 357
-100 4.10 95 6~10 665 4.85 136 152 AL7
- 60 10.0 81 83
- 2 3.90 100 4.25 200 4.20 129 135 I
AL13
- 60 34.0 274 256 18 3.85 90 4.90 505 4.85 128 145 j
AL8
- 50 37.0 298 253 45 3.85 90 4.85 505 4.85 128 144 AL3
- 50 43.0 346 326 21 3.65 85 4.80 640 4.80 121 140 1
ALIS
- 25 78.0 628 329 299 3.75 90 4.80 655 4.30 0.15 123 141 AL6
- 25 79.0 636 328 308 3.60 85 4.75 655 4.30 119 138 AL10 0 127.0 1023 323 699 3.35 115 4.80 690 3.20 0.70 110 135 AL9 50 144.0 1160 372 788 3.06 80 4.35 820 2.15 0.65 100 122 4
AL11 75 170.0 1369 318 1051 2.45 80 3.75 810 81 103 AL12 125 174.0 1401 250 1151 2.05 75 3.35 730 68 90 AL5 250 169.0 1361 326 1035 2.50 55 4.00 775 82 107 AL14 350 158.0 1272 261 1012 1.95 50 3.40 725 65 89 Transverse Orientation AT11
-100 8.0 64 46 19 3.85 85 4.15 125 4.05 128 132 ATS
- 75 5.0 40 33 7
3.90 85 4.15 100 4.05 129 133 AT2
- 75 8.0 64 48 16 3.95 95 4.25 l'45 4.25 131 136 A115
- 60 14.F 113 103 10 4.00 95 4.30 235 4.25 132 137 AT4
- 60 20.0 161 162
- 1 3.85 95 4.65 345 4.65 127 141 ATIS
- 50 24.0 193 195
- 5 3.95 95 4.70 4e5 130 143 AT3
- 50 47.0 378 327 52 3.70 85 4.70 655 4.65 122 139 AT12
- 25 67.0 540 332 208 3.70 100 4.85 665 4.60 123 142 1
AT14
- 25 80.0 644 337 307 3.85 95 4.85 665 4.40 127 144 j
AT6 0
93.0 749 354 395 3.50 90 4.65 730 3.90 0.35 116 135 AT7 50 124.0 998 378 620 3.20 85 4.40 825 2.70 0.90 106 126 ATI 75 160.0 1313 286 1027 3.10 120 4.30 685 103 123 AT8 125 155.0 1248 265 983 2.30 75 3.55 720 76 97 AT10 250 147.0 1184 326 857 2.45 60 4.05 775 81 107 AT9 350 141.0 1135 308 827 2.30 55 3.75 770 77 101 l
i Table 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR BYRON UNJ 1 WELDMETALANDHAZMETALIRRADIATEDAT550*FTOAFLUENCEOF3.50x10'{n/cm2 Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Lead to Yield Load Maximus Load Load Stress Stress 2
(f'-lb/in )
(kips)
(usec)
(kips)
(psec)
(kips)
(kips)
(ksi)
(ksi)
Number (*F)
(ft-Ib) c Weld Metal AW7
-100 15.0 121 120 1
3.70 95 4.25 280 4.25 122 131 AW2
- 50 23.0 185 176 10 3.45 85 4.15 395 4.05 114 126 AW13
- 25 28.0 225 201 25 3.50 85 4.40 435 4.35 115 130 AW9
- 25 33.0 266 234 32 3.55 100 4.30 510 4.20 0.25 118 130 AW6 0
32.0 258 226 32 3.35 90 4.15 510 4.05 111 124 AT1 O
39.0 714 228 86 3.35 95 4.30 505 4.20 1.25 111 127 i
AT10 25 46.0 370 222 148 3.25 95 4.25 505 4.05 1.50 108 124 AW11 25 50.0 403 225 177 3.30 90 4.30 505 4.20 2.20 109 126 AW12 50 44.0 354 219 135 3.30 100 4.20 505 4.00 1.45 108 123 p
AW14 50 54.0 435 218 217 3.30 90 4.20 495 3.85 2.05 109 124 AW15 75 60.0 483 223 260 3.10 100 4.20 520 3.75 2.20 102 120 AW5 125 70.0 564 207 357 2.85 85 3.90 506 95 112 AW3 200 67.0 540 199 340 2.85 100 3.80 505 95 110 AW8 250 70.0 564 223 341 2.70 85 3.80 550 89 107 AW4 300 73.0 588 214 374 2.65 95 3.C0 555 88 104 RAZ Metal AH10
-150 15.0 121 93 27 5.00 120 5.55 205 5.55 165 174 AH15
-150 25.0 201 201 1
4.65 145 5.70 395 5.60 154 172 i
AH2
-100 29.0 234 226 8
4.20 90 5.05 430 5.05 138 153 AH12
-100 111.0 894 380 514 4.45 95 5.50 665 4.25 0.15 147 165 AH1
- 75 32.0 258 239 19 4.40 95 5.45 435 5.40 145 162 AH13
- 75 55.0 443 369 73 4.30 95 5.45 655 5.30 0.15 143 161 AH4
- 50 45.0 362 261 101 3.85 95 5.10 505 4.95 0.20 128 148 AH11
- 50 94.0 757 366 391 4.20 90 5.40 655 4.70 0.65 139 159 AH6 0
69.0 556 343 212 3.75 90 5.05 665 4.80 124 146 AH7 50 58.0 467 253 214 3.70 90 4.95 505 4.40 0.60 123 143 AH5 75 135.0 1087 419 668 3.65 85 5.00 810 120 143 AH9 125 77.0 620 232 388 3.35 85 4.50 500 111 130 AH3 200 89.0 717 344 373 3.10 85 4.50 740 103 126 AH14 300 173.0 1393 349 1044 2.90 85 4.50 770 96 121
i Table 5-5 EFFECT OF 550*F IRRADIATIDN AT 3.50 x 10 m/cm2 (E > 1.0 MEV)
II DN NOTCE TOUCEISSS PROPERTIES OF BYRON UNIT 1 REACTOR VERW-MATannrA Average 35 mil Average Energy Average 30 ft-lb Lateral Espansion Average 50 ft-lb Absorption at Temperature (*F)
Temperature (*F)
Temperature (*F)
Fall Shear (ft-lb)
Material Uelrradiated Irradiated g Uelrradiated Irrediated LI Unirradiated Irradiatd E Unirradiated Irradiated afft-lb)
Forging
-65
-65 0
-53
-53 0
-50
-50 0
168 los 0
~
EP-5033 (Tangential)
~
Forging
-50
-50 0
-32
-32 0
-30
-30 0
145 145 0
y 5P-5G33 g
(Asial)
Weld Metal
-15
-15 0
27 27 0
40 40 0
74 70
-4 EAZ Metal
- 75
-75 0
0 0
0
-27
-27 0
110 110 0
M f
.1
]
Table 5-6 COMPARISDN OF BYRON UNIT 1 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY CUIDE 1.99 REVISION 2 PREDICTIONS 30 ft-lb Transition Temp. Shift Upper Shelf Energy Decrease 4
]
10}gence F
R.G. 1.99 Rev. 2 Capsule U R.C. 1.99 Rev. 2 Capsule U 2
Material n/cm
(*F)
(*F)
(%)
(%)
]
Forging SP-5933 (Tang.)
3.50 22 0
15 0
l Forging SP-5933 (hial) 3.50 22 0
15 0
Weld Metal 3.50 25 0
15 5
a) Cu and Ni values from Table 4-1 were used to determine R.G.1.99 predictions l
w
~
J 1
l 4
5 1
.I
1 Table 5-7 TENSILE PROPERTIES FOR BYRON UNIT @/cmCTy (VESSEL MATERIAL IRRADIATED AT 550*F TO 3.50 x 10 n
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 Material Number M (ksi)
(ksi)
(kip)
(ksil (ksi)
(%)
(%)
(%)
Forging AL2 0
72.8 95.7 2.75 229.6 56.0 11.3 26.7 76 5P-5933 AL1 78 74.4 90.9 2.60 217.0 53.0 10.5 25.5 76 Tang.
