Text

Analysis of Capsule U from Duke Power Co McGuire Unit 2 Reactor Vessel Radiation Surveillance Program
	ML20128C299
Person / Time
Site:	Mcguire
Issue date:	10/31/1992
From:	Shaun Anderson, Chicots J, Madeyski A; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20128C305	List: ML20128C299;
References
	WCAP-13516, NUDOCS 9302030378
	Download: ML20128C299 (187)
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~ WESTINGHOUSE CIASS 3 (Non-Proprietary)

WCAP-13516 ANALYSIS OF CAPSULE U FROM.THE DUKE POWER COMPANY i

MCGUIRE UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM J. M. Chicots S. L. Anderson A. Madeyski October 1992 Work Performed Under Shop Order DXBP-106B Prepared by Westinghouse Electric Corporation for the Duke Power Company l

1.0 MeV)

Fluence 2.02 x 1019 5-2 Charpy V-Notch Impact Data for the McGuire Unit 2 5-9 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 2.02 x 10 n/cm2 (E > 1.0 MeV) 19 5-3 Instrumented Charpy Impact Test Results for the McGuire 5-10, Unit 2 Intermediate Shell Forging-05 Irradiated ht 550*F, n/cm2 (E > 1.0 MeV)
Fluence 2.02 x 1019 5-4 Instrumented Charpy Impact Test Results for the McGuire 5-11 Unit 2 Weld Metal and Heat-Affected-Zone (HAZ)- Metal, 19 2
Irradiated at 550'F, Fluence 2.02 x 10 n/cm (E > 1.0 MeV)-
19 2
5-5 Effect of 550*F Irradiation to 2.02 x 10 n/cm 12 (E > 1.0 MeV) on the Notch Toughness Properties of the McGuire. Unit 2 Reactor Vessel Surveillance Materials iii
LIST OF TABLES (Continued)
Table Title hgg 5-6 Comparison of the McGuire Unit 2 Surveillance Materiai 5-13 1
30 ft-lb Transition Temperature Shifts and Upper Shelf ~
Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-7 Tensile Properties for the McGuire Unit 2 Reactor Vessel 5-14 Surveillance Materials Irradiated at 550'F to 2.02 x 10l9 n/cm2 (E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Parameters at-the 6-15 Surveillance Capsule Center 6-2 Calculated F&st Neutron Exposure Parameters at the Pressure 6-16 Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron-Flux (E > 1.0 MeV)
.6-18 within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux (E > 0.1 MeV)-
6-19 within the Pressure Vessel Wall 6-5 Relative Radial Distributions of Iron Displacement Rate 6-20 (dpa) within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6 6-7 Monthly Thermal Generation During the First Seven Fuel 6-22 Cycles of the W. B. McGuire Unit 2 Reactor iv w
-p y
t LIST 0FTABLES.(Continued)
/*
Table Title Eage
- ~ '
6-8 Measured Sensor Activities and Reactions Rates-6-23 Surveillance Capsule-U 6-9 Measured Sensor Activities and Reactions Rates 25 Surveillance Capsule X 6-10 Measured Sensor Activities and Reactions' Rates 6 Surveillance Capsule V 6-11 Summary of Neutron Dosimetry Results Surveillance Capsule U 6-29 6-12 Summary of Neutron Dosimetry Results Surveillance Capsule X 6-30 t
6-13 Summary of Neutron Dosimetry Results-Surveillance Capsule V 31 6-14 Comparison of Measured and FERRET Calculated Reaction 6-32--
Rates at the Surveillance Capsule Center-Surveillance-Capsule U
~
6-15 Comparison of Measured and FERRET Calculated Reaction -
6-33 Rates at the Surveillance Capsule Center Srrveillance Capsule X l
6-16 Comparison of Measured and-FERRET Calculated Reaction 6 Rates at the Surveillance Capsule' Center Sntveillance Capsule V 6-17 Adjusted Neutron-Energy Spectrum at the Center of 6 [
Surveillance Capsule U l-6-18 Adjusted Neutron Energy Spectrum at the Center of 6-36~
l, Surveillance Capsule-X l
V
LIST OF TABLES (Continued)
Table Title Pjtqt 6-19 Adjusted Neutron-Energy Spectrum at the Center t '
6-37:
Surveillance Capsule V 6-20 Comparison of Calculated and Measured Exposure Levels 6-38 For W. B. McGuire Unit 2 Surveillance Capsules 6-21 Neutron Exposure Projections at Key Locations on the 40-H-
Pressure Vessel Clad / Base Metal litterface 6-22 Projected Fast Neutron Exposure Values 6-41 6-23 Updated Lead Factors for McGuire Unit 2 Surveillance 6-42 Capsules 9
6 e
vi
- =
-LIST OF ILLUSTRATIONS.
Fiaure-Title
Elgg, 4-1 Arrangement of Surveillance Capsules in the McGuire. Unit 2 4-0 Reactor Vessel 4-2 Capsule U Diagram Showing Location of Specimens, Thermal 4-10' Monitors and Dosimeters 5-1 Charpy V-Notch impact Properties for McGuire Unit 2 Reactor 5-15' Vessel Intermediate Shell Forging 05 (Tangential Orientation) 5-2 Charpy V-Notch Impact Properties for McGuire Unit 2 Reactor 5-16 Vessel Intermediate Shell Forging 05 (Axial Orientation).
5-3 Charpy V-Notch Impact Properties for McGuire Unit 2 Reactor 5 Vessel Surveillance Weld Metal 5-4 Charpy V-Notch Impact Properties for McGuire Unit 2 Reactor 5-18:
Vessel Weld Heat-Affected-Zone Metal 5-5 Charpy impact Specimen Fracture Surfaces for McGuire Unit-2 5-19 Reactor Vessel Intermediate Shell Forging 05 (Tangential Orientation) 5-6 Charpy impact Specimen Fracture Surfaces for McGuire Unit 2 5-20 Reactor Vessel Intermediate Shell Forging 05 (Axial Orientation)
O vii
LIST OF ILLUSTRATIONS (Continued)
Fiaure Title P_ang 5-7 Charpy impact Specimen Fracture Surfaces for McGuire Unit _2 5-21 Reactor Vessel Surveillance Weld Metal 5-8 Charpy impact Specimen Fracture Surfaces for McGuire Unit 2 5-22 Reactor Vessel Weld Heat-Affected-Zone Metal 5-9 Tensile Properties for McGuire Unit 2 Reactor Vessel 5-23 Intermediate Shell Forging 05 (Tangential Orientation) 5-10 Tensile Properties for McGuire Unit 2 Reactor Vessel 5-24 Intermediate Shell Forging 05 (Axial Orientation) 5-11 Tensile Properties for McGuire Unit 2 Reactor Vessel 5-25 Surveillance Weld Metal
~
5-12 Fractured Tensile Specimens from McGuire Unit 2 Reactor
.5-26 Vessel Intermediate Shell Forging OC (Tangentiai Orientation) 4 5-13 Fractured Tensile Specimens from McGuire Unit 2 Reactor 5-27 Vessel Intermediate Shell Forging 05 (Axial Orientation) 5-14 Fractured Tensile Specimens from McGuire Unit 2 Reactor 5-28 Vessel Surveillance Weld Metal 5-15 Engineering Stress-Strain Curves for Forging 05 5-29 Tensile Specimens DL1 and DL2 (Tangential Orientation)-
e viii 1
l LIST OF ILLUSTRATIONS (Continued)
Fiaure Title P_tgg 5-16 Engineering Stress-Strain Curve for Forging 05 5-30'
]
Tensile Specimens DL3 (Tangential Orientation) l 5-17 Engineering Stress-Strain Curves for Forging 05 5-31 i
Tensile Specimens DTl and DT2 (Axial Orientation) 5-18 Engineering Stress-Strain Curve for Intermediate Shell 5-32 Forging 05 Tensile Specimen DT3 (Axial Orientation) 5-19 Engineering Stress-Strain Curves for Weld Metal 5-33 Tensile Specimens DW1 and DW2 5-20 Engineering Stress-Strain Curve for Weld Metal 5-34 Tensile Specimen DW3 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-1S 6-2 Axial Distribution of Neutron Fluence (E >1.0 MeV) Along-6-14 the 45 Degree Azimuth I
e ix
SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel :naterihls contained in surveillance capsule U, the third capsule to be removed from the Duke Power Company McGuire Unit 2 reactor pressure vessel, led to the following conclusions:
The capsule received an average fast neutron fluence (E > 1.0 MeV) of o
19 2
2.02 x 10 n/cm after 6.05 EFPY of plant operation, o
Irradiation of the reactor vessel intermediate shell forging 05-Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction (tangential orientation), to 2.02 x 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition 19 temperature increase of 90*F and a 50 ft-lb trane.ition temperature increase of 105'F. This results in a 30 ft-lb transition temperature of 15'F and a 50 ft-lb transition temperature of 45'F for tangentially oriented specimens.
o Irradiation of the reactor vessel intermediate shell forging 05 Charpy specimens, oriented with the longitudinal axis of the specimen normal 19 to the major rolling direction (axial or=lentation), to 2.02 x 10 n/cm2 (E > 1.0 MeV) resulted in a 30 f t-lb transition temperature increase of 85'F and a 50 ft-lb transition temperature increase of 90*F.
This results in a 30 ft-lb transition temperature of 60'F and a 50 ft-lb transition temperature of 115'F for axially oriented specimens.
19 2
o The weld metal Charpy specimens irradiated to 2.02 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 20*F and a 50 ft-lb transition temperature increase of 35*F.
This results in a 30 ft-lb transition temperature of
-30'F and a 50 ft-lb transition temperature of 10*F for the weld metal.
1-1
o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 2.02 x 10 n/cm2 (E >'1.0 MeV) resulted in a 19 30 ft-lb transition temperature increase.of 85'F and a 50 ft-lb-transition temperature increase of 85'F.
This result.t in'a 30 ft-lb transition temperature _of -5'F and a 50 ft-lb transition temperature of 30'F for the weld HAZ metal.
o Irradiation of intermediate shell forging 05 (tangential orientation) to 2.02 x 10 n/cm2 (E > 1.0 MeV) resulted in an average upper 19 shelf energy decrease of 34 ft-lbs, resulting in an upper shelf energy of 122 ft-lbs.
o Irradiation of intermediate shell forging 05 (axial orientation) to 2.02 x 1019 n/cm2 (E > 1.0 MeV) resulted in an average upper shelf energy decrease of 10 ft-lb, resulting in an upper shelf energy of 84 ft-lbs, o
The average upper shelf energy of the weld metal decreased 4 ft-lb I9 after irradiation to 2.02 x 10 n/cm2 (E > 1.0 MeV).
This results in an upper shelf energy of 129 ft-lb for the weld metal.
o The average upper shelf energy of the weld HAZ metal decreased 11 ft-lb after irradiation to 2.02 x 1019 n/cm2 (E > 1.0 MeV). This results in an upper shelf energy of 93 ft-lb for the weld HAZ metal.
o The surveillance capsule U test results indicate that the intermediate shell forging 05 and the surveillance weld material 30 ft-lb transition temperature shift is less than the Regulatory Guide 1.99 Revision 2 predictions, o
The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are.
i expected to maintain an upper shelf energy of no less than 50 ft-lb -
throughout the life (32 EFPY) of the vessel as required by 1GCFR50, Appendix G.
1-2 t
~
- o The' calculated end-of-life '(32 EFPY) maximum neutron fluence (E > 1.0 fleV) for the McGuire Unit 2 reactor vessel-is as follows:-
19 2
Vessel inner radius * - 2.04 x 10
.n/cm Vessel 1/4 thickness - 1.09-x 1019 n/cNi2 18 2-Vessel 3/4 thickness - 2.20 x 10 n/cm
Clad / base metal interface p
4 W
1-3
i SECTION
2.0 INTRODUCTION
This report presents the results of the examination of capsule _U, the third -
a
~
capsule to be removad from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Duke Power company McGuire Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the McGuire Unit 2 reactor pressure vessel materials was designed and reco:nmended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vess; materials is presented in WCAP-9489, entitled " Duke Power Company William B. McGuire Unit No. 2 Reactor Vessel Radiation Surveillance Program" by Koyama and Davidsonlll. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Power Systems personnel were contracted to aid in the preparation of procedures for removing capsule U from the reactor and its shipment to the Westinghouse Science and Technology Center Hot Cell Facility, where, the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance capsule V removed from the Duke Power Company McGuire Unit 2 reactor vessel and discusses the analysis of the data.
W 2-1
SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast?
neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 508 Class 2 (base. material of the McGuire Unit 2 reactor pressure vessel intermediate shell forging 05) are well documented in the literature.
Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and-toughness under certain conditions of irradiation.
A method for performing onalyses to guard egainst fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure,"
Appendix G to Section III of the ASME Boiler and Pressure ' Vessel Codel43
~
The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNGT)*
RTNDT is defined as the greater of either the drop weight nil-ductility-transition temperature (NDTT per ASTM E-208)[5] or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens with the tangential-axis oriented normal -(axial orientation) to the major working direction of the forging.
The'RTNDT Of
given material is used to index that material to a reference stress intensicy.
factor curve (KIR curve) which appears in Appendix G to -the ASME-Code. The KIR curve is a lower bound of dynamic, crack arrest,-and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material-is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature.
'_ Allowable operating limits can then be determined using these allowable stress 4-
-intensity factors.
3-1 e
~,
o- - -
+
._. ~.
i RTNDT and, in turn, the operating limits of nuclear power plants canLbe -
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 McGuire Unit 2 Reactor Vessel Radiation-Surveillance Programill, in which a surveillance capsule is-periodically removed-from the-operating nuclear reactor and the encapsulated specimens tested. Tne increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial +.
ARTNDT) is used to index the material to the KIR curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
S 9
k 3-2
SECTION
4.0 DESCRIPTION
OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure' on the McGuire Unit 2 reactor pressure vessel core region material were inserted in
~
the reactor vessel _ prior to. initial plant start-up.
The six capsules were.
positioned in the reactor vessel between the neutron shielding pads and the vessel wall as shown in Figure 4-1.
The vertical center of the capsules is opposite the vertical center of the core.
Capsule U was removed after 6.05 Effective Full Power Years-(EFPY) of plant operation.
This capsule contained Charpy V-notch, tensile, and 1/2 T compact.
tension (CT) specimens (Figure 4-2) from the submerged' arc weld metal fabricated with the same weld wire and flux as used in the reactor vessel core region girth weld, and Charpy V-notch, tensile, CT, and bend bar specimens (Figure 4-2) from the intermediate shell forging 05.
Capsule Y also contained Charpy V-notch specimens from weld Heat-Affected-Zone (HAZ) material.- All heat-affected zone specimens were obtained from within the HAZ of forging 05.
Test specimens obtained from the intermediate shell forging (after thermal heat treatment and forming of the forging) were taken from at least one forging thickness from the quenched ends of the forging. Test specimens'were machined from the 1/4 thickness location of the forging after performing a simulated postweld stress-relieving treatment on the material and also from weld and heat-affected-zone metal of a stress-relieved weldment joining the lower and intermediate shell forgings. All heat-affected-zone specimens were obt'afned from the weld heat-affected-zone of the intermediate shell forging.
Charpy V-notch specimens from the-intermediate shelll forging were machined in both tangential (longitudinal axis of the specimens-parallel to the major-working direction) and axial (longitudinal axis normal to the-. major working L
