(The Technology Development for Surveillance Test of ...
133
KA E R,/RR-,785/97 KR9800531 (The Technology Development for Surveillance Test of Reactor Vessel Materials) 1997 29-41
Transcript of (The Technology Development for Surveillance Test of ...
KAER,/RR-,785/97 K R 9 8 0 0 5 3 1
(The Technology Development for Surveillance Test
of Reactor Vessel Materials)
1997
2 9 - 4 1
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SUMMARY
I. Project Title
The Technology Development for Surveillance Test of Reactor Vessel
Materials
II. Objective and Importance of the Project
The degradation of material property due to the neutron exposure in
reactor vessel material of pressurized water reactor, that is, radiation
embrittlement phenomenon plays a great role in raising the possibility of
early shutdown of reactor before its design life. Since the 2-loop reactor
vessel such as Kori unit 1 was manufactured with the material including
weld metal in beltline region which is very sensitive to the radiation
embrittlement, this problem becomes a great concern in plant life
management. The evaluation of radiation embrittlement in reactor vessel
material is obtained through the combination of the measurement in
mechanical property change and the neutron fluence irradiated into reactor
vessel material, and the determination and prediction of best-estimated neutron
fluence irradiated into reactor vessel becomes very important in evaluating the
radiation embrittlement. Thus, since USNRC prepared a criteria, Reg. Guide
DG-1053, in order to determine fluence objectively by setting a new
calculation standard, accurate reactor life prediction and setting of safe
operation condition were tried through performing the evaluation of radiation
embrittlement for Kori unit 1 reliably where currently radiation embrittlement
matter is a great concern.
in
III. Scope and Contents of Project
Benchmark test was performed in accordance with the requirement of
USNRC Reg. Guide DG-1053 for Kori unit-1 in order to determine
best-estimated fast neutron fluence irradiated into reactor vessel. Since the
uncertainty of radiation analysis comes from the calculation error due to
neutron cross-section data, reactor core geometrical dimension, core structural
density, temperature and constituting materials, radiation source, mesh density,
angular expansion and convergence criteria, evaluation of calculational
uncertainty due to analytical method was performed in accordance with the
regulatory guide and the proof was performed for entire analysis by
comparing the measurement value obtained by neutron dosimetry located in
surveillance capsule.
Best-estimated neutron fluence in reactor vessel was calculated by bias
factor, neutron flux measurement value/calculational value, from reanalysis
result from previous 1st through 4th surveillance testing and finally fluence
prediction was performed for the end of reactor life and the entire period of
plant life extension. Pressurized thermal shock analysis was performed in
accordance with 10 CFR 50.61 using the result of neutron fluence analysis in
order to predict the life of reactor vessel material and the criteria of safe
operation for Kori unit 1 was reestablished.
IV. Results and Proposal for Applications
A criteria was prepared for neutron transport calculation for Kori unit
1, 2 loop plant, according to a new criterion, USNRC Reg. Guide DG-1053
for determining neutron fluence irradiated into reactor vessel material and
best-estimated calculational method for reactor vessel of fast neutron which
IV
adopted measurement and calculational uncertainty was established.
According to the new criterion, reactor vessel neutron fluence prediction was
performed upto the extended life period of Kori unit 1, and the result
showed that pressurized thermal shock problem for Kori unit 1 will not
occur during the life extension period of 40 EFPY as well as the plant life
target year, 32 EFPY by satisfying the criterion of less than 300° F. This
result can be used to establish the criteria for fast neutron transport
calculation of Korean 3 loop and standard nuclear power plant and to
perform the evaluation of radiation embrittlement of reactor vessel materials
and the life prediction more reliably and objectively.
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Table Title
2-1 Heat Treatment of the Kori Unit 1 Reactor Vessel Beltline RegionMaterials
2-2 Chemical Composition (wt%) of the Unirradiated Kori Unit 1 ReactorVessel Beltline Region Forging Materials
2-3 Weld Metal Data for the Kori Unit 1 reactor Vessel
2-4 Chemical Composition (wt%) of the Unirradiated Kori Unit 1 ReactorVessel Beltline Region Weld Materials
2-5 Copper Concentration for 177-FA Owners' Group Beltline Welds
2-6 Chemistry of B&W Owner's Group 177-FA Plant Reactor VesselBeltline Welds
2-7 Comparison of Chemical Compositions of Reactor Vessel BeltlineRegion Weld Using Linde 80 Flux Between Kori Unit 1 and OtherSimilar Plants
2-8 Nuclear Parameters Used in the Evaluation of Neutron Sensors
2-9 Monthly Thermal Generation During the First 15 Fuel Cycles of theKori Unit 1 Reactor
3-1 Calculated Fast Neutron Exposure Rates at the Surveillance CapsuleCenter
3-2 Calculated Fast Neutron Exposure Rates at the Pressure Vessel 0°Clad/Base Metal Interface
3-3 Relative Radial Distribution within the Pressure Vessel Wall( 0° )
3-4 Measured Sensor Activities and Reaction Rates Surveillance Capsule Vand T Saturated Activites and Derived Fast Neutron Flux
3-5 Measured Sensor Activities and Reaction Rates Surveillance Capsule SSaturated Activites and Derived Fast Neutron Flux
3-6 Measured Sensor Activities and Reaction Rates Surveillance Capsule RSaturated Activites and Derived Fast Neutron Flux
3-7 Summary of Neutron Dosimetry Results Surveillance Capsules V, T, Sand R
3-8 Comparison of Measured and FERRET Calculated Reaction Rates atthe Surveillance Capsule Center Surveillance Capsules V. T. S and R
IX
Table Title
3-9 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule V
3-10 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule T
3-11 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule S
3-12 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule R
3-13 Comparison of Calculated and Measured Neutron Exposure Levels for KoriUnit 1 Surveillance Capsules V, T, S and R
3-14 Neutron Exposure Projection at Vessel 0° Location on the Pressure VesselI.D based on the Cycle 13 thru 15
3-15 Neutron Exposure Projection at Vessel 0° Location on the Pressure VesselClad/Base Metal Interface
3-16 Neutron Exposure Values for the Kori Unit 1 Reactor Vessel
3-17 Neutron Exposure Projection at Capsule 23° Location per Cycle
3-18 Neutron Exposure Projection at Capsule 33° Location per Cycle
3-19 Updated Lead Factors for Kori Unit 1 Surveillance Capsules
3-20 Charpy V-notch Data for the Kori Unit 1 Intermediate Shell Forging124W375 VA1 (Tangential Direction)
3-21 Charpy V-notch Data for the Kori Unit 1 Intermediate Shell Forging124W375 VAKAxial Direction)
3-22 Charpy V-notch Data for the Kori Unit 1 Reactor Pressure Vessel CoreRegion Weld
3-23 Charpy V-notch Data for the Kori Unit 1 Reactor Pressure Vessel WeldHeat-Affected Zone Material
3-24 Effect of Irradiation to 0.440 x 1019n/cm2 (EM.OMeV) on the NotchToughness Properties of the Kori Unit 1 Reactor Vessel SurveillanceCapsule V Materials
3 - 2 5 Effect of Irradiation to 1.055 x lO 'Vcm 2 (EM.OMeV) on the NotchToughness Properties of the Kori Unit 1 Reactor Vessel SurveillanceCapsule T Materials
Table Title
3-26 Effect of Irradiation to 1.415 x lO'Vcnf' (EM.OMeV) on the NotchToughness Properties of the Kori Unit 1 Reactor Vessel SurveillanceCapsule S Materials
3-27 Effect of Irradiation to 2.922 x 1019n/cm2 (EM.OMeV) on the NotchToughness Properties of the Kori Unit 1 Reactor Vessel SurveillanceCapsule R Materials
3-28 Calculation of Average Cu and Ni Weight Percent Values for BeltlineMaterials
3-29 Interpolation of Chemistry Factors from Regulatory Guide 1.99, Revision2, Position 1.1
3-30 Calculation of Chemistry Factors Using Surveillance Capsule Data perRegulatory Guide 1.99, Revision 2, Position 2.1
3-31 Kori Unit 1 Surveillance Capsule Data Calculation of Best-Fit Line asDescribed in Position 2.1 of Reg.Guide 1.99, Rev.2
3-32 Kori Unit 1 Surveillance Capsule Data Evaluation of Credibility asDescribed in Position 2.1 of Reg.Guide 1.99, Rev.2
3-33 Calculation of RTPTS Values for the Kori Unit 1 Reactor Vessel BeltlineRegion Materials
3-34 Peak FluencedO19 n/cm2, EM.OMeV) on the Pressure Vessel Clad/BaseMetal Interface for Kori Unit 1
3-35 Margins for ART Calculation per Reg.Guide 1.99, Rev. 2
3-36 Calculation of ART Values for Kori Unit 1 Reactor Vessel BeltlineRegion Materials at 20EFPY
3-37 ART Values at the 1/4T & 3/4T Locations Used in the CurvesGeneration
XI
2-1 Plan View of Surveillance Capsules in Kori 1 Reactor Vessel
2-2 Weld Specification of Reactor Vessel Beltline Region
2-3 Surveillance Capsule Location at Reactor Vessel Beltline Region
2-4 R & Theta DORT Geometry for Kori Unit 1
2-5 Flow Chart of Neutron Transport Calculation per Reg.Guide DG-1053
2-6 R & Theta DORT Geometry for the Fuel Cycle 4 & 5
3-1 Charpy V-Notch Impact Energy vs. Temperature for Kori Unit 1 ReactorVessel Intermediate Shell Forging 124W375 VA1 (Tangential)
3-2 Charpy V-Notch Lateral Expansion vs. Temperature for Kori Unit 1Reactor Vessel Intermediate Shell Forging 124W375 VA1 (Tangential)
3 3 Charpy V-Notch Percent Shear vs. Temperature for Kori Unit 1 ReactorVessel Intermediate Shell Forging 124W375 VA1 (Tangential)
3-4 Charpy V-Notch Impact Energy vs. Temperature for Kori Unit 1 ReactorVessel Intermediate Shell Forging 124W375 VA1 (Axial)
3-5 Charpy V-Notch Lateral Expansion vs. Temperature for Kori Unit 1Reactor Vessel Intermediate Shell Forging 124W375 VA1 (Axial)
3-6 Charpy V-Notch Percent Shear vs. Temperature for Kori Unit 1 ReactorVessel Intermediate Shell Forging 124W375 VA1 (Axial)
3-7 Charpy V-Notch Impact Energy vs. Temperature for Kori Unit 1 ReactorVessel Weld Metal
3-8 Charpy V-Notch Lateral Expansion vs. Temperature for Kori Unit 1Reactor Vessel Weld Metal
3-9 Charpy V-Notch Percent Shear vs. Temperature for Kori Unit 1 ReactorVessel Weld Metal
3-10 Charpy V-Notch Impact Energy vs. Temperature for Kori Unit 1 ReactorVessel Heat-Affected-Zone (HAZ) Metal
3-11 Kori Unit 1 Reactor Coolant System Heatup Limitations (Heatup rate uplOOT/hr) Applicable for 20EFPY (with Instrument error Margins)
3-12 Kori Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rateup lOOT/hr)Applicable for 20EFPY (with Instrument error Margins)
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Id(r, 0 ,E) = The energy and spatially dependent adjoint function relating the fastneutron flux at geometric location d to meutron sources within theproblem geometry
S(r, 8 ,E) = The energy and spatially dependant neutron source within thepronlem geometry
Ar-g-*V adjoint importance ^Hrfe ^ ^ f f ^ ^ H r <Z>(E>1.0 MeV)
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10CFR50.61(1996.1.18)3 RTprs
4 4 RTFTS
1985. 7. 23«^ US
1991. 5.
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margin
margin^i
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parameter ©1^ RTNDT
beltline *j-3\4\ S.A ^ -g -^^ -^ t f l^ RTFTS ^^8:
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A R T m = [CF] * [ffj
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M = Margin^ Reg.Guide 1.99, Rev.2ofl £<\7) IRTNDT 5t. Cu 4 Ni
fluence ^ ^ K L ^ ^ H -^°fl 7l©l
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Table l(weld)3r Table 2(base metal)^.
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RG 1.99, Rev.2,
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C * Kta + Krr < KIR
Kin, = Stress intensity factor caused by membrane(pressure) stress.
Krr = Stress intensity factor caused by the thermal gradientsthrough the vessel wall.
KIR = Reference stress intensity factor as a function of the metaltemperature T and the metal ART.
C - 2.0 for Level A and Level B service limits for heatup andcooldown.
C =1.5 for hydrostatic and leak test conditions during which thereactor core is not critical.
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Edition, P497H v$& bo ^ b, ^t^r c^g- wj-
n bo + bi S Xi = 2 yi
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S 3-3H1
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241.13 = 4.167 a + 4.476 b
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860.58 = 4.167 a + 4.476 b
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ART &£r Reg.Guide 1.99,
Rev.2
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- 4 3 -
l. ^ ^
10CFR Part 50.6H 7])^ (January 18, 1996)S]jl
Code,
Draft Reg. Guide DG-1053 (April 16, 1997)S. ^^^^<=>1 ^ ^<^1 4 ^ 7l
WestinghouseAf^- ^ - ^ A S Reg.Guide DG-1O53<>11 nf^- Benchmarking^- ^
. fl ? f l^€ #Z}5, -§-71 ^ ^
l~4*} # A H * H tf l^ ^ sg
V, T, S ^
4.401E+18n/cm2(V : 1.13EFPY), 1.055E+19n/cm2 (T :
4.29EFPY), 1.415E+19n/cm2(S : 5.08EFPY) ^ 2.922E+19n/cm2 (R : 6.88EFPY)
«1 (bias factor)^ 0.895 M. ^^-^^.^ *%??: Uncertainty^ 12.8 %
USNRC Reg.Guide DG-1053 Si\ 7l^^l«=l ±20%
*& (best estimated neutron
(13.5EFPY : 1997<d 3Q)7]&*.g. 1.6488E+19n/cm2
S. ^7} S I A I ^ L ^ 24, 32 ^ 40EFPY oflA^ ^^>S.-8-7l
^r 4 4 2.880E+19n/cm2, 3.861E+19n/cm2 ^ 4.874E+19n/cm2
- 4 5 -
] \ # H D ( # ] g l ) ^ 15.51EFPY
: 1999V!) 4
40EFPY
- 4 6 -
2.
2-Loop ^^i<y JLB\ \S.7]^r SS&S. USNRC Regulatory Guide
DG-1053^1
•a-
7} S]^C}. olfe- Af-g-
PWR
^ 20%
15%
13171 ofl
- 4 7 -
1. P.M.Connell, L.Server, W.OldField and F.M.OldField, "Irradiated NuclearPressure Vessel Steel Data Base", EPRI NP-2428U982)
2. Code of Federal Regulations, 10CFR Part 50, Appendix G, "FractureToughness Requirements", Federal Register Vol. 60 No. 243, December 19,1995.
