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==2.1 INTRODUCTION==
==2.1 INTRODUCTION==


2-1 A discrete o rdinates (S N) transport analysis was performed for the Arkansas Nuclear One Unit 2 reactor to determine t he neutron radiation environment within the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant-and fuel-cycle-specific basis. An evaluation of the dosimetry sensor sets from the 284° and 97° surveillance capsules is provided in WCAP-18166-NP  
2-1 A discrete o rdinates (S N) transport analysis was performed for the Arkansas Nuclear One Unit 2 reactor to determine t he neutron radiation environment within the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant-and fuel-cycle-specific basis. An evaluation of the dosimetry sensor sets from the 284° and 97° surveillance capsules is provided in WCAP-18166-NP
[Ref. 21]. The dosimetry analysis documented in WCAP-18166-NP showed that the +/-20% (lcr) acceptance criterion specified in Regulatory Guide 1.190 [Ref. 5] is met. The validated calculations form the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 EFPY. All of the c alculations described in this section were based on nuclear cross-section data derived from the Evaluated Nuclear Data F i le (ENDF) database (specifically, ENDF/B-Vl).
[Ref. 21]. The dosimetry analysis documented in WCAP-18166-NP showed that the +/-20% (lcr) acceptance criterion specified in Regulatory Guide 1.190 [Ref. 5] is met. The validated calculations form the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 EFPY. All of the c alculations described in this section were based on nuclear cross-section data derived from the Evaluated Nuclear Data F i le (ENDF) database (specifically, ENDF/B-Vl).
Furthermore, the neutron transport e v aluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 5]. Additionally, the metho d s used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodolo g y described in WCAP-14040-A, Revision 4 [Ref. 2]. 2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the Arkansas Nuclear One Unit 2 reactor vessel , a s eries of fuel-cycle
Furthermore, the neutron transport e v aluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 5]. Additionally, the metho d s used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodolo g y described in WCAP-14040-A, Revision 4 [Ref. 2]. 2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the Arkansas Nuclear One Unit 2 reactor vessel , a s eries of fuel-cycle
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-00002 Rev O Page 14 of 88 Westinghouse Non-Proprietary Class 3 2-3 The locations of the Arkansas Nuclear One Unit 2 vessel welds and plates are provided in Table 2-1. The axial posit i on of each material is indexed to z = 0.0 cm, which corresponds to the rnidplane of the active fuel stack. These data tabulations include both plant-and fuel-cycle-specific calculated neutron exposures at the end of Cycle 24 , at the end of projected Cycle 25, and at further projections to 54 EFPY. The calculations account for the uprate from 2815 MWt to 3026 MWt that occurred at the beginning of Cycle 16. The projection s are based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 23 , Cycle 24 , and the design of Cycle 25 are representative of future plant operation.
-00002 Rev O Page 14 of 88 Westinghouse Non-Proprietary Class 3 2-3 The locations of the Arkansas Nuclear One Unit 2 vessel welds and plates are provided in Table 2-1. The axial posit i on of each material is indexed to z = 0.0 cm, which corresponds to the rnidplane of the active fuel stack. These data tabulations include both plant-and fuel-cycle-specific calculated neutron exposures at the end of Cycle 24 , at the end of projected Cycle 25, and at further projections to 54 EFPY. The calculations account for the uprate from 2815 MWt to 3026 MWt that occurred at the beginning of Cycle 16. The projection s are based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 23 , Cycle 24 , and the design of Cycle 25 are representative of future plant operation.
The future projections are based on the current reactor power level of 3026 MWt. Selected r e sults from the neutron transport analyses are provided in Table 2-2 through Table 2-7. In Table 2-2 , the calculated maximum fast neutron (E > 1.0 MeV) fluence values for the reactor pressure vessel materials are provided at future projections to 32, 36, 40 , 48 and 54 EFPY. The projections are based on the assumption that the core power distributions and associated plant operating characteristics from Cycles 23-25 are representative of future plant operation.
The future projections are based on the current reactor power level of 3026 MWt. Selected r e sults from the neutron transport analyses are provided in Table 2-2 through Table 2-7. In Table 2-2 , the calculated maximum fast neutron (E > 1.0 MeV) fluence values for the reactor pressure vessel materials are provided at future projections to 32, 36, 40 , 48 and 54 EFPY. The projections are based on the assumption that the core power distributions and associated plant operating characteristics from Cycles 23-25 are representative of future plant operation.
In Table 2-3 , the calculated maximum iron atom displ a cement values for the reactor pressure vessel materials are provided at future projections to 32, 36, 40 , 48 and 54 EFPY. The calcu l ated fast neutron (E > 1.0 MeV) fluence rate , fast neutron (E > 1.0 MeV) fluence , iron atom displacem e nt rate , and iron atom displacements are provided in Table 2-4 through Table 2-7, respectively, for the reactor pressure vessel inner radius at four azimuthal locations , as well as the maximum exposure observed w ithin the octant. The vessel data given in Table 2-4 through Table 2-7 were taken at the clad/base metal interface and represent maximum calculated exposure levels on the vessel. 2.3 CALCULATIONAL UNCERTAINTIES T he uncertainty associated with the calculated neutron exposure of the Arkansas Nuclear One Unit 2 reactor pres s ure vessel materials is based on the recommended approach provided in Regulatory Guide 1.190. In particular , the qualification of the methodology was carried out in the following four s tages: 1. Comparison of calculations with benchmark measurements from the Pool Critical As s embly (PCA) s imulator at the Oak Ridge National Laboratory (ORNL). 2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment.  
In Table 2-3 , the calculated maximum iron atom displ a cement values for the reactor pressure vessel materials are provided at future projections to 32, 36, 40 , 48 and 54 EFPY. The calcu l ated fast neutron (E > 1.0 MeV) fluence rate , fast neutron (E > 1.0 MeV) fluence , iron atom displacem e nt rate , and iron atom displacements are provided in Table 2-4 through Table 2-7, respectively, for the reactor pressure vessel inner radius at four azimuthal locations , as well as the maximum exposure observed w ithin the octant. The vessel data given in Table 2-4 through Table 2-7 were taken at the clad/base metal interface and represent maximum calculated exposure levels on the vessel. 2.3 CALCULATIONAL UNCERTAINTIES T he uncertainty associated with the calculated neutron exposure of the Arkansas Nuclear One Unit 2 reactor pres s ure vessel materials is based on the recommended approach provided in Regulatory Guide 1.190. In particular , the qualification of the methodology was carried out in the following four s tages: 1. Comparison of calculations with benchmark measurements from the Pool Critical As s embly (PCA) s imulator at the Oak Ridge National Laboratory (ORNL). 2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment.
: 3. An a nalytical sensitivity study addressing the uncertainty components resulting from important in p ut parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments. 4. Comparisons of the plant-specific calculations with all available dosimetry results from the Ar kansas Nuclear One Unit 2 surveillance program. WCAP-18 1 6 9-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 15 of 88 Westinghouse Non-Proprietary Class 3 2-4 The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operationa l or geometric variables that impact power reactor calculations. The second phase of the qualification (H.B. Robinson comparisons) addressed uncertainties in these additional a reas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations.
: 3. An a nalytical sensitivity study addressing the uncertainty components resulting from important in p ut parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments. 4. Comparisons of the plant-specific calculations with all available dosimetry results from the Ar kansas Nuclear One Unit 2 surveillance program. WCAP-18 1 6 9-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 15 of 88 Westinghouse Non-Proprietary Class 3 2-4 The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operationa l or geometric variables that impact power reactor calculations. The second phase of the qualification (H.B. Robinson comparisons) addressed uncertainties in these additional a reas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations.
The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Arkansas Nuclear One Unit 2 analysis was established from results of these three phases of the methods qualification.
The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Arkansas Nuclear One Unit 2 analysis was established from results of these three phases of the methods qualification.
The fourth phase of the uncertainty assessment (comparisons with Arkansas Nuclear One Unit 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule or pressure vessel neutron exposures.
The fourth phase of the uncertainty assessment (comparisons with Arkansas Nuclear One Unit 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule or pressure vessel neutron exposures.
Table 2-8 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference  
Table 2-8 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference
: 6. The net calculational uncertainty was determined by combining the individual components in quadrature.
: 6. The net calculational uncertainty was determined by combining the individual components in quadrature.
Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons given in Appendix A of Reference 21 support th e se uncertainty assessments for Arkansas Nuclear One Unit 2. WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 16 of 88 Westinghouse Non-Proprietary C la ss 3 Table 2-1 Pressure Vessel Material Locations Axial Location Ma t e rial R ela ti ve to Core Midplane at O cm (cm) Inl et Nozzle to Upper Shell Welds -Lowest Extent Nozzle I 301.625 (*) Nozzle 2 30 1.625 (*) Nozzle 3 30 1.625 (*) Nozzle 4 301.625 (*) Outlet Nozz l e to Upper Shell We ld s -Lowest Extent Nozzle I 301.625 (*) Nozzle 2 301.625<*> U pp er Shell to Intermediate She ll Ci r cumferent i a l Weld 8-203 248.722 Intermediate S h ell Plates C-8009-1 , -2 , 33.973 to 248.722 Int ermediate S h ell Longitudina l Welds 2-203 A -33.973 to 248.722 2-203 B -33.973 to 248.722 2-203 C -33.973 to 248.722 Intermediate S h ell to Lower Shell C i rcumferentia l Weld 9-203 -33.973 Lower Shell P l ates C-8 010-1 , -2, 3 1 5.892 to -33.973 Lower Shell Longitudinal Welds 3-203 A -3 I 5.892 to -33.973 3-2 03 B -315.892 to -33.973 3-203 C -3 I 5.892 to -33.973 Lower Shell to Bottom Head Circumferential Weld 10-203 -315.892 No te s: Azimut h al Location (degrees) 60 120 240 300 0 180 0 to 360 0 to 360 (b) 90 210 330 0 to 360 0 to 360(c) 90 210 330 0 to 360 (a) This axia l l ocation correspo nd s to the bottom of the vesse l s upport pad of t h e inl et nozzle , instead of the nozzle to upper shell we ld. T hi s provides a bounding fluence for the nozzle to upp er shell we ld. (b) Intermediate shell plates C-8009-1 , -2 , and -3 extend from azimut h al angles of330&deg; to 90&deg;, 90&deg; to 210&deg;, and 210&deg; to 330&deg;, r e spectively. (c) Lower she ll plates C-8 0 1 0-1 , -2, and -3 extend from azimuthal angles of210&deg; to 330&deg;, 330&deg; to 90&deg;, and 90&deg; to 210&deg;, respectively.
Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons given in Appendix A of Reference 21 support th e se uncertainty assessments for Arkansas Nuclear One Unit 2. WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 16 of 88 Westinghouse Non-Proprietary C la ss 3 Table 2-1 Pressure Vessel Material Locations Axial Location Ma t e rial R ela ti ve to Core Midplane at O cm (cm) Inl et Nozzle to Upper Shell Welds -Lowest Extent Nozzle I 301.625 (*) Nozzle 2 30 1.625 (*) Nozzle 3 30 1.625 (*) Nozzle 4 301.625 (*) Outlet Nozz l e to Upper Shell We ld s -Lowest Extent Nozzle I 301.625 (*) Nozzle 2 301.625<*> U pp er Shell to Intermediate She ll Ci r cumferent i a l Weld 8-203 248.722 Intermediate S h ell Plates C-8009-1 , -2 , 33.973 to 248.722 Int ermediate S h ell Longitudina l Welds 2-203 A -33.973 to 248.722 2-203 B -33.973 to 248.722 2-203 C -33.973 to 248.722 Intermediate S h ell to Lower Shell C i rcumferentia l Weld 9-203 -33.973 Lower Shell P l ates C-8 010-1 , -2, 3 1 5.892 to -33.973 Lower Shell Longitudinal Welds 3-203 A -3 I 5.892 to -33.973 3-2 03 B -315.892 to -33.973 3-203 C -3 I 5.892 to -33.973 Lower Shell to Bottom Head Circumferential Weld 10-203 -315.892 No te s: Azimut h al Location (degrees) 60 120 240 300 0 180 0 to 360 0 to 360 (b) 90 210 330 0 to 360 0 to 360(c) 90 210 330 0 to 360 (a) This axia l l ocation correspo nd s to the bottom of the vesse l s upport pad of t h e inl et nozzle , instead of the nozzle to upper shell we ld. T hi s provides a bounding fluence for the nozzle to upp er shell we ld. (b) Intermediate shell plates C-8009-1 , -2 , and -3 extend from azimut h al angles of330&deg; to 90&deg;, 90&deg; to 210&deg;, and 210&deg; to 330&deg;, r e spectively. (c) Lower she ll plates C-8 0 1 0-1 , -2, and -3 extend from azimuthal angles of210&deg; to 330&deg;, 330&deg; to 90&deg;, and 90&deg; to 210&deg;, respectively.
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and Identification Number<*>
and Identification Number<*>
Wt.% Wt.% Wt.% Cu Ni Mn Reactor Vessel Beltline Materials Intermediate Shell Plate C-8 009-1 C8161-3 0.098 0.605 1.35 Intermediate Shell Plate C-8009-2 C8161-l 0.085 0.600 1.37 Intermediate Shell Plate C-8 009-3 C8182-2 0.096 0.580 1.36 Lower Shell Pl ate C-8 010-1 C8161-2 0.085 0.585 1.33 Lower Shell Plate C-8 010-2 B 2545-l 0.083 0.668 1.36 Lower Shell Plate C-80 I 0-3 B2545-2 0.080 0.653 1.36 Intermediate Shell Longitudinal Welds Multiple (d) o.05<d> 1.oo<d> l.16 (d) 2-2 0 3A, B , & C Lower Shell Longitudinal Welds 10120 0.046 (e) 0.08zC 0> l.21 3-203A , B , & C Intermed i ate to Lower She ll Girth 83650 0.045<*) 0.08i 0> 1.24 Weld 9-203 Reactor Vessel Extended Beltline Materials Uooer Shell Plat e G> C-8 008-1 C8182-I 0.13 0.60 1.36 Uppe r Shell Plate C-80 08-2 C7605-l 0.13 0.55 1.36 Uooe r Shell Plate C-8008-3 C8571-2 0.08 0.55 l.29 Upper She ll Longitudinal Welds BOLA 0.02 0.93 1.02 l-203A , B , & C Upper to Intermediate She ll Girth 10137 0.22 0.02 0.94 6329637 0.21 0.11<1) 1.22 Weld 8-203 FAGA 0.03 0.95 1.00 Surveillance Weld Data<il Arka n sas N ucl ear One Unit 2 83650 0.045 0.083 1.33 Ca l vert Cliffs Unit 2 101 37 0.2 1 0.06 ---Millstone Unit 2 0.2 1 0.06 ---J.M. Farley Unit 2 BOLA 0.028 0.89 ---Notes on foll o wing page. WCA P-1 8 1 69-NP Wt.% p 0.010 0.010 0.012 0.009 0.00 8 0.008 o.014<d> 0.01 2 0.006 0.011 0.013 0.014 0.010 0.015 0.011 0.008 0.007 ---------Fracture Toughness Property Initial RT N D T (OF) -1.4 0.5 o.o<s) 1 2.0 -16.7 -22.6 5 6 .40<&) 1 2.2 60.5 27.3 -60 (g) 56 -24<hJ ------------June2018 Revision I (c)
Wt.% Wt.% Wt.% Cu Ni Mn Reactor Vessel Beltline Materials Intermediate Shell Plate C-8 009-1 C8161-3 0.098 0.605 1.35 Intermediate Shell Plate C-8009-2 C8161-l 0.085 0.600 1.37 Intermediate Shell Plate C-8 009-3 C8182-2 0.096 0.580 1.36 Lower Shell Pl ate C-8 010-1 C8161-2 0.085 0.585 1.33 Lower Shell Plate C-8 010-2 B 2545-l 0.083 0.668 1.36 Lower Shell Plate C-80 I 0-3 B2545-2 0.080 0.653 1.36 Intermediate Shell Longitudinal Welds Multiple (d) o.05<d> 1.oo<d> l.16 (d) 2-2 0 3A, B , & C Lower Shell Longitudinal Welds 10120 0.046 (e) 0.08zC 0> l.21 3-203A , B , & C Intermed i ate to Lower She ll Girth 83650 0.045<*) 0.08i 0> 1.24 Weld 9-203 Reactor Vessel Extended Beltline Materials Uooer Shell Plat e G> C-8 008-1 C8182-I 0.13 0.60 1.36 Uppe r Shell Plate C-80 08-2 C7605-l 0.13 0.55 1.36 Uooe r Shell Plate C-8008-3 C8571-2 0.08 0.55 l.29 Upper She ll Longitudinal Welds BOLA 0.02 0.93 1.02 l-203A , B , & C Upper to Intermediate She ll Girth 10137 0.22 0.02 0.94 6329637 0.21 0.11<1) 1.22 Weld 8-203 FAGA 0.03 0.95 1.00 Surveillance Weld Data<il Arka n sas N ucl ear One Unit 2 83650 0.045 0.083 1.33 Ca l vert Cliffs Unit 2 101 37 0.2 1 0.06 ---Millstone Unit 2 0.2 1 0.06 ---J.M. Farley Unit 2 BOLA 0.028 0.89 ---Notes on foll o wing page. WCA P-1 8 1 69-NP Wt.% p 0.010 0.010 0.012 0.009 0.00 8 0.008 o.014<d> 0.01 2 0.006 0.011 0.013 0.014 0.010 0.015 0.011 0.008 0.007 ---------Fracture Toughness Property Initial RT N D T (OF) -1.4 0.5 o.o<s) 1 2.0 -16.7 -22.6 5 6 .40<&) 1 2.2 60.5 27.3 -60 (g) 56 -24<hJ ------------June2018 Revision I (c)
Notes: CALC-AN02-EP-17-00002 R e v O Page 29 of 88 Westinghouse Non-Proprietary Class 3 3-3 (a) The reactor vessel plate and weld material identification and heat numbers were taken from the Arkansas Nuclear One Unit 2 Certified Materia l Test Reports (CMTRs) and/or A-PENG-ER-002  
Notes: CALC-AN02-EP-17-00002 R e v O Page 29 of 88 Westinghouse Non-Proprietary Class 3 3-3 (a) The reactor vessel plate and weld material identification and heat numbers were taken from the Arkansas Nuclear One Unit 2 Certified Materia l Test Reports (CMTRs) and/or A-PENG-ER-002
[Ref. 1 1], unless otherwise noted. (b) All chemistry va l ues obtained from A-PENG-ER-002 and/or the Arkansas Nuclear One Unit 2 CMTRs, u nl ess otherwise noted. Chemistry values for plates are t h e average of all available analyses. Chemistry values for welds are the average of all coated electrode deposit chemistry (CE D C) for the E-8018 stick electrodes or weld flux deposit chemistry (WFDC) for Linde 0091 welds, unless otherwise noted. Where avai l able, additional c h emistry analysis results from BMI-0584 [Ref. 1 2] were also included in the average. The chemistry va l ues for the beltline plates reported in this table are identical to those previously reported in the Arkansas Nuclear One Unit 2 License Renewal Application (LRA) [Ref. 13] and the previous capsule r eport, BAW-2399, Revision I [Ref. 14]. (c) The RT N OT (UJ values for the plates are based on drop-weight data , longitudinally-oriented Charpy V-notch test data and NUREG-0800 , BTP 5-3 Position 1.1(3)(a) and (b) [Ref. 1 0], with the more limiting RTNoT(UJ value being se l ected , un l ess otherwise noted. The RT NOT (UJ values for welds are t h e generic value for Linde 0091 flux type welds (-56&deg;F) per IO CFR 50.61 [Ref. 15], unless otherwise noted. (d) The material Heat numbers for the Intermediate Shell Longitudinal Welds 2-203A, B, & Care unclear in the historical data. For conservatism, the material properties for the Intermediate Shell Longitudinal Welds 2-203A , B , & C reported are the most limiting values from welds relevant to the Intermediate Shell Longitudinal Welds 2-203A , B, & C per A-PENG-ER-002. These welds include Heat# 10120, Flux Type Linde 0091 , Lot# 3999 (sister plant weld), Heat# 10120 , 10120, Heat# AAGC, and the ana l ysis of the in-process weld depos i t chemistry. (e) The Cu and Ni wt. % values for the Lower Shell Longitudinal Welds 3-203A , B , & C and the Intermediate to Lower Shell Girth Weld 9-203 are consistent with BAW-2399 , Revision I [Ref. 14] and were originally taken from CE NPSD-1039, Revision 2 [Ref. 16]. (t) Weld Heat# 6329637 does not contain any WFDC Ni wt.% values , thus t h e bare wire chemical analysis (BWCA) value of 0.11 % from A-PENG-ER-002  