AL3 550 62.6 86.2 2.40 167.2 48.9 9.8 23.3 71 Forging AT3 0
73.3 94.3 2.75 191.6 56.0 11.3 26.1 71 SP-5933 ATI 76 71.3 90.7 2.73 198.6 55.6 11.2 26.1 72 Axial AT2 550 63.2 86.6 2.60 173.3 53.0 9.3 22.7 69 Weld AW1 0
72.8 92.7 3.15 201.5 64.2 13.5 26.0 68 v.
d.
Metal AW2 76 64.2 86.6 2.95 180.9 60.1 12.0 23.3 67 AW3 550 61.1 81.5 3.00 157.4 61.1 10.5 20.3 61 1
E 5
8 4
O O
I l
t l
l Tsble 5-8 HARDNESS (HRB) 0F THE BYRON UNIT 1 REACT 0g VESSg (SURVEILLAN MATERIALS IRRADIATED AT 550*F TO 3.5 x 10 n/cm-E > 1.0 MeV)
Forging SP-5933 Weld Metal Hardness Hardness Sample No.
HRB Sample No.
HRB AL13 90 AW12 90 94 8
93 90 91 AL4 90 AW1 94 87 8
90 92 92 AL6 90 AW8 90 92 92 91 91 e
O 8-13
Curva 754819-A
)
(
- C)
-150 -100 - 50 0
50 100 150 200 250 l
l l
1 I
i 4
8 1
100 3
o-
$ 80
'g 60 2
s C
o M M a) Number associated with data point -
Indicates the number of tests 20 2
naving the same value.
2 I
I i
0 d
^
100 2.5 l
I I
I,
.oI b
I I
I
- E 80
~
- 2.0 E
L5 e M
o
$M L0 E o
0.5 g3 o
0 i
0
'l 200 180 2M 160 o
i 200
_ 10 o
B e
13 1M d
o
~ 100 Code:
120 2 80 o
o - Unirradiated 5
o
. - Irradiated at 550 F 80 60 18 2
3,50 x 10 n/cm 0
~
0 o
o 0
I t
i i
i i
i 0
- 200
-100 0
100 200 300 00 500 Temperature ( F) l Figure 5-1.
Chsrpy impact properties for Byron Unit i reactor vessel shell forging SP-5933 (tangential orientation) a-14
,=-..,,-------w_-
Curvo 754815-A
(
- C)
-150 -100 - 50 0
50 100 150 200 250 _
120 l
t i
l3 I
i i
i i
100
-f 80
'g M o
n e
M M a) Number associated with data point _
indicates the number of tests having the same value.
20 2
2g,2 2
se x
0 15 1%
i i
i i
,i A
i i
i 10
= 80 o
s o
8 L5 e g
LO s
, ),
&w k5 i
i 0
0
=
E i
i i
i i
i i
i i
240 180 160 e*
2 200
_ 140 e
o o
160 g120 100 o
Im a P
= 80 5
Code:
o 80 60 o - Unirradiated
.-Irradiated at 550 F 40 o
40 8
2
- 3. 50 x 10 n/cm 20 o
0 i
s i
i i
i i
i 0
- 200
-100 0
100 200 300 400 500 Temperature (*F)
Figure 5-2.
Charpy impact properties for Byron Unit I reactor vessel shell forging 5P-5933 (axial orientation) j l
5-18
j curve 754818-A
(
- C) 1
-150 -100
- 50 0
50 100 150 200 250 120 1
I 1
i i
i g
2 100
~
~
2 f 80 gw a) Number associated with data point
.c-e M M indicates the number of tests having the same value.
o M
-<2i i
i i
i i
0 2.5 100 2.0
- E 80 2
g L5 g
.0 g
3 NO L0
&N
&5 0
4 i
i i
i i
i 0
200 180 240 Code:
- Unirradiated 160
- -Irradiated at 550 F 200 l
l#
18 2
5
- 3. 50 x 10 n/cm 120 160 g
100
@ 80 2
1N O 1
o
=
a 5
,z e
80 M
0 o'
m e
20 0
i io i
i i
i i
i 0
4
- 200
-100 0
100 200 300 00 500 Temperature ( F)
Figure 5-3.
Charpy impact properties for Byron Unit i rer.ctor vessel weld metal s-is
Curve 754824-A
( C)
-150 -100
- 50 0
50 100 150 200 250 120 2
I i
i i
i 1
100 e
w 3 80
$M
.c M M a) Number associated with data point O
indicates the number of tests 20 having the same value.
'2J 2
i i
i i
i i
0 y
15 lM i
i i
i i
i i
i i
2.0
.E 80 o
E o
L5 s
-M o
_g.
L0" NM o
g O
0.5
-(20 0
t 9
i i
i I
i 0
t 200 i
i i
i i
i i
i i
180 20 160 200 lg
- o 3
~ 120 160 o
~ 100 B
.o im o
l
= 80 o
O Code-80
.o o,.
o - Unirradiated 60 s
- -Irradiated at 550 F 0
o 2
- 3. 50 x 10'8 n/cm 20 o
0 i
i i
i e
i i
i 0
- 200
-100 0
100 200 300 MO 500 Temperature ( F)
Figure 5-4.
Charpy impact properties for Byron Unit I rer.ctor weld heat affected zone metal 5-17 s
l Eg'..
v--
.s-f[~If, ')
f' k.$5.ffi.
.3..:..
ms, m..;
... ;;;;w
- M,-
'kh,:h
.$$[kN'
<.[uyy
- 3' hh
.I.
~
w: -.;
. ;w ; p, ~
~
AL2 AL4 All AL7 AL13
?.Nd
. jd'Mdf
' 8f
.I
' k" h,,4ap,..
M'b
.b
- yh\\;99)$
6.%ty%
wi.
.,u g
- M aa7f
,& L _
7: y.4.'
yX%
l n> ;
% ?.
f?'$O!$,
ll ' $,.i^
?
~"Q;.+
.,W
- n s _,
=--,
AL8 AL3
~ AL15 AL6 Allo
.. q f
?,[.
3 q
- '?
,.f..
.v,,e l W- -
f e*
i-U
~
$7 MJ
& 1-
' -C,
[Q
- f,
-e. :.
N ?g
.1.
s u..
' g).
L a
- b. '
.2, 3;,
AL9 AL11 All2 AL5 AL14 Fig 2re 5-5.
Charpy impact specimen fracture surfaces for Byron Unit I reactor vessel shell forging SP-5933 (tangential orientation) 5-18 F.M-14 711 i
$m. M8
}VW ~-
N. h..
.?ly..,Tj, f$OM).,,
e.4, ;W L9,, k,:,.
.y t
s,O*1!
C
't d A.
N.T"b..v,N"t.c..,;
,3 j=1 c. s -
n4 i
.e
-.... m.
- ' e%.;.g-2 Jps ;
1
.j(t-y;~
3
r
.g2 j
4fjQ-hNQ:y,s.
c,,.h ?
,7. ' S >
_.,gy 6 3..
r3 vii S
'0
i.
v ATil AT5 AT2 AT15 AT4 i
{?p.N ch
'56,. : :4
- f.'
h,,;l,. :[
]
r.? ;%. $,\\;
\\;
D ylW~ ^q);
\\::,M.?
% v-lN;&-:y
]
I'%
?: 5.1 V
j y,,
pm 4
pr$h
,.;..:~.--
W#?
5 )' ' ",
f)'5;
,1 :,,
,r,'s
[.'[
P.
r ;4,Ih,N p
.1 n
e
'9 p
w =.
AT13 AT3 AT12 AT14 AT6
- j w
q.
a a 3 -,--
!=
5 f, s
-J s 4.- 7 t
,xx f fl,
- 5_5, 9:6 pq' m.
v,.
. m..m e-%..
r*
(w.
.%Q ffs4
. A'
~. _.m y;
- f. 'j%
M F(6.,
.c,.
-( V i
~.
sy]
a-g
)
M' t
man.m w'
AT7 AT1 AT8 AT10 AT9 Figure 5-8.
Charpy impact specimen fracture surfaces for Byron Unit I reactor vessel shell forging 5P-5933 (axial orientation) 8-19 D-14712
EB6 ti 6 h;g AW7 AW2 AW13 AW9 AW6 5
- p. w mm i f) [5 M l'Idv,2.
- ph
- $
_J:G_b
_a M AW1 AW10 AW11 AW12 AW14 g,
?-
\\
- e.1 -
t i
.a
'f'.',
kt'l u ~. a g}w w ;,.,--
,j ?h;
',<?