3. Code of Federal Regulations, 10CFR Part 50, Appendix H, "Reactor VesselMaterial Surveillance Program Requirements", Federal Register, Vol 60 No.243.
4. The first surveillance test on the reactor vessel materials of the KoriNuclear Power Plant Unit 1 (Capsule V), Aug. 1980.
5. The second surveillance test on the reactor vessel materials of the KoriNuclear Power Plant Unit 1 (Capsule T), Jan. 1985.
6. The third surveillance test on the reactor vessel materials of the KoriNuclear Power Plant Unit 1 (Capsule S). June. 1986.
7. The fourth surveillance test on the reactor vessel materials of the KoriNuclear Power Plant Unit 1 (Capsule R). May. 1990.
8. The safety (fracture) analysis of the Kori Unit 1 reactor pressure vessel,Sept. 1988.
9. ASME Boiler and Pressure Vessel Code Section XI, Appendix K,"Assessment of reactor vessels with low upper shelf charpy impactenergy levels", 1993.
10. USNRC Draft Regulatory Guide DG-1023, "Evaluation of reactor pressurevessels with charpy upper shelf energy less than 50 ft-lb", Sept. 1993.
11. Integrity assessment of Kori Unit 1 reactor pressure vessel for low uppershelf toughness, KAERI/CR-005/94,Sept. 1994.
12. USNRC 10CFR Part 50.61, "The pressurized thermal shock (PTS) rules",Jan. 1996.
13. USNRC Draft Regulatory Guide DG-1053, "Calculational and dosimetrymethods for determining pressure vessel neutron fluence", April 1997.
14. WCAP-8586, "KORI Unit 1 reactor vessel radiation surveillance program",Westinghouse class3. August 1975.
15. BAW-1799, "B&W 177-FA reactor vessel beltline weld chemistry study",B&W, 1983.
16. ^ H I J A1*.H all 70S " # * H W . 1 9 8 2 9.30.
- 4 8 -
17. 4*M#*la !Al ^ 92-20i "%*}S. «g-3-8-7l 7OVA1A]^ 7 l^" , 1992.12.
18. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests forLight-Water Cooled Nuclear Power Reactor Vessels", E706QF), in ASTMStandards, Section 3, ASTM, Philadelphia. PA. 1993
19. USNRC Standard Review Plan, NUREG-0800, Section 5.3.2,"Pressure-Temperature Limits", Rev.1,1981
20. 10CFR 50, "Analysis of Potential Pressurized Thermal Shock Events", July1985.
21. USNRC 10CFR Part50.61,"Fracture Toughness Requirements for ProtectionAgainst The Pressurized Thermal Shock Events" May 15, 1991.
22. USNRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement ofReactor Vessel Materials", U.S. Nuclear Regulatory Commission, May 1988.
23. USNRC Regulatory Guide 1.154, "Format and Content of Plant-Specific PTSSafety Analysis Report for PWRs". USNRC. 1987.
24. ASME Boiler and Pressure Vessel Code Section XI, Appendix G, "FractureToughness Criteria for Protection Against Failure", 1993.
25. USNRC Regulatory Guide 1.161, "Evaluation of Reactor Pressure Vesselwith Charpy Upper-Shelf Energy Less Than 50ft-lb", June 1995.
26. USNRC 10CFR Part50.66, "Requirements for Thermal Annealing of theReactor Pressure Vessel", December 1995.
27. USNRC Regulatory Guide 1.162, "Format and Content for Report forThermal Annealing of Reactor Pressure Vessels", February 1996.
28. USNRC Draft Regulatory Guide DG-1025,"Calculational and DosimetryMethods for Determunung Pressure Vessel Neutron Fluence", Sept. 1993.
29- ASTM E853-87, "Standard Practice for Analysis and Interpretation ofLight-Water Reactor Surveillance Results", in ASTM Standards, Section 12,American Society for Testing and Materials, Philadelphia, PA, 1993.
30. ASTM E693-79, "Standard Practice for Characterizing Neutron Exposures inFerritic Steels in Terms of Displacements per Atom(dpa)", in ASTMStandards, Section 12, American Society for Testing and Materials,Philadelphia, PA, 1993.
31. RSIC Computer Code Collection CCC-543, "TORT-DORT 2 and3-Dimensional Discrete Ordinates Transport, Version 2.7.3", May 1993.
32. RSIC Data Library Collection DLC-175, "BUGLE-93, Production and Testingof the VITAMIN-B6 Fine Group and the BUGLE-93 Broad GroupNeutron/Photon Cross-Section Libraries Derived from ENDF/B-VI NuclearData". April 1994.
- 4 9 -
33. R.E. Maerker, et al, "Accounting for Changing Source Distributions in LightWater Reactor Surveillance Dosimetry Analysis", Nuclear Science andEngineering, Volume 94, Pages 291-308, 1986.
34. Westinghouse Report, "The Nuclear Design and Core Physics Characteristicsof the Kori Unit 1 Nuclear Power Plant-Cycles 1 through 10" [WestinghouseProprietary Class 2]
35. KWU B324/90/e205, Nuclear Design Report fir KORI 1, Cycle 11, August1990.
36. KAERI NDR for Kori Unit 1, Cycles 12 through 14.
37. KAERI/TR-578/95,"Nuclear Design Report for KORI Unit 1, Cyclel5",December 1995.
38. ASTM Designation E482-89, "Standard Guide for Application of NeutronTransport Methods for Reactor Vessel Surveillance", in ASTM Standards,Section 12, American Society for Testing and Matrials, Philadelphia, PA,1993.
39. ASTM Designation E560-84, "Standard Recommended Practice forExtrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTMStandards, Section 12, American Society for Testing and Materials,Philadelphia, PA, 1993
40. ASTM Designation E706-87, "Standard Master Matrix for Light-WaterReactor Pressure Vessel Surveillance Standard", in ASTM Standards,Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
41. ASTM Designation E261-90, "Standard Practice for Determining NeutronFluence Rate, Fluence, and Spectra by Radioactivation Techniques", inASTM Standards, Section 12, American Society for Testing and Materials,Philadelphia, PA. 1993.
42. ASTM Designation E262-86, "Standard Method for Determining ThermalNeutron Reaction and Fluence Rates by Radioactivation Techniques", inASTM Standards, Section 12, American Society for Testing and Materials,Philadelphia, PA, 1993.
43. ASTM Designation E263-88, "Standard Method for Measuring Fast-NeutronReaction Rates by Radioactivation of Iron", in ASTM Standrad, Section 12,American Society for Testing and Naterials, Philadelphia, PA, 1993.
44. ASTM Designation E264-92, "Standard Method for Measuring Fast-NeutronReaction Rates by Radioactivation of Nickel", in ASTM Standards, Section12, American Society for Testing and Materials, Philadelphia, PA, 1993.
45. ASTM Designation E481-92, "Standard Method for MeasuringNeutron-Fluence Rate by Radioactivation of Cobalt and Silver", in ASTMStandards, Section 12, American Society for Testing and Materials,Philadelphia, PA, 1993.
- 5 0 -
46. ASTM Designation E523-87, "Standard Test Method for MeasuringFast-Neutron Reaction Rates by Radiqactivation of Copper", in ASTMStandards, Section 12, American Society for Testing and Materials,Philadelphia, PA, 1993.
47. ASTM Designation E704-90, "Standard Test Method for Measuring ReactionRates by Radioactivation of Uranium-238", in ASTM Standards, Section 12,American Society for Testing and Materials, Philadelphia, PA, 1993.
4& ASTM Designation E705-90, "Standard Test Method for Measuring ReactionRates by Radioactivation of Neptunium-237", in ASTM Standards, Section12, American Society for Testing and Materials, Philadelphia, PA, 1993.
49. ASTM Designation E1005-84, "Standard Test Method for Application andAnalysis of Radiometric Monitors for Reactor Vessel Surveillance", inASTM Standards, Section 12, American Society for Testing and Materials,Philadelphia, PA, 1993.
50. HEDL-TME79-40, "FERRET Data Analysis Core", F.A.Schmittroth, HanfordEngineering Development Laboratory, Richland, WA, September 1979.
51. AFWL-TR-7-41, Vol. I-IV, "A Computer-Automated Iterative Method ofNeutron Flux Spectra Detemined by Foil Activation", W.N.McElroy. S.Bergand T. Crocket, Air Force Weapons Laboratory, Kirkland AFB, NM, July1967.
52. RSIC Data Library Collection DLC-178, "SNLRML Recommended DosimetryCross-Section Compendium", July 1994.
53. EPRI-NP-2188, "Development and Demonstration of an AdvancedMethodology for LWR Dosimetry Applications", R.E. Maerker, et al., 1981
54. WC AP-14370, "Use of the Hyperbolic Tangent Function for FittingTransition Temperature Toughness Data", T.R.Mager, et al, May 1995.
55. WCAP-14040-NP-A, "Methodology Used to develop Cold OverpressureMitigating System Setpoints and RCS Heatup and Cooldown Curves",Revision 2. J.D.Addrachek, Januarv 1996.
- 5 1 -
Table 2-1 Heat Treatment of the Kori Unit 1 Reactor Vessel Beltline RegionMaterials
Material
Intermediate ShellForging,
Heat No. 124W375 VA1
Weld Metal l "
Temperature (°F)
Austenitizing1550
(843 TC)
Tempered1240
(6711C)
Stress Relief1100/1150
(593/621 -C)
Stress Relief1100/ 1150(593/62112)
Time (hours)
8
14
20-1/3
20-1/4
Coolant
Water-quenchpd
Air-cooled
Furnace-cooled
Furnace-cooled
Notes:
(a) Submerged-arc weldment made from sections of forging 124W375 VA1 and the adjoining lowershell course forging 122X371 VA1 using weld wire representative of that used in the orginalfabrication by Babcock and Wilcox Company.
- 5 2 -
Table 2-2 Chemical Composition (wt%) of the Unirradiated Kori Unit 1 ReactorVessel Beitiine Region Forging Materials
Element
C
Mn
P
S
Si
Ni
Mo
Cr
Cu
Al
Co
Pb
W
Ti
Zr
V
Sn
As
B
N
Nb
Inter.Shell Forging124W375 VA1|a|
0.22
0.63
0.010
0.010
0.26
0.73
0.57
0.32
0.07
N/A
0.010
N/A
<0.01
<0.01
N/A
0.01
0.005
N/A
N/A
0.002
0.01
Inter. Shell Forging124W375 VA11"1
.0.21
0.64
0.006
0.009
0.25
0.70
0.59
0.34
0.05
0.011
0.008
<0.001
<0.001
0.001
0.001
0.01
0.005
0.006
<0.003
0.003
0.01
Lower Shell Forging122X371 VA1W
0.20
0.66
0.007
0.010
0.23
0.76
0.60
0.37
0.04
0.011
<0.001
<0.01
0.001
0.001
0.01
0.008
0.007
<0.003
0.004
0.01
Notes :(a) Kori Unit 1 surveillance program Table A-2.(b) Chemical Analysis by Bethlehem Steel Corporation
- 5 3 -
Table 2-3 Weld Metal Data for the KORI Unit 1 Reactor Vessel
MT-SMAUT-362.
FromWIN
WRD-PEDMaterials Technology284-4438December 12, 1985Weld Metal Data for the KORI Unit 1 Reactor Vessel
E. Schoen/R4D 701-209*
cc: T. A. MeyerK. R. BalkeyG. E. Kubancsek/R&O 701-309
FILE: KOR-108/3* kw/Attachment
The following identification of weld material used in the KORI Unit 1surveillance weld and reactor vessel core region welds is provided perKEPCO's request (see Attachment 1).
Weld Location
Nozzle Shell to Inter. Shell
Inter. Shell to Lower Shell
Inter. Shell to Lower Shell
Surveillance Weldment
B&WWeld Qua!. No.
WF259
WF232
WF233
WF233
Weld WireHeat No.
T29744
8T3914
T29744
T29744
Type
Linde
Linde
Linde
Linde
Flux
80
80
80
80
Lot No
8806
8730
8790
B79O
Attachment 2 provides various test certificates which identify the woUland flux used in the various vessel welds, and includes the chemical coin;,and mechanical properties of the welds.
Please transmit this information to Mr. Lee of KEPCO.
S. EY Yanichko
Structural Materials and Reliability Technology
/sh
Attachment
-54-
Table 2-4 Chemical Composition (wt%) of the Unirradiated Kori Unit 1 ReactorVessel Beltline Region Weld Materials
Element
C
Mn
P
S
Si
Ni
Mo
Cr
Cu
Al
Co
Pb
W
Ti
Zr
V
Sn
As
Nb
N
B
Inter. & Lower ShellWelds(a>
0.10
1.52
0.012
0.15
0.37
0.61
0.48
0.08
0.23
0.011
0.01
0.01
0.01
0.010
0.006
0.01
0.005
Inter. & Lower ShellWelds(b)
0.053
1.60
0.015
0.016
0.44
0.55
0.47
0-22
Inter. & Lower ShellWelds(c)
0.055
1.45
0.011
0.007
0.51
0.69
0.30
0.13
Nozzle Shell &Inter.Shell Welds(d)
0.054
1.60
0.019
0.015
0.40
0.66
0.34
0.21
Notes :
(a) Kori Unit 1 surveillance program Table A-2.(Weld Qual. No.: WF233, Wire Heat No. : T29744, Flux Type & Lot No.
(b) Chemical Analysis by Babcock & Wilcox company.(Weld Qual. No.: WF233, Wire Heat No. : T29744, Flux Type & Lot No.
(c) Chemical Analysis by Babcock & Wilcox company.(Weld Qual. No.: WF232, Wire Heat No. : 8T3914, Flux Type & Lot No.
(d) Chemical Analysis by Babcock & Wilcox company.
(Weld Q'ial. No.: WF259, Wire Heat No. : T29744, Flux Tyne & Lot No.
Linde 80 & 8790)
Linde 80 & 8790)
Linde 80 & 8790)
...irde 80 & 8806)
- 5 5 -
Table 2-5 Copper Concentration for 177-FA Owners' Group Beltline Welds
Copper Concentration for 177-FA Owners' GroupBeltline Welds.
Cu concentration, wt X
Wireheat No.
299L44
72105
406L44
82 IT 44
61782
71249
72442
72445
8T1554
72102
T29 744
8T1762
Weld No.
WT 25SA 1526
WT 70
WF 112WT 154WF 193
WT 182-1WF 200
SA 1135SA 1788
SA 1229SA 1769
WT 67
SA 1585
WT 169-1SA 1174SA 1413SA 1494
WF 29
WT 233
WF 8WT 18SA 14 26SA 1430SA 1493
1P0962 SA 1073
(a)weld-metal auali
Category w
11
1
111
11
11
11
1
1
2222
2
2
33333
3
fication test
/m
0.0.