direction) orientations. The Charpy V-notch specimens from the weld and weld L,
heat-affected-zone metal were machined perpendicular to the weld direction with the notch oriented in the direction of the weld.
4-1
Tensile specimens from.the intermediate snell forging were machined with the tangential axis of the specimen both nsrallel ~and normal to the major working-direction. Welc' specimens were oriented normal to the weld direction.
Bend bar specimens were machined from the intermediate shell forging with the longitudinal axis of the specimen oriented parallel to the working direction of the forging such that the simulated crack would propagate normal to the working direction of the forging. All bend bar specimens were fatigue precracked according to ASTM E-399.
Compact tension test specimens from the intermediate shell forging were machined in both axial and tangential orientations. This was done to obtain toughness data Soth normal and parallel to the major working direction of the forging. Compact tension test ' specimens from the weld were-machined normal to the weld direction with the notch oriented in the direction of the weld. All specimens were fatigue precracked according to ASTM E-399.
The chemical composition and heat treatment of the surveillance material is presented in Tables 4-1 and 4-2.
The chemical analysis and heat treatment reported in Tables 4-1 and 4-2 were obtained from unirradiated material used in the surveillance programfil.
In addition, a chemical analysis performed on four irradiated Charpy specimens from the weld and base metal forging 05 is reported in Table 4-3.
The chemistry results from the NBS certified reference standards are given in Table 4-4.
Capsule U centained dosimeter wires of pure iron, copper, nickel, and aluminum-cobalt (cadmium-shielded and unshielded).
In addition, cadmium-shielded dosimeters of Neptunium (Np237) and Uranium (U238) were contained in the capsule.
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule and were located as shown in Figure 4-2.
S 4-2
The two eutectic alloys and their melting points are:
2.5% Ag, 97.5% Pb Melting Point 579'F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590*F (310*C)
The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule V are shown in Figure 4-2.
a S
f 4-3
TABLE 4-1 HEAT TREATMENT OF THE MCGUIRE UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALSIll Material Temperature (*F)
Time thr)
Coolant Intermediate Shell 1688-1697 3.5
~ Water quenched Forging 05 (HT. 526840) 1229-1238 7.5 Air cooled 1140 1 25 22.0 Furnace cooled Weld Metal 1140 i 25 15.0 Furnace cooled O
' i e
4-4
TABLE 4-2 CHEMICAL COMPOSITION OF THE HCGUIRE UNIT 2 REACTOR VESSEL'
=*
SURVEILLANCEMATERIALSIll Forging 05 (HT. 526840)
Weld Metal [a]
L1 ment (Wt. %)
(Wt. %)
C 0.18 0.055-S 0.017 0.015 N
0.004 0.011 Co 0.019 0.007 Cu 0.16 0.031 Si 0.23 0.29 Mo 0.58 0.55 Ni 0.79 0.73 Hn 0.69 1.81 Cr 0.43 0.030 V
< 0.002
< 0.002
~
P 0.012 0.016 Sn 0.008 0.002 Ti
< 0.001 0.004 Pb 0.001 0.003 W
< 0.002
< 0.002 Zr
< 0.002
< 0.002 As 0.018 0.015 A1 0.016 0.015 B
< 0.003
< 0.003 Sb-
< 0.002
< 0.002 (a)
Surveillance weld specimens were made of the same weld wire and flux as the girth seam weldment between forgings 04 and 05 (Weld Wire Heat-No. 895075 and Grau L.O. Flux Lot No. P46) 4-5
c.
TABLE 4-3 CHEMICAL COMPOSITION OF FOUR McGUIRE UNIT 2 CHARPY SPECIMENS REMOVED FROM' SURVEILLANCE CAPSULE U Concentration in Weicht Percent DW-3 DW-11
-DW-15 DL-8 Betals Weld Weld.
Weld Base Metal Fe MATRIX MATRIX MATRIX HATRIX Co 0.014 0.014 0.014 0.014 Cr 0.045 0.031 0.051 0.410-Cu 0.039 0.036 0.045 0.151 Mn 1.875 1.963 1.882 0.730 Mo 0.590 0.597 0.613 0.628 Ni 0.765 0.747 0.776 0.820 P
0.014 0.015 0.014 0.014 Ti 0.002 0.005 0.005.
<0.001 V
0.002 0.001' O.001 0.003-Al 0.011 0.012 0.012 0.007.
As 0.014 0.014 0.014 0.027 B
0.004 0.004 0.002 0.003 Nb 0.011 0.002 0.002 0.001 Sn
<0.001 0.003 0.005 0.008 W
<0.001 0.001 0.001-0.004 Zr
<0.001 0.001 0.001
<0.001 C
0.063 0.053 0.061 0.167-S 0.007 0.005 0.005 0.012 Si 0.147 0.148 0.146 0.lz9 Analyses Method of Analysis Metals ICPS, Inductively Coupled Plasma Spectrometry Carbon EC-12, LLCO Carbon Analyzer Sulfur Combustion / titration Silicon Gravimetric 4-6
'l
'Y TABLE 4-4 CHEMISTRY RESULTS_FROM THE NBS CERTIFIED REFERENCE STANDARDS HATERIAL 10 Low Alloy Steel:
NIST Control Standard' Concentration in Weicht Percent NBS 361 NBS 362 Metals Certified Beasured Certified Measured Fe 95.6000 (Matrix) 95.3000 (Matrix)
Co 0.0320 0.n31 0.3000 0.305' Cr 0.6940 0.678 0.3000 0.293 Cu 0.0420 0.043 0.5000 0.5 Mn 0.6600 0.668 1.0400 1.053-Mo 0.1900 0.195 0.0680 0.051-Ni 2.0000 2.034 0.5900 0.579 P
0.0140 0.018 0.0410 0.04 Ti 0.0200 0.013 0.0840 0.019 l
V 0.0110 0.011 0.0400 0.0:J Al 0.0210 0.021 0.0950 0.081 As 0.0170 0.033-0.0920 0.134 B
0.0003 0.001 0.0025 0.004 Nb 0.0220 0.027 0.2900 0.022 Sn 0.0100 0.005 0.0160 0.016 W
0.0170 0.018 0.2000 0.180 Zr 0.0090 0.007 0.1900 0.196 C
0.3830 0.382 0.1600 0.158-S 0.0140 0.0360 0.037 Si 0.2220 0.194 0.3900 0.407 BMWWBMWeapSRWWWWWWWWWERABSSSWWWWWhMBBSWWW2WBBBBBBamWEBEEMMWWESzastammaamm 4-7
TABLE 4-4 continued CHEMISTRY RESULTS FROM THE NBS CERTIFIED REFERENCE STANDARDS MATERIAL 10 Low Alloy Steel: NIST Control Standard Concentration in Weiaht Percent NBS 363 NBS 364
1etals Certified figAigrgd Certified liq 1tuted s
Fe 94.4000 (Matrix) 96.7000 (Matrix)
Co 0.0480 0.045 0.1500 0.154 Cr 1.3100 1.243 0.0630 0.066 Cu 0.1000 0.098 0.2490 0.249 Hn 1.5000 1.497 0.2550 0.251 Ho 0.0280 0.027 0.4900 0.494 Ni 0.3000 0.302 0.1440 0.139 P
0.0290 0.034 0.0100 0.010 Ti 0.0500 0.033 0.2400 0.211 V
0.3100 0.299 0.1050 0.105 L
A1 0.2400 0.242 0.0080 0.014 As 0.0100 0.018 0.0520 0.067 8
0.0008 0.002 0.0106 0.013 Nb 0.0490 0.059-0.1570 0.113 Sn 0.1040 0.106 0.0080 0.004 W
0.0460
<0.01 0.1000 J.047 Zr 0.0490 0.032 0.0680 0.062 0.8700 C
0.6200 0.0250 0.024 S
0.0068 0.0650 Si 0.7400 4-8 r
. + ~, - -
J REACTOR VESSEL O'
~
CORE BARREL NEUTRON PAD U
hf F
l Bo' " 56' 50, b,-
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Arrangement of Surveillance Capsules in the McGuire Unit 2 Reactor Vessel 4-9
j c%
aw Crur'ACT CfM*ACT CfwACT Cnri Df90 SAR TD61L ir.
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CORC Td SPECluCN NUMDERING CODE OL - INTEf4CDI ATE DELL FORG!to 05 (TAPGENTI AL.)
r1 - INTERMEDI ATE DELL FOR0!!G OS ( AX1 AL) tN. WLD METAL OH - FEAT. ArfECTED.20rE MATERI AL Y
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ll l Fe Nl ll l i6JLJLJ LJLJLJ f ' I] ['a t==== A 4 -,11R Co (Cd)
(' I' ( ]= t - Al=.itDE Co (C4) 64 1 Ni c wita noa m4 w n sscu TO ICTTOW W W1MiEL Figure 4-2 Capsule U Diagram Showing Location of Specimens.
Thermal Monitors and Dosimeters 1
4-10
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i SECflCN 5.0 1ESTING Of SPEclHENS FROM CAPSULE U 5.1 Qyerview i
The post-irradiation eachanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse Power Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H(2), ASTM Specification E185-82(63, and Westinghouse Remote Meta 11ographic Facility (RMF) Procedure RMF 8402, Revision 2 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-9488Ill. No discrepancies were found.
Examination of the two low-melting point 579'f (304'C) and 590*F (310'C) eutectic alloys indicated no melting of either type of thermal monitor.
Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579'F (304'C).
The Charpy impact tests were performed per AS1H Specification E23-88{73 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358J machine, The tup (striker) of the Charpy machine is instrumented with a GRC 8301 instrumentation system, feeding information into an IBM XT computer. With this system, load-time and energy-tiine signals can be recorded in addition to the standard measurement of Charpy energy (E ).
From the load-time curve D
(Appendix A), the load of general yielding (Pay), the time to general yielding (tgy), the maximum load (P ), and the time to maximum load (t )
M M
can be determined.
Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pp), and the load at which fast fracture terminated is identifici as the arrest load (P )*
A i
5-1
i The energy at maximum load (E ) was determined by comparing the energy-time M
record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen.-
Therefore, the propagation energy for the crack (E ) is the difference p
between the total energy to fracture (E ) and the energy-at maximum load.
D The yield stress (oy) was calculated from the three-point bend formula having the following expression:
oy - Pgy * (L/[B*(W-a)2*C)}
(1) where L - distance between the specimen upports in the impact testing machine; B
the width of the specimen measured parallel to the notch; W = height of the specimen, measured perpendicularly to the notch; a - notch depth, The constant C is dependent on the notch flank angle (d), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending),
in three-point bending a Charpy specimen in which p - 45* and p 0.010", Equation 1 is valid with C - 1.21.
Therefore (for L 4W),
oy - Pay * (L/[B*(W-a)2*1.21]) - (3.3PgyW)/[B(W-a)2]
(2) for the Charpy specimens, B - 0.394 in., W = 0.394 in., and a = 0.079 in.
Equation 2 then reduces to:
oy - 33.3 x Pay (3) where oy is in units of psi and Pay is in units of lbs.
The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.
Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-89(8),
The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
5-2
Tension tests were performed on a 20,000-pound Instron Model 1115, split-console test machine, per ASTM Specification E8-89bl93 and E21-79 (1988)(10), and RMF Procedure 8102, Revision 1.
All pull rods, grips, and pins were made of Inconel 718 hardened to HRC45.
the upper pull rod was
~
connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.
Extension measuremants were made with a linear variable displacement transducer (1VDT) 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 e>tensometer is rated as Class B-2 per ASTM E83-85(12),
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air.
Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.
Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip.
In 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 ovtr the range of room temperature to 550'F (288'C).
The upper grip wa. 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 12'F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diametur and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5-3 I
__ ________________.______~___,j
5.2 Charov V-Notch Imoact Test Results The results of the Charpy V-notch impact tests performed on the various I9 2
materials contained in Capsule U, which was irradiated to 2.02 x 10 n/cm (E > 1.0 MeV), are presented in Tables 5-1 through 5-4 and are ccmpared with unirradiated resultsill 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 the reactor vessel intermediate shell forging 05 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the I9 major rolling direction of the plate (tangential orientation) to 2.02 x 10 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-1) resulted in a 30 ft-lb transition temperature increase of 90'F and in a 50 ft-lb transition temperature increase of 105'F. This resulted in the 30 ft-lb transition temperature of 15'r and a 50 ft-lb transition temperature of 45'F (tangential orientation).
The average Upper Shelf Energy (USE) of the intermediate shell forging 05 Charpy specimens (tangential orientation) resulted in a energy decrease of 34 ft-lb after irradiation to 2.02 x 10 n/cm2 (E > 1.0 MeV) at 550*F.
I9 This results in an average USE of 122 ft-lb (Figure 5-1).
Irradiation of the reactor vessel intermediate shell forging 05 Charpy specimens oriented with the longitudinal axis-of the specimen normal to the I9 major rolling direction of the plate (axial orientation) to 2.02 x 10 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 85'F and in a 50 ft-lb transition temperature increase of 90'F. This resulted in the 30 ft-lb transition temperature of 60'F and a 50 ft-lb transition temperature of Il5'F (axial orientation).
9 i
L 5-4 l-L
i The average USE of the intermediate shell forging 05 Charpy specimens (axial orientation) resulted in an energy decrease of 10 ft-lb after irradiation to 2.02 x 10 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average 19 USE of 84 ft-lb (Figure 5-2).
Irradiation of the reactor vessel core region weld metal Charpy specimens to 2.02 x 10 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-3) resulted in a 19 20'F increase in 30 ft-lb transition temperature and a 50 ft-lb transition temperature increase of 35'F.
This resulted in a 30 ft-lb transition l
temperature of -30*F and the 50 ft-lb transition temperature of 10'F.
The average USE of the reactor vessel core region weld metal resulted in an energy decrease of 4 ft-lb after irradiatica to 2.02 x 10 n/cm2 (E > 1.0 I9 MeV) at 550'F. This resulted in an ave' age USE of 129 ft-lb (Figure 5-3).
Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 2.02 x 10 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-4) resulted in a 19 30 ft-lb transition temperature increase of 85'F and a 50 ft-lb transition temperature increase of 85'F.
This resulted in a 30 ft-lb transition temperature of -5'F and the 50 ft-lb transition temperature of 30'F.
The average USE of the reactor vessel weld HAZ metal experienced an energy decrease of 11 ft-lb after irradiation to 2.02 x 10 n/cm2 (E > 1.0 MeV) 19 at 550'F. This resulted in an average USE of 93 ft-lb (Figure 5-4).
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile or tet:gher appearance with increasing test temperature.
D 5-5
A comparison of the 30 f t-lb transition temperature increases and upper shelf-energy decreases for the various McGuire Unit 2 surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision -
2I3l is presented in Table 5-6.
This comparison iniicates that the transition temperature increase and the upper shelf energy decrease of the 19 intermediate shell forging 05 resulting from irradiation to 2.02 x 10 n/cm2 (E > 1.0 MeV) are less than the Regulatory Guide predictions.
This i
comparison also indicates that the transition temperature increase and the upper shelf energy decrease of the wold metal resulting from irradiation to 2.02 x 10l9 n/cm2 (E > 1.0 MeV) it less than the Regulatory Guide predictions.
The load-time records for the individual instrumented Charpy specimen tests are shown in Appendix A.
5.3 Tension Test Results The results of the tension tests performed on the various materials contained in capsule U irradiated to 2.02 x 10 n/cm2 (E > 1.0 MeV) are presented in 19 Table 5-7 anc tre compared with unirradiated resultslll ar. shown in Figures 5-9 through 5-11.
The results of the tension tests performed on the intermediate shell forging 05 (tangential orientation) indicated that irradiation to 2.02 x 10 n/cm2 (E 19