QL9
%, A
> ~;" h 4iffS
&es
~
\\r igg i
r u
.n l *['[
- L AW15 AWS AW3 AW8 AW4 Figure 5-7.
Charpy impact specimen fracture surfaces for Byron Unit i reactor vessel weld metal 5-20 M-14713
l hlY...
Lh1 j
V
- w. n AH15 A!ilo AH12 AH2 AH13 k-IQ f'
if
\\ w<,:.v.
a~-
~r n
j 2
g% ~,.%
W,,y..
j 8
V, 1 j..
y...
[<.a.m g@yc.g..
- e..
r J7p ty pq r...
AH1 AH11 AH4 AH6 AH7 I.
_O[
,71
... q9' W,
L' q.
h%-,
W.ua l-M*: **
.g,
+
s AH5 AH9 AH3 AH14 Figure 5-8.
Charpy impact specimen fracture surfaces for Byron Unit I reactor vessel weld BAZ metal M *la'714 5-21
l Cur ve 754821-A
- C l
- 50 0
50 100 150 200 250 300 120 1
1 i
i i
'800 110 700 100 5 90 Tensile Strength 5
600 g m
g 1
2 aM
$ 70 r
500 -
m 60
~
400 0,2 % Yield Strength 50 40 i
i i
i i
i 300 Code:
Open Poir.ts - Unirradiated 18 2
Closed Points - Irradiated at 3.5 x 10 n/cm 80 i
i i
i i
70 J
q a
Reduction in Area g
Q a) Number associated with data point
- 50 indicates the number of to.t.
y having the same value.
= 40 r,
Total Elongation 0
E
~
=
1
,2 n
3 2
10 Uniforg Elong9tiod 0-i i
i I
-100 0
100 200 300 400 500 600 Temperature ( F) l Figure 5-9.
Tensile properties for Byron Unit i reactor vessel shell forging 5P-5933 (tangential orientation)
.I J
4 s-as
Curv e 754823-A i
C
- 50 0
50 100 150 200 250 300 i
i i
800 120 110 700 100 Tensile Strength 3 90 4
g _E
\\
~
2 s
'~
gM 500
$ 70 w-vs y
60 0.2 % Yield Strength i
i l
i i
i i-E g
Code:
j Open Pt!nts-Unirradiated 18 2
Closed Points - Irradiated at 3.5 x 10 n/cm 80 i
i i
i i
i i
i 3
70 Reduction in Area
_ 60
[g
_ a) Number associated with data point indicates the number of tests j
y having the same value.
E 40 Total Elongation 5 30 y,.
g o
[ Uniform Elongatio]n 20 e
10 0
i i
1 1
i i
- 100 0
100 200 300 400 500 600 Temperature ( F)
Figure 5-10.
Tensile properties for Byron Unit i reactor vessel shell forging SP-5933 (axial orientation) a-as
4 1
Curve 754822-A
- C
'l 0 50 100 1 50 200 250 300 120 I
I I
I I
i i
i g
110 700 100 5
Tensile Strength 600g
~
i y
a%
~
500 -
E 70 25
~
60 j
400 e
0.2 % Yield Strength 50 40 I
I I
L i
i i-300 Code:
Open Points-Unirradiated
- i 18 2
Closed Points-Irradiated at 3.5x10 n/cm 80 i
i i
i i
i i
i j
70 Reduction in Area
=
S
_M a
e%
xy@
j 30 Total Elongation 3
0
^
20 Y
0 10 Uniform Elongation 0
I I
I 1
I i
-100 0
100 200 303 400 500 600 i
Temperature ( F)
Figure 5-11.
Tensile properties for Byron Unit I reactor vessel weld metal I
s-z4
l I
< ~ 7,
, --- - ( y-3.
a.,
~
$t
+ 3 f:t
,1 3,.
. Iw m,;
,m o :.
1.
%, 't4 ; e 3 0%@
l-i
.:Q s%
b z:
4 e
-kE
.t-g
..g x., g,,..
Specimen AL2 0*F y.
,y c l' < 7.,.,
.i e
\\
r.
g u, 1<
- g uv -.
v
'.ts ;',L. 9 J t
+ ; :f. i,,
ugy a
..x, ua.
n
. qjm
-a -
. ~ _ _
^^39
+
a n
Jn
%* ' i, d-71;,;
.-e t, -,, -
Specimen All 78'F 4
.~
q i
>g L
w-I ff ef
> + g;;
a,
4
- :4p;ll.;
k "y a~ m.. ~.%,.,.
,, n-,,. v w..
m sv m-y
~
r~
i i
F
- e c
w Specimen AL3 550*F Figure 5-12.
Fractured tensile specimens from Byron Unit I reactor vessel shell forging 5P-5933 (tangential orientation) i 2
l I
5-25 R.M-14715 4
l 4,
_. --,, g.
7.-.,.
L b
L _1 L'
A < *; '
T 4
4.
+
- s t.
u
,.s 7,
e,: t, M.:-.,yee
.. }
^ !s a
%4 iw 1
_k j
y j p-
+w
+..
. _,a..
Specimen AT3 0*F
... _...., g (.., _, y ;. y,.,
-+4 ; ;..,7,;:
.,t34 xs n,9
. 1
. ? -<
\\
a.w.
. 3-N.' 6
-pg 7 y
- .; J
-u dAMMAMM';'chbi01M.
(g r
,2
- ryg,
- n;..
g; s.. e i
f
~
l Specimen AT1 76'F l
1
- i,,,
i
- , c p!
> 1 ;
a e
9
- i r
g
=
,w, 4
. t-h
?,'
. x;
..m;: -
W g,,/
Specimen AT2 550*F Figure 5-13.
Fractured tensile specimens from Byron Unit i reactor vessel shell forging 5P-5933 (axial orientation) 5-26 RM-14716 j
1
i y.
.,-. 7.,.s. ;
. _7.,.7.g, y. -. _
. x.:a :.:
d, a :o i 1
- i N
10:4.
Ic
~'
-:7 >
m,c
.g
- 12 i 4 i _ t.
ft. 3 1
. 4
',t.b;& W O d N i:: W Wl 4 l'
t
+;a :
f 1.;
g
'.i
^
u p-(
m.s g
.p wafc;i fa_:..
Specimen AW1 0*F l
[,',.
7.-.
g.
l.
y "'
I
's' 4
4 j 1, '
3 f
i l
-3,
.. s.
m, --.
. l
, ^
. > c 'o 1
~...
1 Specimen AW2 76*F j
l I
l I
)
1
,y,,
o.
3-
's
- ' G / e i '.
g.
,f 1
l w ra g 3
~ g _ ! ; ) ',
x Jr. f c. ' M ;,., > ~
- -, tew 6
4 i
i i
i w
Specimen AW3 550*F
)
Figure 5-14.
Fractured tensile specimens from Byron Unit i reactor vessel weld metal i
i l
5-27 K+.14717
4 t
Cu rve 754820 A 120 100 M
D
[60 h
sn 40 -
20 Specimen Al3
( 550*F) 0 t
i i
1 0
0.05 0.10 0.15 0.20 0.25 Strain,in/in l
1 Figure 5-15.
Stregs-strain curve for tension specimen AL3 (curve is typical for all other tested specimens) 4 23
6.
RADIATION ANALYSIS AND NEUTRON DOSIMETRY t
6.1 INTRODUCTION
Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons.
First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known.
Second, in order to relste the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reantor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis.
This section describes a discrete ordinates S, transport analysis performed for the Byron Unit I reactor to deteralne the fast (E > 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 ac tron expos of the reactor vessel to that of the surveillance 1
capsules Based on the use of spectrum-averaged reaction cross sections dexiicd from this calculation and the Byron Unit 1 power history, the saslysis r: the neutron dosimetry contained in Capsule U is presented.
4.
i 4
I e-1 l
_-_--r
6.2 DISCRETB OEDINATES ANALYSIS A plan view of the Byron Unit 1 geometry at the core midplane is shown in Figure 4-1.
Six irradiation capsules attached to the neutron pad are included in the reactor design to constitute the reactor vessel surveillance program.
The capsules are located at 58.5, 61.0, 121.5, 238.5, 241.0, and 301.5 degrees as shown in Figure 4-1.
A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1.
The stainless steel specimen centainers are 1.182 b'y 1-inch and approximately 56 inches in height.