0.
0.0.0.
0.0.
0.0.
0.0.
0,
0.
0000
0
0
00000
0
rep
TR(-)
2946
27
222019
2126
1729
20.19
.27
.25
.11
.19
.19
.14
.16
.22
.20
.11
.18
.16
.22
.21
ort.
Mean
0.0.
0.
0.0.0.
0.0.
0.0.
0.0.
0.
0.
0.0.0.0
0
0
00000
0
3535
35
313131
2424
2525
2626
24
.21
.18
.18
.18
.18
.23
.29
.29
.29
.29
.29
.29
.29
Stddev' n
0.030.03
0.06
0.020.020.02
0.030.03
0.050.05
0.050.05
0.05
0.03
0.070.070.070.07
0.07
0.07
0.070.070.070.070.07
0.07
-56-
2
5
12
oC
CO
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• " •
a.5u
O
• c
0)
c
*-•
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a
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3
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o u
00IS
"5
Ia>
6CM
t -
o
c
CO
o
o>oo
o
o
co
^oo
o
oo
o
CO
i n
3
-3-
o>c->
c
c
cc
c
o>o
o
o
oo
oo
oo
CO
52
6
o
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o
o.
c
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ooo
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oo
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CO
o
72
10
5
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c
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oo
CO
oo
CO
CO
is
o
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c
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r-,
oo
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oo
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CO
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u.3
o
c
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r~oo
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a.
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o
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o>
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oo
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82-1
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?.o
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o
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ooo
o
oo
r—i
CO
oo
o
cc
c
c
CO
oo
OS
o
oo
_
co
cooc
CO
0 0
61
70
2
c
c
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c
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oc
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780
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229
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769
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u.3
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oo
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u.3
CO
=
o
o
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oo
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oo
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cc
oo
CTs
CO
174
•e
x
c
o
-
GO
oo
o
oo
•*>
oo
CN
oo
CO
<
o
c
o
oo
<r
o
oo
^oo
oo
av
CO
49
4
<-*
o
a
r^
o
oo
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oo
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••o
cooo
o
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u.3
72
10
2
o
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o
CO
oo
o
oo
_
oo
oo
o
CO
s
c
o
c
cs,
o
c
cc
ooo
„
oc
CO
u.
8T
17
62
o
o
o
<-<
o
cr-
o
oo
ooo
oo
o
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oo
=
•(
)
o
o
r-t
o
oo
r~
oo
r " i
Xoo
CO
<
o
=
o
o
(—1
-J
o
oo
r-
oo
•n
CO
oc
CO
or-i
<
c
(I.
o
t v ,
o
o
ooo
r~
o
-
CO
oo
CO
co
493
<
r j
o
o
o
r^
o
;
oo
^oo
^
oo
=
oCO
<
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co
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co
CO
c
c
CC
r>
o
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-o
- 5 7 -BobcockiWilcor.
Table 2-7 Comparison of Chemical Compositions of Reactor Vessel Beltline RegionWeld Using Linde 80 Flux Between KORI Unit 1 and Other Similar Plants
Element
C
Mn
P
S
Si
Cr
Mo
:: w
CU
KORI UNIT 1 i a )
0.05
1.45
0.021
0.015
0.42
0.08
0.44
0.68
0.29
HSST-63W B & W
0.098 0.09
1.65
0.016
0.011
0.63
0.095
0.427
CX685
0.299
1.49
0.016
0.016
0.51
0.06
0.39
0.59
0,28
CE/W-EP23
0.11
1.40
0.008
0.013
0.50
0.04
0.44
:L: :: '0.59.': ' :i
•I t •• 1 0 , 2 3 ;• :
CE/W-EP19
0.12
1.36
0.007
0.013
0.50
0.04
0.44
0.59
0,40
Notes :(a) Precisely Re-tested values by Babcock & Wilcox Co. (BAW-1799, 1983).
- 5 8 -
Table 2-8 Nuclear Parameters Used in the Evaluation of Neutron Sensors
Monitor Material
Copper
Iron
Nickel
Uranium-238*
Neptunium-237*
Cobalt-Aluminum*
Cobalt-Aluminum
Reaction ofInterest
Cue3 (n, a ) C060
Fe54 (n,p) MnM
Nb (n,p) Cosg
U238(n,f) CS)37
NP237 (n,f) CS137
C059 ( n , 7 ) Coeo
C059 (ri,r) Coeo
TargetWeightFraction
0.6917
0.0580
0.6827
1.0
1.0
0.0015
0.0015
Range
Response
E>4.7 MeV
E>1.0 MeV
E>1.0 MeV
E>0.4 MeV
E>0.008 MeV
0.4eV>E>0.015 MeV
E>0.015 MeV
Product
5.271 yrs
312.5 days
70.78 days
30.17 yrs
30.17 yrs
5.271 yrs
5.271 yrs
FissionYield
Half-Life(%)
6.00
6.27
Denotes that monitor is cadmium shielded
- 5 9 -
Table 2-9 Monthly Thermal Generation During the First 15 Fuel Cycles of theKori Unit 1 Reactor
Thermal GenerationYear
1977
1980
1983
Month123456789101112123456789101112123456789101112
L 1
l y o o
r 234
I 5I 6
78910
: 1112
(MW-hr)
35271.110320.0
122048.60
186258.0361458.4
4974.01058834.21137892.61126909.2272707.5
1226157.8916672.3
1150997.61212899.31190000.8872347.2
1186324.11056948.61237775.2213377.8
0530134.5164040.7
1203977.51180484.1888717.6
1202527.41084000.71255368.0745539.7407518.1
1238917.3119056.1
1214066.01285910.21265321.31124504.4380016.2
0478857.2
Thermal GenerationYear
1978
1981
1984
1987
Month12345678910111212345678910111212345678910
(MW-hr)374447.6444220.4640949.2806362.3892218.6635132.0974191.3775672.8883853.7725890.2
00
872783.9000
538733.31050285.41249151.31102588.21012502.5967539.1970363.1
1077518.31116219.9778693.6
1283988.61247285.51165928.81233770.9232374.3
00
902740.0,_11 l__ 1216954.8
1212
r3
U 4
U 5 "!~7p8_I 9
10
h 12 "
1268237.51145186.11128894.31169141.31116072.01247755.61236402.21156514.91272946.01238428.41241592.01237693.51280070.6
Thermal GenerationYear
1979
1982
1985
1988
Month1
234567891011121234567891011121234567891011121
^ 234567
00 CT
>
101112
(MW-hr)835749.3
1038785.7942989.2787189.9769207.4914777.7
1027014.1115634.0
1166085.1870205.0
00
1222142.21152268.21275233.6617164.7
020800.9
962748.61280923.51179493.61176956.61237616.11246676.2826600.5
1071010.41171247.61242319.81249809.71175831.01273554.0556187.1
01137.5
791188.3975962.4465068.4
00000
551940.51239875.51153793.41230105.01216735.31252748.2
- 6 0 -
Table 2-9 Monthly Thermal Generation During the First 15 Fuel Cycles of theKori Unit 1 Reactor (CONTINUED)
Thermal GenerationYear
1989
1992
1995
Month123456789101112123456789101112123456789101112
(MW-hr)1207411.41107057.2112835.2
000
617347.21096023.6
01025014.71234949.91220941.3
120.70
351728.11220580.91033491.61240359.81720700.31281724.01240499.51281835.91239182.71281863.51042689.9984401.1
1086963.21047700.21082142.51047589.81082659.61033636.41036890.31083347.31047533.01086313.4
Thermal GenerationYear
1990
1993
1996
Month123456789
(MW-hr)1259166.61149265.4641777.9
1235804.61069731.71237230.01268187.41277867.01209313.3
10 ! 590726.21112123456789101112123456789101112
00
1281749.71157518.11275981.21025308.4
00
1024446.71267465.31239656.71202941.01236647.41250781.9478690.1
014339.6
1172343.71280335.81236090.81273625.11279924.51238691.51277163.51235516.81278309.6
Thermal GenerationYear
1991
1994
1997
Month123456789101112123456789101112123456789101112
(MW-hr)544539.8
1135229.91274238.71225672.2980111.4
1238427.71268670.11267558.41186418.91269661.01037462.51266693.21246228.41150739.61271591.41236595.71277199.71184639.1562669.3
0617655.9
1281251.6287971.0
01276343.11153878.11023697.0
- 6 1 -
Table 3-1 Calculated Fast NeutronCapsule Center
Cycle
123456789101112131415
CRSD Data
123456789101112131415
CRSD Data
123456789101112131415
CRSD Data
13.0 °
1.529E+11
1.603E+111.743E+111.719E+11
1.493E+111.255E+111.309E+111.156E+111.438E+111.237E+11
1.157E+111.175E+111.299E+111.336E+111.296E+111.915E+11
6.374E+116.683E+117.267E+117.167E+116.224E+115.232E+115.457E+114.819E+115.995E+115.157E+114.824E+114.899E+115.416E+115.570E+115.403E+117.984E+11
Exposure Rates at the Surveillance
Capsule Location23.0 *
0 (E>1.0 MeV) n/cm'-sec8.809E+109.544E+10
9.837E+101.035E+11
8.472E+108.233E+108.021 E+107.970E+109.607E+107.661 E+10
8.538E+107.868E+108.210E+108.508E+108.101 E+101.127E+11
0 (E>0.1 MeV) n/cm*-sec3.340E+113.619E+113.730E+113.925E+113.213E+113.122E+113.042E+113.022E+113.643E+112.905E+113.238E+112.984E+113.113E+113.226E+113.072E+114.274E+11
33.0 °
8.470E+109.255E+10
8.831 E+101.013E+11
7.977E+107.974E+107.220E+107.443E+109.051 E+107.158E+10
8.669E+107.626E+107.474E+107.968E+107.376E+101.060E+11
3.301 E+113.607E+113.441 E+113.948E+113.109E+113.107E+112.814E+112.901 E+113.527E+112.789E+113.378E+11
2.972E+112.913E+113.105E+112.874E+114.131 E+11
Iron Displacement Rate, dpa/sec2.835E-102.972E-103.232E-103.187E-102.768E-102.327E-102.427E-102.143E-10
2.666E-1C2.293E-102.145E-102.178E-102408E-10
2.477E-102.403E-10
! 3.550E-10
1.557E-10
1.687E-101.739E-101.830E-101.498E-10
1.456E-101.418E-10
1.409E-101.699E-1 •:•1.354E-V.1.510E-10 I1.391E-101.452E-10
1.504E-10
1.432E-10 !1.993E-10 !