> 1.0 MeV) at 550*F caused less than a 14 ksi increase in the 0.2 percent offset yield strength and less than a 9 kst increase in the ultimate tensile strength when compared to unirradiated dataill (Figure 5-9).
The results of the tension tests performed on the intermediate shell forging 05 19 (axial orientation) indicated that irradiation to 2.02 x 10 n/cm2 (E >
1.0 MeV) at 550*F caused less than a 19 ksi increase in the 0.2 percent offset yield strength and less than a 15 ksi increase in the ultimate tensile strength when compared to unirradiated dataill (Figure 5-10).
5-6
The results of the tension tests performed on the reactor vessel core region weld metal indicated that irradiation to 2.02 x 10 n/cm2 (E > 1.0 MeV) at I9 550'F caused less than a 12 ksi increase in the 0.2 percent offset yield l
strength and less than a 8 ksi increase in the ultimate tensile strength when compared to unirradiated dataIll (Figure 5-11).
The fractured tension specimens for the intermediate shell forging 05 material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14.
The engineering stress-strain curves for the tension tests a;'e shown in figures 5-15 through 5-19.
5.4 Compact Tension Tests Per the surveillance capsule testing program with the Duke Power Company, the 1/2-T compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center.
f 4
4 1
5-7
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE MCGUIRE UNIT 2
.l INTERME01 ATE SHELL FORGING 05 IRRADIATED AT 550'F, FLUENCE 2.02 x 10 n/cm2 (E > 1.0 MeV) 19 fob f f m$s f
k Samele No._
F$
0 Tangential Orientation 5
3 0.08 5
DL6
-100 10 11 0.28 10 DL9
- 60 17 17 0.43 15 i
DL1
- 25 9
3 0.08 10 DL8
-5 8
6 0.15 10 DL3 0
25 14 0.36 15 DL10 5
47 30 0.76 30 DL12 25 DL13 50 67 45 1.14 50 DL2 75 97 58 1,47 70 DL4 125 80 55 1.40 70 DL14 130 85 59 1.50 85 DL5 150 98 67 1.70 90 DL11 200 127 79 2.01 100 DL15 240 121 79 2.01 100 DL7 275 118 73 1.85 100 Axial Orientation 12 1
11 0.28 10 DT8
-75 6
2 0.05 5
DT7
-50 7
3 0.08 5
DT1
-25 27 18 0.46 15 DT3 20 DTS 35 18 10 0.25 10 DT15 50 33 22 0.50 20 DT11 75 23 14 0.36 15 DT2 75 35 20 0.51 20 DT9 100 38 31 0.79 25 DT14 125 52 42 1.07 45 DT13 165 60 47 1.19 60 DT10 200 74 53 1.35 90 DT12 250 89 57 1.45 100 DT5 285 79 60 1.52 100 DT4 325 84 63 1.60) 100 l
5-8 l
~ ~.
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE MCGUIRE UNIT 2 REACTOR VESSEL WELD METAL AND HAZ METAL 1RRADIATED r
19 AT 550'F, FLUENCE 2.02 x 10 n/cm2 (E > 1.0 MeV)...
Temperature Impact Energy' Lateral Expansion Shear Sagple No.
('F)
('C)
(ft-lb) H),
(mils)
(mm)
(%)
Weld Metal DW1
-100
- 73 8
5 0.13 10 DW5
- 75
- 59 10 12 0.30 29 Dh7
- 50
- 46 28 22 0.56 45 DW15
- 40
- 40 27 20 0.51 45 DW12
- 25
- 32 36 26 0.66 60 DW2 0
- 18 40 31 0.79 65 DW4 5
- 15 48 37 0.94 70 DW13 25 4
60 45 1.14 75 DW8 75 24 89 58 1.47 80 DW11 110 43 105 87 1.70 90 DW9 150 66 120 84 2.13 95 DW10 175 79 116 82 2.08 100 DW3 215 102 118 82 2.08 100 DW14 250 121 141 90 2.29 100' DW6 300 149 141 83 2.11 100 HAZ Metal DH10
-100
- 73 17 23 9
0.23 15 DH9
- 80
- 62 27 37 11 0.28 20 DH13
- 60
- 51 19 26 13 0.33 25, DH5
- 40
- 40 20 27 17 0.43 45 DH3
- 25
- 32 28 38 19 0.48 50 DH11
-5
- 21 28 38 22 0.56 65 DH7 0
- 18 24 33 20-0.51 60 DHS 15 9
53 72 34 0.86 75 DH2 40 4
57 77 42-1.07 80 DH12 70 21 65 88 44 1.12 85 DH4 100 38 74 100 52 1.32 90 DH14 140 60 91 123 60 1.52 100 DH8 175 79 92 125 61 1.55 100 DH15 225 107 98 133 62 1.57 100 DH1 275 l 5 92 125 61 1,55 100 o
G 5-9
-______-_-_________a
il TABLE 5-3 n
3-IhSTRUMENTED CHARDY IMPACT TEST RESULTS FOR THE MCGUIRE UNIT 2 INTERMEDIATE t
SHELL FORGING 05 IRRADIATED AT 550*F, FLUENCE 2.02 X 10 n/cm2 (E > 1.0 MeV) 19 r
i i
Normalised Enerales Test Charpy' Charpy: Maximum Prop Tield Time Maximum Time to Fracture Arrest Yield Flo.
Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stroes S t.re = =
2 Number I'_F),
(ft-lb)
(f t-Ib/in )
(Ibe)
(mece)
(Ibe)
(mece)
(Ibe)
(Ibe)
(be!)
Orei1-Tanaential Orientation i '
DL6
-100 5
40 32 8
3860 0.12 4073 0.13 4C73 62 128 132 i'
DL9
- 60 10 81 56 25 3931 0.14 4099 0.17 4099 66 131 133 DL1
- 25 17 137 120' 17 3004 0.14 4708 0.43 4708 71 130 143 DL8
-5 9
72 35 38 S658 0.13 3845 0.14 3760 121 125 DL3 0-8 04 31 33 3664 0.13 3734 0.14 3734 122 123 DL10 5
25 201 168 33 3690 0.13 4618 0.38 4618 74 123 138 l
.DL12 25 47 378 343 36 3738 0.14 4884 0.70 4864 74
'124 143 DL13 50 67-540 338 203 3627 0.14 4768 0.89 4478 112 120 139 DL2 75 97 781 339 442 3514 0.14 4830 0.68 3971 455 117 139 DL4' 125 80 644 148 496 3148' O.13 4145 0.3M 4056 1085 105 121 i
DL14 130 85 684 390 295 3244 0.13 4420 0.83 3128 971 los 127 i
DL5 150 98 789 318 471 3132 0.14 4549 0.69 2957 1508 104 128 5
DL11 200 127 1023 313 710 3189 0.13 4370 0.69 105 125 DL15 240 121 974 311 663 3168 0.15 4473 0.69 105 127' DL7 275 118 950 322 628 2908 0.13 4239 0.77 97 11e-
?
Aximi Orientation DTS
- 75 12 97 75 22 4098 0.17 4386 0.21-4386 64 136 141-LT7
- 50 6
48 25 23 3032 0.12 3242 0.13 3242 125 101 104 DTI
- 25
'7 56 40 17 3903 0.13 4043 0.15 4043 130 132 DT3 20 27 217 151 66 3739 0.14 4469 0.36 43E2
'124 136 DT6 35 18 145 122
-23 3757 -
O.13 4ll'71 0.30 4271
'55 125 133 DT15 -
SO 33 266 249 17 3676-O.14 4577 0.54 4577 67 123 137 i
DT11 75 23 185 106 SO 8092 O.13 3730 0.31 3730 229 103 113 DT2 75 35 282 161 121 3484 0.13 4437 0.38 4437 366 116 133 DT9
- 100 38 308 241 65 3386 0.13 4433 0.54 4433' 476 112 130-DT14
' 125 52 419-
- 279 140 3450 0.35 4377 0.61 4140 740 115 130 DT13
- 165-80 483 292 191-3056 0.14 3957
. O.54 3633 1631 101 116 DT10 200-74 SOS 300
'295 3201
'O.13 4363 0.69 3380 1753 106 126 DT12 250-89 717
' 229 488 3498 0.22 4266 0.54 116 129 DT5 285-79 636 215 421 3059
'O.13 4171-0.52 102
120 DT4 325-84_
676 209 467 2877
~ O.14 4023 0.52 -
96-115
. Fully ductile fractures no brittle fracture load and no arrest load.
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5-11
4 4
1 TABLE 5-5 II t
EFFECT OF 550*F IRRADIAT10ft TO 2.02 x 10 e (E > 1.0 MeV) f CN ThT fl0TD1 TOUGHNESS PROPERTIES OF THE MCGUIRE UNIT 2 REACTCR VE5SEL SURVEILLA! ICE MATERIALS III III III Average 30 ft-Ib Average 35 mi1 Average 50 ft-lb Average Energy Transitton lateral Expanston Transition Absorption at Temperature (*F)
Temperature (*F)
Temperature (*F7 Full Shea* (ft-Ib)
Material Unirradiated Irradiated AT unirradiated Irradiated AT-unirradiated Irradiated AT unirradiated Irradiated A(ft-151
!g 1
.- Forging 05
- 25 60 85 25 110 85 25 115 93 94 84
- 10
+
( Ar t'al) i i-
{~
' Forging 05-
- 75 15 90
- 70 35 105
- 60 45 105 156
.122
- LA i-
.(Tang-ntial) i Weld Metal
- 50 ;
- 30 20
- 45
-0 45
-- 25 10 35 133 129
- 4 HAZ Metal-
- 90
- 5 85
- 70 30 100
- 55 33 85 :
104 93
- 11 i
1 (1) "AYERAGE" is defined as the value read from the cueve fitted through the data points of the Charpy tests (Figa-es 5-1 through 5-4)..
4 1 12
)
l' s
s r,-
y v-c
-c w,-
,y-m.%,
TABLE 5-6 COMPARISON OF THE MCGUIRE UNIT 2 SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATUR SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PRED 30 ft-lb Transition Tero. Shift _ Upper Shelf EnerQY Decrease Fluence Predicted (a) Measured Predicted (a)
Measured Material Capsule 10 n/m
(.F)
(*F)
(%)
(73) 19 2
Forging 05 V
0.337 80 70 18 10 (Axial)
X 1.45 126 105 26 18 U
2.02 136 85 30 11 Forging 05 V
0.337 80 65 18 14 (Tangential)
X I.45 126 100 26 13 0
2.02 136 90 30 22 Weld Metal V
0.337 34 45 15 0
X 1.45 54 35 21 0
0 2.02 58 20 23 3
HAZ Metal V
0.337 55 6
X I.45 75 C
U 2.02 85 11 1
5-13
~
____ _ l TABLE 5-7 TENSILE PROPERTIES FOR THE MCGUIRE UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS IRRADIATED AT 550*F TO 2.02 X 10 n/cm2 (E > I.0 MeV) 19 Test 0.2% Yield Ultimate Fracture Fr cture Fracture Uniform Total Reduction Sample Temp.
Strength Strength Load Stress Strength Elongation Elongation in Area Material Nue:be r M (k ei)
(kel)
(klo) _
(kei)
(kei)
(%)
(%)
(%)
Forging 05, DL1 70 78.4 96.8 3.00 180.1' 61.1 12.0 25.2 es Heat No. 628840 DL2 180 73.8 92.7 2.90 184.9 59.1 12.0 24.1 es (Tangential)
DI.3 550 87.7-90.7 3.10 139.8 63.2 10.5 21.1 55 Forging 05 DT1 100 80.2 98.8 3.45 165.2 70.3 12.0 22.6 57 Heat No. 526840 DT2 230 75.4 94.1 3.2h 136.8 88.2 11.1 20.6 52 (Axial)
DT3 550 70.3 92.7 3.70 145.4 75.4 9.O-17.3 4s Weld DW1 40 75.4 92.7 2.75 178.6 56.0 13.5 28.4 e9 DW2 200 7c.4 85.6 2.55 167.7 51.9 12.0 25.8 69 DW3 550 72.1 85.6 2.80 154.0 57.4 10.5 20.4 63 5-14
._._._m..
)
(*C)
-150
-100
-50 0
50 100 150 200
~1 l
l i
t.)
t) l l
y 100 7 - +-*
8 80
,/,
/
W 60 40 20 h8W@
I I
0 100 2.5 n
7
)80 E.0
+a,
e p 1.5 m g 60 0
S 9
" 40 O Wr.'
1.0
!,,[
0.5 20 4'~#!f I
I I
I I
O 0
o ta s m itt e men etc n n.ttu tJe x Edu/
200 100 240 160 o n m
o o
200 7
140 120 o
'- *5 160 O
D 100 o
e e
di I
~
l20 D 80 oo U 60 oE 80 40 o
n*r 20
'*'/
0 l
l I
I I
9 0
-200
-100 0
100 200 300 400 tus mm TEMPERATURE ('D Figure 5-1.
Charpy V-Notch Impact Properties for McGuire Unit 2 Reactor Vessel Intermediate Shell Forging 05 (Tangential Orientation) 5-15
l
('C) t
-150 -100
~50 0
50 100 150 200 100 c e e-8 80 fid 60 a
40
,k 20
g9g 0
100 2.5 h80
_ _j '
2.0 3
8-
-M 1.5 60 b
i
-6 1.0 40 e,#
0.5 20 I
I I
d I
O 0
O LMRWWD
.wawaewnnum emx N 120 160 100 120 e
6 2 80 3
b D
80 60 Wr 40
-t
e 40 20 0
O
-200
-100 0
100 200 300 400 500 a#
TEMPERATURE ('f)
Figure 5-2.
Charpy V-Notch Impact Properties for McGuire Unit 2 Reactor Vessel Intermediate Shell forging 05 (Axial Orientation)
.l
)
5-16 1
('C)
-150
-100
-50 0
50 100 150 200
~
I I
I I
I I
I I
- * + * -
100
@ 80 3
Pe 4 60 40 ts
/)6 20 0
2.5 100 g _.,
e,v h80 9f
2.0 1.5 n-60 e,
c.,p[
1.0 40 0.5 3 20 e
0 --I\\ 'I I
I I
I I
0 o nossaiu e waviacwcrtuva enxd%d 160 200 140 -
0 * -
/'
160 120 120 g
q
80 G
80 60 Frm 40 s'ry 20 I
I I
I I
I 0
O h
-200
-100 0
100 200 300 400 TEMPERATURE (*f) figure 5-3.
Charpy V-Notch Impact Properties for McGuire Unit 2 Reactor Vessel Surveillance Weld Metal 5-17
('C)
-150
-100 0 50
-100 150 200 l
e>ehle+e-l l
i I
l l
190
,s'
@ 80 o
g60 o
i5 40 N
I I
I I
i I
0 2.5 100 E
2.0 80 o n o
[?e-o ";
1.5 e 60 e
U 1.0 g*
40 M
O 320 05 O LOTADMTO e InAtata c5rrt ruoct ut x to!%cn' 160 200 140 8
160 120 h100 O
.o 120 f
g
80 o
g a
@ 60 80 m
re, 40 40 o
g 20 I
e i
l I
l l
0-0
-200
-100 0
100 200 300 400 g
TEMPERATURE ('D Figure 5-4.
Charpy V-Notch Impact Properties for McGuire Unit 2 Reactor Vessel Weld Heat-Affected-hne Metal 5-18
g,f] Cj b
f
.c>
g,.,19,,
I u' '
(.
y w ;,.;
"e l E
[.
+
.a s w,
\\@;,3 1
DLO DL9 DL1 DL8 DL3 l
, r.,
.2 L", 's LD,N-
,e-i DL10 DL12 DL13 DL2 DL4 I
r""ses M
DL14 DL5 DL11 DL15 DL7 f
Figure 5-5.
Charpy Impact Specimen Fracture Surfaces for McGuire Unit 2 Reactor Vessel Intermediate Shell Forgir.g 05 (Tangential Orientation) 5-19
2 jjayQ_
7.;
E~
?,I
i, e
w.
v g.
% x Q,,
.y a
, f.
q i
M DT8 DT7 DT1 DT3 DTS
("9;.
I
~,
- T -u -
1
-~
(,.;y.(;,
~
-~
A 4. ;y -
a g
jw
.e,
.ev..y M
l g7;}
g L
"2" J>$(** '
yW,
?
e.;g,,
a ~y<
f,ynt.&
_%,;.7 q
I DT15 DT11 DT2 DT9 DT14 9
N 7
.-q m.m.-=e a'
}
fr4)
_g a;
T' )
((l;f
/ n.;
. ' Q DT13 DTIO DT12 DT5 DT4 Figure 5-6, Charpy Impact Specimen Fracture Surfaces for McGuire Unit 2 Reactor Vessel Intermediate Shell Forging 05 (Axial Orientation) r 5-20 4
.-.r.,
.m.----,.-
-.,--.-,..m.__..--_.---..m.....--,-.._.--.,,,,.----.--------.-.m,.
--.4
a DW1 DW5 DW7 DW15 DW12 i
DW2 DW4 DW13 DW8 DWil
\\
QW 4 + /.
p 6.
4 j
l l
DW9 DW10 DW3 DW14 DWB figure 5-7.
Charpy impact Specimen Fracture Surfaces for McGuire Unit 2 Reactor Vessel Surveillance Weld Metal 5-21 4
..--.,.,._....w-.-
-,-..,....,.,.m.c
,s---,,.._v..,..y-v 3.-,
IN i
_1 )
0" ~ ~,
k-(fj
/
s n.
4 eL
~
3,r;
$.e Fef6 L
9 tva DH10 DH9 DH13 DH5 DH3 i
l DH11 DB7 DH6 DB2 DH12
%.,yy Wh iW s")?
W DH4 DH14 DH8 DH15 DH1 Figure 5-8.
Charpy Impact Specimen fracture Surfaces for McGuire Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal 5-22
l
}
u ~. '
. ~..~i
~
> t
.:n u, =.=-
se
a *f p
, ;,..s
(*
4 y$$h l
l i
WJhd DL6 DL9 DL1 DL8 DL3 l
'.r w
i
' (p,
.jM,
(
- 'k
?-
a f
tQ'* 'a j
t. $Y,
?-
DL10 DL12 DL13 DL2 DL4 A
..~'. f,f
, e m ;*,,,
<,,j ;; g i
g,q DL14 DL5 DL11 DL15 DL7 1
Figure 5-5.
Charpy Impact Specimen Fracture Surfaces for McGuire Unit 2 l
Reactor Vessel Intermediate Shell Forging 05 (Tangential Orientation) l l
5-19 i
-.,. _ _... _... ~. _.. _ _. _ _.
\\
1 I
i D't8 DT7 DT1 DT3 DTS
).
% -l; rj; h pw
^"i}h 1
}]p
? 'h7j:
'hl;K;U anal g-sm A
Ce-1 p.-e r %,4y.
i y
i m
DT15 DT11 DT2 DT9 DT14 N=Am
. :.. m.,
e U
9G j,'
l s.;g". ' 1 ;,.
f 3
,., w,
- p
t N/
"5 Pu.y: r Y
hM$
t,nN.
DT13 DT10 DT12 DT5 DT4 Figure 5-6.
Charpy Impact Specimen Fracture Surfaces for McGuire Unit 2 Reactor Vessel Intermediate Shell Forging 05 (Axial Orientation) 5-20
~
c-f DW1 DW5 Dh7 DW15 DW12 L
DW2 DW4 DW13 DW8
'DW11 1
DW9 DW10 DW3 DW14 DW6 Figure 5-7.
Charpy Impact' Specimen Fracture Surfaces for McGuire Unit-2 Reactor Vessel Surveillance Weld Metal 5-21 i
t ws; V!
06)Di-
"'J;M.. j i
l efi ed
."1
~
.a
.a e M
~r,,a.
~wem -
N W
DH10 DH9 DH13 DH5 DH3 i
f DH11 DH7 DH6 DH2 DH12 i
l
=- w :' y -
w l
DH4 DH14 DH8 DH15 DH1 l
i Figure 5-8, Charpy impact Specimen Fracture Surfaces for McGuire Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal l
5-22
(*C) 0 50 100 150 200 250 300 120 g
l
[
j l
l l-800 110 700 100 NR N "
2 90 A
600 h0 500
~
h 70 O
0_
400 60 a*
On YEO STRDGTH 50 l
l 1
I l
300 s o acmam A 9 RRAMATO AT 550'r, FUDE 22 x1Edm2 80 RDUCTU IN AEA 70 c
M
~
60 3
8 50 3 40 2
$ 30
\\
TUTAL RDGT2 20 1
R n
Q 10 O
uarn nanTo 0
i l
l I
I O
100 200 300 400 500 TEMPERATURE (*F) xam TANG Figure 5-9.
Tensile Properties for McGuire Unit 2 Reactor Vessel Intermediate Shell Forging 05 (Tangential Orientation) l 5-23
(*C) 0 50 100 150 200 250 300 120 l
l l
l l-800 11 0
~
100 h
600 4
ut.TmAtt TDcu sTRDCTH O" 80 i
I N-m 70 500 E
ea YHU STFDCTH 60 O
400
-0 50 O
l l
l l
l 300 40 A O UNIRRAMAfD A 2 EFE 2 F ssc'r, Rioct tm x!Pn/m2 80 70 acum a m 60 O
e 50 3 40 i:
$30 W.Taig anc e 20 s
e O
10 f-O U
tg, a,7, 0
0 100 200 300 400 500 mm TEMPERATURE (*D Figure 5-10.
Tensile Properties for McGuire Unit 2-Reactor Vessel Intermediate Shell Forging 05 (Axial Orientation) 5-24
_ _ _ _ _ - _ _ _ _ - - - _ - _ _ _ - _ _ _ ~
('C) 0 50 100 150 200 250 300 120 l
l j
l l
l-800
!!0 700 100 ULINTE TDCRI ITRDETH
?n 90 600
^
e 80 m
'y m 70 500
L 60 c2x Tu u sTRocTH 400 50 40 A O LAMAMATO A 9 RRAKATO AT 550'F, RLDE 2J2 x1M'n/cn2 80 70
-0 0
Rotets m AIA 60 8 50 C
3 40
$30 q_
imAL nacAin
___3
_g
~~
20 e R_
a m
10
- a yta 0
0 100 200 300 400 500 i
viu TEMPERATUPI (*F)
Jam i -
i Figure 5-11.
Tensile Properties for McGuire Ur.it 2 Reactor Vessel i
Surveillance Weld Metal 5-25 l
2 G n Qvnbenp i
41;65l,7y4
,s-i;'.
1 1 3 3
Nad205K $1A
~i.
b -}Qfiflpb iA[ 3l;y-? f Gl 6'
Lj2
)
4-4,
..s
.m Specimen DL1 70*F k$
e i
5 : =.
,,j
5. '
9,-:
l s
Specimen DL2 180*F 4 j ayag gg q,v;gy
- o' ~ y] [,.; i, '.
~No.C305Bic O f
( R,l4
. _. Y. p:
y. N ME:L8Q ff S CS' s1.E 3 9;4f:
11.a4 c
y i
I i
l c
4' ;m.'
,J t.,.
[ ;.pj;f)),p,
(
167%
. /. '.
.-l.k-t
+
_e Specimen DL3 550*F i
i l
Figure 5-12.
Fractured Tensile Specimens from McGuire Unit 2 Reactor Vessel I
Intermediate Shell Forging 05 (Tangential Orientation) t 5-26 i
l l
i i
v 7
-- !-l d-2 I 3. 74 d emn,.;- m e acq1 g g $ 3 '5 y
16 -
=
-
w Q. 3 0 5 K, mS,3 5:.
. - w-f 4 -
s >.i n,;2(wu nEN..ffB M s
.3
b