The containers are positioned axially such that the specimens are centered on the core aldplane, thus spanning the central 5 feet of the 12-foot-high reactor core.
From a neutron transport standpoint, the surveillance capsule structures 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 neutron pad and the reactor vessel.
In order to properly determine the neutron environment at the test specimen locations, the capsules thinsolves aust be included in the analytical model. This requires at least a two-dimensional calculation.
In the naalysis of the neutron environment within the Byron Unit 4
1 reactor geometry, two sets of transport calculations were carried out.
The first, a single computation in the conventional forward mode, was used primarily to obtain spectrun-averaged reaction cross sections and radial gradient information for the pressure vessel. The second set of calculations consisted of a series of adjoint analyses relating the fast j
neutron (E > 1.0 MeV) flux at the surveillance capsule location and selected location on the reactor vessel inner wall to the power distributions in the reactor core. These adjoint importance functions, 4
when combined with cycle-specific core power distributions, yield l
I plant-specific f ast neutron exposure at the surveillance capsule and pressure vessel locations for each operating fuel cycle. Both the forward and adjoint calculations used an S angular quadrature.
8 4
e-s l
i f
i a e The forward transport calculation was carried out-in R, 8 geometry using the DDT two dimensional discrete ordinates code (4) and the SAILOR cross-section library.(5) The SAILOR library is a 47 group, ENDF/B-IV based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with a P3 *Xpansion of the cross-sections. The energy group structure used in the analysis is j
listed in "able 6-1.
The design basis core power distribution utilised in the forward analysis was derived from statistical studies of long-tors operation of Westinghouse 4-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 k uncertainty derived from the j
statistical evaluation of plant to plant and cycle to cycle variationa in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal + k level for a
~
large number of fuel cycles, the use of this design basis distribution is expected to yield somewhat conservative results. This is especially true in cases where low leakage fuel management has been employed.
[
The adjoint analyses were also carried out using the P cross 3
l section approximation from the SAILOR library. Adjoint source locations I
were chosen at the center of each of the surveillance capsules as well j
as at positions along the inner diameter of the pressure vessel. Again, j
these calculations were run in R, 8 geometry to provide power l
distribution importance functions for the exposure parameters of interest. Having the adjoint importance functions and appropriate core
)
power distributions, the response of interest is calculated as 1
l R
I(R,8,E) F(R,#,E) dERdRd8 (6-1)
R,8 * ' r F E i
i s-a i
.- - -l
where:
R
= Response of interest (e.g., p (E > 1.0 MeV)) at R,8 radius R and asinuthal angle 8.
I(R,8,E) = Adjoint importance function at radius R and asinuthal angle 8 for neutron energy group E.
F.(R,8,E) = Full power fission density at radius R and asinuthal angle 8 for neutron energy group E.
The fission density distributions used reflect the burnup-dependent inventory of fissioning actinides, including U-235, U-238, Pu-239, Pu-240 and Pu-241.
Core power distributions for use in the plant specific evaluation for Byron Unit 1, cycle 1 were obtained from WCAP-10315.(6)
The data extracted from reference 6 represented cycle averaged relative assembly powers. Therefore, the adjoint results applicable to capsule U represent the neutron flux averaged over cycle 1 which when multiplied by the cycle length yields the incremental fast neutron fluence for that cycle.
The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oak Ridge National Laboratory (ORNL) Poolside critical Assembly (PCA) facility as well as against the Westinghouse power reactor surveillance capsule data base. (7) The benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produce flux levels that tend to be conservative by 7 to 22%. When plant specific power distributions are used with the adjoint importance functions, the benchmarking studies show that fluence predictions are within *15% of measured values at surveillance capsule locations. The analysis is consistent with established ASTM standards.(8-12) 6.8 RADIOMBTRIC MONITORS The passive radiometric moniters included in the Byron Unit 1 surveillance program are listed in Table 6-2.
The first five reactions d
n d
e-4
u
-.aa.:
a-.
u e
+
4 in Table 6-2 are used as fast neutron monitors to determine neutron i.
fluence (E > 1.0 MeV). Exposure as measured by fluence (E > 1.0 MeV) is the primary parameter currently used to correlate sensured material property changes. However, the data provided by the five monitors sufficiently covers the neutron energy range so that other parameters proposed for damage evaluation (such as displacement per ston, dpa) may be determined with an acceptable degree of uncertainty.
In addition, bare and cadmium shielded cobalt-aluminua monitors are included in order to provide a measure of the thermal and resonance region neutron fluxes at the monitor location. These latter neutron energy regions are not currently thought to make any significant contribution to irrvilation damage in steel, but may be used for evaluations in the future. In addition, the thermal and resonance neutron measurements allow an evaluation of burn-in and burn-out, and impurity effects on the fast neutron reaction measurements.
l 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 j
in holes drilled in spacers at several axial levels within the capsules.
I The cadmium-shielded neptunium and uranius fission monitors are I
accommodated within the dosimeter block located near the axial center of the capsule. All monitors are located radially et the center of the i
capsule and asinuthally within *0.23 degrees of the capsule center.
d The use of passive monitors such as those listed in Table 6-2 does not yield a direct measure of the energy-dependent neutron flux 1
level at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux
)
level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well j
known. In particular, the following variables are importanti j.
)
i e-s
_~
- The operating history of the reactor
- The energy response of the monitor
- The neutron energy spectrum at the monitor location i
- The physical characteristics of the monitor The analysis of the passive monitors and the subsequent i
derivation of the average neutron flux require two operations. First, the disintegration rate of product nuclide per unit mass of monitor must be determined. Second, in order to define a suitable spectrun-averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated.
The specific activity of each monitors is determined using established ASTM procedures.(13-21) Following sample preparation, the activity of each monitor is determined by means of a lithium-drifted germanium, Ge(Li), samma ray spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed from Byron Unit 1, the overall 2a deviation in the measured data is determined to be plus or j
minus 10 percent. The neutron energy spectrum at the monito; location is determined analytically using the method described in para. graph 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 P
-At[e-At 3
g gd FY a (E) HE) dB 2 p
, -e g = N,g g g, E A
1 (6-2) g j=1 max 5
e-o
1
~
J where A
= induced product activity (dps per gram) g N,g
= number of target element atoms per gram l-F
= weight fraction of the target nuclide in the target i
g material Y
= number of product atoms produced per reaction g
a (E) = energy dependent reaction cross section y
p(E) = energy dependent neutron flux at the monitor location with the reactor at full (reference) power P)
= average core power level during irradiation period j P,,
= maximum or refersace core power level A
= decay constant of the product nuclide g
t)
= length of irradiation period j t
= decay time following irradiation period j i
d i
n
= total number of irradiation periods I
This equation may be simplified by defining R, the reactions g
per second, with reaction i occurring at full power.
R is sometimes g
referred to as the saturated activity or reaction rate.
A R
M l
g= N,gtgfp
- 3..,A gi3
_g g,
FY
,1-e
]
p j
3 j=1 max Substituting in Equation 6-2, j
R a (E) p (E) dB (6-4) g='E g
It should be noted that an assumption is made in the above equations j
that the ratio of p(E) to measured power (P)) does not vary with time.
This assumption is accurate for cases where fresh fuel is loaded on the periphery and no major fuel management changes are made. In other.
cases, corrections to the shorter lived reaction products must be made I
and in general, uncertainities in fluence derived from these products i
aust be increased.
i e-r I
a
,--,--J
Two methods were used to derive the neutrou flux based on Equation 6-4.
The first of these involves an average based on each reaction measurement and is described next. The second method involves a fit to the data using a least squares technique to minimise the uncertainty. The least sqaares method is described afterwards.
Because the neutron flux distributions o.re calculated using multigroup transport methods and because the main interest is in the f ast (E > 1.0 WeV) neutron flux, spectrum-averaged reaction cross sections are defined such that the integral term in Eq'untion 6-4 is replaced by the following relation.
P a(E) p(E) dB = 8 pf where p
7 a(E) p(E) dB a, p8 y,b p1 M
8 p(E)dB p
8
'1MeV g=1
.a 8
i p=
p(E)dE =
p g
8
' IWeY g=1 g = groupe number from Table 6-1.