1.517E-10
1.658E-101.582E-101.814E-101.429E-101.428E-101.293E-10
1.333E-101.624E-1O1.282E-101.553E-101.366E-101.339E-10
1.427E-10
1.321E-101.898E-10
- 6 2 -
Table 3-2 Calulated Fast Neutron Exposure Rates at the Pressure Vessel 0*Clad/Base Metal Interface
Cycle
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
CRSD Data
0 (E>1.0 MeV)n/cm2-sec
4.814E+10
4.933E+10
5.439E+10
5.274E+10
4.785E+10
3.938E+10
4.227E+10
3.429E+10
4.503E+10
4.137E+10
3.531 E+10
3.553E+10
4.296E+10
4.336E+10
4.307E+10
5.932E+10
' "0 (E>0.1 MeV)
n/cm2-sec
1.304E+11
1.336E+11
1.473E+11
1.429E+11
1.296E+11
1.067E+11
1.145E+11
9.288E+10
1.220E+11
1.121E+11
9.564E+10
9.624E+10
1.164E+11
1.174E+11
1.167E+11
1.607E+11
Iron AtomDisplacement Rate,
dpa/sec
7.837E-11
8.031E-11
8.855E-11
S.586E-11
7.790E-11
6.411E-11
6.882E-11
5.582E-11
7.331 E-11
6.735E-11
5.748E-11
5.784E-11
6.994E-11
7.059E-11
7.012E-11
9.657E-11
- 6 3 -
Table 3-3 Relative Radial Distribution Within the Pressure Vessel Wall (0°)
Radius (cm)
167.961"
168.19
168.80
169.67
170.85
172.17""
173.45
174.90
176.3810'
177.5
178.95
180.5810'
181.55
182.52
183.98
184.791"
0 (E>1.0 MeV)
1.00
0.985
0.944
0.869
0.762
0.647
0.547
0.450
0.366
0.311
0.254
0.198
0.169
0.145
0.111
0.101
0 (E >0.1 MeV)
1.00
1.005
1.006
0.987
0.943
0.882
0.818
0.744
0.669
0.612
0.541
0.463
0.417
0.372
0.299
0.276
d pa/sec
1.00
0.989
0.955
0.895
0.811
0.719
0.637
0.553
0.477
0.424
0.365
0.302
0.268
0.237
0.189
0.174
NOTES:(a) Base Metal Inner Radius(b) Vessel 1/4 thickness(c) Vessel 1/2 thickness(d) Vessel 3/4 thickness(e) Base Metal Outer Radius
- 6 4 -
Table 3-4 Measured Sensor Activities and Reaction Rates Surveillance CapsuleV and T Saturated Activities and Derived Fast Neutron Flux
Monitor
CU-63(n,a) Co-60
Averages
Fe-54 (n,p) Mn-54
Averages
Ni-58 (n,p) Co-58
Averages
U-238 (n,f) Cs-137 (Cd)
Np-237 (n,f) Cs-137 (Cd)
Co-59 (n , r ) Co-60 (Cd)
Averages
Capsule V
Reaction Rate(rps/nucleus)
6.871 E-17
8.101 E-15
7.936E-15
3.529E-14
3.990E-13
5.022E-12
LocationFactor
1.13
0.95
1.15
0
0
0.98
AdjustedReaction Rate(rps/nucleus)
7.764E-17
7.696E-15
9.126E-15
3.529E-14
3.990E-13
4.922E-12
Capsule T
ReactionRate
(rps/nucleus)
5.168E-17
5.806E-15
6.320E-15
2.236E-14
2.057E-13
3.248E-12
LocationFactor
1.11
0.95
1.13
0
0
0.96
AdjustedReaction Rate(rps/nucleus)
5.736E-17
5.516E-15
7.142E-15
2.236E-14
2.057E-13
3.118E-12
- 6 5 -
Table 3-5 Measured Sensor Activities and Reaction Rates Surveillance Capsule SSaturated Activities and Derived Fast Neutron Flux
Monitor andAxial Location
CU-63(n, a) Co-60
86-4101 TOP
86-4102 MID
Averages
Fe-54 (n,p) Mn-54
86-4201 TOP
86-4202 TOP-MID
86-4203 MID
86-4204 MID-BOT
86-4205 BOTAverages
Ni-58 (n,p) Co-58
86-4301 MID
Averages
U-238 (n,f) Cs-137 (Cd)
86-4601 MID
Np-237 (n,f) Cs-137 (Cd)86-4701 MID
Co-59 (n, r ) Co-60
86-4401 TOP
86-4402 BOT
Averages
Co-59 (n,r) Co-60 (Cd)86-4501 TOP
Measured Activity(dis/sec-gm)
1.119E+05
1.133E+05
1.126E+05
1.524E+06
1.467E+06
1.508E+06
1.520E+06
1.552E+061.514E+06
4.901 E+06
4.901 E+06
3.444E+05
1.726E+06
2.579E+07
2.688E+07
2.634E+07
1.062E+07
Averages 1.062E+07
Saturated Activity(dis/sec-gm)
2.825E+05
2.861 E+05
2.843E+05
3.390E+06
3.263E+06
3.355E+06
3.381 E+06
3.452E+063.368E+06
4.991 E+07
4.991 E+07
3.290E+06
1.649E+07
6.512E+07
6.787E+07
6.650E+07
2.682E+07
2.682E+07
Reaction Rate(rps/nucleus)
4.814E-17
5.116E-15
7.982E-15
2.168E-14
1.035E-13
4.208E-12
1.697E-12
1.697E-12
- 6 6 -
Table 3-6 Measured Sensor Activities and Reaction Rates Surveillance Capsule RSaturated Activities and Derived Fast Neutron Flux
Monitor andAxial Location
CU-63(n, a) Co-60
88-4101 TOP
88-4102 MID
Averages
Fe-54 (n,p) Mn-54
88-4201 TOP
88-4202 TOP-MID
88-4203 MID
88-4204 MID-BOT
88-4205 BOT
Averages
Ni-58 (n,p) Co-58
88-4301 MID
Averages
Co-59 (n, y) Co-60
88-4401 TOP
88-4402 BOT
Averages
Measured Activity(dis/sec-gm)
2.317E+05
1.987E+05
2.152E+05
3.130E+06
2.618E+06
3.079E+06
3.035E+06
3.112E+06
2.995E+06
3.387E+07
3.387E+07
4.842E+07
5.748E+07
5.295E+07
Saturated Activity(dis/sec-gm)
4.779E+05
4.098E+05
Reaction Rate(rps/nucleus)
4.439E+05 ' 7.652E-17
5.209E+06
4.357E+06
5.124E+06
5.051 E+06
5.179E+06
4.984E+06
7.592E+07
7.592E+07
9.987E+07
1.186E+08
1.092E+08
7.571 E-15
1.247E-14
6.983E-12
- 6 7 -
Table 3-7 Summary ofV, T, S, and
Reaction
Neutron Dosimetry ResultsR
Flux
Calculation of Measured Fluence for Capsule V
Meas Fluence < 0.414 eV
Meas Fluence > 0.1 MeV
Meas Fluence > 1.0 MeV
dpa
1.241E+11
5.503E+11
1.238E+11
2.396E-10
Calculation of Measured Fluence for Capsule T
Meas Fluence < 0.414 eV
Meas Fluence > 0.1 MeV
Meas Fluence > 1.0 MeV
dpa
6.992E+10
3.019E+11
7.795E+10
1.410E-10
Calculation of Measured Fluence for Capsule S
Meas Fluence < 0.414 eV
Meas Fluence > 0.1 MeV
Meas Fluence > 1.0 MeV
dpa
5.500E+10
3.444E+11
8.839E+10
1.577E-10
Calculation of Measured Fluence for Capsule R
Meas Fluence < 0.414 eV
Meas Fluence > 0.1 MeV
Meas Fluence > 1.0 MeV
dpa
1.382E+11
5.735E+11
1.346E+11
2.530E-10
Time
3.555E+07
3.555E+07
3.555E+07
3.555E+07
1.353E+08
1.353E+08
1.353E+08
1.353E+08
1.601E+08
1.601E+08
1.601 E+08
1.601 E+08
2.171 E+08
2.171 E+08
2.171 E+08
2.171 E+08
Surveillance Capsules
Flunce
4.412E+18
1.956E+19
4.401 E+18
8.518E-03
9.460E+18
4.085E+19
1.055E+19
1.908E-02
8.804E+18
5.513E+19
1.415E+19
2.524E-02
3.000E+19
1.245E+20
2.922E+19
5.492E-02
Uncertainty
±80%
±16%
±8%
±11%
±80%
±24%
±13%
±17%
±54%
±23%
±14%
±18%
±50%
±25%
±16%
±19%
- 6 8 -
Table 3-8 Comparison of Measured and FERRET Calculated Reaction Rates atthe Surveillance Capsule Center Surveillance Capsules V, T, S and R
Reaction
Surveillance Capsule V
FE-54 (n,p) Mn-54
Ni-58 (n,p) Co-58
U-238 (n,f) Cs-137
Np-237 (n,f) Cs-137
Surveillance Capsule T
Cu-63 (n, a) Co-60
Fe-54 (n,p) Mn-54
Ni-58 (n,p) Co-58
Surveillance Capsule S
Fe-54 (n,p) Mn-54
Ni-58 (n,p) Co-58
Co-59(n, 7 )Co-60
Co-59 (n,r) Co-60 (Cd)
Surveillance Capsule R
Fe-54 (n,p) Mn-54
Co-59(n, 7)Co-60
Measured
7.70E-15
9.13E-15
3.53E-14
3.99E-13
5.74E-17
5.52E-15
7.14E-15
5.12E-15
7.98E-15
4.21 E-12
1.70E-12
7.57E-15
6.98E-12
AdjustedCalculation
7.36E-15
9.51E-15
3.79E-14
3.84E-13
5.64E-17
5.49E-15
7.30E-15
5.25E-15
7.77E-15
2.96E-12
2.96E-12
7.59E-15
6.98E-12
C/M
0.96
1.04
1.07
0.96
0.98
1.00
1.02
1.03
0.97
0.70
1.74
1.00
1.00
- 6 9 -
Table 3-9 Adjusted Neutron Energy Spectrum at the Center of SurveillanceCapsule V
Group
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
Energy(MeV)
1.733E+01
1.492E+01
1.350E+01
1.162E+01
1.000E+01
8.607E+01
7.408E+01
6.065E+01
4.966E+01
3.679E+01
2.865E+01
2.231 E+01
1.738E+01
1.353E+01
1.108E+01
8.208E-01
6.393E-01
4.979E-01
3.877E-01
3.020E-01
1.832E-01
1.111E-01
6.738E-02
4.087E-02
2.554E-02
1.989E-02
1.503E-02
28 9.119E-03
Adjusted Flux(n/cm'-sec)
7.209E+06
1.560E+07
5.911E+07
1.653E+08
3.811E+08
6.836E+08
1.679E+09
2.685E+09
5.920E+09
7.191E+09
1.380E+10
1.849E+10
2.555E+10
2.947E+10
5.180E+10
5.996E+10
6.546E+11
4.461 E+10
6.761E+10
7.213E+10
7.245E+10
5.306E+10
4.182E+10
2.360E+10
2.664E+10
1.379E+10
2.310E+10
2.645E+10
Group
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
49
50
51
52
53
Energy(MeV)
5.531 E-03
3.355E-03
2.839E-03
2.404E-03
2.035E-03
1.234E-03
7.485E-04
4.54E-04
2.75E-04
1.670E-04
1.013E-04
6.144E-05
3.727E-05
2.260E-05
1.371E-05
8.31 E-06
5.043E-06
3.059E-06
1.855E-06
1.125E-06
6.826E-07
4.140E-07
2.511E-07
1.523E-07
9.237E-08
i
Adjusted Flux(n/cm2-sec)
3.074E+10
9.715E+09
9.379E+09
9.197E+09
2.742E+10
2.738E+10
2.548E+10
2.286E+10
2.520E+10
2.665E+10
2.635E+10
2.608E+10
2.534E+10
2.443E+10
2.354E+10
2.255E+10
2.159E+10
2.115E+10
2.081 E+10
1.529E+10
1.557E+10
1.570E+10
1.383E+10
1.211E+10
5.251E+10
- 7 0 -
Table 3-10 Adjusted Neutron Energy Spectrum at the Center of SurveillanceCapsule T
Group
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
28
Energy(MeV)
Adjusted Flux(n/cm2-sec)
1.733E+01 7.347E+06
1.492E+01
1.350E+01
1.162E+01
1.000E+01
8.607E+01
7.408E+01
6.065E+01
4.966E+01
3.679E+01
2.865E+01
2.231 E+01
1.738E+Q1
1.353E+01
1.108E+01
8.208E-01
6.393E-01
4.979E-01
3.877E-01
3.020E-01
1.832E-01
1.111E-01
6.738E-02
4.087E-02
2.554E-02
1.989E-02
1.503E-02
1.589E+07
5.896E+07
1.613E+08
3.639E+08
6.262E+08
1.487E+09
2.210E+09
4.404E+09
4.928E+09
9.182E+09
1.188E+10
1.572E+10
1.708E+10
2.876E+10
3.210E+10
3.427E+10
2.341E+10
3.485E+10
3.766E+10
3.693E+10
2.748E+10
2.172E+10
1.234E+10
1.385E+10
7.357E+09
1.247E+10
9.119E-03 | 1.423E+10
Group
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
49
50
51
52
53
Energy(MeV)
5.531 E-03
3.355E-03
2.839E-03
2.404E-03
2.035E-03
1.234E-03
7.485E-04
4.540E-04
2.754E-04
1.670E-04
1.013E-04
6.144E-05
3.727E-05
2.260E-05
1.371E-05
8.31 E-06
5.043E-06
3.059E-06
1.855E-06
1.125E-06
6.826E-07
4.140E-07
2.511E-07
1.523E-07
9.237E-08
Adjusted Flux(n/cm2-sec)
1.666E+10
5.266E+09
5.078E+09
4.976E+09
1.483E+10
1.474E+10
1.369E+10
1.231 E+10
1.342E+10
1.431E+10
1.405E+10
1.376E+10
1.345E+10
1.304E+10
1.256E+10
1.203E+10
1.157E+10
1.136E+10
1.121E+10
8.377E+09
8.435E+09
1.425E+10
1.331E+10
1.245E+10
2.991E+10
- 7 1 -
Table 3-11 Adjusted Neutron Energy Spectrum at the Center of SurveillanceCapsule S
Group
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
Energy(MeV)
1.733E+01
1.492E+01
1.350E+01
1.162E+01
1.000E+01
8.607E+01
7.408E+01
6.065E+01
4.966E+01
3.679E+01
. 2.865E+01
2.231 E+01
1.738E+01
1.353E+01
1.108E+01
8.208E-01
6.393E-01
4.979E-01
3.877E-01
3.020E-01
1.832E-01
1.111E-01
6.738E-02
4.087E-02
2.554E-02
1.989E-02
1.503E-02
28 I 9.119E-03
Adjusted Flux(n/cmJ-sec)
5.493E+06
1.188E+07
4.443E+07
1.237E+08
2.859E+08
5.117E+08
1.272E+09
2.018E+09
4.344E+09
5.182E+09
1.017E+10
1.357E+10
1.854E+10
2.050E+10
3.459E+10
3.851 E+10
4.058E+10
2.711 E+10
3.943E+11
4.155E+10
3.989E+11
2.914E+10
2.261E+10
1.261E+10
1.407E+10
7.362E+10
1.218E+10
1.400E+10
Group
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
49
50
51
52
53
Energy(MeV)
5.531 E-03
3.355E-03
2.839E-03
2.404E-03
2.035E-03
1.234E-03
7.485E-04
4.54E-04
2.75E-04
1.670E-04
1.013E-04
6.144E-05
3.727E-05
2.260E-05
1.371 E-05
8.31 E-06
5.043E-06
3.059E-06
1.855E-06
1.125E-06
6.826E-07
4.140E-07
2.511E-07
1.523E-07
9.237E-08
Adjusted Flux(n/cm2-sec)
1.617E+10
5.100E+09
4.906E+09
4.785E+09
1.415E+10
1.394E+10
1.279E+10
1.133E+10
1.230E+10
1.242E+10
1.281E+10
1.267E+10
1.245E+10
1.215E+10
1.179E+10
1.136E+10
1.093E+10
1.075E+10
1.059E+10
7.797E+09
7.557E+09
1.206E+10
1.086E+10
9.897E+09
2.218E+10
- 7 2 -
Table 3-12 Adjusted Neutron Energy Spectrum at the Center of SurveillanceCapsule R
Group
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
28
Energy(MeV)
1.733E+01
1.492E+01
1.350E+01
1.162E+01
1.000E+01
8.607E+01
7.408E+01
6.065E+01
4.966E+01
3.679E+01
2.865E+01
2.231 E+01
1.738E+0.1
1.353E+01
1.108E+01
8.208E-01
6.393E-01
4.979E-01
3.877E-01
3.020E-01
1.832E-01
1.111E-01
6.738E-02
4.087E-02
2.554E-02
1.989E-02
1.503E-02
9.119E-03
Adjusted Flux(n/cm2-sec)
7.064E+06
1.530E+07
5.817E+07
1.637E+08
3.809E+08
6.916E+08
1.720E+09
2.785E+09
6.225E+09
7.690E+09
1.508E+10
2.044E+10
2.826E+10
3.204E+10
5.558E+10
6.353E+10
6.849E+10
4.612E+10
6.898E+10
7.274E+10
7.139E+10
5.266E+10
4.135E+10
2.330E+10
2.631 E+10
1.365E+10
2.292E+10
2.631 E+10
Group
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
49
50
51