[Ref. 1 1], unless otherwise noted. (b) All chemistry va l ues obtained from A-PENG-ER-002 and/or the Arkansas Nuclear One Unit 2 CMTRs, u nl ess otherwise noted. Chemistry values for plates are t h e average of all available analyses. Chemistry values for welds are the average of all coated electrode deposit chemistry (CE D C) for the E-8018 stick electrodes or weld flux deposit chemistry (WFDC) for Linde 0091 welds, unless otherwise noted. Where avai l able, additional c h emistry analysis results from BMI-0584 [Ref. 1 2] were also included in the average. The chemistry va l ues for the beltline plates reported in this table are identical to those previously reported in the Arkansas Nuclear One Unit 2 License Renewal Application (LRA) [Ref. 13] and the previous capsule r eport, BAW-2399, Revision I [Ref. 14]. (c) The RT N OT (UJ values for the plates are based on drop-weight data , longitudinally-oriented Charpy V-notch test data and NUREG-0800 , BTP 5-3 Position 1.1(3)(a) and (b) [Ref. 1 0], with the more limiting RTNoT(UJ value being se l ected , un l ess otherwise noted. The RT NOT (UJ values for welds are t h e generic value for Linde 0091 flux type welds (-56&deg;F) per IO CFR 50.61 [Ref. 15], unless otherwise noted. (d) The material Heat numbers for the Intermediate Shell Longitudinal Welds 2-203A, B, & Care unclear in the historical data. For conservatism, the material properties for the Intermediate Shell Longitudinal Welds 2-203A , B , & C reported are the most limiting values from welds relevant to the Intermediate Shell Longitudinal Welds 2-203A , B, & C per A-PENG-ER-002. These welds include Heat# 10120, Flux Type Linde 0091 , Lot# 3999 (sister plant weld), Heat# 10120 , 10120, Heat# AAGC, and the ana l ysis of the in-process weld depos i t chemistry. (e) The Cu and Ni wt. % values for the Lower Shell Longitudinal Welds 3-203A , B , & C and the Intermediate to Lower Shell Girth Weld 9-203 are consistent with BAW-2399 , Revision I [Ref. 14] and were originally taken from CE NPSD-1039, Revision 2 [Ref. 16]. (t) Weld Heat# 6329637 does not contain any WFDC Ni wt.% values , thus t h e bare wire chemical analysis (BWCA) value of 0.11 % from A-PENG-ER-002
[Ref. 11] was used. (g) Th i s RT N OT (U) value for the surveillance plate, Intermediate Shell Plate C-8009-3 , is based on drop-weig h t data, transverse orientation Charpy V-notch test data taken from the baseline capsule test report , TR-MCD-002  
[Ref. 11] was used. (g) Th i s RT N OT (U) value for the surveillance plate, Intermediate Shell Plate C-8009-3 , is based on drop-weig h t data, transverse orientation Charpy V-notch test data taken from the baseline capsule test report , TR-MCD-002
[Ref. 17] and ASME Code Section Ill Subarticle NB-2331 [Ref. 9]. The RT N OT (UJ value for the two weld materials, Heat numbers 83650 and BOLA, is based on drop-weight data , Charpy V-notch test data taken from A-PENG-ER-002 and ASME Code Section Ill Subarticle NB-233 1 [Ref. 9]. (h) Drop-weight test data is not available for this E-8018 weld heat. Therefore, to assign an RT N OT (UJ value to this E-8018 stick electrode weld, the data in Table 8 of A-PENG-ER-002 was analyzed. The average T NOT value for the 17 E-8018 weld heats is -57"F with a s tandard deviation of 16.5"F. This yields a bounding value of -24"F using a mean plus two sigma model; therefore, a value of -24"F is acceptable for the initial RT NOT value of this weld material , with consideration that its Charpy impact energy at IO"F, which i s less than TNoT + 60"F , was greater than 100 ft-lb. Furthermore, -24&deg;F bounds all of the E-8018 stick electrode T NOT values present in A-PENG-ER-002  
[Ref. 17] and ASME Code Section Ill Subarticle NB-2331 [Ref. 9]. The RT N OT (UJ value for the two weld materials, Heat numbers 83650 and BOLA, is based on drop-weight data , Charpy V-notch test data taken from A-PENG-ER-002 and ASME Code Section Ill Subarticle NB-233 1 [Ref. 9]. (h) Drop-weight test data is not available for this E-8018 weld heat. Therefore, to assign an RT N OT (UJ value to this E-8018 stick electrode weld, the data in Table 8 of A-PENG-ER-002 was analyzed. The average T NOT value for the 17 E-8018 weld heats is -57"F with a s tandard deviation of 16.5"F. This yields a bounding value of -24"F using a mean plus two sigma model; therefore, a value of -24"F is acceptable for the initial RT NOT value of this weld material , with consideration that its Charpy impact energy at IO"F, which i s less than TNoT + 60"F , was greater than 100 ft-lb. Furthermore, -24&deg;F bounds all of the E-8018 stick electrode T NOT values present in A-PENG-ER-002
[Ref. 11). (i) Surveillance data exists for weld Heat# 83650, # IO 137 , and# BOLA from multiple sou r ces; see Section 4 for more detai l s. The data for Arkansas Nuclear One Unit 2 weld metal Heat# 83650 was taken as the average of the data available from MCD-002 , as well as the subsequent analyses completed during testing of the first capsule , BMl-0584 [Ref. 12]. The data for the Calvert Cliffs Unit 2 weld metal Heat# IO 137 was taken from Table 4-3 of WCAP-1750 I-NP [Ref. 18). The data for the Millstone Unit 2 weld metal Heat# 10137 was taken from Table 4-1 ofWCAP-16012  
[Ref. 11). (i) Surveillance data exists for weld Heat# 83650, # IO 137 , and# BOLA from multiple sou r ces; see Section 4 for more detai l s. The data for Arkansas Nuclear One Unit 2 weld metal Heat# 83650 was taken as the average of the data available from MCD-002 , as well as the subsequent analyses completed during testing of the first capsule , BMl-0584 [Ref. 12]. The data for the Calvert Cliffs Unit 2 weld metal Heat# IO 137 was taken from Table 4-3 of WCAP-1750 I-NP [Ref. 18). The data for the Millstone Unit 2 weld metal Heat# 10137 was taken from Table 4-1 ofWCAP-16012
[Ref. 19]. The data for the J.M. Farley Unit 2 weld metal Heat# BOLA was taken from Table 4-1 of WCAP-16918, Revision I (Ref. 20]. U) Upper Shell Plate C-8008-1 shares the same material heat number as t h e Arkansas Nuclear One Unit 2 surveillance p l ate material , Intermediate Shell Plate C-8009-3; therefore, surveillance program test results apply to this material as well. WCAP-18169-NP June 2018 Revision I Table 3-2 CALC-AN02-EP-17-00002 Rev O Page 30 of 88 Westinghouse Non-Proprietary Class 3 3-4 Summary of Arkansas Nuclear One Unit 2 Reactor Vessel Closure Heads, Vessel Flange and Balance of RCS Initial RT NDT Values Reactor Vessel Material Initial RT NDT Methodology (OF) Current C lo sure Head 10 BTP 5-3, Paragraph B1.1(3)(a) and (b) [Ref. 10] (Heat# 125B404) Re p lacement Closure Head<*J -22 ASME Code, Section III , Subsection NB-2300 (Heat# R378 l/R3782) [Ref. 9] Vessel Flange 30 BTP 5-3, Paragraph B1.1(3)(a) and (b) [Ref. 10] (Heat# 122A440) Balance of Rcs Cb> 50 Note (b) N otes: (a) A replacement, si ngle forging, closure head has been fabricated; however , it has not yet been insta lled at Arkansas Nuclear One Unit 2. The vessel flange initial RT N oT value is higher than both the current and replacement clo sure head initial RT NDT values. Thus , the results contained herein are conservative for the current and replacement closure heads. (b) 50&deg;F was conservatively assigned to all RCS material not specifically te s ted p e r Section 5.2.4.3 of the Arkansas Nuclear One Unit 2 updated Final Saf ety Analysis Report (FSAR). WCAP-18169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 31 of 88 Westinghouse Non-Proprietary C l ass 3 4 S U RVEILLANCE DATA 4-1 Per Regulatory Guide 1.99, Revision 2 [Ref. 1 ], calc ul ation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, data from survei llance programs at other plants which include a reactor vessel beltline or extended beltline material sho uld also be considered when calculating Position 2.1 chemistry factors. Data from a surveillance program a t another plant is often called 'sister plant' data. The survei ll ance caps ule plate material for Arkansas Nuclear One Unit 2 is from Intermediate Shell Plate C-80 09-3. Surveillance results from this plate also apply to Upper Shell Plate C-8008-1 , because the two plates wer e made from the same heat of material (Heat # C8182). The surveillance capsule weld material for Arkansas Nuclear One Unit 2 is Heat # 83650, which is applicable to the intermediate to lower shell girth weld. Table 4-1 summarizes the Arkansas Nuclear One Unit 2 surveillance data for the plate materia l an d we ld ma ter ia l (Heat # 83650) that will be used in the calc ulation of the Position 2.1 chemistry factor val ues for these materials.