Thus, Equation 6-4 is rewritten Rg=6pg i
or, solving for the fast (E > 1.0 WeV) neutron flux, R
p=
(6-5) g a
e-s
The total fast (E > 1.0 W V) neutron fluences is then given by n P pf=pf 2p t)
(6-8) j=1 max where total effective full power seconds of t3 = reactor operation up to the time of capsule j=1 max removal An assessment of the potential for product nuclide burnout may be made using the bare and cadalua shielded cobalt measured activities and published data for the 2200 m/s absorption cross-section and the j
resonance integral. This is done by rewriting Equation 6-2 in terms of l
i '
a monitor 2200 m/s neutron flux and a monitor resonance flux as follows:
R f2200 + Ry(7,,
(6-7) bare "
2200 i
Red =
R p,,,
(6-8) y l
ihere Rb ue = bare Co reaction rate 59 cadmium shielded Co reaction rate R
=
ed 60 2200
- Published 2200 m/s absorption cross-section for Co a
60 published epicadmium dilute resonance integral for Co RI
=
p2200 = a nit r 2200 m/s neutron flux to be determined from measured activities p
a nitor resonance neutron flux to be determined from
=
res
~
measured activities 6
1 s-e
i I
Equations 6-7 and 6-8 are solved for (2200 ""d f using the ree average measured bare and cadmium shielded cobalt activities at the l
nonitor location, i
The total loss rate of a product nuclide may then be expressed.
as the sua of its radioactive decay rate and the neutron absorption rate in that nuclide while the reactor is at power. The product nuclide I
neutron absorption rate may be estimated from the published data for and EI and the monitor fluxes determined above.
If the neutron 2200 absorption rate is small when compared to the decay rate then there is no concern regarding burnout, Similar corrections can be determined for other reactions a
affecting the measured results.
Of most significance are corrections l
for U fission in the U dosimeter and bufissionaftercapture I0 i
of neutrons in U.
The latter two step reaction only becomes 1
important for very long irradiations.
.t The least squares analysis technique is performed with the FERRET (22) code. This code employs a los normal least-squares algoritha which weights all the calculated fluzes, group cross sections, and 4
measured values with assigned uncertainties and correlations. All the quantities in the series of Equation 6-4 are simultaneously adjusted to i
give the minlaum weighted least squares error. The lognormal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
For the analysis of dosimetry data, the continuous quantities i
i l
(i.e., fluxes and cross sections) were approximated in 53 groups. The calculated 47 group flux-spectrum was expanded and contracted into the j
53 group structure using the SAND II code.(23) This procedure is carried out by first expanding the spectrum into the SAND II 620-group i
structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The high end of the spectrum is extrapolated using a fission spectrum form. The 620-point spectrum is then easily collapsed to the 53 groups.
i i
e-lo
The cross sections were also collapsed into the 53 energy-group structure using SAND II with the calculated spectrum (as expanded to 620 groups) as the weighting function. The cross sections were taken from the ENDF/B-Y dosimetry file. Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section.
Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.
For each set of data or a priori values, the inverse of the corresponding relative'covariance matrix M is used as a statistical weight. In some cases, as for the cross sections, a multigroup' covariance matrix is used.
For the fluxes, a simple parameterised form is used:
Es,=Rh+R,R,,p M
Eg' where R specifies an verall fractional normalisation uncertainty N
(i.e., complete correlation) for the corresponding set of values.
The fractional uncertainties R,specuy M Monal randon uncednind u for 1
group g that are correlated with a correlation matrix.
1
(
.) 2-gg, = (1 - 6) 6
, + 6 exp p
The first tern specifies purely randon uncertainties while the second tera describes short-range correlations over a range 7 (6 specifies the strength of the latter ters).
For the a priori calculated fluxes, a short-range correlation of 7 = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 6 is close to 1.
Strong long~ range correlations (or anticorrelations) are justified based on detailed studies carried out at ORNL.(24) For the integral reaction rate covariances, simple normalisation and randon uncertainties were combined e-11
as deduced from experimental uncertainties. Specific assignments of uncertainty values were made as follows:
l Flux normalisation uncertainty 30%
Flux group uncertainty 30% (E > 0.1 WeV)
Flux group uncertainty 50%-100% (E < 0.1 WeV)
Short range correlation fraction #
0.9 Reaction rate uncertainties See Table 6-7 The magnitude of the uncertainty assignments are based on general agreement of calculation and measurements for similar cases.
Since the uncertainties assigned to the fluxes are in general larger than those for the measurements, the solution is little affected by the specific values assumed.
6.4 NBUT10N ft1NSF0tf ANALYSIS LESULTS Results of the discrete ordinates transport calculations for the Byron Unit i reactor are summarised in this section.
In Figure 6-2, the calculated maximum f ast (E > 1.0 MeV) neutron flux levels at the radius of the surveillance capsule center, the reactor vessel inner radius, the reactor vessel 1/4 thickness location, and the reactor vessel 3/4 thickness location are presented as a function of asinuthal angle.
In Figure 6-3, the radial distribution of maximum f ast (E > 1.0 MeV) neutron flux through the thickness of the reactor vessel is shown. The relative axial variation of f ast neutron flux within the reactor vessel is given in Figure 6-4.
Absolute axial variations of f ast neutron flux may be obtained by multiplying the levels given in Figure 6-2 or 6-3 by the appropriate values from Figure 6-4.
Table 6-3 provides the i
calculated f ast neutron exposure parameters for the Byron Unit i reactor vessel.
Tsble 6-4 provides the calculated f ast neutron exposure parmasters and updated lead factors for all of the Byron Unit 1 4
s-ta
i j
I i
~
i surveillance capsules. The lead factor is defined as the ratio of the f ast (E > 1.0 MeV) neutron flux at the dosimeter block location (capsule center) to the maximum f ast neutron flux at the reactor vessel inner radius.
l In order'to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required.
The calculated neutron energy spectrum at the center of the Byron Unit 1 surveillance Capsule U is listed in Table 6-5.
6.5 D0SIMITRY RESULTS The irradiation history of the Byron Unit I reactor up to the j
time of removal of Capsule U is listed in Table 6-6.
These values were used to calculate the reaction rates according to Equation 6-3.
All the i
measured activities and reaction rate results are presented in Table j
s l
i-6-7.
l Neutron fluxes derived using the spectrum averaged cross sections are listed in Table 6-8.
A value for the f ast neutron flux is derived using the five fast neutron monitors. Yalues for resonance and l
2200 m/s thermal fluxes are derived based on the bare and cadmium 1
covered cobalt measurements. The thermal and resonance fluxes were used 1
to estimate absorption rates in the radioactive product nuclides. Burn out of these nuclides was found to have less than a 0.1% effect for all 58 except the Ni(n,p)58Co reaction where the correction is about 0.3%.
i Results from the FERRET calculation are given in Tables 6-9 j
through 6-11.
Table 6-9 presents the a priori and adjusted neutron
]
spectra in the 53 group structure. Adjusted spectral uncertainties are j
also given. Table 6-10 presents the ratios of calculated to measured reaction rates for all the reactions used both for the a priori spectrua j
and the adjusted spectrum. All the reaction rates agree with j.
calculation within the uncertainty. Derived values for integtol fluxes i
and dpa are presented in Table 6-11.
The uncertainty in the fi,nal results are also given as derived by FERRET from the input assumed 3
e-as 1
d
.i
,m r.-
,,m-
.m,
.v,m--._,--------%..m,-..-
t uncertainties and correlations. The uncertainties are therefore 4
estimates of the actual uncertainty in the flux and fluence results, and may be used for guidance in setting upper limit values.
Excellent agreement was obtained between the two flux evaluation methods. This follows from the consistency of the a priori spectral shape with the measurements. Minor differences between the two measured flux evaluations are due partly to the different weighting given the reaction rates and the different group structures used.
Based on the analysis, an average value of the two flux evaluation methods was chosen. This value is 9.63 x 1010 (,7,, ?_,,,,
E > 1.0 MeV) with an uncertalaty of 8% as evaluated by FERRET. This value may be compared with the design basis calculated flux value of 11 2
1.20 x 10 n/cm -see and a plant specific calculated flux value for the Byron Unit 1, Cycle 1 fuel loading which is 9.85 x 1010,je,2-sec.
The measured value is in excellent agreement with the latter value.
A sn===7 of measured and calculated current f ast (E > 1.0 MeV) neutron exposures for Capsule U is presented in Table 6-12.