:. 1%gh)mn.
cI g
4 k
/..
^ &.x
'; W.;4e n:s,.b n
y
,,s Specimen DL1 70*F mp a,
g 3
V E
~-
^
q --
f?-[,
d
^
$,0
,4
~.
Specimen DL2 180*F
- y 3;.7 <f 2.
y y; q, Qil
'8 Qy " ~1 :L 3M } 'f'- py):]{
NofC305
~
. y o-
r.
-81.9-
.... -. 6k 5k. 4 V
a ; n ';R v:t:.
9, eun ng y,g q,A q:.:
3pj.;,ggy r-y Specimen DL3 550*F Figure 5-12.
Fractured Tensile Specimens from McGuire Unit 2 Reactor Vessel Intermediate Shell Forging 05 (Tangential Orientation) 5-26 h-.-.._.-.
I i
,., y n p : t., c.a $4 f 1_,.gr.$.g, p..
--' i --
Pr
]fL
,;,.g f
t,
.t g
l
_ g ppr
.Q
. g aggy k ['. ?
,; L -3nf 0$y~ysq
?-
7.$ gjd$ r f idi^ !K f
W l
vs
>a
.c>
..g 2-Specimen DT1 100*F
o y
.. a,.
5 % '
c e 4 e,7 c
q g<
- y :
Y.
L
- ;;es;c acM '
i
, _e
+.
. i
...2 :.. ' 3. & f... K M,., S 3
.+
"2 41 l
? t:;7. f W,9,;2;4!WI M,lk
= _
I l
x.-
.:. ? ~p p -
1 i
A-J, I I,'
s; t
.s nu
.m-
_2.,
l-Specimen DT2 230*F
.._._y...._y..
s _
..g s = c,w c o
. a.-_-,
']2
k.'
e f
.f
'+'l
'NVi.QQkelUg(?;l]Ne d$ 0 ih 4 q.g y
~
L.
1.
m
..~: :-.
+
Specimen DT3 550*F Figure 5-13.
Fractured Tensile Specimens from McGuire Unit 2 Reactor Vessel Intermediate Shell Forging 05 (Axial Orientation) i 5-27 r
1-l f
$(*r M,] 'O iy Q, t 2]M9y {TS$
7 [*.7
,e
v. M.O *J 0 5 R..s..f1 (8/J9.c;[lijsp{3f4PM6fk;$A, e ahJ
'.g;yewtmiy9 n?p~yggg.
., pqypyy w
2
RAM
- Wwu i
I Specimen DW1 40*F l
i
.r w
m#
c..7 g:ur
,j.-
,p.
n
's:< 1 1
- R;: 3 058 D
,_:.e 3:
El 7 Se i
sM
4., 4 #.. -a rjgQR l
J
. M w ts w
~
1 I
t Specimen DW2 200'F i
... M =.-
.8 r ;e.
3-
.6 41 !
- !6:0 558
{-
1, '
9.7 3, 4 -'.,~ : 5, ;?5. _ s u.
t s
" ((..
,;-s
. S
I i 7,'r !
A!
I
$4,,
I I
I i
4 m.
,,_4 g
Specimen DW3 550*F i\\
i Figure 5-14.
Fractured Tensile Specimens from McGuire Unit 2 Reactor Vessel Surveillance Weld Metal 5-28 I
t.
_ - - - _ - - -... ~.... -.. - -
g
(; p$ e
.l 1 Z 31 4# T,6
// = ';Q e
.p
^
No; C 305 R.
f.
3' 4) k 6f.-?.-![8[;
e9 1
2,
'm 3,
H ;. :4 @. n;4p ey nc sr w:
a ns Specimen DW1 40*F s,
4: 5 e-
?. k,.
Y
. : I ~;
y J v.
, _.. a; Id !d euwuwvm.
?-
Specimen DW2 200'F
,.;. S -
,. 3 -
3 Ql'
_ c
i:e
. ",,;; T ;g j
d:,. "#.} i
.(
Specimen DW3 550*F
. l Figure 5-14.
Fractured Tensile Specimens from McGuire Unit 2 Reactor Vessel Surveillance Weld Metal 5-28
, (:s y g V 's..
,y 2
o
' i'-
t-u;.c 3 %
.Rl M
.;y [ b ;!flggia
$1,,6jd ty}jdkE!,;i
- =.
3, :_.
y,
- e =
100*F Specimen DT1 2
4 o, e 7's W
If
,.M,,, a
..t n
'i "4
(
g-
,I
.e
,,. c
-+ 4!:i 0;%Qc.
,7
'A(
230*F Specimen DT2 r
v y
f q.
.6'.
6,,1 I.
,,, i } ;
q..
g b -
! ' b i4,
et JTi.
550*F Specimen DT3 Fractured Tensile Specimens from McGuire Unit 2 Reactor Vessel Figure 5-13.
Intermediate Shell Forging 05 (Axial Orientation) 5-27
%.m-_
-.1
,q-
.. y _
100,00 q
]
M.oo-80.00- W N.1. -.
.p-
[+.
70.00-cn 60.00-
-50.00-
=-
g g
40.00-g 30.00-20.00-CL1' 10.00-70 F a:
o.00-i o.oo
= 0.10
-- o.20 0.30-t
. STRAIN, IN/IN -
100.00
~
80.00-70.00-
. cn 00.00-
-x
.ui O
50.00-g ac g -
40.'oo-
-30.00-20.00-
. DL2 -
10.00-180 F:
o.00 o.oo 0.10--
o.20-
0.3o1
~
L STRAIN,- IN/IN 9
. Figure 5-15.
Engineering' Stress-Strain Curves =forfForging 05; Tensile)-
Specimens.DL1;and DL2 (Tangential" Orientation) 5-29 o4 w
e n
a y
-q
s e
100.00 90.00-80.00-70.00-N
-cn v
i 60.00-50.00-Wg 40.00-a 30.00-20.o0 DL3 10.00-550 F o.oo o.Oo 0.1 o -
O.20 STRAIN,-IN/IN
~
m.
Figure 5-16.
Engineering Stress-Strain Curve-for Forging 05 TensileiSpecimens DL3 (Tangential. Orientation).
5-30 Y
_.___.__.__..___i'_._____1_
_.2.
__1_.1__._.._i_:__1.
100.00 N
90.00-N g
- s\\
80.00- ~'
\\
70.00-CD
~
60.00-(6 50.00-Cy 40.00-30.00-20.00-DT1 10.00-100F O.00 0.00 0.10 0.20 STRAIN, IN/lN 100.00-90.00-80.00-e--
70.00-CO 60.00-00.00-g Cy 40.00-30.00-20.00-DT2 10.00-230 F O.00 O.00 0.1 O O.20 STRAIN, IN/lN Figure 5-17.
Engineering Stress-Strain Curves for Forging 05 Tensile o-Specimens DTI and DT2 (Axial Orientation) 5-31
t
.:s.
100.00 90.00-80.00-70.00-u)x
80. m -
Vi e
sO.m-C'g
>. m-aO.m-20.00-DT3:
10.00-550F 0.000.00 0.04
~.
O,08-
' O.12.
- 0.16 '
STRAIN; IN/IN Figure 5-18.
Engineering Stress-Strain Curve for Intermediate Shell-. Forging 05 ' Tensile Specimen DT3 (Axial Orientation).
5-32 rii g -
3
f4 00.00
,90,00-
/y = 7-s5 m
/
%i 60.00- &
'%-s,-
- =
70.00-
'$4 m --
v.
-- x 60.00-
-Y, en O
'50.00-g-
ap 40.00-
-30.00-DWri
-20.00-
40 F-10.00-0.00 o.co -
0.10 0.20-o.301 STRAIN, IN/IN 90.00 N
80.00-N s
70.00-N s%(
55 s0.00-N M-g of 50.00-cn 1
-w C
40.00-l-- :
en 30.00-20.00-DW2!
10.00-200F o.oc o.00
. o.10
. o.20 t o.so STRAIN, IN/IN.
a Figure 5-19.
. Engineering Stress-St' rain Curves for. Weld Metal _ Tensile Specimens DW1 and DW2 5-33 4
o f
p f%7-g 9
e>wera af m
y er
-e E
a
90.00 g --~~ %,,
,g/
kx 80.00-gM
\\
70.co-s cn 60.00-M ui 50.00-cn tu cr 40.00-w W
30.00-e 20.00-DW3 10.00-550 F 0.00 O.OO O.10 0.'20 i
STRAIN, IN/IN Figure 5-20.
Engineering Stress-Strain Curve for Weld Metal Tensil.e Specimen -
DW3 5-34
SECTION 6.0 RADIATION ANALYSIS ANO llEUTRON DOSIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons.
First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (enorgy spectrum, flux, fluence) to which the test specimens were exposed must be known.
Second, in order to relate the changes observed in the test specimens.to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.
The latter information is derived solely from analysis.
~
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition, in recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold flueria toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to ptovide a data base for future 6-1
reference. The energy dependent dpa function to be used for this evaluation' is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposurcs in Ferritic Steels in Terms of Displacements per Atom." The application of-the dpa parameter to the assessment of embrittlement gradients through the Lthickness of the pressure vessel wall has already been promulgated in-Revision 2 to the Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials."
This section provides the results.of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance capsule U, withdrawn at the end of the seventh fuel cycle. Also included are updates of dosimetry contained in capsule X, withdrawn at the end of Cycle 5, and capsule V, withdrawn at the conclusion of Cycle 1.
These updates are based on current methodology and nuclear data; and together with the capsule U results provide a consistent up to date data base for-use in evaluating materials properties of the W. B. McGuire Unit 2 reactor vessel.
In all cases, fast aeutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itsel f.
Also unc%inties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided.
6.2 Discrete Ordinates Analysis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1.
Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the reactor vessel surveillance program. The capsules-are located at azimuthal angles of 56.0, 58.5*, 124.0*, 236.0',
238.5*, and 304.0* relative to the core cardinal axes 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 containers are 1.182 by
'l-inch and approximately 56 inches in height.
The containers are positioned 6-2
u axially such -that the specimens are ces.tered' onl the core midplane, 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 t fect 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 themselves must be included in the analytical model.
In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout-the reactor geometry as well as to establish relative radial distributions of exposure parameters {p(E > 1.0 Mev), d(E > 0.1 Mev), and dpa) through the vessel wall.
The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsulc as well as for the determination of exposure parameter ratios; i.e.,
dpa/p(E > 1.0 MeV), within the pressure vessel geometry.
The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e.,
the 1/4T,1/2T, and 3/4T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule ~ positions, and soveral azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core.
The importance functions generated from thase adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement.
These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of ' interest for each cycle of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles.
It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within 6-3
the _ reactor' core;1but, f also accounted forLthe effects ofL: varying neutron: yield' per fission:and-fission _--spectrum introduced by the build-up of ' plutonium. as;the burnup of individual fuel assemblies increased.
The absolute cycle specific data = from the adjoint evaluations _together with relative neutron energy spectra and radial distribution _information from the
- forward calculation provided the means to:'
1.
Evaluate neutron dosimetry obtained from surveillance _ capsule-locations.
2.
Extrapolate dosimetry results to key locations at-the -inner radius; and through the thickness of the pressure vessel wall.
3.
Enable a direct comparison of analytical prediction with measurement.
4.
Establish a_ mechanism for projection of-pressure vessel exposure as the design of each.new fucl cycle evolves.
The forward transport calculation for the reactor model summarized in Figures.
4-1 and 6-1 was carried out in R, 0 geometry using the D0T two-dimensional discrete ordinates codell43 and the SAILOR cross-section library (15). The SAILOR library is a 47 group ENOFB-IV based data set produced specifically-for light water reactor applications.
In these analyses anisotropi_c scattering was treated with a P expansion of _ the cross-sections.and the angular.
3 discretization was.modeled with an S _ order of-angular quadrature.
8 The reference-core power distribution utilized in the forward analysis was derived from statistica11 studies of long-term operation of Westinghouse 4-loop.
plants.. Inherent in the _ development.of this reference core power distribution =
is the use of an out-in fuel management strategy; i.e,- fresh fuel on the core -
periphery.
Furthermore, for the peripheral fuel assemblies, a 20-uncertainty derived from the statistical evaluation!of plant to plant and cycle) to cycle variations:in peripheral power was used. fSince_ it-is unlikely'that. an i
l~
single reactor would have' a power distribution:at the nominal +2a 6 1 1
level for' a large number of. fuel cycles,1the use of.this_ reference distribution 1 is expected to yield somewhat' conservative results.
' All adjoint. analyses were also carried out using an S _ order:of angular
1.. -
8
. quadrature and the P. cross-section approximation from the SAILOR library.
3 Adjoint source _ locations were chosen at several azimuthal location _s alongj the-pressure vessel inner radius as well as 'the geometric center of each-surveillance capsule.- Again, these calculations were_ run-in-R, 9. geometry 1 to provide neutron source distribution-importance functions for the exposureL parameter of interest; in this case, p V > 1.0 MeV).- Having the importance functions = and appropriate core nurce distributjons, the resp ~onse of interest could be calculated as:
g33
}
R (r, 0) -gr I.I I(r, 0, E) S (r, '0, E) r dr de dE O E where:
R (r, 0)
- p (E > 1.0 MeV) at radius:r and azimuthal' angle-0
+
I (r, 0, E)
- Adjoint _importance function at-radius, r, azimuthal angle 0, and neutron source energy E.
/
~*
S (r, 0, E)
- Neutron source strength at_ core location r,-0 and'-
energy E.
Although the adjoint importance functions used'in the analysis were based on a e
response function defined by the threshold neutron flux (E > 1.0 MeV), prior-1 calculations have shown that, while the implementation = of low leakage.. loading-;
patterns significantly impect the magnitude and the spatial distribution"of the*
7 neutron field, changes in the relative neutron energy' spectrum are!of second:
order.
Thus, for-a given location the ratio of dpa/d (E > '1.0 MeV)-is.
insensitive to changing core source distributions ~ _ In the application of..these 4 adjoint important' functions to the Trojan reactor,;therefore, the iron displacement rates (dpa) and the neutron flux'(E > 0.1-MeV):were computed on a; R
-cycle specific basis by_using dpa/p-(E > 1.0 MeV) and p (E. > 0.1 MeV)/p (E >:1.0 MeV) ratios from the forward _ analysis.in
},
conjunction with the cycle specific p-(E > 1.0 MeV) solutions-from the L
individual adjoint evaluations.
F 6-5 a.
The reactor core power-' distribution used in the' plant specific adjoint calculations was taken from the fuel cycle design reports for the first seven -
.l operating cycles of W. B. McGuire Unit 2(16 through 22),
j Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5.
The data listed in these tables establish the means for. absolute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutt i exposure of the pressure vessel wall.
In Table 6-1, the calculated exposure parameters [p (E > 1.0 MeV),
((E > 0.1 MeV), and dpa] are given at the geometric. center of the two surveillance capsule positions for both the design basis and the plant specific core power distributions.
The plant specific data, based on the adjof-t transport analysis, are meant to establish the absolute-comparison of measurement with analysis.
The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the cycles 1 through 7 plant specific power distribution.
It is important to note that, the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself.
Radial gradient information for neutron flux (E > 1.0 MeV),
neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tables-6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected -
exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.
6-6
T i For-example,Kthe neutron flux (E > l' 0 MeV) atithe 1/4T.' position on the 45'
~
azimuth;is-given by:.
. p(220.27, 45') F ;(225.75, 45');
4 '/4T(45')
=
1 Projected neutroniflux at theLl/4T= positionion -
where:
4 4T(45')~
=
1/
the~45' azimuth l Projected or calculatedLneutron flux at the2 p '(220.27,45')
vessel inner radius on the 45' azimuth..
~
Relative radial 1 distribution function 1from1 F (225.75, 45*)
=
Table-6-3.-
Similar expressions apply for exposure parameters in. terms:of ( (E-> 0.1 MeV) and dpa/sec.
6.3 Neutron Dosimetty
+
4
~
The. passive neutron sensors included in the W. B. McGuire Unit 2' surveillance-g program are listed in Table'6-6. Also-given in Table 6-6 arelthe pr.imary-nuclear reactions and associated nuclear const' ants that were used in the' evaluation of. the neutron energy spectrum within:the= capsule and1the subsequent-determination of the various ' exposure parameters of. interest-
[p (E > 1.0 Mev), p-(E > 0.l~MeV),-dpa].
The relative locations of the neutron sensors;within the-capsules. are shownTin-
. Figure 4-2.
The iron, nickel, copper, and cobalt-aluminum monitors,cin wire' form, were' placed in ' holes drilled in spacers at-several axial'levelsiwithin" e;
Ther cadmium-shielded neptunium' and uranium: fission monitorsTwere'
' the capsules.
accommodated within the dosimeter block located near the~ center of-the capsule.
1.
kt L
6-7 5
,).
Gt
r
,%,,e k
w w
nr g
g-4 4
$i*N.m
J
-The use of passive monitors such as those-listed in Table 6-6 does not yield a direct measure of the energy dependent flux. level at the point of interest.
Rather, the activation or fission process is a measure of the integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.
In particular, the following variables are of interest:
o The specific activity of each monitor, o
The operating history of the reactor.
o The energy response of the monitor, o
The neutron energy spectrum at the monitor location.
o The physical characteristics of the monitor.
The specific activity of each of the neutron mons' ors was determined u..ng established ASTM procedures [23 through 36].
Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer.
The irradiation history of the W. B. McGuire Unit 2 reactor during cycles 1 through 7 was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable period.
The irradiation history applicable to capsules U, X, and V is given in Table 6-7.
Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Tables 6-8 through 6-10 for capsules U, X, and V, respectively.
Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code [37].
The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded-to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data.
The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra.
6-8
In the FERRET evaluations, a log-normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations.
In general, the measured values f are linearly related to the flux p-by some response matrix A:
(s,a)
(s)
(a)-
f
-I A
(
9 19 9
where i indexes the measured values belonging to a single data set s, g designates the energy group and a delineates spectra that may be simultaneously adjusted. For example, I
a p
R i
g ig g
relates a set of measured reaction rates Rg to a single spectrula p by g
the multigroup cross section ojg.
(In this case, FERRET also adjusts the cross-sections.) The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.
In the FERRET analysis of the dosimetry data, the continuous quantities (i.e.,_
fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code [38).
This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620' group structure using a SPLINE interpontion procedure for interpolation in regions.where group boundaries do not coincide.
The 620-point spectrum was then easily collapsed 4
to the group scheme used in FERRET.
The cross-sections were also collapsed into the 53 energy-group structure using St.ND II with calculated spectra (as expanded to 620 groups)- as weighting functions. The cross sections were taken from the ENDF/B-V msimetry file.
Uncertainty estimates and 53 x 53 covariance matrices were constructed for-each 6-9
cross section. Correla'tions between cross-sections were neglected due to.d'ata-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. More often,-
a simple parameterized form is used:
gg,=Rh+R M
R.P g g gg, where RN specifies.an overall fractional normalization uncertainty (i.e.,
complete correlation) for the corresponding set of values.
The fractional uncertainties R specify additional random uncertainties for group g that are g
correlated with a correlation matrix:-
Pgg, = (1 - 0) 6gg, + 0 exp (~
]
The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 7 (0 specifies _the strength of the latter term).
For the a priori calculated fluxes, a short-range correlation of 1'- 6 groups was used. This choice implies.that neighboring groups are strongly correlated when 0 is close to 1.
Strong-long-range corralations (or anticorrelations) were justified based on information presented by R.E.
Maerker[39]. Maerker's results are. closely duplicated when 7 - 6.
For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.
4 I
6-10
Results of the FERRET evaluations of the capsules U,:X, _ and V dosimetry are.
given in Table-6-11 through 6-13. The data summarized in these tables include fast neutron exposure evaluations in terms of fluence (E > 1.0 MeV), fluence (E
> 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit of the j
adjusted spectrum to the measurements for each of the three capsules are provided in Tables 6-14 through 6-16.
In general, excellent 'results.were achieved in the fits of the adjusted spectra to the individual experimental reaction rates.
The adjusted spectra from the least squares evaluations'are-given in Tables 6-17 through 6-19 in the FERRET 53 energy group structure.
A summary of the measured and calculated neutron exposure of capsules V, X, and V is presented in Table 6-20.
The agreement between calculation and measurement falls within i 20% for all fast neutron exposure parameters listed.
The thermal neutron exposure calculated for the exposure period undepredicted the measured value by about 60 percent.
Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-21.
Along with the current (6.05 EFPY) exposure derived from the set of capsule measurements, projections are also provided for an exposure period of 16 EFPY and to end of vessel design life (32 EFPY).
Representations of the axial distribution of the maximum vessel exposure (45' azimuth) are shown in Figure 6.2 for all three exposure periods.
In computing these vessel exposures, the calculated values from Table 6-2 were scaled by the average measurement / calculation ratios observed from evaluations-of dosimetry from capsules V, X, and V.
The resultant best estimate exposure rates were then used to compute the integrated exposure of the-vessel beltline.
This procedure resulted in the following bias factors being applied to the analytical results:
Flux (E > 1.0 MeV)
Bias - 1.218 Flux (E > 0.1 MeV)
Bias - 1.208 dpa/sec Bias - l.168 6-11
Projections for future operation were based on the assumption that the time averaged exposure rates for the first seven cycles of cperation would continue to be applicable throughout plant life.
In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the W. B. McGuire Unit 2 reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were also employed.
Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-21.
In order to access RTNDT vs. fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/41 and 3/4T positions were defined by the relations d' (1/4T)
- 4 (Suriace) (dpa (5 ae d' (3/4T) 4 (Surface) (dpa ( u f e}} Using this approach results in the dpa equivalent fluence values listed in Table 6-22. In Table 6-23 updated lead factor; a.e listed for each of the W. B. McGuire Unit 2 surveillance capsules. These date may be used as a guide in establishing future withdr wal..hedules for the remaining capsules. W 6-12 I
gCHARPY SPECIMEN r'> r ' ' ' ' < f / l ////// ////J7 b %N N % N N T T % N N % TN N N N N NEUTRON PAD g N NA A N N NN NN N \\\\ \\ \\\\ \\ \\ s 4 Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 f
1.0E+20 / 1.0E+19 f s s f R i, / ) s E i j ( t f 1.0E+18 f g 8 '/ \\' E / \\ 2 l \\- u. 8 / \\ 2: 1.0 E+ 17 Sz 1 1.0 E+ 16 .t 6.05 EFPY --32 EFPY 1.0 E+ 15 -8 6 -4 2 0 2 4 6 8 Distance From Core Midplane.(ft) Figure 6-2 Axial Distribution of Neutron Fluence (E > 1.0 MeV) Along the 45 Degree Azimuth 6-14
TABLE 6 t CALCULATED FAST NEUTRON EXPOSURE PARAMETERS I' AT THE SURVEILLANCE CAPSULE CENTER i ((E > 1.0MeV) ((E > 0.1Mev) Iron Displacement Rate 2 2 In/cm -sec1 Ideafsec1 In/cm -sec1 31.5* 34.0* 31.5* 34.0* 31.5* 34.0* ll Il Il II 2.21 X 10-10 2.62 X 10-10 DESIGN BASIS 1.11 X 10 1.29 X 10 4.88 X 10 5.93 X 10 1 a Il Il 1.66 X 10-10 1.93 X 10-10 10 3.66 X 10 4.37 X 10 10-9.50 X 10 CYCLE 1 8.33 X 10 r ll 1.93 X 10-10 2.25 X 10-10 II 5.10 X 10 10 ll 4.26 X 10 CYCLE 2 9.69 X 10 1.11 X 10 10 10 3.69 X'10ll II 1.'67 X 10-10 1.97 X 10-10 CYCLE 3' 8.40 X 10 9.73 X 10 4.47 X 10 10 10 11 Il 1.47 X 10-10 1.69 X'10-10 CYCLE' 4 7.38 X 10 8.34 X 10 3.24 X 10 3.83 X 10 10 10 Il Il 1.41 X 10-10 1.60 X 10-10 CYCLE 5 7.09 X 10 7.89 X 10 3.12 X 10 3.63 X 10 10 II Il 1.37 X 13-10 1.54 X'10-10 . CYCLE-6 6.87 X 1010 7.59 X 10 3.02 X 10 3.49 X 10 10 10 ll II 1.40 X 10-10 1.59 X 10-10 CYCLE 7 7.06 X 10 7.86 X.10 3.10 X'10 3.61 X 10 i 6-15 4
i TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 2 (IE > 1.0MeV) fn/cm _igrJ L.O.* lla' 10 d'. 10.0 '. 10 1.69 X 1010 2.44 X 1010 DESIGN BASIS 1.45 X 1010 2.21 X 10 10 CYCLE 1 1.08 X 1010 1.62 X 1010 1.27 X 1010 1.80 X 10 10 1.92 X 1010 1.48 X 1010 2.11 X 1010 CYCLE 2 1.27 X 10 10 1.28 X 1010 1.86 X 1010 CYCLE 3 1.03 X 1010 1.58 X 10 10 1.57 X 1010 CYCLE 4 1.06 X 1010 1.57 X 1010 1.14 X 10 10 3,47 y 3o10 CYCLE 5 1.02 X 1010 1.54 X 1010 1.10 X 10 CYCLE 6 1.11 X 1010 1.62 X 1010 1.08 X 1010 1.42 X 1010 CYCLE 7 1.04 X 1010 1.56 X 1010 -1.10 X 1010 1,47 X 1010-2 d(E > 0.lMeV) In/cm.,,yp_gy ._0_a* ILO' 10.d.' iLO'. DESIGN BASIS 3.02 X 1010 4.66 X 1010 4.?5 X 1010 6.11 X 1010 10 CYCLE 1 2.25 X 1010 3.42 X 1010 3.19 X 1010 4.51 X 10 10 3.72 X 1010 5.28 X 1010 CYCLE 2 2.65 X 1010 4.05 X 10 CYCLE 3 2.15 X 1010 3.33 X 1010 3.22 X 1010 4.66 X 1010 10 CYCLE 4 2.21 X 1010 3.31 X 1010 2.87 X 1010 .3.93 X 10 10 10 10 3.68 X 10 CYCLE 5 2.12 X 1010 3.25 X 10 2.77 X 10 CYCLE 6 2.31 X 10 3.42 X 1010 2.72 X 1010 3.56 X 1010 10 CYCLE 7 2.17 X 1010 3.29 X 1010 2.77 X 1010 3.68 X 1010 6-16 w -- -=y.,
TABLE 6-2 (continued) CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE HETAL INTERFACE f ron Atom Disolacement JRate Idoa/sec]_ _Ad* lid' 3 0. 0 *. ALQ', DESIGN BASIS 2.25 X 10-11 J.41 X 10-I1 2,,73 X 10-11 3.88 X 10-11 CYCLE 1 1,68 X 10-11 2.50 X 10-11 2.05 X 10-11 2,86 X 10-11 CYCLE 2 1.97 X 10-I1 2.96 X 10-Il 2.39 X 10-11 3.35 X 10-11 CYCLE 3 1.60 X 10-11 2.44 X lo-ll 2.07 X 10-Il 2.96 X 10-1I CYCLE 4 1.65 X 10-11 2.42 X 10-11 1.84 X 10-11 2.50 X 10-11 CYCLE 5 1.58 X 10-II 2.38 X 10-11 1.78 X 10-1I 2.34 X 10-11 CYCLE 6 1.72 X 10-II 2.50 X 10-11 1.75 X 10-11 2.26 X-10-11 CYCLE 7 1,62 X 10-11 2.41 X 10-11 1.78 X 10-11 2.34 X 10-11 9 e e 4 I 6-17 i 1-
TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL ' Radius h O' 15' 35' 45' 220.27(I) 1.00 1.00 1.00 1.00 220.64 0.979 0.979 0.980 0.979' 221.66 0.891 0.891 0.893 0.889 222.99 0.771 0.769 0.773 0.766 224.31 0.655 0.652 0.658 0.648 225.63 0.552 0.549 0.555 0.543 226.95 0.463 0.459 0.467 0.452 228.28 0.387 0.383 0.390 0.376 229.60 0.322 0.318 0.326 0.311 230.92 0.268 0.263 0.271 0.257 232.25 0.222 0.218 0.225 0.211 233.57 0.183 0.180 0.187 0.174 234.89 0.151 0.148 0.155 0.142 236.22 0.125 0.121 0.128 0.116 237.54 0.102 0.0992 0.105 0.0945 238.86 0.0831 0.0807 0.0862 0.0762 240.19 0.0673 0.0650 0.0703 0.0608 241.51 0.0539 0.0512 0.0567 0.0472 242.17(2) 0.0508 0.0477 0.0536-0.0438 NOTES:
1) Base Metal Inner Radius