52
53
Energy(MeV)
5.531 E-03
3.355E-03
2.839E-03
2.404E-03
2.035E-03
1.234E-03
7.485E-04
4.54E-04
2.75E-04
1.670E-04
1.013E-04
6.144E-05
3.727E-05
2.260E-05
1.371E-05
8.31E-06
5.043E-06
3.059E-06
1.855E-06
1.125E-06
6.826E-07
4.140E-07
2.511E-07
1.523E-07
9.237E-08
Adjusted Flux(n/cm2-sec)
3.068E+10
9.734E+10
9.433E+10
9.289E+10
2.782E+10
2.793E+10
2.613E+10
2.353E+10
2.600E+10
2.816E+10
2.724E+10
2.676E+10
2.598E+10
2.502E+10
2.406E+10
2.303E+10
2.204E+10
2.161E+10
2.126E+10
1.563E+10
1.635E+10
2.755E+10
2.600E+10
2.443E+10
6.024E+10
i
- 7 3 -
Table 3-13 Comparison of Calculated and Measured Neutron Exposure Levels forKori Unit 1 Surveillance Capsules V, T, S and R
Calculated Measured M/C
Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule V
Fluence (E> 1.0 MeV), n/cm*
Fluence (E> 0.1 MeV), n/cm*
dpa
5.432E+18
2.265E+19
1.007E-02
4.401 E+18
1.956E+19
8.518E-02
0.810
0.864
0.846
Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule T
Fluence (E> 1.0 MeV), n/cm*
Fluence (E> 0.1 MeV), n/cm*
dpa
1.261E+19
4.782E+19
2.230E-02
1.055E+19
4.085E+19
1.908E-02
0.837
0.854
0.856
Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule S
Fluence (E> 1.0 MeV), n/cm*
Fluence (E> 0.1 MeV), n/cm*
dpa
1.393E+19
5.428E+19
2.494E-02
1.415E+19
5.513E+19
2.524E-02
1.016
1.016
1.012
Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule R
Fluence (E> 1.0 MeV), n/cm*
Fluence (E> 0.1 Mev), n/cm*
dpa
3.180E+19
1.326E+20
5.898E-02
2.922E+19
1.245E+20
5.492E-02
0.919
0.939
0.931
- 7 4 -
Table 3-14 Neutron Exposure Projection at Vessel GVesse
* Location on the PressureI I.D based on the Cycle 13 thru 15
Best Estimate Exposure at the Pressure
Fulecycle
12345678910111213141516171819202122232425*
2627282930313233*#3435363738394041
42
EFPY
1.131.882.693.414.295.085.866.887.558.499.4010.4711.4712.4813.5014.5015.5116.5217.5318.5319.5420.5421.5522.5623.5724.0024.5725.5826.5927.6028.6029.6130.6231.6332.0032.6333.6434.6535.6636.6637.6738.6839.6940.0040.70
Cumulativeirradiation time
3.55500E+075.93700E+078.49300E+071.07490E+081.35300E+081.60080E+081.84880E+082.17080E+082.38110E+082.67770E+082.96380E+083.30180E+083.61840E+083.93490E+084.2551 OE+084.57280E+084.89050E+085.20820E+085.52590E+085.84360E+086.16130E+086.47900E+086.79670E+087.11440E+087.4321 OE+08
7.74980E+088.06750E+088.38520E+088.70290E+089.02060E+089.33830E+089.65600E+089.97370E+08
1.02914E+091.06091E+091.09268E+091.12445E+091.15622E+091.18799E+091.21976E+091.25153E+09
1.28330E+09
Cumulative timeinteg. response
1.71150E+182.88662E+184.27675E+185.46648E+186.79707E+187.77302E+188.82131E+189.92546E+181.08725E+191.20996E+191.31099E+191.43108E+191.56710E+191.70432E+191.84233E+191.97925E+192.11627E+192.25329E+192.39031 E+192.52733E+192.66435E+192.80137E+192.93839E+193.07541 E+193.21243E+19
3.34945E+193.48647E+193.62349E+193.76051 E+193.89753E+194.03455E+194.17157E+194.30859E+19
4.44561E+194.58263E+194.71965E+194.85667E+194.99369E+195.13071E+195.26773E+195.40475E+19
5.54177E+19
Vessel Inner Radius
Best estimated fluence
I.D1.E318E+182.5835E+183.8277E+184.8925E+186.0833E+186.9568E+187.8951 E+188.8833E+189.7309E+181.0829E+191.1733E+191.2808E+191.4026E+191.5254E+191.6488E+191.7714E+191.8941 E+192.0167E+192.1393E+192.2620E+192.3846E+192.5072E+192.6299E+192.7525E+192.8751 E+192.8804E+192.9978E+193.1204E+193.2430E+193.3657E+193.4883E+193.6109E+193.7336E+193.8562E+193.8607E+193.9788E+194.1014E+194.2241E+194.3467E+194.4694E+194.5920E+194.7146E+194.8373E+194.8741E+194.9599E+19
1/4T1.0370E+181.7490E+182.5913E+183.3122E+184.1184E+184.7098E+185.3450E+186.0140E+186.5878E+187.3312E+187.9432E+188.671 OE+189.4956E+181.0327E+191.1162E+191.1992E+191.2823E+191.3653E+191.4483E+191.5314E+191.6144E+191.6974E+191.7804E+191.8634E+191.9464E+191.9500E+192.0295E+192.1125E+192.1955E+192.2786E+192.3616E+192.4446E+192.5276E+192.6106E+192.6136E+192.6936E+192.7766E+192.8597E+192.9427E+193.0258E+193.1088E+193.1918E+193.2749E+193.2998E+193.3579E+19
- 7 5 -
Table 3-15 Neutron Exposure Projection at Vessel 0* Location on the PressureVessel Clad/Base Metal Interface
Best Estimate Exposure at the Pressure Vessel Inner Radius
Total Irradiation Time
EFPY
13.5
20.0
24.0
32.0
40.0
Bias:
Vessel Inner RadiusBest Estimate Neurton Fluence
E > 1.0 MeV
1.649E+19
2.390E+19
2.880E+19
3.861 E+19
4.874E+19
0.895
E > 0.1 MeV
4.581E+19
6.640E+19
8.001 E+19
1.073E+20
1.354E+20
0.918
DPA
2.732E-2
3.960E-2
4.771 E-2
6.397E-2
8.075E-2
0.911
- 7 6 -
Table 3-16 Neutron Exposure Values for the Kori Unit 1 Reactor vessel
Location 20 EFPY 24 EFPY 32 EFPY
Fluence Based on E > 1.0 MeV Slope
Surface
1/4T
3/4T
2.390E+19
1.546E+19
4.732E+18
2.880E+19
1.863E+19
5.702E+18
3.861E+19
2.498E+19
7.645E+18
Fluence Based on dpa Slope
Surface
1/4T
3/4T
2.390E+19
1.718E+19
7.218E+18
2.880E+19
2.071 E+19
8.69SE+18
3.861 E+19
2.776E+19
1.166E+19
- 7 7 -
Table 3-17
Fule cycle
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
' Neutron Exposure Projection t
Best Estimate Exposure at the I
EFPY
2.02
3.37
5.19
7.11
8.95
10.72
12.18
14.07
15.56
17.23
19.02
20.93
22.84
24.82
26.73
28.66
30.60
32.52
Cumulativeirradiation time
3.55500E+07
5.93700E+07
8.49300E+07
1.07490E+08
1.35300E+08
1.60080E+08
1.84880E+08
2.17080E+08
2.38110E+08
2.67770E+08
2.96380E+08
3.30180E+08
3.61840E+08
3.93490E+08
4.2551 OE+08
4.57280E+08
4.89050E+08
5.20820E+08
at Capsule 23* Location per cycle
'ressure Vessel Inner Radius
Cumulative timeintegrated response
3.13175E+18
5.40519E+18
7.91956E+18
1.02550E+19
1.26110E+19
1.46512E+19
1.66405E+19
1.92069E+19
2.12272E+19
2.34995E+19
2.59421 E+19
2.85405E+19
3.11398E+19
3.38324E+19
3.64262E+19
3.90545E+19
4.16829E+19
4.43112E+19
Best estimatedfluence
2.80292E+18
4.83765E+18
7.08801 E+18
0.91782E+19
1.12868E+19
1.31128E+19
1.48932E+19
1.71902E+19
1.89983E+19
2.10321 E+19
2.32182E+19
2.55437E+19
2.78701 E+19
3.02800E+19
3.26014E+19
3.49538E+19
3.73062E+19
3.96585E+19
- 7 8 -
Table 3-18 Neutron
Fule cycle
1
2
3
4
5
6
7
8
g
10
11
12
13
14
15
16
17
18
19
20
21
22
23
Exposure Projection (
Best Estimate Exposure at the
EFPY
1.95
3.26
4.84
6.73
8.40
10.15
11.52
13.29
14.68
16.25
18.07
19.96
21.70
23.56
25.29
27.07
28.84
30.63
32.40
34.18
35.96
37.73
39.51
Cumulativeirradiation time
3.55500E+07
5.93700E+07
8.49300E+07
1.07490E+08
1.35300E+08
1.60080E+08
1.84880E+08
2.17080E+08
2.38110E+08
2.67770E+08
2.96380E+08
3.30180E+08
3.61840E+08
3.93490E+08
4.2551 OE+08
4.57280E+08
4.89050E+08
5.20820E+08
5.52590E+08
5.84360E+08
6.16130E+08
6.47900E+08
6.79670E+08
Dapsule 33* Location per cycle
'ressure Vessel Inner Radius
Cumulative timeintegrated response
3.01110E+18
5.21568E+18
7.47290E+18
9.75763E+18
1.19760E+19
1.39519E+19
1.57423E+19
1.81390E+19
2.00425E+19
2.21657E+19
2.46458E+19
2.72235E+19
2.95897E+19
3.21116E+19
3.44733E+19
3.68900E+19
3.93062E+19
4.17256E+19
4.41390E+19
4.65554E+19
4.89718E+19
5.13883E+19
5.38047E+19
Best estimatedtluence
2.69493E+18
4.66803E+18
6.68825E+18
8.73308E+18
1.07185E+19
1.24870E+19
1.40894E+19
1.62344E+19
1.79380E+19
1.98383E+19
2.20580E+19
2.43650E+19
2.64828E+19
2.87399E+19
3.08536E+19
3.30166E+19
3.51790E+19
3.73444E+19
3.95044E+19
4.16671E+19
4.38298E+19
4.59925E+19
4.81552E+19
- 7 9 -
Table 3-19 Updated Lead Factors for Kori Unit 1 Surveillance Capsules
Capsule
V ( 13* )
T ( 2 3 - )
S ( 33' )
R* ( 13* )
P ( 23" )
N ( 33- )
Lead Factor
2.87
1.73
2.03
3.29
1.97
1.83
Withdrawn
EOC 1
EOC 5
EOC 6
EOC 8
* Basis for this analysis
- 8 0 -
Table 3-20 Charpy V-notch Data for the Kori Unit 1 Intermediate Shell Forging124W375 VA1 (Tangential Direction)
Capsule
Unirrad.
V
T
Fluence(xiO'fycm2)
0.0
0.4401
1.0547
SpecimenI.D1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
KL-9
KL-10
KL-11
KL-12
KL-7
KL-8
KL-5
KL-6
KL-1
KL-2
KL-3
KL-4
KL-44
KL-45
KL-42
KL-43
KL-46
KL-47
KL-41
KL-48
KL-40
KL-37
KL-39
KL-38
Temperature( *F)
-100-100-100-25
-25
-25
10
10
10
60
60
60
100
100
100
210
210
210
-40
-40
-20
-20
0
0
40
40
76
100
210
210
-80
-80
-40
-40
-20
-20
0
0
40
75
165
210
Energy(ft-lbl
5
6
5
21
46
86
125
107.5
22
170
169
175
166
169
164
178
167
181
11.5
14.0
81.0
74.0
46.0
88.5
106.5
101.0
151.5
153.0
160.0
158.0
6.0
4.5
54.5
15.5
42.5
91.0
103.0
Lateral Exp.(mils)
1
1
1
13
31
59
82
72
17
85
85
86
87
88
85.5
93
87
84
10
12
63
61
37
67
54
77
88
97
98
95
6
2
42
13
35
73
79
84.0 65
124.0 80
159.0 93
155.0 96
142.5 94
Shear(%)
3
3
3
14
20
35
65
35
30
100
100
100
100
100
100
100
100
100
0
0
30
30
10
50
100
100
100
100
100
100
0
0
2.5
0
20
50
70
30
80
100
100
100
- 8 1 -
Table 3-20 Charpy V-notch Data for the Kori Unit 1 Intermediate Shell Forging124W375 VA1 (Tangential Direction) (CONTINUED)
Capsule
S
R
Flue nee(xio'fycm*)
1.4149
2.9219
SpecimenI.D
KL-28
KL-26
KL-31
KL-27
KL-25
KL-30
KL-29
KL-35
KL-32
KL-34
KL-33
KL-36
KL-15
KL-18
KL-16
KL-14
KL-20
KL-13
KL-23
KL-19
KL-24
KL-22
KL-17
KL-21
Temperature( *F)
-76
-40
-20
0
0
15
39
75
100
167
212
550
-40
-27.4
1.4
17.6
37.4
64.4
86
123.8
167
260.6
325.4
550.4
Energy(ft-lbl
10.0
9.0
52.0
105.0
7.0
82.5
107.5
107.0
163.0
164.0
169.5
158.2
8.0
7.0
60.0
96.0
90.0
114.0
135.0
149.0
148.0
149.0
155.5
164.0
Lateral Exp.(mils)
13
14
39
67
13
59
76
67
44
24
46
5.5
4.3
41.3
59.1
60.6
68.1
52.8
56.7
41.7
81.1
64.6
Shear(%)
10
5
10
50
40
40
30
70
100
100
100
100
0
0
40
30
30
60
55
85
90
100
100
63.1 100
- 8 2 -
Table 3-21 Charpy V-notch Data for the Kori Unit 1 Intermediate Shell Forging124W375 VA1 (Axial Direction)
Capsule
Unirrad.
V
T
Fluence ; Specimen(xio'fycm*) . |.D
0.0
0.4401
1.0547
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
KT-9
KT-10
KT-11
KT-12
KT-7
KT-8
KT-5
KT-6
KT-1
KT-2
KT-3
KT-4
KT-43
KT-44
KT-45
KT-46
KT-42
KT-48
KT-47
KT-41
KT-37
KT-40
KT-38
KT-39
Temperature( *F)
-100
-100
-100
-40
-40
-40
0
0
0
40
40
40
90
90
90
210
210
210
-40
-40
-20
-20
0
0
40
40
76
100
210
210
-40
-40
-20
-20
0
0
20
Energy(ft-lbl
5.5
10
7.5
70
87
57.5
60
101
110
120
174
101
168
168
171.5
155
162
166
21.0
13.0
58.5
72.0
33.0
77.5
48.0
103.0
115.0
115.0
136.5
147.0
6.0
26.0
41.0
14.5
90.5
72.5
84.5
40 i 89.0
75
165
210
250
124.5
143.5
149.0
145.0
Lateral Exp.(mils)
0
2.5
0.5
503
62
38
43
71
72
76
85
67
87
89
88
81
81
83
19
11
49
56
29
68
42
81
82
92
98
98
6
23
34
14
74
59
71
71
83
96
95
93
Shear(%)
0
0
0
20
29
20
20
51
53
59
100
53
100
100
100
100
100
100
0
0
20
30
10
50
10
100
100
100
100
100
0
5
10
0
50
30
50
50
80
100
100
100
- 8 3 -
Table 3-21 Charpy V-notch Data for the Kori Unit 1 Intermediate Shell Forging124W375 VA1 (Axial Direction) (CONTINUED)
Capsule
S
R
Flue nee(x10'9n/cm*)
1.4149
2.9219
SpecimenI.D
KT-30
KT-31
KT-27
KT-28
KT-26
KT-29
KT-35
KT-25
KT-34
KT-33
KT-32
KT-36
KT-19
KT-13
KT-14
KT-17
KT-15
KT-22
KT-23
KT-24
KT-t6
KT-21
KT-18
KT-20
Temperature( *F)
-40
-20
-7
25
25
39
75
126
167
212
261
550
-41.8
1.4
14
17.6
36.6
64.4
123.8
167
212
260.6
325.4
550.4
Energy(ft-iby
3.0
2S.0
22.5
82.0
83.5
37.0
112.2
136.3
151.5
147.2
151.0
206.0
5.0
32.0
24.0
16.0
56.0
78.0
113.0
128.0
143.0
145.0
139.5
133.5
Lateral Exp.(mils)
4
22
21
61
70
32
70
68
5
66
68
-
6.3
25.2
16.1
15.7
41.3
57.9
46.5
43.7
34.3
85.4
83.5
83.5
Shear(%)
0
20
30
50
40
60
70
80
95
100
100
100
0
0
10
5
20
30
60
85
100
90
100
100
- 8 4 -
Table 3-22 Charpy V-notch Data for the Kori Unit 1 Reactor Pressure Vessel CoreRegion Weld
Capsule
Unirrad.