The corresponding end-of-life (BOL) fluences are miso presented. Based on the data given in this analysis, the best estimate f ast (E > 1.0 MeV) i neutron exposure of the Byron Unit 1 Capsule U is:
4 = 3.50 x 1018,7,,2 (E > 1.0 MeV) at 1.15 EFPY j
i s-14
=-
i
?
i Table 6-1 SAILOR 47 NEUTRON ENERGY GROUP STRUCTURE Group Group Energy Lower Energy Energy I,ower Energy Group (MeV)
Group (MeV) 1 14.19(*)
25 0.183 2
12.21 26 0.111 3
10.00 27 0.0674 4
8.61 28 0.0409 5
7.41 29 0.0318 6
6.07 30 0.0261 7
4.97 31 0.0242 8
3.68 32 0.0219 9
3.01 33 0.0150 10 2.73 34 7.10 x 10-3 11 2.47 35 3.36 x 10-3 12 2.37 36 1.59 x 10-3 13 2.35 37 4.54 x 10~4 14 2.23 38 2.14 x 10~4 15 1.92 39 1.01 x 10-4 16 1.65 40 3.73 x 10-5 17 1.35 41 1.07 x 10-5 18 1.00 42 5.04 x 10-6 19 0.821 43 1.86 x 10-6 20 0.743 44 8.76 x 10-7 21 0.608 45 4.14 x 10~7 22 0.498 46 1.00 x 10~7 23 0.369 47 0.00 24 0.298
(*)The upper energy of group 1 is 17.33 MeV.
e-1s
i Table 6-2 i
NUCLEAR CONSTANTS FOR RADIOMETRIC WONITORS C0hTAINED IN THE BYRON UNIT 1 SURVEILLANCE CAPSULES Reaction Target Fission of Welsht Product Yield Monitor Waterials Interest Fraction Half-life 00 Iron wire Fe" (n,p) Mn" 0.058 312.2 dy Nickel wire N158 (n,p) Co68 0.6827 70.91 dy 63 (n,a) Co" 0.6917 5.272 yr Copper wire Cu Uranium-238(*) in U 0 38 (n,f) Cs #
1.0 30.17 yr 5.99
)
I 38 Neptunium-2'37(*) in Np0 Np237(n,f) Cs13I 1.0 30.17 yr 6.50 2
Cobalt-aluminum (*) wire Co69 (n,7) Co 0.0015 5.272 yr l
60 Cobalt-aluminum wire Co69 (n,7) Co60 0.0015 5.272 yr
.I
(*) Denotes that the monitor is cadmium-sbielded.
4 6-16
1
- Table 6-3 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS FOR THE PEAK LOCATION OF THE BYRON UNIT 1 REACTOR VESSEL i
Iron Radial Location Fast Neugron Flux Displacement within thu (n/cm -sec)
Rate ReactorVessel(")
(E > 1.0 WeV_1 OB > 0.1 WeV)
(dpa/see)
Ir ner Surf ace 3.14 x 10 7.61 x 10 4.99 x 10-11 10 10 (R = 86.500 inches) 4 1/4 Thickness 1.73 x 10 6.70 x 10 3.25 x 10-11 10 10 (R = 88.673 inches) 1/2 Thickness 7.82 x 10 4.72 x 10 1.92 x 10-11 8
10 (R = 90.846 inches) 3/4 Thickness 3.27 x 10 2.98 x 10 1.09 x 10-11 8
10 (R = 93.020 inches)
Outer Surface 1.35 x 10 1.50 x 10 5.24 x 10-12 8
10 (R=95.193 inches)
(*)The peak is located at 26.5 degrees asinuthally and on the core i
midplane. The peak exposure occurs in octants with a 12.5* neutron 1
pad and is calculated using design basis power distributions, i
I I
l G
e i
J l
e-17
1 l
I i
Table 6-4 CALCULATED FAST NEUTRON EXPOSURE PARAETERS AND LEAD FACTORS FOR TE BYRON UNIT 1 SURVEILLANCE CAPSULES Iron Asinuthg)
Fast Negtron Flux Displacement (n/ca -sec)
Rate Lead Capsule Location I.D.
(Derrses)
(B > 1.0 MeV)
(P > 0.1 VfQ (d9a/sec)
Factor (b) 11(*)
11 0.27 x 10-10 3.85 U
58.5 1.20 x 10 5.51 x 10 11 11 Y
61 1.13 x 10 5.14 x 10 2.12 x 10-10 3.65 I
11 11 1
W 121.5 1.20 x 10 5.51 x 10 2.27 x 10-10 3.85 11 11 I
238.5 1.20 x 10 5.51 x 10 2.27 x 10-10 3.85 11 11 10 Y
241 1.13 x 10 5.14'.c 10 3.12 x 10 3.65 11 11 Z
301.5 1.20 x 10 5.51 x 10 2.27 x 10-10 3.85
(*)The radius of the surveillance capsule center is 81.625 inches.
]
(b)The lead f actor is the ratio of the f ast (E > 1.0 MeV) neutron flux at the center of the surveillance capsule to that at the peak location on the reactor vessel inner surface.
(*)A plant specific evaluation for gele 1 81*** "" "#'8' II"*
2 value for capsule U of 9.85 x 10 n/cm -sec (E > 1.0 MeV).
1 1
I 4
i 4
i 5
6-18
Table 6-5 CALCULATED (*) NEUTRON ENERGY SPECTRUM AT THE CENTER OF BYROF UNIT 1 SURVEILLANCE CAPSULE U Energy Neuttgn Flux Energy NeutrgaFlux Group (n/cm -rec)
Croup (n/cm -sec) 7 10 1
2.25 x 10 25 6.77 x 10 7
10 2
8.25 x 10 26 7.03 x 10 8
10 3
2.85 x 10 27 5.66 x 10 8
10 4
5.17 x 10 28 3.97 x 10 8
10 5
8.55 x 10 29 1.20 x 10 6
1.88 x 10 30 6.44 x 10' 8
7 2.57 x 10' 31 1.74 x 1010 8
5.20 x 10' 32 1.11 x 1010 9
4.73 x 10' 33 2.01 x 1010 8
I0 10 3.96 x 10 34 2.96 x 10 8
10 11 4.75 x 10 35 5.01 x 10 12 2.37 x 10' 36 4.51 x 1010 8
10 13 7.30 x 10 37 6.30 x 10 14 3.67 x 10' 38 3.43 x 1010 8
10 15 9.84 x 10 39 3.81 x 10 10 10 16 1.33 x 10 40 5.15 x 10 10 10 17 2.07 x 10 41 6.08 2 10 10 10 18 4.61 x 10 42 3.37 x 10 j
10 10 19 3.52 x 10 43 3.85 x 10 10 10 20 1.67 x 10 44 2.39 x 10 10 10 21 6.07 x 10 45 1.86 x 10 10 10 22 4.75 x 10 46 2.63 x 10 10 10 23 5.94 x 10 47 3.27 x 10 10 24 5.75 x 10
(*) Design basis power distribution.
l I
e-se 4
.l Table 6-6 IRRADIATION HISTORY OF BYRON UNIT 1 SURVEIILANCE CAPSULE U Monthly p
Generation max Irradiation Time Decay Time P P Wonth Year (WW-Hr)
(MWt) 3 j,,
gp,y,)
gn,y,)
3 1985 81105 109.0 0.0320 31 822 4
1985 689973 958.3 0.2809 30 792 l
5 1985 980792 1318.3 0.3865 31 761
)
6 1985 1584137 2200.2 0.6450 30 731 7
1985 568273 763.8 0.2239 31 700 l
8 1985 1720850 2313.0 0.6781 31 669 9
1985 774891 1076.2 0.3155 30 639 10 1985 1764810 23; L1 0.6954 31 608 l
11 1985 0
0.0 0.0000 30 578 12 1985 799491 1074.6 0.3150 31 547 1
1 1986 2029647 2728.0 0.7998 31 516 l
2 1986 1211423 1802.7 0.5285 23 488 3
1986 2121673 2851.7 0.8360 31 457 4
1986 2336975 3245.8 0.9516 30 427 5
1986 2141879 2878.9 0.8440 31 396 j
6 1986 1777104 2468.2 0.7236 30 366 7
1986 615385 827.1 0.2425 31 335 8
1986 2191871 2946.1 0.8637 31 304 9
1986 2168895 3012.4 0.8831 30 274 10 1986 2018377 2712.9 0.7953 31 243 l
11 1986 2292247 3183.7 0.9334 30 213 12 1986 2301166 3093.0 0.9068 31 182 1
1987 1687879 2268.7 0.6651 31 151 2
1987 523101 1556.8 0.4564 14 137
.l (I)Dacay time is referenced to 7/1/87.