[Ref. 19]. The data for the J.M. Farley Unit 2 weld metal Heat# BOLA was taken from Table 4-1 of WCAP-16918, Revision I (Ref. 20]. U) Upper Shell Plate C-8008-1 shares the same material heat number as t h e Arkansas Nuclear One Unit 2 surveillance p l ate material , Intermediate Shell Plate C-8009-3; therefore, surveillance program test results apply to this material as well. WCAP-18169-NP June 2018 Revision I Table 3-2 CALC-AN02-EP-17-00002 Rev O Page 30 of 88 Westinghouse Non-Proprietary Class 3 3-4 Summary of Arkansas Nuclear One Unit 2 Reactor Vessel Closure Heads, Vessel Flange and Balance of RCS Initial RT NDT Values Reactor Vessel Material Initial RT NDT Methodology (OF) Current C lo sure Head 10 BTP 5-3, Paragraph B1.1(3)(a) and (b) [Ref. 10] (Heat# 125B404) Re p lacement Closure Head<*J -22 ASME Code, Section III , Subsection NB-2300 (Heat# R378 l/R3782) [Ref. 9] Vessel Flange 30 BTP 5-3, Paragraph B1.1(3)(a) and (b) [Ref. 10] (Heat# 122A440) Balance of Rcs Cb> 50 Note (b) N otes: (a) A replacement, si ngle forging, closure head has been fabricated; however , it has not yet been insta lled at Arkansas Nuclear One Unit 2. The vessel flange initial RT N oT value is higher than both the current and replacement clo sure head initial RT NDT values. Thus , the results contained herein are conservative for the current and replacement closure heads. (b) 50&deg;F was conservatively assigned to all RCS material not specifically te s ted p e r Section 5.2.4.3 of the Arkansas Nuclear One Unit 2 updated Final Saf ety Analysis Report (FSAR). WCAP-18169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 31 of 88 Westinghouse Non-Proprietary C l ass 3 4 S U RVEILLANCE DATA 4-1 Per Regulatory Guide 1.99, Revision 2 [Ref. 1 ], calc ul ation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, data from survei llance programs at other plants which include a reactor vessel beltline or extended beltline material sho uld also be considered when calculating Position 2.1 chemistry factors. Data from a surveillance program a t another plant is often called 'sister plant' data. The survei ll ance caps ule plate material for Arkansas Nuclear One Unit 2 is from Intermediate Shell Plate C-80 09-3. Surveillance results from this plate also apply to Upper Shell Plate C-8008-1 , because the two plates wer e made from the same heat of material (Heat # C8182). The surveillance capsule weld material for Arkansas Nuclear One Unit 2 is Heat # 83650, which is applicable to the intermediate to lower shell girth weld. Table 4-1 summarizes the Arkansas Nuclear One Unit 2 surveillance data for the plate materia l an d we ld ma ter ia l (Heat # 83650) that will be used in the calc ulation of the Position 2.1 chemistry factor val ues for these materials.
The results of the last withdrawn and tested surveillance capsule , Caps ul e 28 4&deg;, were documented in WCAP-18166-NP  
The results of the last withdrawn and tested surveillance capsule , Caps ul e 28 4&deg;, were documented in WCAP-18166-NP
[Ref. 21]. Appendix D of WCAP-18166-NP concluded that the surveillance plate and weld (Hea t # 83650) data are credi bl e; therefore , a reduced margin term will b e utilized in t he ART calculations contained in Section 7. The Arkansas Nuclear One Unit 2 reactor vessel upper to intermediate shell girth weld seam was fab r i cated using weld Heat# 10137. Weld Heat # 10137 is contained in the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs.
[Ref. 21]. Appendix D of WCAP-18166-NP concluded that the surveillance plate and weld (Hea t # 83650) data are credi bl e; therefore , a reduced margin term will b e utilized in t he ART calculations contained in Section 7. The Arkansas Nuclear One Unit 2 reactor vessel upper to intermediate shell girth weld seam was fab r i cated using weld Heat# 10137. Weld Heat # 10137 is contained in the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs.
Thus, the Ca l vert C li ffs Unit 2 and Millstone Unit 2 data will be used in the calculation of the Position 2.1 chemistry factor val u e for Arkansas Nuclear One Unit 2 weld Heat # 10 13 7. Note that no s urveillance data is available for the other two Heats (# 6329637 and # FAGA) which were also used to make the Upper to Intermediate Shell Girth Weld 8-203. Table 4-2 su mmarize s the applica bl e surveillance capsule data pertaining to weld Heat # 10137. The combined surve illanc e data is deemed credi ble per Append i x D; however, as a result of the Mi ll stone Unit 2 surve illanc e data including both weld Heat # 10137 and 90136, the Position 2.1 chemistry factor calc ulation s for wel d Heat# 10137 will utilize a full margin term for conservatism.
Thus, the Ca l vert C li ffs Unit 2 and Millstone Unit 2 data will be used in the calculation of the Position 2.1 chemistry factor val u e for Arkansas Nuclear One Unit 2 weld Heat # 10 13 7. Note that no s urveillance data is available for the other two Heats (# 6329637 and # FAGA) which were also used to make the Upper to Intermediate Shell Girth Weld 8-203. Table 4-2 su mmarize s the applica bl e surveillance capsule data pertaining to weld Heat # 10137. The combined surve illanc e data is deemed credi ble per Append i x D; however, as a result of the Mi ll stone Unit 2 surve illanc e data including both weld Heat # 10137 and 90136, the Position 2.1 chemistry factor calc ulation s for wel d Heat# 10137 will utilize a full margin term for conservatism.