2) Base Metal Outer Radius t
6-18 t
TABLE 6-4 i RELATIVE RADIAL DISTRIBUTIONS Of NEUTRON FLUX'(E > 0.1 MeV) WITHIN THE PRESSURE VESSEL WALL Radius
Iggi, 0*
15' 35' 45' 220.27(3) 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 221.66 1.00 1.00 1.00 0.995 222.99 0.974 0.966 0.982 0.956 224.31 0.928 0.915 0.938 0.902 225.63 0.875 0.859 0.886 0.843 226.95 0.819 0.802 0.832 0.782 228.28 0.762 0.743 0.777 0.722 229.60 0.705 0.686 0.721 0.663 230.92 0.649 0.529 0.665 0.605 232.25 0.594 0.575 0.611 0.549 233.57 0.540 0.522 0.558 0.495 234.89 0.488 0.470 0.506 0.443 236.22 0.436 0.421 0.455 0.392 237.54 0.386 0.373 0.406 0.343 238.86 0.337 0.326 0.358 0.296 240.19 0.290 0.280 0.310 0.248 241.51 0.244 0.232 0.261 0.201 242.17(2) 0.233 0.219 0.249 0.188 l l NOTES:
1) Base Metal Inner Radius

2) Base Metal Outer Radius 1
k L-6-19 l L 9
i i TABLE 6-5 l RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL WALL -) Radius ltal_ O' 15' 35' _ 45' 220.27(I) 1.00 1.00 1.00 1.00 220.64 0.982 0.982 0.986 0.984 221.66 0.911 0.910 0.923 0.915 222.99 0.813 0.812. 0.837 0.821 224.31 0.721 0.718 0.751 0.730 225.63 0.637 0.633 0.673 0.646 226.95 0.562 0.558 0.602 0.572 228.28 0.496 0.491 0.539 0.505 229.60 0.438 0.433 0.481 0.447 230.92 0.387 0.381 0.430 0.394 232.25 0.341 0.335 0.383 0.347 233.57 0.300 0.295 0.341 0.305 234.89 0.263 0.258 0.302 0.266 236.22 0.230 0.225 0.267 0.231 237.54 0.199 0.195 0.234 0.199 238.86 0.171 0.168 0.203 0.169 240.19 0.145 0.142 0.174 0.140 241.51 0.121 0.117 0.146 0.113 242.17(2) 0.116 0.110 0.140 0.106 NOTES:
1) Base Metal Inner Radius