(2) Total irradiation time is 3.63 x 10 effective full power 7
~
see nds (EFPS) or 1.15 effective full power years (EFPY).
(3)P) is the average core power durins the irradiation period.
(4)P,,, is the rated thermal power, 3411 WWth, r.-so
i l
l Table 6-7 l
V.dASURED RADICMETRIC MONITOR ACTIVITIES AND AEACTION RATES FOR BYRON UNIT 1 SURVEILLANCE CAPSULE U j
Monitor and Activity Reaction Rate Uncertainty'*)
Aximi Location dis /see-na reactions /see-ston 5
l 23Sgg,,f)137Cs l
middle 1.50 x 10 3.84 x 10-14 i
2 b) 2 3.26 x 10-14 15 corrected 1.28 x 10 l
237,g,,f)137Cs 3
middle 1.30 x 10 3.06 x 10-13 10 3
be(n.nik
~
3 6.03 x 10-15 top 1.46 x 10 3
1 middle 1.35 x 10 5.57 x 10-15 3
botton 1.36 x 10 5.62 x 10-15 average 5.74 x 10-15 5
58 58 ggg,,,3 Co 4
top 1.04 x 10 7.68 x 10-15 3
middle 9.55 x 10 7.05 x 10-15 3
bottos 9.55,t 10 7.06 x 10-15 7.26 x 10-15 average 630u tn.o} 60Co 1
top 5.37 x 10 6.32 x 10'17 1
middle 4.87 x 10 5.73 x 10-17 1
bottos 4.93 x 10 5.80 x 10~17 5.95 x 10~17 5
average s-st l
l
i Table 6-7 (cont'd.)
Monitor and Activity Reaction Rate Uncertainty (*)
r-Axial I,ocation gg/sec-as reactions /sve-stos 59 60 Cofn.91 Co(Cd) top 5.16 x 10 2.60 x 10-12 3
middle 5.35 x 10 2.69 x 10-12 3
1 botton 5.20 x 10 2.62 x 10-12 3
2.64 : 10-12 10 average 59Co fn.9)"Co(bare) top 1.07 x 10 5.38 x 10-12 4
middle 1.08 x 10 5.43 x 10-12 4
i botton 1.06 x 10 5.33 x 10-12 4
average 5.38 x 10-12 10 i
(*)Uncerta.inties are based uncertainties except (1) isbtimat.d absolute radiometric counting i
uncertainty is increased due to uncertaintise in corrections for other fissions in the monitor and 1
variations in other surveillance capsule results, (2) Ni uncertaintyisggeressedduetotimehistorguncertaintiesbecause
)
of the shorter Co hsif-life, and (3) the Co reactions have increased uncertainty due to greater spatial variations of the low energy neutron flux.
(b) Corrected by 15% for U fission and other effects.
235 i
I e-as j
i Table 6-8 RESULTS OF NEUTRON DOSIMETRY FOR BYRON UNIT 1 SURVEILLANCE CAPSULE U Reaction Reaction Rate a (barne)(*)
Flux (n/ce _,',,)
2 FAST NEUTRONS (E > 1.0 MeV) 54Fe(n,p)b 5.74 x 10-1b 5.83 x 10-2(b) 9.85 x 1010 583g(,,p)58Ni 7.26 x 10-15 7.90 x 10-2(b) 9.19 x 1010 63Cu(n,a)60Co 5.95 x 10~17 7.00 x 10-4(b) 8.50 x 1010 238 (,,f)1370s 3.26 x 10-14 3.20 x 10-1(b) 10.19 x 1010 g
0) 10 p [,,g)137Cs 3,06 x 10-13 3.30 237 9.27 x 10 p
10 Average 9.40 x 10 RESONANCE & THERMAL NEUTRONS 10(e) 59Co(n,1)60Co (Cd) 2.64 x 10-12 7.55 x 10-23(c) 3.50 x 10 59Co(n,7)60Co (bare) 5.38 x 10-12 Bare Minus Cd 2.74 x 10-12 3.72 x 10-23(d) 7.37 x 1010(f) i i
(*) Cross sections are from ENDF/B-V dosimetry file (b) Cross section for f ast neutrons (E > 1.0 MeV)
(*)Resonanceintegral (d) Thermal (2200 m/s) cross section
(*) Resonance flux I
(I)2200 m/s flux i
i I
]*
i j
6-3s
"Qf R, %
C z
s E'
/
Table 0 '
4 l
FERRET-SAND II RESULTS FOR BYRON UNIT l1 SURVEILLANdB OAPSULE'U Energy A Pri 51 Flux
- E Adjttstjd Flux'
.N' Uncertain Group (MeV)
(n/cm -see)
'(n/cm -seeb
'(1~Std) 1 1.753E+01 1.014E+07 8.630E+06 e 22 2
1.492E+01 2.298E+07
- 1.05SE+07 P 20 3
1.350E+01 8.826E+07
. 7.555B+07 '
17 4
1.162E+01 1.9468+08
- 1.681E+08-14 5
1.000E+01-4.201E+08 3.085B+08 12 6
8.607E+00 6.936B+03 0.243E+08.
11 7
7.4088+00 1.5308409;
' 1'. 415B+00 -
10 8
6.065E+00 2.000E+00.
1.994E+09; 10 9
4.966E+00 4.200E+C9-4.139Et 0.
10 0
10 3.679E+00 6.377B+09 5.380E+00.
10 11 2.Sb5E+00 1.090E+10
-l'1073+10:
11 12 2.231E+00 1.482B+10.
11.499E+10 12 13 1.738E+00 2.084E+10 2.080E+10:
13 14 1.353E+00 2.320B+10 2.310E+10 14 15 1.108E+00 4.273E+10 4'.215E+10t 15 16 8.2088-01 4.938B+10-
'4.813E+10 17
~
17 6.393E-01 5.194B+10-5.003E410?
19 18 4.9793-01 3.817E+10 3.640E+10
'21 19 3.877E-01 5.458E+10 5'154S+10.
23 20 3.0203-01 5.687E+10 5.3383410 25 21 1.8323-01 5.697E+10 5.336E+10 27 22 1.111E-01 4.599E+10
'4.206E+10-29 23 6.738B-02 3.218B+10 3.010E+10 30 24 4.087E-02 1.828E+10
.1.715E+10-31 25 2.554E-02 2.405E+10 2.2668+10L 32 26 1.989E-02 1.186E+10 l ' '122E +10 '
33 27 1.503E-02 1.506E+10 1.430E+10'
-33 28 0.119E-03 2.169E+10 2.066E+10 60 29 5.531B-03 2.813E+10 2.087E+10
~ 60 Ob 3.355E-03 8.7468+09 8.376E+00-60 31 2.839E-03 8.3298+09 7.988B+00' 59 32 2.404B-03 8.003E+09 7.682B+00:
59 33 2.035E-03 2.251E+10 2.101E+10 58 34 1.234E-03 2.073E+10 1.989E+10 56 35 7.485B-04 1.919E+10 1.841E+10-53 36 4.540E-04 1.825E+10 1.760B+10 51 37 2.754E-04 1.961E+10 1.8703+10 48 38 1.070E-05 2.075E+10 1.992S+10 16 39 1.013E-05 2.100E+10 2.0213+10 47 I
8-24 m..
t 9[-
1 J'
l.;
1 Table 6-9 (cont'd.)
Energy A Pri 51 Flux
- Adjustgd Flux
% Uncertain Group (MeV)
(n/cm -sec)
(n/cm -see)
(1 Std)-
' -1 40 6.144E-05 2.090E+10 2.013E+10 49 41 3.727E-05 2.028E+10 1.967E+10 52 42 2.260E-05 1.953E+10 1.909E+10 54 43 1.371E-05 1.883E+10 1.854E+10 56 44 8.315E-06 1.786E+10 1.772B+10 57 45 5.043E-06 1.641E+10 1.644E+10 58 46 3.059E-06 1.488E+10 1.505E+10 58 47 1.855E-06 1.3292+10 1.357E+10 59 48 1.125E-06 1.150E+10.