See Appendix D for details. The Arkan s as Nuclear One Unit 2 reactor vesse l upper she ll lon gitudinal we ld seams were fabricated using weld Heat # BOLA. Weld Heat # BOLA is contained in the J.M. Farley Uni t 2 surveillance program. Thus, the J.M. Farley Unit 2 data will b e used in the calculation of the Position 2.1 chemistry facto r va lu e for Arkansas Nuclear One Unit 2 weld Heat# BOLA. Table 4-3 summarizes the applicab l e surve illanc e capsule data pertaining to weld Heat # BOLA. Per Appendix D of WCAP-1 6918-NP, Revision 1 [Ref. 20], the J.M. Farley Unit 2 survei ll ance weld data is deemed non-credible. Since the J.M. Farley Unit 2 survei llance weld is not analyze d with any a dditional surveillance capsule material herein, this credibility conclusion is applicable to the Arkansas N ucl ear One Unit 2 weld Heat # BOLA. Therefore, a full margin term will be utilized in the ART calc ulation s containe d in Section 7. WCAP-1816 9-NP June2018 Re visio n 1 Westinghouse Non-Proprietary C l ass 3 4-2 Table 4-1 Arkansas Nuclear One Unit 2 Surveillance Capsule Data Material Capsule(*> Intermediate Shell Plate C-8009-3 (Longitudina l) 97&deg; 284&deg; 97&deg; Intermediate Shell Plate C-8009-3 (Transverse) 104&deg; 284&deg; 97&deg; Surveillance Weld Material (Heat# 836 50) 104&deg; 284&deg; No t e: (a) Surveillance data was tak e n from Table 5-10 of WCAP-18166-NP  
See Appendix D for details. The Arkan s as Nuclear One Unit 2 reactor vesse l upper she ll lon gitudinal we ld seams were fabricated using weld Heat # BOLA. Weld Heat # BOLA is contained in the J.M. Farley Uni t 2 surveillance program. Thus, the J.M. Farley Unit 2 data will b e used in the calculation of the Position 2.1 chemistry facto r va lu e for Arkansas Nuclear One Unit 2 weld Heat# BOLA. Table 4-3 summarizes the applicab l e surve illanc e capsule data pertaining to weld Heat # BOLA. Per Appendix D of WCAP-1 6918-NP, Revision 1 [Ref. 20], the J.M. Farley Unit 2 survei ll ance weld data is deemed non-credible. Since the J.M. Farley Unit 2 survei llance weld is not analyze d with any a dditional surveillance capsule material herein, this credibility conclusion is applicable to the Arkansas N ucl ear One Unit 2 weld Heat # BOLA. Therefore, a full margin term will be utilized in the ART calc ulation s containe d in Section 7. WCAP-1816 9-NP June2018 Re visio n 1 Westinghouse Non-Proprietary C l ass 3 4-2 Table 4-1 Arkansas Nuclear One Unit 2 Surveillance Capsule Data Material Capsule(*> Intermediate Shell Plate C-8009-3 (Longitudina l) 97&deg; 284&deg; 97&deg; Intermediate Shell Plate C-8009-3 (Transverse) 104&deg; 284&deg; 97&deg; Surveillance Weld Material (Heat# 836 50) 104&deg; 284&deg; No t e: (a) Surveillance data was tak e n from Table 5-10 of WCAP-18166-NP
[Ref. 21]. WCAP-18169-NP Capsule Fluence<*> (x 10 1 9 n/cm2, E > 1.0 MeV) 0.303 3.67 0.303 2.15 3.67 0.303 2.15 3.67 Measured 30 ft-lb Transition Temperature Shift<*> (&deg;F) 23.5 85.7 33.4 52.9 85.6 13.2 16.1 12.0 June 2018 Revi sio n 1 (") l> r (") )> z 0 "' I m I .... -.,j c:, 0 0 0 "' ::0 CD < 0 Ill IQ CD (,) "' 0 -co co Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 Ca l vert Cliffs Unit 2 and Millstone Unit 2 S u rveillance Capsule Data for We l d Heat# 10137 Ca p s ul e<*> C a p s ul e F lu e n ce<*> Meas ur ed 3 0 ft-l b Tra n siti on In l et Te mp eratu r e(bl Tem p e r at u re Material (x 10 19 n/cm2, E > 1.0 MeV) Temperat u re Shift<*> (&deg;F) Adj u st m ent<<) (&deg;F) {&deg;F) 263&deg; 0.825 72.7 550 -1.0 Calvert Cliffs Unit 2 Data 97&deg; 1.95 82.9 549 -2.0 104&deg; 2.44 69.7 548 -3.0 97&deg; 0.324 65.93 544.3 -6.7 Millstone Unit 2 104&deg; 0.949 52.12 547.6 -3.4 Data 83&deg; 1.74 56.09 548.0 -3.0 Notes: (a) For surveillance weld Heat# 10137 , data pertaining to Calvert Cliffs Unit 2 were taken from Table 5-10 of WCAP-17501-NP [Ref. 18]. Data pertaining to Millstone Unit 2 were taken from Table 5-10 ofWCAP-16012  
[Ref. 21]. WCAP-18169-NP Capsule Fluence<*> (x 10 1 9 n/cm2, E > 1.0 MeV) 0.303 3.67 0.303 2.15 3.67 0.303 2.15 3.67 Measured 30 ft-lb Transition Temperature Shift<*> (&deg;F) 23.5 85.7 33.4 52.9 85.6 13.2 16.1 12.0 June 2018 Revi sio n 1 (") l> r (") )> z 0 "' I m I .... -.,j c:, 0 0 0 "' ::0 CD < 0 Ill IQ CD (,) "' 0 -co co Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 Ca l vert Cliffs Unit 2 and Millstone Unit 2 S u rveillance Capsule Data for We l d Heat# 10137 Ca p s ul e<*> C a p s ul e F lu e n ce<*> Meas ur ed 3 0 ft-l b Tra n siti on In l et Te mp eratu r e(bl Tem p e r at u re Material (x 10 19 n/cm2, E > 1.0 MeV) Temperat u re Shift<*> (&deg;F) Adj u st m ent<<) (&deg;F) {&deg;F) 263&deg; 0.825 72.7 550 -1.0 Calvert Cliffs Unit 2 Data 97&deg; 1.95 82.9 549 -2.0 104&deg; 2.44 69.7 548 -3.0 97&deg; 0.324 65.93 544.3 -6.7 Millstone Unit 2 104&deg; 0.949 52.12 547.6 -3.4 Data 83&deg; 1.74 56.09 548.0 -3.0 Notes: (a) For surveillance weld Heat# 10137 , data pertaining to Calvert Cliffs Unit 2 were taken from Table 5-10 of WCAP-17501-NP [Ref. 18]. Data pertaining to Millstone Unit 2 were taken from Table 5-10 ofWCAP-16012
[Ref. 19). (b) Inlet temperatures were calculated as the average inlet temperature from all the previously completed cycles at the time of capsule withdrawal. (c) Temperature adjustment=
[Ref. 19). (b) Inlet temperatures were calculated as the average inlet temperature from all the previously completed cycles at the time of capsule withdrawal. (c) Temperature adjustment=
l .O*(T capsu t e -T p 1 a n1), where T p tan, = 55 l .0&deg;F for Arkansas Nuclear One Unit 2 (applied to the weld ~T N DT data for each of the Calvert Cliffs Unit 2 and M i llstone Un i t 2 capsules in the Position 2.1 chemistry factor calculation  
l .O*(T capsu t e -T p 1 a n1), where T p tan, = 55 l .0&deg;F for Arkansas Nuclear One Unit 2 (applied to the weld ~T N DT data for each of the Calvert Cliffs Unit 2 and M i llstone Un i t 2 capsules in the Position 2.1 chemistry factor calculation  
Line 190: Line 190:
= 12.1&deg;F Notes: (a) f= fluence. (b) FF= fluence factor= t<0*2 s-o.i o*i osf)_ (c) tiRT N DT values are the meas u red 3 0 ft-lb shift va l ues. The tiRT NDT values are adjusted using t h e ratio p roced u re to ac c ount for diffe r ences in the s u rveillance weld c h emistry and t h e reacto r vesse l we l d chemistry (pre-adj u sted values a r e listed in parentheses and were taken from Table 4-1 of this report). Ratio applied to the Arkansas N u clear One Unit 2 surveillance data= CF v ess el W e ld/ Cf surv. Weld= 34.1 &deg;F / 33.7&deg;F = 1.0 I. WCA P-1 8 1 6 9-NP June 20 1 8 R evisio n 1 _j Table 5-3 Weld Metal Heat# 10137 Calvert Cliffs Unit 2 Data Millstone Unit 2 Data (dJ Notes: (a) f= tluence. CALC-AN02-EP-17  
= 12.1&deg;F Notes: (a) f= fluence. (b) FF= fluence factor= t<0*2 s-o.i o*i osf)_ (c) tiRT N DT values are the meas u red 3 0 ft-lb shift va l ues. The tiRT NDT values are adjusted using t h e ratio p roced u re to ac c ount for diffe r ences in the s u rveillance weld c h emistry and t h e reacto r vesse l we l d chemistry (pre-adj u sted values a r e listed in parentheses and were taken from Table 4-1 of this report). Ratio applied to the Arkansas N u clear One Unit 2 surveillance data= CF v ess el W e ld/ Cf surv. Weld= 34.1 &deg;F / 33.7&deg;F = 1.0 I. WCA P-1 8 1 6 9-NP June 20 1 8 R evisio n 1 _j Table 5-3 Weld Metal Heat# 10137 Calvert Cliffs Unit 2 Data Millstone Unit 2 Data (dJ Notes: (a) f= tluence. CALC-AN02-EP-17  
-00002 Rev O Page 36 of 88 Westinghouse Non-Proprietary Class 3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat# 10137 Using Surveillance Capsule Data Capsule t<*> L\.RTNDT (c) FF*L\.RTNDT Capsule FF<bl (x 10 19 n/cmZ, E > 1.0 MeV) (OF) (OF) 263&deg; 0.8 25 0.9460 73.1 (72.7) 69.19 97&deg; 1.95 1.1825 82.5 (82.9) 97.58 104&deg; 2.44 1.2401 68.0 (69.7) 84.37 97&deg; 0.324 0.6902 60.4 (65.93) 41.70 104&deg; 0.949 0.9853 49.7 (52.12) 48.97 83&deg; 1.740 1.1523 54.2 (56.09) 62.40 SUM: 404.20 CF we ld Heat# 10137 = }.:(FF
-00002 Rev O Page 36 of 88 Westinghouse Non-Proprietary Class 3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat# 10137 Using Surveillance Capsule Data Capsule t<*> L\.RTNDT (c) FF*L\.RTNDT Capsule FF<bl (x 10 19 n/cmZ, E > 1.0 MeV) (OF) (OF) 263&deg; 0.8 25 0.9460 73.1 (72.7) 69.19 97&deg; 1.95 1.1825 82.5 (82.9) 97.58 104&deg; 2.44 1.2401 68.0 (69.7) 84.37 97&deg; 0.324 0.6902 60.4 (65.93) 41.70 104&deg; 0.949 0.9853 49.7 (52.12) 48.97 83&deg; 1.740 1.1523 54.2 (56.09) 62.40 SUM: 404.20 CF we ld Heat# 10137 = }.:(FF
* LlRT ND T).;.. l:(FF 2) = (404.20).;..  