V
Fluence(xiO'Vcm2)
0.0
0.4401
T 1.0547
SpecimenI.D1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
KW-1
KW-2
KW-3
KW-4
KW-5
KW-6
KW-7
KW-6
KW-9
KW-10
KW-11
KW-12
KW-37
KW-45
KW-46
KW-38
KW-39
KW-40
KW-41
KW-42
KW-43
KW-44
KW-47
KW-48
Temperature( °P)
-25
-25
-25
10
10
10
40
40
40
100
100
100
210
210
210
76
100
165
165
210
210
250
250
325
325
400
400
75
165
165
210
210
250
325
325
400
400
450
450
Energy(ft-lbl
20
24
19
36
39
41
44
48
50
60
57
58
67
66
66
12.5
16.5
28.5
24.0
32.0
35.0
39.5
42.0
44.0
41.0
47.0
48.5
13.5
24.0
27.5
32.0
35.0
42.0
45.0
41.5
41.5
42.5
40.0
40.0
Lateral Exp.(mils)
17
18
16
33
33
38
37
42
44
65
55
57
71
73
69.5
13
19
30
26
34
38
42
40
44
46
45
48
10
27
28
32
36
42
48
41
46
45
45
45
Shear(%)
10
12
10
32
34
45
48
60
60
80
65
75
100
98
100
0
10
20
25
45
50
60
60
60
80
100
100
5
25
30
50
60
100
100
100
100
100
100
100
- 8 5 -
Table 3-22 Charpy V-notch Data for the Kori Unit 1 Reactor Pressure Vessel CoreRegion Welds (CONTINUED)
Capsule
S
Fluence(x1019n/cm*)
1.4149
I!
R 2.9219
SpecimenI.D
KW-28
KW-35
KW-27
KW-34
KW-33
KW-26
KW-32
KW-25
KW-31
KW-30
KW-29
KW-36
KW-14
KW-13
KW-24
KW-21
KW-20
KW-23
KW-19
KW-22
KW-18
KW-17
KW-16
KW-15
Temperature( *F)
16
75
126
167
212
212
261
284
325
351
399
550
17.6
64.4
125.6
167
212
212
284
284
325.4
350.6
399.2
550.4
4.5
15.5
19.0
23.0
26.0
32.0
36.5
42.0
45.5
49.0
46.5
45.0
10.5
11.5
14.0
18.0
24.0
24.0
41.0
41.3
39.0
40.5
41.0
38.0
Lateral Exp.(mils)
9
13
22
26
30
37
26
37
43
38
39
36
2
7.1
13.4
•0.4
-8.3
-9.4
39.4
33.5
34.6
31.5
36.2
35.8
Shear(%)
5
30
50
80
90
95
90
95
100
100
100
100
0
5
10
20
50
50
90
100
90
100
100
100
- 8 6 -
Table 3-23 Charpy V-notch Data for the Kori Unit 1 Reactor Pressure Vessel WeldHeat-Affected-Zone Material
Capsule
Unirrad.
V
T
Fluence(xio'fycm*)
0.0
0.4401
1.0547
SpecimenI.D1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
KH-11
KH-12
KH-9
KH-10
KH-7
KH-8
KH-5
KH-6
KH-1
KH-2
KH^3
KH-4
KH-44
KH-45
KH-42
KH-43
KH-41
KH-39
KH-38
KH-40
KH-37
KH-46
KH-47
KH-48
Temperature( *F)
-100
-100
-100
•50
-50
-50
0
0
0
40
40
40
100
100
100
210r 210
210
-100
-100
-40
-40
0
0
40
40
76
100
210
210
-80
-80
-40
-40
0
40
75
165
210
•
-
-
Energy(tt-lbl
31
31
50
40
146
43
72
103
101
88
110
122
76
93
97
108
188
204
121.0
51.5
87.0
123.0
139.0
80.0
76.0
176.0
170.5
107.0
115.5
75.5
27.0
105.0
90.5
62.5
115.0
161.0
163.0
99.0
110.0
-
-
-
Lateral Exp.(mils)
18
15
Shear(%)
14
14
22 | 20
23 ! 35
75
26
37
55
56
53
64
63
58
65
64
72
86
71
73
29
53
75
81
49
50
93
98
65
73
60
17
64
53
37
72
89
93
60
73
-
-
-
60
20
45
50
48
57
75
78
-
94
-
98
100
100
100
10
100
100
100
100
50
100
100
100
100
35
5
90
75
20
50
100
100
80
100
-
-
-
- 8 7 -
Table 3-23 Charpy V-notch Data for the Kori Unit 1 Reactor Pressure Vessel WeldHeat-Affected-Zone Material (CONTINUED)
Capsule
S
R
Fluence(x1019n/cm*)
1.4149
2.9219
SpecimenI.D
KH-27
KH-28
KH-30
KH-31
KH-26
KH-29
KH-35
KH-25
KH-34
KH-33
KH-32
KH-36
KH-16
KH-15
KH-14
KH-13
KH-21
KH-22
KH-17
KH-19
KH-20
KH-18
Temperature( -F)
-101
-76
-60
-20
0
39
75
100
167
212
399
550
-56.2
•0.4
32
64.4
125.6
167
260.6
325.4
399.2
550.4
Energy(1Mb!
29.5
94.0
98.0
108.0
66.0
94.0
150.0
89.0
140.5
154.0
189.0
167.5
104
31
52
134
149
138
152
94.5
151
228
Lateral Exp.(mils)
27
55
50
65
35
54
66
26
-
72
71
-
52.4
14.2
26.4
65.8
36.6
40.2
66.5
60.6
85.4
67.7
Shear(%)
0
10
40
50
50
100
100
90
100
100
100
100
50
30
30
100
70
70
85
75
100
100
- 8 8 -
Table 3-24. Effect of Irradiation to 0.440 xiO^n/cm' (E>1.0 MeV) on the Notch Toughness Properties of the Kori Unit 1Reactor Vessel Surveillance Capsule V Materials
Material
IntermediateShell Forging124W375 VA1
(Tangential)
IntermediateShell Forging124W375 VA1
(Axial)
Weld Metal
— - 1
HAZ Metal
Average 30 ft-lbTr-T.;-;i.'ion Temperature (°F)
Unirrad.""
-35.5
-74.46
-6.13
-223.12
Irrad.
-44.78
43.08
184.84
-179.15
AT
-9.27
31.37
190.97
43.97
Average 35 mil LateralExpansion Temperature (°F)
Unirrad.11"
-21.27
-49.83
19.97
Irrad.
-28.68
-24.2
202.17
AT
-7.4
25.62
182.19
Average 50 ft-lbTransition Temperature (°F)
Unirrad.10'
-18.61
-48.06
49.24
Irrad.
-20.64
-11.27
AT
-2.02
36.79
Average Energy Absorptionat Full Shear (ft-lb)
Unirrad.'"
171.0
166.4
66.5
lrrad. (C)
155.6
123.3
47.8
AE
-15.4
-43.1
-18.7
NOTES ;(a) The values were obtained from the hyperbolic tangent curve-fitting program CVGRAPH, Versions 4.0 and 4.1(b) The unirradiated values differ from those reported in WCAP-8586 since those previously reported were developed from hand fit curves
using engineering judgement. The values reported here were determined from curves generated by CVGRAPH Versions 4.0 and 4.1.(c) Values determined per the definition of "upper shelf energy" given in ASTM E185-82.
I
sI
Table 3-25. Effect of Irradiation to 1.055 xicTn/cnV' (E>1.0 MeV) on the Notch Toughness Properties of the Kori Unit 1Reactor Vessel Surveillance Capsule T Materials
Material
IntermediateShell Forging124W375 VA1
(Tangential)
IntermediateShell Forging124W375 VA1
(Axial)
Weld Metal...
HAZ Metal
Average 30 ft-lbTransition Temperature (°F)
Unirrad.10'
-35.5
-74.46
-6.13
-223.12
Irrad.
-46.55
-31.07
180.84
-179.15
AT
-11.05
43.39
186.97
43.97
Average 35 mil LateralExpansion Temperature (°F)
Unirrad.'"'
-21.27
-49.83
19.97
Irrad.
-34.05
-17.08
207.48
AT
-12.77
32.75
187.51
Average 50 ft-lbTransition Temperature (°F)
Unirrad.""
-18.61
-48.06
49.24
Irrad.
-29.14
-9.13
AT
-10.52
38.93
Average Energy Absorptionat Full Shear (ft-lb)
Unirrad.1"
171.0
166.4
66.5
Irrad.1"
152.2
145.8
41.8
AE
-18.8
-20.6
-24.7
NOTES.;(a) The values were obtained from the hyperbolic tangent curve-fitting program CVGRAPH, Versions 4.0 and 4.1(b) The unirradiated values differ from those reported in WCAP-8586 since those previously reported were developed from hand fit curves
using engineering judgement. The values reported here were determined from curves generated by CVGRAPH Versions 4.0 and 4.1.(c) Values determined per the definition of "upper shelf energy" given in ASTM E185-82.
I
Table 3-26. Effect of irradiation to 1.415 x101vn/cml! (E>1.0 MeV) on the Notch Toughness Properties of the Kori Unit 1Reactor Vessel Surveillance Capsule S Materials
Material
IntermediateShell Forging124W375 VA1
(Tangential)
IntermediateShell Forging124W375 VA1
(Axial)
Weld Metal
HAZ Metal
Average 30 ft-lbTransition Temperature (°F)
Unirrad.""
-35.5
-74.46
-6.13
-223.12
Irrad.
-32.38
-17.1
202.19
-179.15
AT
3.11
57.35
208.32
43.97
Average 35 mil LateralExpansion Temperature (°F)
Unirrad.w
-21.27
-49.83
19.97
Irrad.
-17.58
1.02
264.07
AT
3.69
50.86
244.09
Average 50 ft-lbTransition Temperature (°F)
Unirrad.""
-18.61
-48.06
49.24
Irrad.
-0.98
14.66
AT
17.63
62.73
Average Energy Absorptionat Full Shear (ft-lb)
Unirrad.tC)
171.0
166.4
66.5
lrrad.'C)
163.7
168.1
46.5
AE
-7.3
1.7
-20.0
NOTES ;(a) The values were obtained from the hyperbolic tangent curve-fitting program CVGRAPH, Versions 4.0 and 4.1.(b) The unirradiated values differ from those reported in WCAP-8586 since those previously reported were developed from hand fit curves
using engineering judgement. The values reported here were determined from curves generated by CVGRAPH Versions 4.0 and 4.1.(c) Values determined per the definition of "upper shelf energy" given in ASTM E185-82.
ItoINS
I
Table 3-27.
Material
IntermediateShell Forging124W375 VA1
(Tangential)
IntermediateShell Forging124W375 VA1
(Axial)
Weld Metal
HAZ Metal
Effect of Irradiation to 2.922 xiO^n/cnV (E>1.0 MeV) on the Notch Toughness Properties of the Kori Unit 1Reactor Vessel Surveillance Capsule R Materials
Average 30 ft-lbTransition Temperature (°F)
Unirrad.TO
-35.5
-74.46
-6.13
-223.12
Irrad.
-22.7
14.66
228.94
-179.15
AT
12.8
89.13
235.07
43.97
Average 35 mil LateralExpansion Temperature (°F)
Unirrad.""
-21.27
-49.83
19.97
Irrad.
-2.08
71.62
AT
19.19
121.45
Average 50 ft-lbTransition Temperature (°F)
Unirrad.1"'
-18.61
-48.06
Irrad.
•0.91
40.47
AT
17.7
88.54
Average Energy Absorptionat Full Shear (ft-lb)
Unirrad.10'
171.0
166.4
66.5
Irrad.1"
156.2
138.7
40.2
AE
-14.8
-111
-26.3
NOTES ;(a) The values were obtained from the hyperbolic tangent curve-fitting program CVGRAPH, Versions 4.0 and 4.1.
(b) The unirradiated values differ from those reported in WCAP-8586 since those previously reported were developed from hand fit curvesusing engineering judgement. The values reported here were determined from curves generated by CVGRAPH Versions 4.0 and 4.1.
(c) Values determined per the definition of "upper shelf energy" given in ASTM E185-82.
Table 3-28. Calculation of Average Cu and Ni Weight Percent Values for BeltlineMaterials
Ref.
(a)
(b)
(c)
(d)
(e)
(t)
(g)
(h)
(i)
Avg.
Inter. ShellForging
124W375 VA1Cu
0.07
0.02
0.11
0.08
0.02
0.06
Ni
0.73
0.73
0.70
0.70
0.715
Inter. ShellForging
124W375 VA1Cu
0.05
Ni
0.70
Lower ShellForging
124W375 VA1Cu
0.04
Ni
0.76
SurveillanceWeld Metal
Cu
0.23
0.18
0.30
0.19
0.18
0.29
0.23
Ni
0.61
0.76
0.65
0.63
0.68
0.67
Inter. & LowerShell Weld
Cu
0.22
0.13
Ni
0.55
0.69
Notes ;
(a) Surveillance program material (Kori 1 Surveillance program WCAP-8586, Table A-2)(b) Chemical analysis by Babcock & Wilcox(c) Weld qualification No. WF233, Wire heat No. T29744, Flux type Linde 80, Lot No. 8790(d) Weld qualification No. WF232, Wire heat No. 8T3914, Flux type Linde 80, Lot No. 8790(e) The 1st surveillance test result(f) The 2nd surveillance test result(g) The 3rd surveillance test result(h) The 4th surveillance test result(i) ^e-evaluation by Babcock & Wilcox
- 9 3 -
Table 3-29. Interpolation of Chemistry Factors from Regulatory Guide 1.99,Revision 2,Position 1.1
Material
Intermediate Shell Forqinq 124W375 VA1
Given Cu wt% - 0.060
Intermediate Shell Forqinq 124W375 VA1
Given Cu wt% - 0.050
Lower Shell Forqinq 122X371 VA1
Given Cu wt% - 0.040
Weld Metal w
Given Cu wt% - 0.23
Weld Metal w
Given Cu wt% -.0.29
Ni, wt%
0.715
0.70
0.76
0.67
0.68
Chemistry Factor, (°F)
37.0
31.0
26.0
180.9
203.4
Note :(a) Based on the total averaged value(b) Based on the re-evaluation value by Babcock & Wilcox Co.