2) Base Metal Outer. Radius 6-20 h
i TABLE 6-6 NUCLEA.R PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Weight
Response
Product Yield Material Interest Fraction Rance Hal f-Life (%) Copper Cu63(n,a)Co60 0.6917 E > 4.7 MeV 5.272 yrs . Iron Fe54(n,p)Mn54 0.0582 E > 1.0 MeV 312.2 days l Nickel N158(n p)CoS8 0.6830 E > 1.0 MeV 70.90 days Uranium-238* -U238(n,f)Csl37 1.0 E > 0.4 MeV 30.12 yrs 5.99 ~ Neptunium-237* Np237(n.f)Csl37 1.0 E > 0.08 MeV 30.12 yrs 6.50 Cobalt-Aluminum
CoS9(n,0)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs l
Cobalt-Aluminum CoS9(n,0)Co60 0.0015 E > 0.015 MeV 5.272 yrs != l
Denotes that monitor is cadmium' shielded.
l 6-21
~ TA8LE 6-7 MONTHLY 1HERMAL GENERATION DURING THE FIRST SEVEN FUEL CYCLES OF THE W. B. McGUIRE UNIT 2 REACTOR THERMAL THERHAL THERMAL THERHAL GENERATION GENERATION GENERATION GENERATION EQH[1] WW-hr) BOK[!i (HW-hr) BQHill (MW-hr) K0fGJ (MW-hr) 5/83 29804 8/85 1698280 11/87 2091145-2/90 2283648 6/83 281343 9/85 2451943 12/87 24*9205 3/90 2469938 7/83 503 10/85 2342207 1/88 2375936 4/90 2446697 8/83 713998 11/85 2366518 2/88 2372171 5/90 2474847 9/83 1283803 12/85 1328243 3/88 2528325 6/90 2359097 10/83 1407029 1/86 2284248 4/88 2446326 7/90 2492907 11/83 1677739 2/86 2262594 5/88 1965065 8/90 2298172 12/83 1637288 3/86 1070655 6/88 0 9/90 5074 1/84 430628 4/86 0 7/88 182814 10/90 0 2/84 1908578 5/86 0 8/88 2312603 11/90 0 3/84 2306794 6/86 98860 9/88 2447331 12/90 55107 4/84 2301882 7/86 2419839 10/88 2518180 1/91 2425825 5/84 2131577 8/86 2295334 11/88 2395542 2/91 2231028 6/84 2382476 9/86 2455840 12/88 2528858 3/91 2538082 7/84 1558900 10/86 2268723 1/89 2475080 4/91 2444210 8/84 614384 11/86 834670 2/89 2280475 5/91 2523168 9/84 2272550 12/86 2539926 3/89 2248654 6/91 2372591 10/84 2260438 1/87 1942205 4/89 1929466 7/91 2191309 11/84 1822772 2/87 2094930 5/89 2259109 8/91 2528077 12/84 1719551 3/87 2537082 6/89 2365375 9/91 2315536 1/85 1948246 4/87 2391293 7/89 290705 10/91 2156131 2/85 0 5/87 20798 8/89 0 11/91 2185092 3/85 0 6/87 0 9/89 745948 12/91 2500590 4/85 0 7/87 1968987 10/89 2483669 1/92 628307 5/85 1613921 8/87 2304777 11/89 2400111 6/85 2080510 9/87 2209664 12/89 2469574 7/85 949838 10/87 2534464 1/90 2458034 e 6-22
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES i SURVEILLANCE CAPSULE U Measured Saturated Reaction I i Monitor and Activity Activity Rate Axial Location (dis /sec-am) (dis /sec-am) (RPS/NUCLEQ1). Cu-63 (n a) Co-60 5 3.35 x 105 Top 1.54 x 10 Middle 1,63 x 10 3.55 x 105 5 Bottom 1.56 x 10 3.39 x 105 5 Average 1.58 x 10 3.43 x 105 5.23 x 10-17' 5 Fe-54(n,p) Mn-54 6 6 Top 1.75 x 10 3.19 x 10 6 6 ~ Middle 1.86 x 10 3.39 x 10 6 6 Bottom 1.70 x 10 3.10 x 10 6 6 5.13 x 10-15 Average 1.77 x 10 3.22 x 10 Ni-58 (n.p) Co-58 7 4.90'x 107 Top 1.16 x 10 7 7 Middle 1.19 x 10 5.03 x 10 7 4.98 x 107 Bottom 1.18 x 10 7 7 7.09 x 10-15 Average 1.18 x 10 4.97 x 10 U-238 (n,f) Cs-137 (Cd) 5 5.58 x 106 3.69 x 10-14' Middle 7.03 x 10 6-23
TABLE 6-8 HEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd SURVEILLANCE CAPSVLE U Measured Saturatod Reaction Monitor and Activity Activity Rate Axial location (dis /sec-am) (dis /sec-am) (RPS/NVCLEUS) Np-237(n f) Cs-137 (Cd) Hiddle 6.48 x 106 5.15 x 107 3.12 x 10-13 Co-59 (n,0) Co-60 i Top 3.27 x 107 7.10 x 107 Middle 3.27 x 107 7.10 x 10 7 Bottom 3.10 x 10 7.19 x 107 7 7 7 4.65 x 10-12 Average 3.28 x 10 7.13 x-10 Co-59 (n,a) Co-60 (Cd) Top 2.11 x 107 a.59 x 107 7 Middle 1.95 x 107 4.24 x 10 7 Bottom 2.05 x 107 4.46 x 10 Avorage 2.04 x 107 4.43 x 107 2.89 x 10-12 O e 6-24
TABLE 6-9 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE X HeasuriA Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /sec-am) (dis /sec-atnl (RPS/ NUCLEUS) Cu-63 (n.a) Co-60 i Top 1.23 x 105 3.42 x 105 ] 5 5 3.67 x 10 Middle 1.32 x 10 Bottom 1.28 x 105 3.56 x 105 5 5.42 x 10'17~ 5 3.55 x 10 Average 1.28 x 10 Fe-54(n.p) Mn-54 6 6 3.31 x 10 Top 1.58 x 10 Middle 1.71 x 10 3.58 x 106 6 6 6 Bottom 1.67 x 10 3.49 x 10 6 5.51 x 10-15 6 3.46 x 10 Average 1.65 x 10 Ni-58 (n.p) Co-58 Top 6.20 x 10 5.22 x 107 6 7 6 5.52 x 10 Middle 6.55 x 10 6 5.43 x 107 Bottom 6.44 x 10 Average 6.40 x 10 5.39 x 107 7.69 x 10-15 6 U-238 (n,f) Cs-137 (Cd) 5 6 3.82 x 10-14 Middle 5.14 x 10 5.78 x 10 6-25
TABLE 6-9 MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd SVRVEILLANCE CAPSULE X Measured Saturated Reaction Monitor and Activity Activ4ty Rate Axial location (dis / soc-om) (dis /see-am) (RPS/ NUCLEUS) Np-237(n,f) Cs-137 (Cd) 7 3.13 x 10-13' 6 5.17 x 10 Middle 4.60 x 10 Co-59 (n,B) Co-60 Top 3.10 x 107 8.61 x 107 Middle 2.92 x 107 8.12 x 107 Bottom 2.99 x 107 0.30 x 107 Average 3.00 x 107 8.35 x 107 5.44 x 10-12 Co-59 (n,6) Co-60 (Cd) 7 7 4.95 x 10 Top 1.78 x 10 Middle 1.67 x 107 4.65 x 107 7 Bottom 1.69 x 107 4.70 x 10 7 3.11 x 10-12 Average 1.71 x 107 4.77 x 10 e 6-26 ll
TABLE 6-10 MEASURED SENSOR AC11VITIES AND REACTION RATES SURVEILLANCE CAPSULE V Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /sec-om) (dis / soc-om) (RPS/ NUCLEUS 1 Cu-63 (n.o) 00-60 4 5 Top 3.84 x 10 3.38 x 10 Middle 4.03 x 104 3.55 x 105 Bottom 3.76 x 10 3.31 x 105 4 Average 3.88 x 104 3.42 x 105 5.21 x 10~II Fe-54(n.p) Mn-54 Top 9.69 x 105 3.23 x 106 6 6 Middle 1.01 x 10 3.37 x 10 5 0 Bottom 9.87 x 10 3.29 x 10 5 3,30 x 10 5.25 x 10-15 6 Average 9.89 x 10 Ni-58 (n.p) C0-58 6 5.03 x 107 Top 3.96 x 10 6 7 Middle 3.58 x 10 4.55 x 10 6 7 Bottom 4.04 x 10 5.13 x 10 6 4.90 x 107 7.00 x 10-15 Average 3.86 x 10 U-238 (n,f) Cs-137 (Cd) 5 4.61 x 106 3.45 x 10*l4 Middle 1.20 x 10 6-27
TABLE 6-10 MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd SURVEILLANCE CAPSULE V Measured Saturated Reaction Monitor and Activity Activity Rate Axial location (dis /sec-am) Idis/sec-am) JRPS/ NUCLEUS) Np-237(n,f) Cs-137 (Cd) Middle 1.20 x 10 5.24 x 107 3.17 x 10-13 6 Co-59 (n,B) Co-60 Top 1.01 x 10 8.90 x 107 7 Middle 8.33 x 106 7.34 x 107 7 Bottom 8.86 x 106 7.81 x 10 Average 8.89 x 10 7.83 x 107 5.11 x 10-12 6 Co-59 (n,a) Co-60 (Cd) Top 5.04 x 10 4.44 x 107 6 Middle 4.72 x 106 4.16 x 107 Bottom 4.74 x 10 4.lb x 107 6 Average 4.83 x 106 4.26 x 107 2.78 x 10-12 i 6-28
TABLE 6-11
SUMMARY
Of NEUTRON DOSIMElRY RESULTS SURVEILLANCE CAPSULE U IEC AVERAGED EXPOSURE RATES p (E > 1.0 MeV) (n/cm -sec) 1.06 x 10ll 2 i 8% -[ p (E > 0.1 MeV) (n/cm -sec) 4.84 x 10Il i 15% 2 dpa/sec 2,07 x 10-10 i 11% 4 (E < 0.414 eV) {n/cm -sec) 7.29 x 1010 1 25% 2 INTEGRATED CAPSVLE EXPOSURE t (E > 1.0 MeV) (n/cm ) 2.02 x 1019 2 i 8% t (E > 0.1 MeV) {n/cm ) 9.24 x 1019 2 i 15% dpa 3.95 x 10~2 i 11% 2 19 + (E < 0.414 eV) {n/cm ) 1.39 x 10 1 25% r NOTE: Total Irradiation Time - 6.05 EFPY f 6-29
TABLE 6-12
SUMMARY
OF NEUTRON DOSIMETRY RESULTS SURVEILLANCE CAPSULE X TIME AVERAGED EXPOSURE RATES p (E > 1.0 MeV) {n/cm -sec) 1.10 x 10ll 2 i 8% ( (E > 0.1 MeV) (n/cm -sec) 4.92 x 10ll i 15% 2 dpa/sec 2.13 x 10-10 i 11% ( (E < 0.414 eV) {n/cm -sec) 9.55 x 1010 2 1 23% INTEGRATED CAPSULE EXPOSURE 4 (E > 1.0 MeV) (n/cm ) 1.45 x 1019 8% 2 4 (E > 0.1 MeV) {n/cm ) 6.46 x 10I9 i 15% 2 dpa 2.80 x 10-2 i 11% f (E < 0.414 eV) {n/cm ) 1.25 x 1019 1 23% 2 NOTE: Total Irradiation Time - 4.16 EFPY 6-30 i 'I
1 TABLE 6-13
SUMMARY
OF NEUTRON DOSIMETRY _RESULTS SURVEILLANCE CAPSULE V ~ TIME AVERAG10 EXPOSURE RATES-p (E > 1.0 MeV) {n/cm -sec) 1.04 x 10I3 E i 8% p (E > 0.1 MeV) (n/cm -sec) 4.62 x 10ll' i 15% 2 dpa/sec 2.04 x 10-10 g gjg_ p (E < 0.414 eV) {n/cm -sec) 9.63 x 1010 i 22%- 2 INTEGRATED CAPSULE EXPOSURE f (E > 1.0 MeV) {n/cm ) 3.37 x 1018 2 i 8% t f (E > 0.1 MeV) {n/cm ) 1.50 x 1019 i 15%- 2 dpa 6.48 x 10~3 i 11% + (E < 0.414 eV) (n/cm ) 3,3; x ig,18 1 22% 2 NOTE: Total Irradiation Time i 1.03 EFPY 6 O J 6-31
TABLE 6-14 COMPARISON Of HEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER SURVEILLANCE CAPSULE U Adjusted Reaction Measured Calculation CIM Cu-63 (n,a) C0-60 5.23x10'I7 5.20x10'I7 1.00 Fe-54 (n.p) Mn-54 5.13x10-15 5.27x10-15 1.03 Ni-58 (n.p) C0-58 7.09x10-15 7.30x10-15 1.03 U-238 (n,f) Cs-137 (Cd) 3.69x10~I4 -3.23x10-14 0.88 Np-237 (n,f) Cs-137 (Cd) 3.12x10-I3 3.28x10"I3 1.05 00-59 (n,8) Co-60 (Cd) 2.89x10-12 2.90x10'12 0.99 Co-59 (n,a) Co-60 4.65x10-12 - 4.62x10-12 1.00 e 6 9. 4. 6-32
TABLE 6-15 COMPARISON OF MEASURED AND FERRE1 CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER SURVEILLANCE CAPSULE X Adjusted Beaction Measured Calculation .GM Cu-63 (n.o) 00-60 5.42x10'I7 5.42x10'I7 1.00 Fe-54 (n.p) Hn-54 5.51x10-15 5.61x10-15 -1.02 Ni-56 (n.p) 00-58 7.69x10-15 7.76x10-15 1.01 U-238 (b f) Cs-137 (Cd) 3.82x10"I4 3.38x10"I4 0.89 Np-237 (n.f) Cs-137 (Cd) 3.13x10-13 3.34x10-13 1.07 Co-59 (n,8) Co-60 (Cd) 3.lix10-12 3.12x10-12 3,og Co-59 (n,a) Co-60 5.44x10-12 5.40x10-12 o,gg 6 1 i 4 4 6-33
TABLE 6-16 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSVLE CENTER SVRVEILLANCE CAPSVLE V Adjusted Reaction Measured Calculation [jj Cu-63 (n,o) Co-60 5.21x10-17 5.29x10*l7 1.01-fe-54 (n,p) Mn-54 5.25x10-15 5.26x10-15 1.00 Ni-58 (n p) C0-58 7.00x10-15 7.19x10-15 1.03 U-238 (n,f) Cs-137 (Cd) 3.45x10~I4 3.17x10-14 0.92 Np-237 (n,f) Cs-137 (Cd) 3.17x10-33 3.26x10-13 1.03 Co-59 (n,6) 00-60 (Cd) 2.78x10-12 2.79x10-12 1.00 Co-59 (n,8) Co-60 5.lix10-12 5.07x10-12 0.99 9 6-34
k TABLE 6-17 ADJUSTED NEUIRON ENERGY SPECTRUM AT CENTER OF SURVEILLANCE CAPSULE U Ene y AdjusgedFlux Energy Adjusgedflux (n/cm sec) Group (Mev (n/cm -sec) Group (Mev) 6 28 9.12x10'3 2.18x1010 1 1.73x101 4.36x10 2 1.49x101 1.06x10 29 5.53x10-3 2.82x1010-7 7 30 3.36x10~3 8.89x109 3 1.35x101 4.89x10 4 1.16x101 1.24x108 31 2.84x10*3 8.54x109 5 1.00x10I 2.99x10 32 2.40x10-3 8.27x109 8 8 33 2.04x10-3 2.33x1010 0 5.40x10 6 8.61x10 7 7.41x10 1.29x109 34 1.23x10'3 2.13x1010 0 9 35 7.49x10'4 1.97x1010 0 1.87x10 8 6.07x10 9 4.97x10 3.98x10 36 4.54x10-4 1.88x1010 0 9 0 5.36x109 37 2.75x10~4 2.02x1010 10 3.68x10 11 2.87x100 1.14x1010 38 1.67x10-4 2.19x1010 0 10 39 1.0lx10-4 2.17x1010 12 2.23x10 1.60x10 0 10 40 6.14x10-5 2.14x1010 13 1.74x10 2.29x10 14 1.35x10 2.57x1010 43 3,73y,go-5 2.09x1010 0 15 1.11x10 4.77x1010 42 2,26x10-5 2.0lx1010 0 16 8.21x10-1 5.52x1010 43 1.37x10-5. 1.95x1010 17 6.39x10~I 5.79x1010 44 8.32x10-6 1.84x1010 18 4.98x10-I 4.22x10 45 5.04x10-6 1.60x1010 10 19 3.88x10-1 6.07x1010 46 3.06x10-6 1.55x1010 20 3.02x10~l 6.03x1010 47 1.86x10-6 1.41x1010 21 1.83x10-l 6.03x1010 48 1.13x10-6 1.06x1010 22 1.11x10-1 4.81x1010 49 6.83x10~7 1.24x1010 23 6.74x10-2 3.30x1010 50 4.14x10-7 1.49x1010 24 4.09x10-2 1.86x1010 51 2.51x10~7 1.39x1010-25 2.55x10-2 2.53x'1010 52 1.52x10-7 1.24x1010 26 1.99x10~2 1.19x1010 53 9.24x10-8 3.18x1010 27 1.50x10-2 1.49x1010 NOTE: Tabulated energy levels represent the upper energy of each group. 6-35 l
TABLE 6-18 ADJUSTED NEUTRON ENERGY SPECTRUM AT CENTER OF SURVEILLANCE CAPSULE X Energy AdjusgedFlux Energy AdjusgedFlux Group (Hev) (n/cm-sec) Group (Hev) (n/cm -sec) I 1.73x10 4.50x106 28 9.12x10-3 2.27x1010 I 2 1.49x101 1.01x107 29 5.53x10-3 2.95x1010 I 5.07x107 30 3.36x10-3 9.32x109 3 1.35x10 8 31 2.P4x10-3 8.97x109 4 1.16x10I 1.30x10 5 1.00x101 3.13x10 32 2.40x10'3 8.70x109 8 6 8.61x10 5.68x108 33 2.04x10'3 2.46x1010 0 0 1.37x10 34 1.23x10-3 2.25x1010 9 7 7.41x10 8 6.07x10 2.00x109 35 7.49x10-4 2.09x1010 l 0 0 4.26x10 36 4.54x10-4 1.99x1010 9 9 4.97x10 10 3.68x10 3.72x109 37 2.75x10-4 2.14x1010 0 11 2.87x10 1.21x1010 38 1.67x10~4 2.36x1010 0 12 2.23x100 1.68x1010 39 1.0lx10-4 2.31x1010' 0 10 40 6.14x10-5 2.28x1010 13 1.74x10 2.37x10 0 2.64x1010 41 3.73x10-5 2.22x1010 14 1.35x10 0 4.87x1010 42 2.26x10-5 2.15x1010 15 1.11x10 10 43 1.37x10-5 2.08x1010 16 8.21x10-1 5.59x10 17 6.39x10-I 5.84x1010 44 8.32x10-6 1.97x1010 18 4.98x10-1 4.25x1010 45 5.04x10-6 1.80x1010 19 3.88x10-1 6.10x1010 46 3.06x10-6 1.66x1010 20 3.02x10-1 6.07x1010 47 3.'86 x10-6 1.51x1010 21 1.83x10-1 6.09x1010 48 1.13x10-6 1.14x1010 22 1.11x10-1 4.88x1010 49 6.83x10-7 1.39x1010-23 6.74xlC-2 3,37xiol0 50 4.14x10-7 1.74x1010 10 51 2.51x10-7 1.70x1010 24 4.09x10-2 1.90x10 10 52 1.52x10-7 1.58x1010 25 2.55x10-2 2.60x10 26 1.99x10-2 1.23x1010 53 9.24x10-8 4.54x1010 27 1.50x10-2 1.55x1010 NOTE: Tabulated energy levels represent the upper energy'of each group. 6-36 7 I --,----,.,,---,.-n, er
f TABLE 6-19 ADJUSTED NEUTRON ENERGY SPECTRUM AT CENTER OF SURVEILLANCE CAPSULE V j Energy Adjus}edFlux Energy AdjusgedFlux (n/cm sec) Group (Mev) (n/cm sec) Group (Mev) F 6 28 9.12x10-3 2.03x1010 1 1.73x101 7.53x10 2 1.49x10I 1.69x10 29 5.53x10-3 2.62x1010 7 9 3 1.35x101 6.49x107 30 3.36x10~3 8.20x10 8 9 1 1.44x10 31 2.84x10-3 7.85x10 4 1.16x10 9 l 8 32 2.40x10~3 7.58x10 1 3.18x10 5 1.00x10 6 8.61x100 5.45x108 33 2.04x10-3 2.14x1010 0 9 34 1.23x10-3 1.98x1010 7 7.41x10 1.26x10 8 6.07x100 1.81x109 35 7.49x10-4 1.85x1010 0 3.88x109 36 4.54x10~4 1.77x1010 9 4.97x10 9 37 2.75x10-4 1.91x1010 0 5.23x10 10 3.68x10 0 10-11 2.87x10 1.12x1010 38 1.67x10-4 2.10x10 0 1.59x1010 39 1.Clx10~4 2.07x1010 4 12 2.23x10 0 2.26x1010 40 6.14x10-5 2.05x1010 13 1.74x10 14 1.35x10 2.54x1010 41 3.73x10-5 1.99x1010 0 0 10 15 1.llx10 4.67x1010 42 2.26x10-5 1.92x10 10 16 8.21x10~I 5.34x1010 43 1.37x10-5 1.86x10 10 44 8.32x10-6 1.77x1010 17 6.39x10-I 5.54x10 18 4.98x10-1 4.00x1010 45 5.04x10-6 -1.62x1010-10-i 19 3.88x10~I 5.59x1010 46 3.06x10-6 1.51x10 10 20 3.02x10-1 5.70x1010 47 1.86x10-6 1.39x10 21 1.83x10~l 5.61x1010 48 1.13x10-6 1.03x1010 10 49 6.83x10-7 1.31x1010 22 1.11x10-I 4.45x10 10 23 6.74x10-2 3.07x1010 50 4.14x10~7 1.72x10 10 51 2.51x10-7 1.69x1010 24 4.09x10-2 1.73x10 10 52 1.52x10-7 1.60x1010-25 2.55x10-2 2.26x10 L.- 26 1.99x10-2 1.11x1010 53 9.24x10-8 4.63x1010 27 1.50x10-2 1.41x1010 NOTE: Tabulated energy levels represent the upper energy of each group. 6-37 .,i
TABLE 6-20 COMPARISON OF CALCULATED AND MEASUP.ED EXPOSURE LEVELS FOR W. B. MCGUIRE UNIT 2 SURVEILLANCE CAPSULES ) J CAPSULE O
-l i
Calculated Measured CM 1 f(E > 1.0 MeV) {n/cm ) 1.67 x 1019 2.02 x 1019 0.83 2 f(E > 0.1 MeV) {n/cm ) 7.67 x 1019 9.24 x 1019 '0.83 2 dpa 3.39 x 10-2 3.95 x 10-2 0.86 i t i f(E < 0.414 eV) {n/cm ) 6.54 x 1018 1.39 x 1019 0.47-2 i CAPSULE X C3Jsid.gind Measured C3 f(E > 1.0 MeV) {n/cm ) 1.21 x 1019 1.45 x 1019 0.83-2 f(E > 0.1 MeV) (n/cm ) 5.56 x 1019 6.46 x 1019 0.86 2 dpa 2.46 x 10-2 2.80 x 10-2 0.88 f(E < 0.414 eV) (n/cm ) 4,74 x 3018 1.25 x 1019 0.38 2 l l 6-38 + -..m,.
TABLE 6-20 (Continued) COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR W. B. MCGUIRE UNIT 2 SURVE!LLANCE CAPSULES CAPSULE V Calculated Measured [fd 2 18 3.37 x 1018 0.80 f(E > 1.0 MeV) (n/cm ) 2.70 x 10 t(E > 0.1 MeV) (n/cm ) 3,39 x go19 1.50 x 1019 0.79 2 dpa S.39 x 10-3 6.48 x 10-3 0.83 2 18 18 0.41 t(E < 0.414 eV) {n/cm ) 1.29 x 10 3.12 x 10 = F G 6-39
TABLE 6 < NEUTRON LXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE AZIMUTHAL ANGLI-O' 1 30' 45' 6.05 EFPY 8: f(E>1.0 MeV) 2.52 x 10 3.76 x 10 2.79 x 10 3.85 x 10 2 (n/cm ) 13 4(E>0.1 MeV) 5.21 x 10 7.86 x 10 6.96 x 10 9.56 x 10 2 (n/cm ) dpa 3.75 x 10 5.56 x 10 4.31x15-3 5.86 x 10 ~3 ~ ~ 16.0 EFPY f(E>1.0 MeV) 6.66 x 10 9.94 x 10 7.38 x 10 1.02 x-10 ' 18 18 0 (n/cm'd) 4(E>0.1 MeV) 1.38 x 10 ' 2.08'x 10 ' 1.84 x 10 ' - 2.53-x 10 ' 2 (n/cm ) -2 ~ dpa 9.92 x 10~ 1.47 x 10 1.14 x 10~ 1.55 x'10 32.0 EFf_1 b*.l.i HeV) 1.33 x 10 ' 1.99 x 10 1.48 x 10 2.04 x'10 2 (n/cm) f(E>0.1 MeV) 2.76 x 10 4.16 x 10 ' 3.68 x 10 ' 5.06 x 10 2 (n/cm ) dpa 1.98 x 10~ 2.94 x 10~ 2.28 x 10~ 3.10 x 10~ 6-40
TABLE 6-22 PROJECTED FAST NEUTRON EXPOSURE VALUES 16 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2 2 (n/cm ) (equivalent n/cm ) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 18 17 18 I8 18 18 3.62 x 10 7.79 x 10 6.66 x 10 4.20 x 10 1.46 x 10 O' 6.66 x 10 18 18 18 18 18 15" 9.94 x 10 5.38 x 10 1.12 x 10 9.94 x 10 6.22 x l'+' 2.13 x 10 l8 18 18 17 18 4.92 x lb ' 1.88 x'!O 30" 7.38 x 10 4.04 x 10 8.86 x 10 7.38 x 10 19 18 1.10 x 10 1.02 x 10 6.52 x 10 ,2.23 x'1018 18 19 16 45*(a) 1.02 x 10 5.46 x 10 32 EFPY .NFUTRON FLUENCE (E > 1.0 MeV) SLOPE 'doa SLOPI' 2 2 (n/cm ) (equivalent n/cm ) Surface 1/4 T 3/4 T Surface 1/4 T '3/4 T 18 18 18 19 8.38 x 10 2.91_.x 10 18 1.56 x 10 !.33 x 10 19 7.24 x 10 0* 1.33 x 10 18. 19 19 18 19 19 '4.26 xL 10 15' l.99 x 10 1.08 x 10 2.25 x 10 1.99 x 10 1.25 x 10 l9 18 18 I9 18 3.77'x 1018 30* 1.48 x 10 8.10 x 10 1.78 x 10 1.48 x 10 9.87 x 10 19 19 18 19 19 18 45*(a) 2.04 x 10 1.09 x 10 2.20-x 10 2.04 x 10 1.30 x 10 4.47 x.10 (a) Maximum point' on the pressure vessel 6-41