* LlRT ND T).;.. l:(FF 2) = (404.20).;..
(6.606) = 61.2&deg;F 5-3 FF 2 0.895 1.398 1.538 0.476 0.971 1.328 6.606 (b) FF= tluence factor= t<02 B-o.io*io g I)_ (c) ti.RT NDT val ues ar e the measured 30 ft-lb s hift va lues. The ti.RT N DT values are adju s ted first by the difference in operating temperature then using the ratio procedure to account for differen ces in the s urv eilla nce weld chemistry a nd the reactor vessel weld chemistry (pre-adjuste d values are listed in parenthe ses and were taken from T a ble 4-2 of this report). The temperature adjustments are listed i n Table 4-2. Ratio applied to the Calvert Cliffs Unit 2 surveilla nc e data = CF vesse l Weld / CFsurv. We l d = 98.5&deg;F / 96.8&deg;F = 1.02. Ratio applied to the Millstone Un it 2 surveillance data = CF vesse l We l d I CF s urv. We ld= 98.5&deg;F / 96.8&deg;F = 1.02. (d) Mi ll s tone Uni t 2 s urv eillance data contain s s pecimens from both weld H eat# 10137 and weld Heat# 90136. However , thi s inclusion of a n additio nal heat is not expected to negatively impact the s ub seq uent r eactor vessel integrity calculation results, as additional conservatisms are in place. See Appendix D for more det a il s. WCAP-18169-NP June 2018 Re visio n 1 Table 5-4 CALC-AN02-EP-17-00002 Rev O Page 37 of 88 Westinghouse Non-Proprietary Class 3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat# BOLA Using Surveillance Capsule Data {c) 5-4 Weld Metal Capsule f'3> ~RT NDT FF*~RTNDT Heat#BOLA Capsule (x 10 19 n/cm2, E > 1.0 MeV) FF{bl (OF) (OF) FF 2 u 0.605 0.8593 o.o<d> (-28.4) 0.00 0.738 w 1.73 1.1508 o.o<d) (7.0) 0.00 1.324 J.M. Farley X 2.98 1.2891 o.o<d) (-15.6) 0.00 1.662 Unit2 z 4.92 1.3992 2.2 (10.2) 3.08 1.958 y 6.79 1.4579 61.1 (69.1) 89.08 2.125 V 8.73 1.4960 47.5 (56.5) 71.06 2.238 SUM: 163.22 10.046 CF Heat# BOLA= I:(FF * ~RT N DT) I:(FF 2) = (163.22) -c-(10.046) = 16.2&deg;F Notes: (a) f= fluence. (b) FF= fluence factor= t<0*2 s-O.IO'l ogfl_ (c) ~RT NDT va lues are the measured 30 ft-lb shift values. The ~RT NOT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the s urv eillance weld chemi s try and the reactor vessel weld chemistry (pre-adjusted va lu es are li s ted in parenthe ses and were taken from Table 4-3 of this report). The temperature adj u stments are listed in Table 4-3. A ratio of 1.00 was conservatively applied to the J.M. Farley Unit 2 surveillance data , s inc e CF vessel Weld< CFsurv. W e ld* (d) A negative ~RT NOT value was calculated afte r temperature adj u stment. Phy s ically , thi s s hould not occur; thus a conservative va lue of0.0&deg;F was u sed. WCAP-18169-NP June2018 Revision 1 CA LC-AN02-EP-17-00002 Rev O Page 38 of 88 Westinghouse Non-Proprietary Class 3 5-5 Table 5-5 Summary of Arkansas Nuclear One Unit 2 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Chemistry Factor (&deg;F) and Identification Number Heat Number Position 1.1 <*> Position 2.1 Reactor Vessel Beltline Materials Intermediate Shell Plate C-8009-1 C8161-3 63.6 ---Intermediate Shell Plate C-8009-2 C8161-l 54.5 ---Intermediate Shell Plate C-8009-3 C8182-2 62.2 55.8 (b) Lower Shell Plate C-8010-1 C8161-2 54.5 ---Lower Shell Plate C-8010-2 B2545-l 53.1 ---Lower Shell Plate C-8010-3 B2545-2 51.0 ---Inte rmedi a t e Shell Longitudinal Welds Multiple 68.0 ---2-203A, B, & C Lower Shell Longitudinal Welds 3-203A, B, & C 10120 34.0 ---Intermediate to Lower S hell Girth Weld 9-203 83650 34.1 12.1 (c) Reactor Vessel Extended Beltline Materials Upper Shell Plate C-8008-1 C8182-l 91.0 55.8 (b) Upper Shell Plate C-8008-2 C7605-l 89.5 ---Uppe r Shell Plate C-8008-3 C8571-2 51.0 ---Upper Shell Longitudinal Welds BOLA 27.0 16.2 (d) l-203A , B, & C 10137 98.5 6 l .2(e) Upper to Intermediate Shell Girth Weld 8-203 6329637 100.8 ---FAGA 41.0 ---Surveillance Weld Data Arkansas Nuclear One Unit 2 83650 33.7 ---Ca lv ert C liffs Unit 2 96.8 ---10137 Millstone Unit 2 96.8 ---J.M. Farley Unit 2 BOLA 38.2 ---Notes: (a) Position I.I ch e mistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of th i s report and Tables I and 2 of Regulatory Guide 1.99, Revi sion 2. (b) Position 2.1 chemistry factor was taken from Table 5-1 of this report. As discussed in Section 4, the surve ill a nc e plate data is credible a n d applicable to both Intermediate Shell Plate C-8009-3 and Upper Shell Plate C-8008-1. (c) Position 2.1 chemistry factor was taken from Table 5-2 of this report. As discussed in Section 4 , the surve illan ce weld data for Heat # 83650 is credi bl e. (d) Position 2.1 chemistry factor was taken from Table 5-4 of this report. As discussed in Section 4, the s urveillanc e weld data fo r Heat# BOLA is not credible. (e) Position 2.1 chemistry factor was taken from Table 5-3 of this report. As discussed in Section 4 , the survei llanc e weld data for Heat# l O 137 is credible; however no reduction in the margin term wi ll b e taken. WCAP-18169-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 3 9 of 88 Westinghouse Non-Proprietary Class 3 6 CRITE RI A F O R A L L O WA BL E P RE SSURE-TEM P ERAT URE RELATI O NSHIPS 6.1 O VERALL APPROACH 6-1 The ASME approach for calculating the allowa b le limit curves for vario u s heatup and coo l down rates specifies that the total stress intensity factor , K1, for the com b ined therma l and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress i n tensity factor, K, c , for the metal temperature at that time. K1c is o b tained from the refere n ce fracture to u ghness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Ref. 3]. The K1c curve is given by the following eq u ation: where , K 1 c (ksi-V in.) K l e =33 .2+ 20. 734 *e[0.02 (T-RT ND T )] (1) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ducti l ity temperature RT NDT This K1c curve is based on the l ower bound of static critica l K 1 val u es measured as a f unction of t emperature on s pecimens of SA-533 Grade B Class 1 , SA-508-1, SA-508-2, and SA-508-3 steel. 6.2 METHO DO LOGY F OR PRESSURE-TEM P E RAT URE LIMIT CU R VE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: w h e re, C C W C AP-1 81 6 9-NP stre s s intensity factor c a used by membrane (pres s ure) s tress stress intensity factor caused b y the therma l gradients (2) reference stress intensity factor as a function of the meta l temperature T and the metal r efe rence nil-ductility temperature RT N D T 2.0 for Leve l A and Level B service limits 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical June 2 018 Revi s ion I J CALC-AN02-EP-17-00002 Rev O Page 40 of 88 Westinghouse Non-Proprietary C l ass 3 For mem b rane tension, the corresponding K 1 for the postulated defect is: Kim= Mmx(pR/t) where, Mm for an inside axial surface flaw is given by: Mm 1.85 for Ji < 2 , Mm 0.926 Ji for 2::; Ji ::; 3.464, Mm 3.2 1 for Ji > 3.464 and, Mm for an outside axial surface flaw is given by: Mm 1. 77 for Ji < 2 , Mm 0.8 93 Ji for 2::; Ji ::; 3 .464, Mm 3.09 for Ji > 3.464 Similarly , Mm for an inside or an outside circumferential surface flaw is given b y: Mm 0.89 for Ji < 2, Mm 0.443 Ji for 2::; Ji ::; 3 .464 , Mm 1.53 for Ji > 3.464 Where: 6-2 (3) p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in). For bending stress, the corresponding K 1 for the postulated axial or circumferential defect is: K 1 b = Mb