- 9 4 -
Table 3-30. Calculation of Chemistry Factors Using Surveillance Capsule Data PerRegulatory Guide 1.99, Revision 2, Position 2.1
Material
Intermediate ShellForging124W375 VA1(Tangential)
Intermediate ShellForging124W375 VA1(Axial)
Weld Metal""
Capsule
V
T
S
R
V
T
S
R
Capsule1 "
0.4401
1.0547
1.4149
2.9219
0.4401
1.0547
1.4149
2.9219
FF1"
0.772
1.015
1.096
1.284
0.772
1.015
1.096
1.284
ARTNDT16'
-7.40
-11.05
3.11
12.80
25.62
43.40
57.35
89.13
SUM
FF« ARTNDT
-5.71
-11.22
3.41
16.44
19.78
44.05
62.86
114.44
214.13
FF2
0.596
1.030
1.201
1.649
0.596
1.030
1.201
1.649
4.476
CFimemHrtiae Shdl Foxing - Z ( F F * A R T N D T ) ^ Z ( F F 2 ) - 53.9 'F
V
T
sR
0.4401
1.0547
1.4149
2.9219
0.772
1.015
1.096
1.284
182.20
186.97
208.32
235.07
SUM
140.66
189.77
228.32
301.83
860.58
0.596
1.030
1.201
1.649
4.476
CFw*. Met* - Z(FF*ARTNDT) -Z(FF2) - 192.3 'F
NOTES:
(a) f-fluence (101* n/cm2, E>1.0MeV). All updated fluence values were taken from the Capsule Ranalysis.
(b) FF-fluence factor - p***1"** »(c) ARTNDT values were obtained from the Capsule V, T, S & R analysis, rounded to two-decimal.(d) The reactor vessel intermediate to lower shell circular weld seam was made with the same weld
wire and flux as the surveillance weld specimens.(e) Base metal CF value was calculated from the axial direction materials.
- 9 5 -
Table 3 - 3 1 . KORI Unit 1 Surveillance Capsule Data Calculation of Best-Fit Line asDescribed in Position 2.1 of Regulatory Guide 1.99, Revision 2
Material
Intermediate ShellForging124W375 VA1(Tangential)
Intermediate ShellForging124W375 VA1(Axial)
Weld Metal
Capsule
V
T
S
R
Capsulef""
0.4401
1.0547
1.4149
2.9219
2. M
V
T
S
R
0.4401
1.0547
1.4149
2.9219
Z i-i
V
T
S
R
0.4401
1.0547
1.4149
2.9219
Z i-i
F F .W
(X)
0.772
1.015
1.096
1.284
4.167
0.772
1.015
1.096
1.284
4.167
0.772
1.015
1.096
1.284
4.167
ARTNOT ( C )
(y)
-7.40
-11.05
3.11
12.80
-2.54
25.62
43.40
57.35
89.13
215.5
182.20
186.97
208.32
235.07
812.56
F F ' A R T N O T
(XV)
-5.71
-11.22
3.41
16.44
2.92
19.78
44.05
62.86
114.44
241.1
140.66
189.77
228.32
301.83
860.58
FF2
(x2)
0.596
1.030
1.201
1.649
4.476
0.596
1.030
1.201
1.649
4.476
0.596
1.030
1.201
1.649
4.476
NOTES:
(a) f-fluence (1019 n/cm2, E>1.0MeV). All updated fluence values were taken from the Capsule Ranalysis.
(b) FF-f!uence factor(c) ARTNDT values were obtained from the Capsule V, T, S & R analysis, rounded to two-decimal.
- 9 6 -
Table 3-32. KORI Unit 1 Surveillance Capsule Data Evaluation of Credibility asDescribed in Position 2.1 of Regulatory Guide 1.99, Revision 2
Material
Intermediate ShellForging124W375 VA1(Tangential)
Intermediate ShellForging124W375 VA1(Axial)
Weld Metal
Capsule
V
T
S
R
V
T
S
R
V
T
S
R
Capsulef
0.4401
1.0547
1.4149
2.9219
0.4401
1.0547
1.4149
2.9219
0.4401
1.0547
1.4149
2.9219
FF
0.772
1.015
1.096
1.284
0.772
1.015
1.096
1.284
0.772
1.015
1.096
1.284
ARTNOT
(30ft-lb)-7.40
-11.05
3.11
12.80
25.62
43.40
57.35
89.13
182.20
186.97
208.32
235.07
Best FitARTNOT
21.6
50.8
60.5
83.1
176.6
200.66
208.68
227.28
Scatter ofARTNOT
3.96
-7.4
-3.2
6.1
5.58
13.7
0.36
7.79
** Base Metal : < 17*F215.15 - 4a + 4.167b241.13 - 4.167a + 4.476ba - -70.9, b - 119.9
Y - 119.9 (X) - 70.9
** Weld Metal : < 28'F812.56 - 4a + 4.167b860.58 - 4.167a + 4.476ba - 100.23, b - 98.95Y - 98.95 (X) + 100.23
- 9 7 -
Table 3-33. RTPTS Calculation for the Kori 1 Reactor Vessel Beltline Region Materials
EFPY
32
40
Material
nter. Shell Forging
Using S/C data
Weld Metal
Using S/C data
Inter. Shell Forging
Using S/C data
Weld Metal
Using S/C data
CF
37
53.9
180.9|a|
203.41"1
192.3
37
53.9
180.91"
2O3A™
192.3
f
3.861
3.861
3.861
3.861
3.861
4.874
4.874
4.874
4.874
4.874
FF
1.349
1.349
1.349
1.349
1.349
1.397
1.397
1.397
1.397
1.397
1
30
30
-20
-20
-20
30
30
-20
-20
-20
M
34
17
56
56
28
34
17
56
56
28
ARTVre
49.9
72.7
244.0
274.4
259.4
51.7
75.3
252.7
284.1
268.6
RTVrs
113.9
119.7
280.0
310.4
267.4
115.7
122.3
288.7
320.1
276.6
NOTE :(a) : Based on the total average value(b) : Based on the re-evaluation value by Babcock & Wilcox
Table 3-34. Peak Fluence (1019 n/cm2, E>1.0 MeV) on the Pressure Vessel Clad/BaseMetal Interface for Kori Unit 1
EFPY
13.5
20.0
24.0
32.0
40.0
Vessel 0° Location
1.649
2.390
2.880
3.861
4.874
- 9 9 -
Table 3-35. Margins for Adjusted Reference Temperature (ART) Calculation perRegulatory Guide 1.99, Revision 2
Material Properties Surv. Capsule Data NOT Used Surv. Capsule Data Used
Plates or Forgings
Measured IRTNDT
Generic IRTNOT
34
48
17
38
Weld Metal
Measured IRTNDT
Generic IRTNDT
56
66
28
44
- 1 0 0 -
Table 3-36. Calculation of ART Values for the Kori Unit 1 Reactor Vessel BeltlineRegion Materials at 20EFPY
Material ICF flat LD) f(1/4,3/4T) FF ,w M ARTNDT ART
1/4T Calculations
Inter. Shell Forging
Using S/C data
Inter. Shell Forging
Lower Shell Forging
Weld Metal
Using S/C data
37.0
53.9
31.0
26.0
180.9
192.3
2.390
2.390
2.390
2.390
2.390
2.390
1.618
1.618
1.618
1.618
1.618
1.618
1.133
1.133
1.133
1.133
1.133
1.133
30
30
30
30
-10
-10
34.0
17.0
34.0
29.5
56.0
28.0
41.9
61.1
35.1
29.5
205.0
217.9
105.9
108.1
99.1
89.0
251.0
235.9
3/4T Calculations
Inter. Shell Forging
Using S/C data
Inter. Shell Forging
Lower Shell Forging
Weld Metal
Using S/C data
37.0
53.9
31.0
26.0
180.9
192.3
2.390
2.390
2.390
2.390
2.390
2.390
0.742
0.742
0.742
0.742
0.742
0.742
0.916
0.916
0.916
0.916
0.916
0.916
30
30
30
30
-10
-10
33.9
17.0
28.4
23.8
56.0
28.0
33.9
49.4
28.4
23.8
165.7
176.1
97.8
96.4
86.8
77.6
211.7
194.1
NOTES:
(a) Initial RTNDT values are measured values.
- 1 0 1 -
Table 3-37. ART Values at the 1/4T and 3/4T Locations Used in the Curves Generation
Material
Inter. Shell Forging
Using S/C Data
Inter. Shell Forging
Lower Shell Forging
Weld Metal
Using S/C Dataf)
1/4T ART
105.9
108.1
99.1
89.0
251.0
23S.9
3/4T ART
97.8
96.4
86.8
77.6
211.7
194.1
(*) Used in the generation of the heatup/cooldown curves.
- 1 0 2 -
VESSEL
THERMALSHIELD
SURVEILLANCECAPSULE
S,
* Different from FSAR and WCAP 8586
Fig. 2-1. Plan View of Surveillance Capsules in Kori 1 Reactor Vessel
- 1 0 3 -
Weld WF-259(Linde 80)
Intermediate Shell (Forging)SA508-2124W375VA1
Weld WF 232 / WF 233(Linde 80)
Lower Shell (Forging)SA508-2122X371VA1
Weld
Fig. 2-2. Weld Specification of Reactor Vessel Beltline Region
- 1 0 4 -
Inlet Nozzle Weld
Weld WF-259(Linde 80)
63Intermediate ShellSA508-2(Forging)'124W375VA1
Capsule
Weld WF 232 / WF 233(Linde 80)
Lower ShellSA508-2(Forging),122X371VA1
Thermal Shield —
Weld
-21"
-10"
57-3/4"
61-1/4"
CORE
-14-3/4"
62-3/8"
66"72.5"
-39-1/2"
144"
VLcv
CAPSULE
0.030" STAINLESSSTEEL
— • CORESPACERSCHARPY SPECIMEN
WELD
Fig. 2-3. Surveillance Capsule Location at Reactor Vessel Beltline Region
- 1 0 5 -
EGADS
I
g
1.1 X I997 /07 /28 3028452797231KOR! Unit 1 (R.Theta OORT Geometry)
100 150 200 250X(em)
Fig. 2-4. R & Theta DORT Geometry for Kori Unit 1
SORCERY
(4*SC) 'sc' Capsule locationvessel 0, 15. 30 . 45
CRSD • CRSD source<SC) - survelltance Capsule
DORT forward(CRSD)
DORT adjoint
importance function KR.T.E) * ;
ADIOS CRSD(4.SC)
TADLIB (4»SC)
T/ importance function I(R,T,E!*•*.».. (pin & box)
< overall tolas fa,- Kj> ADIOS (4*SC>
^ Reference Pin PowerGradient
Plant Specific PowerdisU for all fuel cycles
Calculate*(E>1.0>for all cycles
FCALC (SC)
mea, reaction rales "with uncertainties
SAND (SC)
Power history1 Measured fail data
U-Z3S, Pu-239 and ( r.1)correction
[ cross section
"'" dosim, c.i, <S3g) ""*~-col. flux and acti. (forward)
^ correction factor *-
FERRET (SC)•<E>1.OV
• ^ covar, matrix for c.s
•(E>0.l dpa
•<E>1.0 •<E>1.0
adj. n spectrum and c.s
adj. cal. reaction ratesadj. mea- •(E>1.0)
_ with uncertainties _
dpa, n fluxes, fluencewith uncertainties
SensitivityStudy
^overall uncertainty*^
Fig. 2-5. Flow Chart of Neutron Transport Calculation per Reg.Guide DG-1053
-107 -
EGADS 1.1 X1997/09/23 2835114527974KOR! Unit 1(R, Theta DORT Geometry, cycle 4 - 5 )
X (cm)
Fig. 2-6. R & Theta OORT Geometry for the Fuel Cycle 4 & 5
KORI UNIT 1 (BASE METAL : KL)
Curve Fluence LSE d-LSE USE
Results
d-USE T o 30 d-T e 30 T e 50 d-T e 50
300"
04.401E+181.055E+191.415E+19Z922E+19
2.192.192.192J92.19
171155.6152.19163.69156.19
0-15J39-18.8-7.3-14.8
-35.5-44.78-4655-3236-22.7
0 -18.61-927 -20.64-11.05 -29.14
3.11 - A£ 7 9 -.91
0-2.02-1052
17.6317.7
-200 -100 0 100 200 300 400
Temperature in Degrees F
Curve Legend
2 O 3 ^ 4 * 5^T
500 600
Curve Plant Capsule
Data Set(s) Plotted
Material Ori. HeatflKORUKORUKORUKORUKORIl
UN1RRVTSR
FORGING SA508CL2FORGING SA508CL2FORGING SA50BCL2FORGING SA50BCL2FORGING SA508CL2
KL I24I375VMKL 124W375VA1KL 124W375VA1KL 124W375VA1KL 124W375VA1
Fig. 3-1. Charpy V-Notch Impact Energy vs. Temperature for Kori Unit 1 ReactorVessel Intermediate Shell Forging 124W375 VA1 (Tangential)
- 1 0 9 -
KORI UNIT 1 (KL : Lateral Exp.)
rve12345
Fluence0
4.4O1E+18L055E+19U15E+19a922E+l*
USE86.6994594.376
6959
Results
d-USE07.87.6
-10.69-17.09
ToLE35-2127-28.68-34.05-1758-2.0B
d-T e LE350
-7.4-12.77
3.6919.19
CD
Exp
mil
CO
ate
200"!
150
100
50~
0~
r
sf-
&~ ...