TABLE 6-23 UPDATED-LEAD FACTORS FOR HCGUIRE UNIT 2 SURVEILLANCE-CAPSULES CARigla lead Factor. U 5.28(a) X 5.28(a) W 5.28 Z 5.28 V 4.62(a)- Y 4.68 (a) Plant specific evaluation 9 e 6-42
'SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASlM E185-82 and.is recommended for future capsules to be' removed from the McGuire Unit 2 reactor vessel: Capsule Estimated Locction Lead Fluence Capsule (deg.) Factor Removal Time (a) -(n/cm ). 2 V 58.5 4.62 1.03 (Removed)l 3.06 x 1018 (Actual) X 236.0-5.28 4.16 (Removed) 1.45 x'1019 (Actual) U 56.0 5.28 6.05 (Removed) 2.02 x 1019 (Actual) W 124.0 5.28 9 3.03'X 10I9 Z 304.0 5.28 Standby Y 238.5 4.68 Standby (a) Effective Full Power Years (EFPY) from plant startup. l 7-1
SECTION
8.0 REFERENCES
1. K. Koyama and J. A. Davidson, " Duke Power Company William B. McGuire Unit i;o. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9489, May 1979. 2. Code of Federal Regulations,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 Mtterials", U.S. Nuclear Regulatory Commission, May 1988. 4. Section III of the ASME Boiler and Pressure Vessel Code, Appendix G, " Protection Against Nonductile Failure." S. ASTM E208, " Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels." 6. ASTM E185-82, " Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)." 7. ASTM E23-88, " Standard Test Methods for Notched 8ar Impact Testing of Metallic Materials." 8. ASTM A370-89, " Standard Test Methods and Definitions for Mechanical-Testing of Steel Products." 9. ASTM E8-89b, " Standard Test Methods of Tension Testing of Metallic Materials."
10. ASTM E21-79(1988), " Standard Practice for Elevated Temperature Tension
~ - Tests of Metallic Materials." 8-1
.~ i --11.JASTM E83-85, '" Standard ' Practice for: Verification and Classification' of Extensometers." 112 _ S. E.- Yanichko,- et al.,;" Analysis of Capsule V from ~the Duke Power Company ; t McGuire Unit 2 Reactor Vessel Radiation Surveillance Program", WCAP-11029, y January.1986. j 13. E. Terek, et al., " Analysis-of Capsule X from the Duke Power Company-McGuire Unit 2 Reactor-Vessel Radiation Surveillance _ Program",-~WCAP-12556,_ April 1990. 14. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler,L " Nuclear - Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates. Transport - Technique", WANL-PR(LL)-034,' Vol. 5, August 1970. - 15. "0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled'SelfeShielded, _47_ Neutron, 20 Gamma-Ray, P3, Cross Section Library for. Light Water-Reactors". 16. V. A. Perone, et al., "The Nuclear Design and Core Physics. Characteristics-of the W. B. McGuire Unit 2 Nuclear Power Plant - Cycle 1", WCAP-10182,. September 1982. (Proprietary) a ~ 17. C. R. Savage, et al., "The Nuclear Design' of the McGuire Unit 2: Nuclear Power Plant - Cycle 2", WCAP-10747, March 1985. --(Proprietary) 18. P. D. Banning,.et, al., "The Nuclear Design of the' McGuire Unit 2 Nuclear ~ Power Plant - Cycle 3", WCAP-11048, March 1986. (Proprietary)- 19. P. D. Banning, et. al., "The Nuclear Design of the McGuire' Unit 2 Nuclear Power Plant - Cycle 4",.WCAP-11530, June _~1987. -(Proprietary) 20. J. R. Lesko, _et. al., "The Nuclear Design of the McGuire Unit 2 Nuclear Power Plant - Cycle 5", WCAP-ll891, July 1988._ (Proprietary)- 8-2 a y c -:pewm-aet-a-
21. M, A. Kotun, et. al., "The Nuclear Design of _the McGuire Unit 2 Nuclear Power Plant - Cycle 6", WCAP-12316, August 1989. (Proprietary) 22. J. R. Lesko, et. al., "The Nuclear Design of the McGuire Unit 2 Nuclear Power Plant - Cycle 7", WCAP-12736, November 1990. (Proprietary) 23. ASTM Designation E482-89, " Standard Guide for Application of Neutron. Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991. 24. ASTM Designation E560-84, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, SectiorL 12, American Society for Testing and Materials, Philadelphia, PA,1991. 25. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991. 26. ASTM Designation E706-87, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", ir. ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991. 27. ASTM Designation E853-87, " 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,1991. 28. ASTM Designation E261-90, " Standard Method for Determining Neutron Flux, l Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,- 1991. 8-3 I
29. ASTM Designation E262-86, Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section_12, American Society for Testing and Materials, Philadelphia, PA,-1991.
30. ASTM Designation E263-88, " Standard Method' for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12,-
American Society for Testing and Materials, Philadelphia, PA,1991. -
31. ASTM Designation E264-87, " Standard Method for Determining Fast-Neutron flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
32. ASTM Designation E481-86, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelph'ia, PA, 1991. 33. ASTM Designation E523-87, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTM _ Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991. 34. ASTM Designation E704-90, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
35. ASTM Designation E705-90, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237", -in ASTM Standards, _ Section 12, American Society for Testing and Materials, Philadelphia,- PA,1991.
36. 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,1991. 8-4
37. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-THE 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979. 38. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative ligthod of Neutron Flux Snectra Determined by Foil Activation, AFWL-TR-7-41, Vol.1-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967. 39. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al.,1981. o-W 8-5 l
4 ~ APPENDIX A Load-Time Records for Charpy Specimen Tests-O 9 l l l A-0
n n y 0 O a A O . e O L L E T A S U E T R C A A A R = F p A W P p / D A O g l i i i I 8 L M d U ro M I ce X r A M e [ m M I i. t g T P d = a o 1 l I I I 1 I I I I iI ' l I 1 I l = dez i lae d I I 1 -A er W, u g i m F t L A R D EN A E O G L O L y l 1 i l i j l r E gl y I P Y g t o403 >4
rCculet s2 *u' nec OL6 OL6 .g i - a a i 7e A4 a g. n T- ^ y-e i ^ i .D 4.0 8.0 sk.0 16.0 20.0 TINC ( PCEC ) stess-t MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U TANG
DL6 nceu Rc et u-rice oL9 tts s
s s ?=- AJ m e = 3-a w w tu - .0 4.0
8. 0 12.0 16.0 20.0 TINC
( MSEC ) stems-t MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U TANG
DL9 Figure A T.
Load-time records for Specimens DLG and DL9 A-2
k '~ nCCutrC e2 u# nCC DLt - 0L1 i a a a g. 7m p4 ] .4 33_ M w a N o s .m A L 4 i a i .D .0 1.6 2.4. 3.2
4. 0.
TIMC ( PCCC ) MCGUIRE #2 "U" Batch:MCG.DL1,_ NCCutRC a2 "U' ttCC DL9 DLS i i i i J 7 m _. 4. o Q. en v v .a of _ \\s-o . s s m e i .0 2.0 4.0 6.0 8.0 10.0 TINC ( MSEC ) s10=s.t MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" TANG
DL8
- * = Figure A-3. Load-time records.for Specimens.DL1 and DL8 A-3
4 NCutRC e2 *U* -NCC OL3-ON 4 i. i a 6 ? e_ a# e a = +_ - 4 'j n v .qu mut : gN - g .e N-i (A .D .8 1.6 2.4 3.2 4.0 TIMC ( MSCC ) MCGUIRE #2 "U" Batch:MCG_DL3._ MCCUIRC 82 *U* MCC OL10 DL10 i s i i g ? m-a4 7 m e_ ci w w ~ . A, ^_ o e .0 2.0 4.0 6.0 8.0 10.0 TIMC ( MICC 3 s10se-1 MCGUIRE #2 "U" Note the change in MCGUIRE #2 the horizontal scale CAPSULE "U" TANG
DL10 Figure A-4.
Load-time records for Specimens DL3 and DL10 A-4 - l
NCCu!RC e2 *U* PCC DL12 OLi2 g i i a 7 e-A m w w_ l q-A_ A .D .8 1.6 2.4 3.2 4.0 TitC ( tCtc ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U TANG
DL12 r1CCu!RE e2 *U' MCC OL13 DL13 i
4 a i g. m- +- 2 ^- S 9-n v v-o 4 8 a q-ks \\ o i 4 .D .8 8.6 2.4 3.2 4.0 TIMC ' < fnEC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" - -TANG
DL13-Figure-A-5. Load-time records for Specimens DL12 and DL13 A-5
N CutRC st-'U* 'MCC OL2 OL2' o i i i i j
8. -
a A $_d-w l ~. - i N- .D .8 1.6 2.4 3.2 4.0 TtriE ( n:CC ) MCGUIRE #2."U" MCGUIRE #2 CAPSULE U TAND
DL2 MCCU!RC e2 *U*
NCC OL4 OL4 i i e i g e_ g,. n 4-n v w a m- -m .D ,8 1.6 2.4- - 3. 2 4.0-TIME < MSEC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" TANG
DL4, Figure A-6.
Load-time records for Specimens DL2 and DL4 A-6 i
EU -NCCUIRE s2 'U'. MCC DLtd-OLtd i a i g_ 7=- A - i v w-a l q- +- o-i i i i .0 .8 1.6 2.4 3.2 4.0 T!r4 ( MSEC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" TANG
DL14 MCCUIRC s2 "U*
MCC OLS DL5 i i g i 7 m. . i 4 ~ ^ S.9 - 0 'T"- a '.M. - \\ l o i .0 1.2 2.4 3.6 4.8
6. 0 TIME
( MSEC ) l. MCGUIRE #2 "U"' MCGUIRE #2 CAPSULE U ( TANG
DL5 ll Figure A-7.
Load-time records-for' Specimens DL14 and DL5-t l~ A-7 I l ? I t
,NCCulE'c 02 'u' NCC OLil' OLI1-i a e i _j 7e E4 ~ 1 m e, _ a v-a n-o i i i s .0 .8 1.6 2.4 3.2 4.O i:nc < n:cc > MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U TANG
DL11 MCCUIRC e2 'U*
MCC OL15 DL15 i i i i g ~ E4 n w e v a m. N o i i i ,D .0 1.6 2.4 3.2 4.0 TINC ( MSEC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" TANG
DL15 Figure A-8.
Load-time records for Specimens DL11 and DL15 A-8 I e
NCCu!FC e2 u' NCC Ot> ' OL? ..g. 2 i i i ?
a. -
E# _e S 9-n w v l g4 .a q-l l I o i i i i .D .0 1.6 2.4 3.2 4.O TIMC ( tt:EC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" TANG
DL7 NCCU1RC 62
'U" MCC 078 0T8 .g i i e i 7m ~ &J + n 39-e v .J 9-e i .D 4.0 . 9.0 12.0 16.0 20.0-TIME < MSEC ) atone-1 MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U AXIAL
DT8 Figure A-9.. Load-time records for Specimens DL7 and DT8 A-9 l
l
NCCult'C at "U' NCC CTP ' 017 i i i i g $ 9~ -. e a n-S ' *. - n w v. N-e .0
2. 0 _
4.0
6. 0 0.0 10.0 TINC (N CC 3 a1Cae.1 MCGUIRE /2 "U"
MCGUIRE #2 CAPSULE "U" AXIAL
DT7 MCCUIRC s2 U"
MCC 071 0T1 i e i a g \\ 0 m_ a4 3 *, - n w a n-o i .0 4.0 8.0 12.0 16.0 20.0 TIMC ( MSCC ) 510am-l MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U AXIAL
DT1 Figure A-10.
Load-time records for Specimens DT7:a.nd DT1 A-10 l
NCu!FC e2 *U* MCC 013 Of3 i i 1 .t ,j 7m a4 4 3 4, - n w .a N-i i i U 2.0 4.0 6.0 8.0 10.0 TINC ( ftSCc ) atOso-1 MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U AXIAL
DT3 NCCU1RC 62 *U*
NCC OT6 OT6 4 a a i g 7m &J 7 3., _ m-v .a = n,. A o i i _ i i .D 2.0 4.0 6.0
8. 0 10.0-TIMC
( M3EC ) =10==-1 MCGUIRE #2 "U" MCGUIRE-#2 CAPSULE "U" AXIAL
DT6 Figure-A-11.
Load-time records for Specimens DT3 and DT6 .A-11
MCCutRC e2 'U* r1CC 0145 - 0715 e i i ..j ?.. &4 m, y n w s. .a N, = .V .e 1.6 2.4 3.2 4.0 T!rtC ( MCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U AXIAL
DT15 11CCU!RC s2 "U*
PtCC Off! 0T11 i i i i e ?=_ a4 7 i l ^ l n w N-l l I l f , e A~ j .D .8 1.6 2.4 3.2 4.0 .] T!?1C t MEC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" AXIAL
DT11 Figure A-12.
Load-time recordc. for Specimens DT15 and DT11 A-12 I i
rtCCultC et "U" rtCC 072 Ot3 i i i i g 7 e- &J 7 33-j e, - N ew - , he A A w. i i i o .0 .0 1.6 2.4 3.2 4.0 Titic ( ft3CC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" AXIAL
DT2 tiCCulRC 62
'U" t1CC DT9 DT9 i i i i ?.- i# 3
e. -
n w a eu, e. ~- .0 ,e 1.6 2.4 3.2 4.0 TIMC ( F1 SCC ) MCGUIRE #2 "U" Batch MCG.DT9 - Figure A-13. Load-time records for Specimens DT2 and DT9 A-13
71CCutRC at.'U* NCC DTid - of te - i i i .4 7.- &4 a 3 e,.. n w v. .2
a -
A m o-e s .0 1.2 2.4 3.6 4.8 6.0 Tlt1C ' ( nSCC ) MCGUIRE #2 "U" MCGUIRE #2 -i CAPSULE U AXIAL
DT14 f1CCUIRE e2 "U*
f1CC 0T13 0T13 i i i g ?. E4 ~ ~ ".1 * - n w v- .J q_ _. ~. - .0 .8 1.6 2.4 3.2 - 4. 0 ftt1C . ( rtSCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" . AXIAL
DT13 Figure A-14.
Load-time records for Specimens DT14 and DT13 A-14 i l'
.PCCU!EC 42 *U* M"C 0710 -0T80: i i_ i. i 4 s ?._ Ei a n .r v o i i .D .e 1.6 2.4 3.2 4.0 T!ric ( MstC > MCGUIRE /2 "U" MCGUIRE #2-CAPSULE U -AXIAL
DT10 NCCUIRE #2 *U*
MCC GT12 0T12 i e i ? a. h a g a_ n v A I .a q_ .D .8 1.6 2.4 3.2 4.0 TIME t MSCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U AXIAL.
DT12-Figure A-15.
Load-time records -for Specimens DT10 and DT12 - A-15
7=- ttCCulpt e2 *U* ttCC 015 013 a a a i g 7=- a4 A e R-n w w-o i i i .0 .8 1.6 2.4 3.2 4.O T!!1C ( ttTCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" AXIAL
DTS ftCCU!RC 32 *U*
ttCC OT4-074 i i s a g ? m- &4 ~ a 3 g-n v a N, - I e i a i .V .8 1.6 2.4 3.2 4.0 TIME ( ttSCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" AXIAL
DT4 Figure A-16.
Load-time records foz Specimens DT5 and DT4 s A-16
MCCulRC s2 'U* MCC N1 out i i i i j 7 e. 5 a ~ n v A y_ A o i e i .9 4.0 8.0 12.0 16.0 20.0 TIME ( MSEC > uloss-t MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U WELD
DW1 necutRc s2 u-nec ous ous i
i i i g 7._ h a 3._ n. v 9N S q_ ~ i f f .M A _ _ (. o i i i .D .8 1.6 2.4 3.2 4.0 TINC ( MSEC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" WELD
DWS Figure A-17.
Load-time records for Specimens DW1 and 'DW5 A-17 i
1CCulRC s2 'U*
PCC Ou? our i i i g 7.. 5# A a 4 w .4 y A._ m o e i i i .9 1.2 2.4 3.6 4.8 6.0' TIMC ( MstC > MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U WELD
DW7 MCCUIRC a2 *U*
HCC Cut 5 Out5 i i i g 7m 4 9 g-n w q .0 4.0
8. 0 -
12.0 16.0 20.0 TIMC ( ttstC ) stone-1 MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" WELD
DW15 Figure A-18.
Load-time recc.;.s for Specimens DW7 and DW15 A-18
o W oulet at 'U' NCC Dutt 7 a g u = k N~ ~ T
.. f N
4 w o, ~ = e= .9 .9 1.6 E.4
3. P.
4.0 fif1C ( M3CC 3 HCGUIRE #2 "U" Note the change in Batch:McG_DW12 the vertical scale ecoutre et au-nce oaa our g d' ?=_ &4 M e 4, N w v, - N- ,.. I ~g_ % ^ _- .0 .e 1.6 t.4 3.2 4.0 FINE ( ft3CC ) MCGUIRE #2 "U" Datch:MCG_DW2_ Figure A-19. Load-time records for Specimens DW12 and DW2 A-19 l l m
< w iec., v. <c cu. cu. i i i i A4 I 3 *.
D
~ n v v, N.. .9 .8
1. t, r.4 3.2 4.0 flht
( MtEC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" WELD
DW4 nccutet er *u-nec tuis tuta i
i i j 7 e.
h.
34m w
w_ J N, - = .9 .0 1.6 2.4 3.2 4.0 TinC ( MitC > MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U WELD
DW13 Figure A-20.
Load-time records for Specimens DW4 and DW13 A-20 _.w
I t i l Mtculpt se *u' Tc (us tus i i i g 7 m. i &4 _i. ^ S*- n w 1 i M i i .0 .s 16 8.4 ' 3.0 4.0 fint < n:tc > MCGUIRE #2 "U" Batch!MCG_DW8_ l nec tuli ouis - . _ necutet er 'u' 7e S4 T e m e. ei w w. a 4 9. ~ -v_- .0 .0 1.6 2.4 3.2 4.0 Tint ( nStc ) MCGUIRE #2 "U" - -c MCGUIRE #2 CAPSULE "U" WELD - : DW11. Figure A-21. Load-time records for Specimens DW8 and DW11-t i A-21
NCCutet et 'vo Mcc tus (nre - 1 i i j i i 7._ e i \\ i e *_ ei X w i a N e i .0 1.2 2.4 3.6 4.8 6.0 finc ( nstc > MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U WELD
DW9 5
necutet se *u-nec cute outo i i i i ?... ~ g g_ n a w-o e i i .D .8 t.6 2.4 3.2 4.0 11ht ( nsCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U WELD
DW10 Figure A-22.
Load-time records for Specimens _ DW9 and DW10 A-22
TCulrt et 'U' 'CC (W3 to) i i i g i I *.- I a li .- ) s.- - - 1 d-a W I a' N-o .9 .0 1.6 2.4 3.2 4.0 flNC ( MSCC 1 - MCGUIRE #2 "U" MCGUIRE /2 CAPSULE "U" WELD
DW3 necutRt er u.
ecc out4 ous4 i g._ a e 3e n w e i. N e i i e i .0 1.2 2.4 3.6 4.8 6.0 Tit 1C ( M$CC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U WELD
DW14 Figure A-23.
Load-time records for Specimens DW3 and DW14 A-23 t
FCCultt of *ua K C Du6 Du6 i i i j 7.. Si ~ ~ 3 "*~ w s l a h .3 ,3 1.6 2.4 3.2 4.0 TINC ( MSCC ) MCGUIRE #2 "U" 1DyJIRE /2 t;p SULE "U" t WELD iDW6 f MCCUlRC 02 'U* fCC DHl0 OH10 i i 4 j 7._ E4 ~ e e e, w ~ w y_ e i i i .0 4.0 8.0 12.0 16.0 20.0 flNC ( MSCC ) aloss-1 MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U IIAZ
Dillo Figure A-24.
Load-time records for Specimens DW6 and Dillo I A-24 4
nCCutK 82 'U' NCC tatt (>t1 e i i i 4 7e Se . ) I 5*- n w 4 y.,.. _w A..A n --- 4. .9 .8 16 2.4 3.3
4. 0 -
-TINC ( MSCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" HAZ
D119 neculet sa u-nec tsis osis i
i i =- 8# a S *- M s M M8'*%-- A A. m_ A .0 4.0 8.0 12.0 16.0 20.0 T!NC ( MSCC ) stone. MCGUIRE #2 "UH-MCGUIRE #2 CAPSULE U HAZ
DH13
. Figure A-25. Load-time records for Specimens DH9 and DH13 A-25 9 l
PCQUItt 8i *O' NCC nd mf5 i i i i j a t, e' 1 ~ ~ e w ~ ) a ) \\ n-j ~
e. o '