(6.606) = 61.2&deg;F 5-3 FF 2 0.895 1.398 1.538 0.476 0.971 1.328 6.606 (b) FF= tluence factor= t<02 B-o.io*io g I)_ (c) ti.RT NDT val ues ar e the measured 30 ft-lb s hift va lues. The ti.RT N DT values are adju s ted first by the difference in operating temperature then using the ratio procedure to account for differen ces in the s urv eilla nce weld chemistry a nd the reactor vessel weld chemistry (pre-adjuste d values are listed in parenthe ses and were taken from T a ble 4-2 of this report). The temperature adjustments are listed i n Table 4-2. Ratio applied to the Calvert Cliffs Unit 2 surveilla nc e data = CF vesse l Weld / CFsurv. We l d = 98.5&deg;F / 96.8&deg;F = 1.02. Ratio applied to the Millstone Un it 2 surveillance data = CF vesse l We l d I CF s urv. We ld= 98.5&deg;F / 96.8&deg;F = 1.02. (d) Mi ll s tone Uni t 2 s urv eillance data contain s s pecimens from both weld H eat# 10137 and weld Heat# 90136. However , thi s inclusion of a n additio nal heat is not expected to negatively impact the s ub seq uent r eactor vessel integrity calculation results, as additional conservatisms are in place. See Appendix D for more det a il s. WCAP-18169-NP June 2018 Re visio n 1 Table 5-4 CALC-AN02-EP-17-00002 Rev O Page 37 of 88 Westinghouse Non-Proprietary Class 3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat# BOLA Using Surveillance Capsule Data {c) 5-4 Weld Metal Capsule f'3> ~RT NDT FF*~RTNDT Heat#BOLA Capsule (x 10 19 n/cm2, E > 1.0 MeV) FF{bl (OF) (OF) FF 2 u 0.605 0.8593 o.o<d> (-28.4) 0.00 0.738 w 1.73 1.1508 o.o<d) (7.0) 0.00 1.324 J.M. Farley X 2.98 1.2891 o.o<d) (-15.6) 0.00 1.662 Unit2 z 4.92 1.3992 2.2 (10.2) 3.08 1.958 y 6.79 1.4579 61.1 (69.1) 89.08 2.125 V 8.73 1.4960 47.5 (56.5) 71.06 2.238 SUM: 163.22 10.046 CF Heat# BOLA= I:(FF * ~RT N DT) I:(FF 2) = (163.22) -c-(10.046) = 16.2&deg;F Notes: (a) f= fluence. (b) FF= fluence factor= t<0*2 s-O.IO'l ogfl_ (c) ~RT NDT va lues are the measured 30 ft-lb shift values. The ~RT NOT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the s urv eillance weld chemi s try and the reactor vessel weld chemistry (pre-adjusted va lu es are li s ted in parenthe ses and were taken from Table 4-3 of this report). The temperature adj u stments are listed in Table 4-3. A ratio of 1.00 was conservatively applied to the J.M. Farley Unit 2 surveillance data , s inc e CF vessel Weld< CFsurv. W e ld* (d) A negative ~RT NOT value was calculated afte r temperature adj u stment. Phy s ically , thi s s hould not occur; thus a conservative va lue of0.0&deg;F was u sed. WCAP-18169-NP June2018 Revision 1 CA LC-AN02-EP-17-00002 Rev O Page 38 of 88 Westinghouse Non-Proprietary Class 3 5-5 Table 5-5 Summary of Arkansas Nuclear One Unit 2 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Chemistry Factor (&deg;F) and Identification Number Heat Number Position 1.1 <*> Position 2.1 Reactor Vessel Beltline Materials Intermediate Shell Plate C-8009-1 C8161-3 63.6 ---Intermediate Shell Plate C-8009-2 C8161-l 54.5 ---Intermediate Shell Plate C-8009-3 C8182-2 62.2 55.8 (b) Lower Shell Plate C-8010-1 C8161-2 54.5 ---Lower Shell Plate C-8010-2 B2545-l 53.1 ---Lower Shell Plate C-8010-3 B2545-2 51.0 ---Inte rmedi a t e Shell Longitudinal Welds Multiple 68.0 ---2-203A, B, & C Lower Shell Longitudinal Welds 3-203A, B, & C 10120 34.0 ---Intermediate to Lower S hell Girth Weld 9-203 83650 34.1 12.1 (c) Reactor Vessel Extended Beltline Materials Upper Shell Plate C-8008-1 C8182-l 91.0 55.8 (b) Upper Shell Plate C-8008-2 C7605-l 89.5 ---Uppe r Shell Plate C-8008-3 C8571-2 51.0 ---Upper Shell Longitudinal Welds BOLA 27.0 16.2 (d) l-203A , B, & C 10137 98.5 6 l .2(e) Upper to Intermediate Shell Girth Weld 8-203 6329637 100.8 ---FAGA 41.0 ---Surveillance Weld Data Arkansas Nuclear One Unit 2 83650 33.7 ---Ca lv ert C liffs Unit 2 96.8 ---10137 Millstone Unit 2 96.8 ---J.M. Farley Unit 2 BOLA 38.2 ---Notes: (a) Position I.I ch e mistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of th i s report and Tables I and 2 of Regulatory Guide 1.99, Revi sion 2. (b) Position 2.1 chemistry factor was taken from Table 5-1 of this report. As discussed in Section 4, the surve ill a nc e plate data is credible a n d applicable to both Intermediate Shell Plate C-8009-3 and Upper Shell Plate C-8008-1. (c) Position 2.1 chemistry factor was taken from Table 5-2 of this report. As discussed in Section 4 , the surve illan ce weld data for Heat # 83650 is credi bl e. (d) Position 2.1 chemistry factor was taken from Table 5-4 of this report. As discussed in Section 4, the s urveillanc e weld data fo r Heat# BOLA is not credible. (e) Position 2.1 chemistry factor was taken from Table 5-3 of this report. As discussed in Section 4 , the survei llanc e weld data for Heat# l O 137 is credible; however no reduction in the margin term wi ll b e taken. WCAP-18169-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 3 9 of 88 Westinghouse Non-Proprietary Class 3 6 CRITE RI A F O R A L L O WA BL E P RE SSURE-TEM P ERAT URE RELATI O NSHIPS 6.1 O VERALL APPROACH 6-1 The ASME approach for calculating the allowa b le limit curves for vario u s heatup and coo l down rates specifies that the total stress intensity factor , K1, for the com b ined therma l and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress i n tensity factor, K, c , for the metal temperature at that time. K1c is o b tained from the refere n ce fracture to u ghness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Ref. 3]. The K1c curve is given by the following eq u ation: where , K 1 c (ksi-V in.) K l e =33 .2+ 20. 734 *e[0.02 (T-RT ND T )] (1) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ducti l ity temperature RT NDT This K1c curve is based on the l ower bound of static critica l K 1 val u es measured as a f unction of t emperature on s pecimens of SA-533 Grade B Class 1 , SA-508-1, SA-508-2, and SA-508-3 steel. 6.2 METHO DO LOGY F OR PRESSURE-TEM P E RAT URE LIMIT CU R VE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: w h e re, C C W C AP-1 81 6 9-NP stre s s intensity factor c a used by membrane (pres s ure) s tress stress intensity factor caused b y the therma l gradients (2) reference stress intensity factor as a function of the meta l temperature T and the metal r efe rence nil-ductility temperature RT N D T 2.0 for Leve l A and Level B service limits 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical June 2 018 Revi s ion I J CALC-AN02-EP-17-00002 Rev O Page 40 of 88 Westinghouse Non-Proprietary C l ass 3 For mem b rane tension, the corresponding K 1 for the postulated defect is: Kim= Mmx(pR/t) where, Mm for an inside axial surface flaw is given by: Mm 1.85 for Ji < 2 , Mm 0.926 Ji for 2::; Ji ::; 3.464, Mm 3.2 1 for Ji > 3.464 and, Mm for an outside axial surface flaw is given by: Mm 1. 77 for Ji < 2 , Mm 0.8 93 Ji for 2::; Ji ::; 3 .464, Mm 3.09 for Ji > 3.464 Similarly , Mm for an inside or an outside circumferential surface flaw is given b y: Mm 0.89 for Ji < 2, Mm 0.443 Ji for 2::; Ji ::; 3 .464 , Mm 1.53 for Ji > 3.464 Where: 6-2 (3) p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in). For bending stress, the corresponding K 1 for the postulated axial or circumferential defect is: K 1 b = Mb
* Maximum Stress, where Mb is two-thirds of Mm (4) The maximum K 1 produced by radial thermal gradient for the postulated axial or circumferential inside surface defect ofG-2120 is: K11 = 0.953x10-3 x CR x t 2*5 (5) where CR i s the cooldown rate in &deg;F/hr., or for a postulated axial or circumferential outside su r face defect K11 = 0.753x10-3 x HU x t 25 where HU is the heatup rate in &deg;F /hr. WCAP-18169-NP (6) June2018 Revis i on 1 CALC-AN02-EP-17-00002 Rev O Page 41 of 88 Westinghouse Non-Propr ie tary Class 3 6-3 The throu g h-wall temperature difference associated with the maximum thermal K 1 can be determined fromASME Code, Section XI , Appendix G, Fig. G-2214-1.