1.185E+10 59 49 6.826E-07 9.531E+09 1.122E+10 103 50 4.140E-07 7.682E+09 1.022E+09 100 51 2.511E-07 7.142E+09 1.097E+10 93 52 1.523B-07 8.387E+09 1 471E+10 85 53 9.237E-08 2.461E+10 6.321E+10 32
- The a priori flux was taken from a generic 4 loop plant calculat u.and normalised to give approximately the correct e(n,p) reaction rate for the Byron Unit 1 capsule.
The thermal flux, which is severely depressed at the center oftggcapsuleisseparatelyadjustedtoapproximatelyfit the Co(n,7) reaction.
6-25
fic l
Table 6-10 COMPARISON OF MEASURED AND CALCULATED REACTION RATES
.1 USED IN THE ANALYSIS OF BYRON UNIT 1 CAPSULE U j
Reaction Rate Ratio Cale/ Mess Reaction Me aured Uncertainty 00 A Priori Adjusted j
De(n,p) 5.74 x 10 5
1.00 0.98
-15 58
-15 Ni(n,p) 7.27 x 10 8
1.07 1.02 Cu(n,a) 5.95 x 10-17 5
1.16 1.01 63 238 (n,f) 3.26 x 10 15 0.97 0.96
-14 U
237
-13 Np(n,f) 3.06'x 10 10 1.06 1.03
-12 Co(n,7)Cd 2.64 x 10 10 1.04 1.01 59
-12 Co(n,7) Bare 5.38 x 10 10 0.74 0.98 I
e 6-26
~. -....
I 1
Table 6-11 INTEGRAL NEUTRON FLUX RESULTS DERD'ED FROM ADJUSTED SPECTRUM Adjust d Yalue (n/cm$-see)
Uncertainty (la)
Parameter 10 Neutron Flux (B > 1.0 MeV) 9.86 x 10 8%
11 Neutron Flux (B > 0.1 MeV) 4.28 x 10 15%
10 Neutron Flux (E < 0.414 MeV) 9.91 x 10 26%
12 Total' Neutron Flux 1.03 x 10 15%
-10
. Iron Displacement (dpa/s) 1.89 x 10 11%
18 Neutron Fluence (E > 1.0 MeV) 3.58 x 10 8%
-3 Iron Displacements (dpa) 6.86 x 10 11%
4 5
d e
1 6-27 i
_,. _ _ _ _ _, _ _ _,. _.... _. _.... _.., _. _.. _ _. _.. _. _ _... -. - _ _ _. ~ _.... _....,. _.,. _. _,
Table 6-12 SUH! TiY OF BYRON UNIT 1 FAST (E > 1.0 MeV) NEUTRON
~
J
'!J;O'CE RESULTS BASED UPON SURVEILLANCE CAPSULE U j
End-of-Life Current Fast (E > 1.0 MeV)
Fast (E > 1.0 MeV)
Neutron Fluence (*)
Neutron Fluence (b)
(n/c=2)
(,fc,2)
{
Location Measured Calculated Measured Calculated 18 18 Capsule U 3.50 x 10 3.58 x 10 17 17 19 19 Vessel IR 9.09 x 10 9.30 x 10 2.53 x 10 3,14 x 10 17 17 19 19 g
Vessel 1/4T 5.00'x 10 5.11 x 10 1.39 x 10 1.72 x 10 10 18 18 18
,t Yessel 3/4T 9.45 x 10 9.66 x 10 2.63 x 10 3.24 x 10 h
,~
(") Current fluences are based on operation at 3411 Mwt for 1.15 EFPY and use a plant specific calculated fluence.
0)EOL fluences based on operation at 3411 Mwt for 32 EFPY and use design basis calculated fluences.
1 l
4 J
1 3
e 6.38
l 1
1 (TYPICAL)
C 0
0
- 61.0 N
- 58.5 s
Cu Y
T(
F 7
4 ll Mfh/////h l-81.625 e ^ ^
N.' \\
h kkkk'N O
Figure 6-1.
Plan view of a dual reactor vessel surveillance capsule 6-29
l curve 753074-A 20.0 i
i 10.0
- 8. 0 Surveillance Capsule Geometry
- 6. 0 0
S, Reactor Vessel IR
~
e X
E J2.0 5
2 S
LO i
1/4 T 1.ocation
- 0. 8 A
- 0. 6 E
~
~
3/4 T 1.ocation 0.2 0.1 i
i I
I O
20 40 60 80 100 Azimuthal Angle ( Degreel Figure 6-2.
Calculated asinuthal distribution of maximum f ast (E > 1.0 MeV) neutron flux within the reactor vessel-surveillance capsule geometry a
6-30
_.. ~ _ ___._
curve 753073-A 100 80 219.71 40 e.2 M5. 8 x
i 20 IR "Ta 230.75 ui 4T
[ 10 yS 2 % Z1 6
~
^
1/2T t
g 4
C 241.79 3/4 T l
2 1
Oh I
i i
i i
200 210 220 230 240 250 Radius (cm)
Figure 8-3.
Calculated radial distribution of maximum fast (E > 1.0 MeV) neutron flux within the reactor vessel 6-31
i l
t17565 i
1.000 0.700 0.500 0.200 0.100 5 0.070 e
0.050
~
4 s>=
- S
=
$ 0.020 C<
d i
e O.010 0.007 0.005 CORE MIDPLANE 0.002 TO VESSEL
' CLOSURE HE AD 0.001 300 200 100 0
100 2 00 300 DISTANCE FROM CORE MIDPLANE (cm)
Figure 8-4.
Relative axial variation of f ast (E > 1.0 MeV) neutron flux within the reactor vessel 6-32 l
7.
SURVEILLANCE CAPSULE REMOVAL SCIIEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the Byron Unit i reactor vessel:
Capsule Estimated Location Lead Capsule (den.)
Factor Removal Time (*) Fluenge (n/cm )
18 U
58.5 3.85 1.15 (Removed) 3.50 x 10 19(b) j Y
241 3.65 5
1.80 x 10 19(c)
V 61 3.65 9
3.20 x 10 19 I
238.5 3.85 15 5.60 x 10 W
121.5 3.85 Standby Z
301.5 3.85 Standby
(*) Effective full power years from plant startup.
Approximate fluence at 1/4 thickness reactor vessel wall at end of life.
(*) Approximate fluence at reactor vessel inner wall at end of life.
7-1 i
- - ~,
.r--
-, ~
~
i 8.
REFERENCES 1.
J. A. Davidson, ' Commonwealth Edison Company Byron Station Unit No. 1 Reactor Vessel Radiation Surveillance Program', WCAP-9517, July 1979.
2.
Code of Federal Regulatiou, 10CFR50, Appendix G, ; Fracture Toughness Requirements', and Appendix H, ' Reactor Vessel Material Surveillance Program Requirements', U.S. Nuclear Regulatory Commission, Washington, D.C.
3.
Regulatory Guide 1.99, Proposed Revision 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.
J. V. Alexander, Jr., et al., 'The Nuclear Design and Core Physics Characteristics of the Byron Unit 1 Nuclear Power Plant, Cycle l',
WCAP-10315 Rev.1, December,1984. (Proprietary) 7.
S. L. Anderson and K. C. Tran, ' Benchmark Testing of Westinghouse Neutron Trusport Analysis Methodology-PCA Evaluations', WCAP-11428, September 1987.
8.
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 Materials, Philadelphia, PA, 1984.
9.
ASTM Designation E560-77, ' Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
s-1
l 10.- ASTM Designation E693-79, ' Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 11. ASTM Designation E706-81a, ' Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 12. 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, 1984.
- 13. ASTM Designation B261-77, ' 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, 1984.
- 14. ASTM Designation E262-77, ' Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
l
- 15. ASTM Designation E263-82, ' Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron', in ASTM l
Standards, Section 12, American Society for Testing and Materials, i
Philadelphia, PA, 1984.
1
- 16. ASTM Designation B264-82, ' 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, 1984.
- 17. 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, 1984.
- 18. 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,
' Philadelphia, PA, 1984.
l i
I s-a
- 19. 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, 1984, 20.
ASTM Designation E705-79, ' Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237', in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
)
i
- 21. 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, 1984.
22.
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