4. c' 6.c'

e. o' t o. o o
tinc c nstc > mio.. i HCGUIRE #2 "U" HCGUIRE #2 CAPSULE "U" HAz DH5 necutet se u-Nec oss ao a A M v v a N ^___ .9 .8 1.6 a.4 3.2 4.0 finC ( nttC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U HAZ
DH3 Figure A-26.
Loa.d-time records for Specimens DHS and DH3 A-26
... = -. _.. T.Culst of 'U* PCC Dill (Hit i i i e g 7e E. t 3 i-a V e-a q-1 m _. 0 .0 l.6 2.4 3.2 4.0 f!NC ( M:tc ) MCGUIRE #2 "U" HCGUIRE #2 CAPSULE "U" .HAZ
DH11 necutat er *u-nec av av y
e-3* a m 59-M w ~ v ID 2.o'
4. 0'

6. 0 '

8. 0' 10.o TInc
( MSEC ) eiCan-1 MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" HAZ
DH7 Figure A-27.
Load-time records for Specimens Dilll-and D117 A-27
MCculPC et *U* MCC 3 4 34 4 u e a e ?. 04 T " i-s n s v .e q .9 .s 86 2.4 3.2 4.0 TINC ( PCEC > MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" HAZ
DH6 MCCUIRC et 'U' MCC DH2 OH2 g
a i i i 7=- &J A 53-n w e, .a N-i .9 .8 1.6 2.4 3.2 4.0 flMC ( ICEC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U HAZ
DH2 Figure A-28.
Load-time records for Specimens DH6 and DH2-A-28 y e
NCul&C of "U* TC D4tf D4 2 4 i a ?.- I + s... l t ~_ = .o .e 1.6 a.4 3.e 4.o TIE - ( MSCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" HAZ
DH12 neculet en au-nec OH4 DH4 4
] e, _ ~ s ^ 3*- a w v ~ \\ , N, o i i .0 .8 1,6 2.4 3.2 4.0 TIME ( MSCC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE U HAZ
DH4 Figure A-29.
Load-time records for Specimens DH12 and DH4 A _ _
elk st u TC DHte 04:4 e i i i 4 7&4 ~ r $ n* - e )- q._ i ~ o i i .0 .8 1.6 2.4 3.2 4.0 FINE ( n:CC ) MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" HAZ
DH14 necutar. et u-nec mis use i
i i ?. &4 M 3,_ n v v gd y_ 1 I .0 .8 1.6 2.4 3.2 4.0 l TIMC ( MSCC ) l MCGUIRE #2 "U" l MCGUIRE #2 CAPSULE U l. HAZ
DH8 Figure A-30.
Load-time records for Specimens DH14 and DH8 l A-30
NCCultt et 'Us NCC DOS OHis i i i i g 7.- h a *. n v 1 i N. o .9 .9 1.6 2.4 3.2 4.0 TINC ( MSCC 3 MCGUIRE #2 "U" MCGUIRE #2 CAPSULE "U" HAZ
DH15 NCCU!RE st
'U* NCG OH1 OH1 .g ,i i e i g; A 3 4,. n w v. a N-o .0 1.2 2.4 3.6 4.0 6.0 TIME ( MSEC 3 MCGUIRE #2 "U" MCG;1 IRE #2 CAPLULE U HAZ
DH1 Figure A-31.
Load-time records for Specimens DH15 and DH1
t APPENDIX B + 4 HEATVP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION FOR HCGUIRE UNIT 2 P er W B-0 l ~
l TABLE OF CONTENTS j Section Ilth Eagg LIST OF ILLUSTRATIONS B-2 LIST OF TABLES B-2 B-1 INTRODUCTI.N B-3 B-2 FRACTURE TOUGHNESS PROPERTIES B-3 B-3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS B-4 B-4 HEATUP AND C00LDOWN PRES $URE-TEMPERATURE LIMIT CURVES B B-S CALCULATION OF ADJUSTED REFERENCE TEHPERATURE B-9 ~ B-6 REFERENCES B-15 ATTACHMENT B1: DATA POINTS FOR HEATUP AND COOLDOWN CURVES B WITH MARGINS OF 12'F AND 70 PSIG FOR INSTRUMENTATION ERRORS ATTACHMENT B2: DATA POINTS FOR HEATUP AND COOLDOWN CURVES B-24 WITHOUT MARGINS FOR INSTRUMENTATION ERRORS e B-1
LIST OF ILLUSTRATIONS Fioure Ij_th hgg B-1 McGuire Unit 2 Reactor Coolant System lleatup Limitations B-13 (Heat up rates up to 60*F/hr) Applicable for the First 16 EFPY (With Margins of 12*F and 70 psig For Instrumentation Errors) B-2 McGuire Unit 2 Reactor Coolant System Cooldown Limitations B-14 (Cooldown Rates up to 100'F/hr) Applicable for the First 16 EFPY (With Margins of 12'F and 70 psig For instrumentation j Errors) LIST OF TABLEi, lahh 11.th hat B-1 McGuire Unit 2 Reactor Vessel Toughness Table B-10 (Unitradiated) B-2 Summary of Adjusted Reference Temperatures (ART's) at 1/4T B-ll and 3/4T Locations for 16 EFPY B-3 Calculation of Adjusted Reference Temperatures at 16 EFPY B-12 for the Limiting McGuire Unit 2 Reactor Vessel Material - Lower Shell forging 04 O B-2
l B-1. INTRODUCTION l Heatup and cooldown limit curves are calculated using the most limiting value f of RTNDT (reference nil-ductility temperature) corresponding to the limiting t,eltline region material for the reactor vessel. The most limiting RTNDT Of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture toughness properties and estimating the radiation-induced ARTNDT. RTNDT is designated as the higher of either the drop weight n!1-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F. increases as the material is exposed to fast-neutron radiation. RTNDT Therefore, to find the most limiting RTNDT at any time period in the due to the radiation exposure associated with that reactor's life, ARTNDT time period must be added to the original unirradiated RTNDT. The extent of the shift in RTNDT i s enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)[B1). Regulatory Guide 1.99, Revision 2 is used for the calculation of ART values at 1/4T nd 3/4T locations (T is the thickness of the vessel at the beltline region). B-2. FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in_ accordance with the NRC Regulatory-Standard' Review Plan [B2). The pre-irradiation fracture-toughness properties of the McGuire Unit 2 reactor vessel are presented in Table B-1, B-3 L i**9-+ g- ,r... .-,-e. .y - -, -.,,,e,..e.--,--- ,:-,-,.m..m. m -.-- -- +
4
.<- + - -..
,.,-*-.=---e,
B-3. CRITERIA TOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3 The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIRs for the metal temperature at that time. Kjg is obtained from the reference frac'" n toughness curve, defined in Appendix G to the ASME Code [B3). The KIR ' is given by the following equation: KIR - 26.78 + 1.223 exp [0,0145 (T-RTNDT + 160)) (1) where klR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTHDT Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Coda [B3) as follows: C*Kg+ KIT $KIR (2) i where r K g - stress intensity factor caused by membrane (pressure) stress i KIT - stress intensity factor caused ej the thermal gradients i KIR - function of temperature relative to the RTNDT of the material-C , 2.0 for level A and Level B service limits C - 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical B-4
l At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the l reference flaw are computed. From equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. for the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of i the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curv? in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the maturial temperature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation, it follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR at the 1/4 T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KIR exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various B-5 u
intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. i Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the irside of the wall that alleviate the tensile stresses produced by-internal pressure. The metal temperature at the crack tip lags the coolant temperature; for the 1/4 T crack during heatup is lower than the KIR therefore, the KIR for the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especial?r at the end of the transient, conditions may oxist so that the effects of compre,sive thermal stresses and lower KIR's f do rot offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar mves for finite heatup rates when the 1/4 T flaw is considered. Therefore. both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of the ~ pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses'are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each neatup rate must be analyzed on an individual basis.- Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by-c constructing a composite curve based on a point-by-point comparison of the B-6
steady-state and finite heatup rate data. At any given temperat se, the allowable pressure is taken to be the lesser of the three values taken from the cury. under cesideration. The use of the composite curve is necessary to set conser 2tive heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion. Finally, the 1983 Amendment to 10CFR50[843 has a rule which addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for McGuire Unit 2). Table B-1 indicates that the limiting unirradiated RTNDT of l'F occurs in the closure head flango of McGuire Unit 2, so the minimum allowable temperature of this region is 121'F. These limits plus additional margins of 12'F and 70 'psig are shown in Figures B-1 and B-2 whenever applicable. B-4. HEATUP AND C00LDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor pressure vessel have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Section B-3. If pressure readings are measured at locations other than the limiting beltline region, the pressure differences between the pressure transmitter and the limiting beltline region must be accounted for when using the pressure-temperature limit curves herein. Figures B-1 presents the heatup curves using heatup rates up to 60'F/hr applicable for the first 16 EFPY. Figure B-2 presents the cooldown curves using cooldown rates up to 100*F/hr applicable for the first 16 EFPY. Margins _of 12*F and 70 psig for possible-instrumentation errors are included. in the development of heatup and cooldown curves. B-7 l
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures B-1 and B-2. This is in addition to other criteria which must be met before the reactor is made critical. The leak limit curve shown in Figure B-1 represents minimum temperature requirements at the leak test pressure specified by applicable codes (B2,B3), The leak test limit curve was determined by methods of References B2 and B4. The criticality limit curve shown in Figure B-1 specify pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference B4. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissable temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section B-3. The minimum temperature for the inservice hydrostatic leak test for the McGuire Unit 2 reactor vessel at 16 EFPY is 251'F. A vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel. Figures B-1 and B-2 define limits for ensuring prevention of nonductile failure for the McGuire Unit 2 reactor vessel. The data points used to develop the heatup and cooldown pressure-temperature limit curves shown in Figures B-1 and B-2 are presented in Attachment Bl. Attachment B2 contains the data points for the 16 EFPY heatup and cooldown pressure-temperature limit curves generated without margins for instrumentation errors. O B-8 i
5. CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2[Bl] the adjusted reference temperature (ART) for each material in the beltline is given by the following expression: ART = Initial RTNDT + ARTNDT + Margin (3) Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section 111 of the ASME Boiler and Pressure Vessel Code. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class. ARTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows: ARTNDT = [CF]f(0.28-0.10 log f) (4) To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth. f(depth X) " fsurface(e ' 4X) (5) where x (in inches) is the depth into the vessel wall measured from the vessel clad / base metal interface. The resultant fluence is then put into equation (4) to calculate ARTN?T at the specific depth. CF (*F) is the chemistry factor, obtained from Tables in Reference B1, using the mean values of the copper and nickel content as reported in Table-B-1. All materials in the beltline region of McGuire Unit 2 were considered in determining the limiting material. The results of the ART's at 1/4T and 3/4T are summarized in Table B-2. From Table B-2, it can be seen that the limiting material is the lower shell forging 04, for heatup and cooldown curves-applicable up to 16 EFPY. Sample calculations to determine the ART values for-the lower shell-forging at 16 EFPY are shown in Table B-3. B-9
I l TABLE B-1 HCGUIRE UNIT 2 REACTOR VESSEL TOUGHNESS TABLE (Unirradiated) ~ CU N1 1-RINDT Material Description (%) (%) (*F) 1 (a) Closure Head Flange - 4 (a) Vessel Flange Intermediate Shell Forging 05* 0.153 0.793 - 1 (b) Lower Shell Forging 04* 0.15 0.88 -30 (b) Circumferential Weld 0.039 0.73 -6B (b) (a) Initial RTNDT values were estimated per U.S. NRC Standard Review Plan [B2). These values are used for considering flange requirements for the heatup/cooldown curves [B4), (b) The initial RTNDT values for the plates and welds are measured values. The mean values of copper and nickel content determined as indicated below: Copper Nickel Material Data Source (wt. %) (wt. %) forging 05 Original Hill Test Report 0.16 0.85 Surveillance Program [B5] 0.16 0.79 Capsule V Report (B6) 0.14 0.71 See Table 4-3, Section 4.0 p_dM _ 0.82_ Hean Value 0.153 0.793 Forging 04 Original Hill Test Report 0.15 0.88 Circ. Weld Original Hill Test Report 0.05 0.70 Surveillance Program [B5) 0.031 0.73 Capsule V Report (B6) 0.03 0.66 See Table 4-3, Section 4.0 0.039 0.765 See Table 4-3, Section 4.0 0.036 0.747 See Table 4-3, Section 4.0 0.045 0.776 ~ Mean Value 0.039 0.730 B-10 l
TABLE B-2
SUMMARY
OF ADJUSTED REFERENCE TEMPERATURES (ART's) IN *F AT 1/4T and 3/4T LOCATIONS FOR 16 EFPY 4 Material Description 1/4-T 3/4-T Intermediate Shell Forging 05 ~131 100 (86) (63) Lower Shell' Forging 04 104* 73* Circumferential Weld 33 19 (-14) (-22) ART numbars within ( ) are based on chemistry factors: calculated using surveillance capsule data. These ART numbers were used to generate heatup and cooldown curves. 4 u e B-11
TABLE B-3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES AT 16 EFPY FOR THE LIMITING MCGUIRE UNIT 2 REACTOR VESSEL MATERIAL - L')WER SHELL FORGlWG 04 i Reaulatory Guide 1.99 - Revision 2 16 EFPY Parameter 1/4 T 3/4 i L Chemistry Factor, Cr (*F) 115.8 115.8 n/cm)(a) .614 .222 I9 2 Fluence, f (10 Fluence Factor, ff .863 .595 ARTNDT CF x ff ('F) 100 69 Initial RTNDT, I (*F) -30 -30 Margin, H (*F) (b) 34 34 Revision 2 to Regulatory Guide 1.99 i Adjusted Reference Temperature, 104 73 ART = Initial RTNDT + ARTNDT + Margin l9 2 (a) Fluence, f, is based upon fsurf (10 n/cm, E>l Mov) = 1.02 at 16 EFPY. The McGuire Unit 2 reactor vessel wall thickness is 8.465 inches at the beltline region. 2 0.5 The standard deviation (b) Margin is calculated as, M = 2 ( og2+y 3 for the initial RTNDT margin term, og, is assumed to be.0'F since the initial RTHDT is a measured value. The standard deviation for ARTNOT term, oA, is 17'F for the plate, except that oA need not exceed 0.5 times the C, is 8.5'F for the plate (half the value) when mean value of ARTNDT-A surveillance data is used. B-12
i b MATERIAL PROPERTY BASIS i. LlHITING MATERIAL: LOWER SHELL FORGING 04 i LIMITING ART AT 16 EFPY: 1/4T, 104*F 3/4T, 73'F 0 7.13,g 3,_ i;;i j iiii s I r n i! Ii !I J j T
50 l l!
LEAX TEST LIMIT f l l i i,q 1 1 1 \\ l* I f f i i I J J 2000 i l l l r r 4 A I i 1750 UNACCEPTABLE l / 6 OPERATION-1 1 i m l l 1500 ) l I / t '2 ACCEPTA8LE 2 I ii OPERATION p E 1250 j' 32 / I a r r 5 ) 8 1000 / ^ Tc, r I te ] l 0: ~HEATUP RATES 8 750 UP TO I E --60'F/HR / 2 m e ._2:: CRITICALITY LIMIT $00 BASED ON INSERVICE HYDROSTATIC TEST 1 TEMPERATURE (251'F) -FOR THE SERVICE 250 PERIOD UP TO 16 EFPY l \\' I O 50 100 150 200 250 300 350 400 450 500 l Beltline Region Fluid Temperature (DEO. F) Figure B-1 McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heat up rates up to 60'F/hr)' Applicable for the First 16 EFPY-(With Margins of 12*F and 70 psig For Instrumentation Errors) B-13 i. l-
MATERIAL PROPERTY BASIS j LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART AT 16 EFPY: 1/4T, 104'F f 3/4T, 73*F F t4! !l l -l ll lll i il li j, ; i ^ I i i,i! I I'i
250
~i i i F +t-1 i i i I. I _i ! } e ii 2000 l l j l 4 llil 'i i 1 f i bEACCEPTA8LE:/
75c OPERATION I
= I f G' i i~n / ACCEPTABLE l00 6 OPERATION [ 8 i ? i j 1250 / "i! i o o \\ {1000 l / o 3
n a
u Jr i M i 750 o fiY .E Z COOLDOWN RATES a:"7 2 FIHR l (jf
C0 Oswq 20" w r
40 " /' r 250 ~l ~~~ i i i 0 O 50 100 150 200 250 300 350 400 450 500 l Beldine Region Fluid Temperature (DEG. F) Figure B-2 McGuire Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100*F/hr) Applicable for the First 16 EFPY (With Margins cf 12'F and 70 psig for Instrumentation Errors) B-14 1 e
B-6. REFERENCES [Bl] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,
1988,
[B2] " Fracture Toughness Requirements", Branch Technical Position HTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981. [B?) ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G. Protection Against Nonductile Failure", pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New (ork, 1986. t [B4] Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements", U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Vol. 48 No.104, May 27,1983. ~ ^ [B5] K. Koyama and J. A. Davidson, " Duke Power Company William B. McGuire Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9489, May 1979. [B6)
5. E. Yanichko, et al., " Analysis of Capsule V from the Duke Power Company Mcguire Unit 2 Reactor Vessel Radiation Surveillance Program', WCAP-11029, January 1986.
4 w B-15
.-a. r ATTACHMENT-B1 LAfA POINTS FOR HEATUP AND C00LDOWN CURVES (With Margir.s of 12*F and 70 psig for Instrumentation Errors) O 'l B-16
i s 2 9 /9 2 / 90 ) Y P F ) E T 0 N S 0 I 0 M. 0 2 ) E 1T T F. 6 K R. 6 7 1 7 G K 1 O 2 4
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09/29/92 DBP COOLDOWN CURVES REG. GUIDE 1.99,REV.2 WITH MARGIN THE FOLLOWING DAT A WERE PLOT TED F OR COOLDOWN PROFILE 2( 20 DEG-F / HR CCOLDOWN ) 16.000 EFP YEARS 1RRADIATION PERIOD = FLAW DEPTH = A0 WIN 7 l INDICATED INDICATED INDICATED INDICATED [NDICATEO INDICATED l TEMPE9ATURE FRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE l (DEG.r) (PSI) (DEG.F) (PSI) (DEG.F) (OSI) 1 87.000 489.40 9 127.000 GM 17 167.000 813.80 18 172.000 849.39 2 92 000 500.46 to 132.000 42" 29 3 97.000 512.40 11 137.000 646.74 19 177.000 887.70 4 102.000 525.10 12 142.000 669.79 20 182.000 928.80 5 107.000 538.93 13 147.000 G94.54 21 187.000 973.02 6 112.000 993-09 14 152.000 721.06 22 192.000 1020.54 fB 157.000 749.87 23 197.000 1071.67 j 7 117.000 $69-et 47 .0 16 162.000 780.58 9 8 122,000 "c0 % Finge retyiswc& 551 ps.'. CX2 1 N N 5 I
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1 3.

An analysis is performed that conservatively demonstrates, making' appropriate allowances for all encertainties, the existence of equivalent margins of safety for continued operation. s The issue of low upper shelf toughness has been a subject of active concern with the NRC for several years. ?' 1981, NUREG'0744 was; issued. This document-tuggested an analysis procedure for determining the. margins;of safety:present in reactor vessels. However, no definite criteria for acceptance margins were given. In 1982 a formal request was issued to the ASME Code Section XI committee to develop more specific criteria. In =1989f the ASME completed development of alternative criteria to defiae acceptable reactor vessel integrity when the Charpy upper shelf energy;dropsibelow'50 ft-lb. These criteria were transmitted to the NRC.for their evaluation'and incorporation into the regulations governing reactor 1 vessel integrity, Upper shelf energy values for each of the beltline region 1 materials 11n the-Mcguire Unit 2 reactor vessel were' calculated for 32 and 48 effective ~ full power-years (EFPY) of operation. The results show that all of the' beltline 1 - region materials will iemain above the_50-ft-lb screening; criteria at 32-and 48- - EFPY as indicated in Table C-1. m a C-2 t v ww u y m -- e n-s TABLE C-1 MCGUIRE UNIT 2 VPPER SHELF ENERGY VALUES (ft-lbs) Material Description Initial USE 32 EFPY 48 EFPY Lower Shell Fcrging 04 141 106 103 Inter. Shell Forging 05 94 71 69 Circumferential Weld 133 106 104 4 4' C-3 . _ _.}}

ML20128C299

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