-300 -200 -100 0 100 200 300 400
Temperature in Degrees FCurve Legend
2 O" 3 O • 4 5 ^
500 600
Curve Plant Capsule
DaU Set(s) Plotted
Material OrL Heat |K0RI1K0RI1KORUK0RI1KORU
UNIRRVTSR
FORGING SA506CL2FORGING SA508CL2FORGING SA508C12FORGING SA508CL2FORGING SA508CL2
KL 124W375VA1KL 124W375VA1KL 124W375VA1KL 124W375VA1KL 124W375VA1
Fig. 3-2. Charpy V-Notch Lateral Expansion vs. Temperature for Kori Unit 1 ReactorVessel Intermediate Shell Forging 124W375 VA1 (Tangential)
-no-
KORI UNIT 1 (KL : /.SHEAR)
Curve12345
Fluence0
4.401E+181D55E+191.415E+I9Z922E-ri9
ResultsTe 50/. Shear
10.155.620
473460
d-T *> 50:/. Shear0
-4.52-10.15
37.1649.84
cti<vA
o
100"
«o
60
40
<
tj1
D /Jb,/
/1
-300 -200 -100 0 100 200 300 400
Temperature in Degrees F500 600
Curve12345
2O-
PlantK0RI1K0RI1K0R11K0RI1K0RI1
a
CapsuleUNIRR
VTSR
Curve Legend
Data Set(s) Plotted
MaterialFORGING SA508CL2FORGING SA508CL2FORGING SA508CL2FORGING SA508CL2FORGING SA508CL2
Ori.KLKLKLKLKl,
Heattf124W375VA1124W375VA1124W375VA1124W375VA1124W375VA1
Fig. 3-3. Charpy V-Notch Percent Shear vs. Temperature for Kori Unit 1 Reactor VesselIntermediate Shell Forging 124W375 VA1 (Tangential)
- 1 1 1 -
KORI UNIT 1 (BASE METAL : KT)
Curve Fluence LSE d-LSE USE
Results
d-USE T o 30 d-T o 30
CO
£
O!
w
>
04.401E+181.055E+191.415E+19Z922E+19
awZ192J9Z192A9
0 166350 123.10 14530 168.;0 138.6G
o-43.09-2059
1.7-27J69
-74.46-43.06-31.07- 1 ' . !14J66
03137433957.3569.13
T9_50-48.06-1127-9.1314.6640.47
d-T o 500
36.7938536Z738854
J00
250
200
150
100
bO
0
i(
ao (
a e
v/y
7t
n •
*-±-
-300 -200 -100 0 100 200 300 400
Temperature in Degrees FCurve Legend
2 O 3 ^ 4 A 5TT
500 600
Curve Plant Capsule
Data Set<s) Plotted
Material Ort HeatjK0RI1K0RI1K0R11K0RI1K0RI1
UNIRRVTSR
FORGING SA508CL2FORGING SA508CL2FORGING SA506CL2FORGING SA508CL2FORGING SA508CL2
KT 124W375VA10 124W375VA1CT 124W375VA10 124W375VA10 124W375VA1
Fig. 3-4. Charpy V-Notch Impact Energy vs. Temperature for Kori Unit 1 Reactor VesselIntermediate Shell Forging 124W375 VA1 (Axial)
-112-
KORI UNIT 1 (KT : Lateral Exp.)
Curve12345
Fluence0
4.401E+1B1.055E+191.415E+192.922E+19
USE64.990.1994.696884
Results
d-USE0
52»9.73
-If .3-.9
To LE35
-49.83-242-17.08
1.0271.62
d-T e LE350
25.62327550.86121.45
200
CO
n3 150"
X
ater
a
100"
-300 -200 -100 0 100 200 300 400
Temperature in Degrees FCurve Legend
2 O 3 ^ 4 ^ 5T?
500 600
Curve Plant Capsule
Data Set(s) Plotted
Material Ori. Heat |K0RI1K0R11K0R11K0RI1K0R11
UNIRR.VTSR
FORGING SA508CL2FORGING SA508CL2FORGING SA508O2FORGING SA508CL2FORGING SA508CL2
KT 124W375VA1KT 124W375VA1KT 124W375VA1CT 124I375VA1KT 124W375VA1
Fig. 3-5. Charpy V-Notch Lateral Expansion vs. Temperature for Kori Unit 1 ReactorVessel Intermediate Shell Forging 124W375 VA1 (Axial)
-113-
KORI UNIT 1 (KT : v. Shear)
Curve12345
Fluence0
4.401E+18L055E+191.415E+192922E+I9
Results
T o 507. Shear7.9625.3128.1235.15102B3
d-T o 5ft': Shear0
173420.1527.1894.86
in
ouCO
a,
-200 -100 0 100 200 300 400
Temperature in Degrees FCurve Legend
500 600
Curve Plant Capsule
Data Set(s) Plotted
Material Ori. Heat#K0R11KORUKOMIKORUKOMI
UN1RR.VTSR
FORGING SA508C12FORGING SA508CI2FORGING SA508CL2FORGING SA508CI2FORGING SA508CI2
KT 124W375VA1KT 124W375VA1KT 124W375VA1KT 124W375VA1KT 124W375VA1
Fig. 3-6. Charpy V-Notch Percent Shear vs. Temperature for Kori Unit 1 Reactor VesselIntermediate Shell Forging 124W375 VA1 (Axial)
- 1 1 4 -
KORI UNIT 1 (WELD METAL :KW)
Curve Fluence
1 02 4.401E+18
3 1.055E+194 1.415E+K-5 2.922E+19
LSE d-I.SE USE
Results
d-USE T o 30 d-Te30 T e 50
ai9
22
22
22
22
66.547.7941.79465402
0-18.7-24.7-20
-2629
-6.1318-4.84
3E.19228.94
0190.9,
186.97
2O8X.
235.07
4924
d - T e 50
0
300
en 250
T+->&H 2 0 °
5H 1 5 0CO
W
* m
5cr
0^
r—
,##
-300 -200 -100 0 100 200 300 400
Temperature in Degrees F
Curve Legend
2 O . 3 ^ _ 4 ^ 5 ^
500 600
Curve Plant Capsule
Data Set(s) Plotted
Material On. Heat)?K0RI1K0RI1K0RI1KORUK0RI1
UNIRRVTSR
WELD SA508CL2WELD SA508CL2WELD SA506CL2WELD SA508CL2WELD SA508CL2
KW 124W375VA1KW 124W375VA1KW 124W375VA1KW 124W375VA1KW 124W375VA1
Fig. 3-7. Charpy V-Notch Impact Energy vs. Temperature for Kori Unit 1 Reactor VesselWeld Metal
-115-
KORI UNIT 1 (KW : Lateral Exp.)
Curve12345
Fluence0
4.401E+18L055E+19L415E+192922E+19
USE703465
445939
34.4
Results
d-USE0
-208-25.7-3L3-355
T o L E 3 51937
202.17207.48264.07
d-T o LE350
182.1918751244.09
s—<
3a
ax)
lbO
100
50~
0mmsmmMmam i iwg
_ ^ —
1
-300 -200 -100 0 100 200 300 400
Temperature in Degrees FCurve Legend
2
500 600
Curve Plant Capsule
DaU Sel(s) Plotted
Material Ort HeatjfKORUKORUKORUKORUKORU
mm.vTsR
WELD SA508C12WELD SA508CL2WELD SA508CL2WELD SA5D8CL2WELD SA508CL2
CT 124I375VMKW 124W375VA1KW 124W375VA1KW 124W375VM
KW 124W375VA1
Fig. 3-6. Charpy V-Notch Lateral Expansion vs. Temperature for Kori Unit 1 ReactorVessel Weld Metal
-116-
KORI UNIT 1 (KW : v. Shear)
Curve1234D
Fluence0
4.401E+18L055E+I9L».15E+19
Z922E+19
ResultsT e 50/. Shear
39.B42325197.46
nao6210.46
d-T e 50/ Sliear0
192 5157*1
170?:
cd
in
CO
oCD
PU
100~
ao
60
41)
a)
o
/ (
7j
'iIt.
1
[ I
I///
Ifi
t
-300 -200 -100 0 100 200 300 400
Temperature in Degrees FCurve Legend
2 O . 3 0 4 * 5 ^
500 600
Curve Plant Capsule
Data Set|s) Plotted
Material O n HeatjlKORUK0RI1K0R11K0RI1K0RI1
UNIRR.VTSR
WELD SA508CL2WELD SA508CL2WELD SA508CL2WELD SA508CL2WELD SA508CL2
KW 124W375VMKW 124W375VA1KW 124W375VMKW 124W375VA1KW 124W375VA1
Fig. 3-9. Charpy V-Notch Percent Shear vs. Temperature for Kori UnK 1 Reactor VesselWeld Metal
- 1 1 7 -
en
1
S-i0)
w
u
Curve12345
300
250
200
150
100
5CT
- 3
I D
Fluence0
4.401E+181.055E+191.415E+192922E+19
00 -2
Curve12345
USE2192222219219
00 -1
KORI
d-LSE00000
D
> o
/ * .
00
UNIT 1 (HAZ :
Results
USE d-USE278.37 0118.07 -160312927 -149.127053 -7.8427354 -4.83
_|
-\
0 o
0
Y ° [
0 1
y-
DO
Temperature
2O-
PlantKORUKORUKORUKORUKORU
—
CapsuleUNIRR.
VTSR
T e 30-223J2-179.15-126.74-498.69-416.68
2
D
D
y
.—
o
00
KH)
d-T e 300
43.9796.38
-27556-19355
y
- ^ >_-——
300
in DegreesCurve Legend
3 4 -
Data Set(s) Plotted
MaterialHEAT AFFD ZONEHEAT AFFD ZONEHEAT AFFD ZONEHEAT AFFD ZONEHEAT AFFD ZONE
SA508CL2SA508CL2SA508CI2SA508CL2SA508CL2
5^
Ori. Heat#KH 124W375VA1KH 124W375VA1KH 124W375VA1KH 124W375VA1KH 124W375VA1
T e 50-11281-14853-6276-291.9-22L46
400
F
d-T e 500
-35.7250.04
-179.09-10&65
500 600
Fig. 3-10. Charpy V-Notch Impact Energy vs. Temperature for Korl Unit 1 ReactorVessel Heat-Affected-Zone (HAZ) Metal
-118-
LIMITING MATERIAL : WELD METAL (USING SURVEILLANCE DATA)
LIMITING ART VALUES AT 20 EFPY : 1/4T : 235.93/4T : 194.1
2500 -,
b« 2 2 5 0 -
enft, 2 0 0 ° -
t-t3 1500 -
v 1250 -
1000 -
^ 7 5 0 -
° 5 0 0 -
S 250 -k—<
0 -(
i
3 !
In
L I A K T E S T L I M I T
DNACCEPTABLEOPERATION
KEATVP RATEOP TO «O I / I i .
HSATUP RATEVT TO 1 0 0 F / « r .
50 10d i c {
s
I1
•?
//
ti V
r-i
//
14I
. A
CXITICALITY LIMIT • 4ft f t 0*l a t i R T i c i K T i a o i T A t i e t t i if i l T i c i r i « i i > ur t o t t . i ttry
oi te
50 200 250 300 3J
d T e m p e r a t u r e
* • - -
-A
zz- 4
zl1i~L----4,
i
/r
ri
iJl
fr
itt
Ei tnzTACOP
L
/1
f
cK
-
•T8t
E P T A B L ER A T I O N
H i ^ . i i i i i i —
C H I T . L I K I Trot so r/Hr.C U T . L IHI TFOB 1-0 0 r / H r .
)0 4C
(B)0 4f
eg •
1
50 5(
F))0
Rg. 3-11. Kori Unit 1 Reactor Coolant System Heatup Limitations (Heatup rate up1OO*F/hr) Applicable for 20EFPY (with Instrument error Margins)
-119-
LIMITING MATERIAL : WELD METAL (USING SURVEILLANCE DATA)
LIMITING ART VALUES AT 20 EFPY : 1/4T : 235.93/4T : 194.1
2 5 0 0 -T-
'""-• 9 o <; n .
00 7 o n o -
1 7 5 0 - -o>
•^ 1 s n nonCO
<u 1 2 5 0 - -
1 A n n .
OJ -T* J 7 5 0 - ;«J
^>
._ 500 - -
Q
I'I 1 1 1 1 1 1t t l U M I T
c o oSAT
LOOTNISr .
r 0to40« O
10 0
Ml
WnACCEPTlBLE0PIB1TION
> /r/
—taJffr
iL
•4
1
11
11
11
t•r- J
. 1 1
1
" 7
A C C E P T A B L E tO P E R A T I O N L
6 5 0 1 0 0 1 5 0 2 0 0 2 5 0 3 0 0 3 5 0 4 0 0 4 5 0 5 (
I n d i c a t e d T e m p e r a t u r e ( D e g . F )
Fig. 3-12. Kori Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rate up 1OO'F/hr)Applicable for 20EFPY (with Instrument error Margins)
- 1 2 0 -
>H *1 ^ &. <$ q^r *J 71 -Q:
KAERI/RR-1785/97
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lliE., •§:£ ^ "T1"^^-^. lo1"A|-Ai-^^ O-?]JL Mesh density, Angular expansion,
Convergence criteria ^-3] Tfl + fJ- <>fl 3^3: J5- j-°ll 7l^]*}-P15. Tf^l7]^c>)] n)-
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BIBLIOGRAPHIC INFORMATION SHEETPerforming Org.
Report No.Sponsoring Org.
Report No.Standard
Report No. IMS Subject Code
KAERVRR-1785/97Title/Subtitle The Technology Development for Surveillance Test of
Reactor Vessel Materials
Project Manager / Dept. Chang,Kee-Ok (Reactor Vessel Surveillance Test Dept.)
Researcher and Dept.Byoung Chul Kim, Sam Lai Lee, Sun Pil Choi,Day Young Park, Kwen Jai Choi(Reactor Vessel Surveillance Test Dept)
Pub. Place TaejonlPub.Org.] KAERI 1997. 12. 30
Page P. 132 Table Yes(X ), No( ) 3. 7]
Note
Classified. Open(V), Outside( ), Class ReportType
ResearchReport
Sponsoring Org. KAERI ContractNo.
Abstract (About 300 words)
Benchmark test was performed in accordance with the requirementof USNRC Reg. Guide DG-1053 for Kori unit-1 in order to determinebest-estimated fast neutron fluence irradiated into reactor vessel. Sincethe uncertainty of radiation analysis comes from the calculation error dueto neutron cross-section data, reactor core geometrical dimension, corestructural density, temperature and constituting materials, radiationsource, mesh density, angular expansion and convergence criteria,evaluation of calculational uncertainty due to analytical method wasperformed in accordance with the regulatory guide and the proof wasperformed for entire analysis by comparing the measurement valueobtained by neutron dosimetry located in surveillance capsule.
Best-estimated neutron fluence in reactor vessel was calculated bybias factor, neutron flux measurement value/calculational value, fromreanalysis result from previous 1st through 4th surveillance testing andfinally fluence prediction was performed for the end of reactor life andthe entire period of plant life extension. Pressurized thermal shockanalysis was performed in accordance with 10 CFR 50.61 using theresult of neutron fluence analysis in order to predict the life of reactorvessel material and the criteria of safe operation for Kori unit 1 wasreestablished.Subject Keywords(about 10 words)Reactor Pressure Vessel, Radiation Embrittlement, Neutron transport calculation,Ptessurized Thermal Shock(PTS)