* Maximum Stress, where Mb is two-thirds of Mm (4) The maximum K 1 produced by radial thermal gradient for the postulated axial or circumferential inside surface defect ofG-2120 is: K11 = 0.953x10-3 x CR x t 2*5 (5) where CR i s the cooldown rate in &deg;F/hr., or for a postulated axial or circumferential outside su r face defect K11 = 0.753x10-3 x HU x t 25 where HU is the heatup rate in &deg;F /hr. WCAP-18169-NP (6) June2018 Revis i on 1 CALC-AN02-EP-17-00002 Rev O Page 41 of 88 Westinghouse Non-Propr ie tary Class 3 6-3 The throu g h-wall temperature difference associated with the maximum thermal K 1 can be determined fromASME Code, Section XI , Appendix G, Fig. G-2214-1.
Line 228: Line 228:
-100&deg;F/hr. T (OF) P (psig) T (&deg;F) P (psig) T (&deg;F) P (psig) T (&deg;F) P (psig) 60 0 6 0 0 6 0 0 6 0 0 60 622 6 0 62 2 6 0 622 6 0 5 76 65 6 22 65 622 65 6 22 65 5 94 70 622 7 0 6 2 2 7 0 622 7 0 613 75 622 75 622 75 622 75 622 8 0 622 8 0 62 2 8 0 6 22 8 0 622 85 622 85 622 85 622 8 5 6 22 90 6 22 9 0 6 22 9 0 622 90 622 95 622 95 622 9 5 622 95 622 1 00 6 22 100 622 1 0 0 622 100 6 22 1 05 6 22 10 5 622 105 622 10 5 622 1 10 622 110 6 2 2 110 622 llO 6 22 115 622 11 5 622 115 622 11 5 622 1 20 6 22 1 2 0 622 1 20 62 2 1 2 0 622 125 622 1 25 622 1 25 622 1 2 5 622 130 6 22 1 3 0 622 13 0 62 2 1 3 0 62 2 135 622 1 35 622 135 622 1 35 622 1 40 622 140 6 22 1 4 0 622 140 622 145 622 14 5 622 145 622 145 622 150 622 1 5 0 622 1 50 622 1 5 0 622 150 1 321 1 5 0 1 32 1 150 1 32 1 1 5 0 1 3 21 1 55 1 393 1 55 1 393 1 55 1 393 15 5 13 93 160 14 74 1 6 0 14 7 4 160 147 4 1 60 14 7 4 1 65 1 562 1 65 1 5 62 165 1 562 1 6 5 15 6 2 1 70 1 660 1 7 0 1 66 0 1 70 1 66 0 1 7 0 1 66 0 1 75 1 768 17 5 1768 175 1 768 1 75 1 768 1 80 1 888 1 8 0 1888 1 80 1 888 1 8 0 1 888 1 85 2020 1 8 5 2 0 2 0 1 85 202 0 1 8 5 2 0 2 0 1 90 2 1 66 1 9 0 2 1 6 6 190 2 1 66 1 9 0 2 1 6 6 1 95 2 3 28 1 95 2 3 28 195 2328 1 95 2328 8-8 WCA P-1 8 1 69-NP Jun e 2 01 8 R evisio n 1 T a ble 8-3 WCAP-18 1 6 9-NP CALC-AN02-EP-17-00002 Rev O Page 60 of 88 Wes t inghouse No n-Proprie t ary C l ass 3 A rkansas Nucl e ar One Unit 2 54 EFPY lns e rvice H y dro s tatic and Le ak T e st Cur ve Data Point s u s ing the 1998 through the 2000 A ddenda A pp. G Methodolog y (w/ K 1c, w/ Flange and LST Requirements , a nd w/o M argins for Instrumentation E rrors) T (&deg;F) P (psi g) 60 0 60 622 65 622 70 622 75 622 80 622 85 622 90 622 95 622 100 622 105 622 110 622 115 622 120 622 125 622 130 622 135 622 140 622 145 622 150 622 150 176 1 155 1858 160 1965 162 2000 165 2083 170 22 1 4 175 2358 179 2 485 8-9 J une 20 1 8 Revision l CA L C-A N02-EP-17  
-100&deg;F/hr. T (OF) P (psig) T (&deg;F) P (psig) T (&deg;F) P (psig) T (&deg;F) P (psig) 60 0 6 0 0 6 0 0 6 0 0 60 622 6 0 62 2 6 0 622 6 0 5 76 65 6 22 65 622 65 6 22 65 5 94 70 622 7 0 6 2 2 7 0 622 7 0 613 75 622 75 622 75 622 75 622 8 0 622 8 0 62 2 8 0 6 22 8 0 622 85 622 85 622 85 622 8 5 6 22 90 6 22 9 0 6 22 9 0 622 90 622 95 622 95 622 9 5 622 95 622 1 00 6 22 100 622 1 0 0 622 100 6 22 1 05 6 22 10 5 622 105 622 10 5 622 1 10 622 110 6 2 2 110 622 llO 6 22 115 622 11 5 622 115 622 11 5 622 1 20 6 22 1 2 0 622 1 20 62 2 1 2 0 622 125 622 1 25 622 1 25 622 1 2 5 622 130 6 22 1 3 0 622 13 0 62 2 1 3 0 62 2 135 622 1 35 622 135 622 1 35 622 1 40 622 140 6 22 1 4 0 622 140 622 145 622 14 5 622 145 622 145 622 150 622 1 5 0 622 1 50 622 1 5 0 622 150 1 321 1 5 0 1 32 1 150 1 32 1 1 5 0 1 3 21 1 55 1 393 1 55 1 393 1 55 1 393 15 5 13 93 160 14 74 1 6 0 14 7 4 160 147 4 1 60 14 7 4 1 65 1 562 1 65 1 5 62 165 1 562 1 6 5 15 6 2 1 70 1 660 1 7 0 1 66 0 1 70 1 66 0 1 7 0 1 66 0 1 75 1 768 17 5 1768 175 1 768 1 75 1 768 1 80 1 888 1 8 0 1888 1 80 1 888 1 8 0 1 888 1 85 2020 1 8 5 2 0 2 0 1 85 202 0 1 8 5 2 0 2 0 1 90 2 1 66 1 9 0 2 1 6 6 190 2 1 66 1 9 0 2 1 6 6 1 95 2 3 28 1 95 2 3 28 195 2328 1 95 2328 8-8 WCA P-1 8 1 69-NP Jun e 2 01 8 R evisio n 1 T a ble 8-3 WCAP-18 1 6 9-NP CALC-AN02-EP-17-00002 Rev O Page 60 of 88 Wes t inghouse No n-Proprie t ary C l ass 3 A rkansas Nucl e ar One Unit 2 54 EFPY lns e rvice H y dro s tatic and Le ak T e st Cur ve Data Point s u s ing the 1998 through the 2000 A ddenda A pp. G Methodolog y (w/ K 1c, w/ Flange and LST Requirements , a nd w/o M argins for Instrumentation E rrors) T (&deg;F) P (psi g) 60 0 60 622 65 622 70 622 75 622 80 622 85 622 90 622 95 622 100 622 105 622 110 622 115 622 120 622 125 622 130 622 135 622 140 622 145 622 150 622 150 176 1 155 1858 160 1965 162 2000 165 2083 170 22 1 4 175 2358 179 2 485 8-9 J une 20 1 8 Revision l CA L C-A N02-EP-17  
-0 0002 Rev O Page 61 of 88 Westinghouse Non-Proprietary Class 3 9 REFERENCES 9-1 1. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research , Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials ," May 1988. 2. Westin g house Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpr e ssure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004. 3. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure." 4. Code o f Federal Regulations, 10 CFR Part 50, Appendix G , "Frac ture Toughness Requirements," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19 , 1995. 5. Regula t ory Guide 1.190 , "Ca lculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, March 2001. 6. Westin g hou se Report WCAP-16083-NP, Revision 1, " Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Do s imetry," April 2013. 7. RSICC Data Library Collection DLC-1 8 5 , "BUGLE-96:
-0 0002 Rev O Page 61 of 88 Westinghouse Non-Proprietary Class 3 9 REFERENCES 9-1 1. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research , Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials ," May 1988. 2. Westin g house Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpr e ssure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004. 3. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure." 4. Code o f Federal Regulations, 10 CFR Part 50, Appendix G , "Frac ture Toughness Requirements," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19 , 1995. 5. Regula t ory Guide 1.190 , "Ca lculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, March 2001. 6. Westin g hou se Report WCAP-16083-NP, Revision 1, " Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Do s imetry," April 2013. 7. RSICC Data Library Collection DLC-1 8 5 , "BUGLE-96:
Coupled 47 Neutron, 20 Gamma-Ray Group Cross-S ection Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applic a tions ," March 1996. 8. NRC Regulatory Issue Summary 2014-11, " Information on Licensing Applicat i ons for Fracture Toughness Requirements for Ferri tic Reactor Coolant Pressure Boundary Components," U.S. Nuclear Regulatory Co mmission , October 2014. [Ag e nc ywide Do cu m ent Management System (ADAMS) Accessi o n Numbe r ML14149Al65}  
Coupled 47 Neutron, 20 Gamma-Ray Group Cross-S ection Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applic a tions ," March 1996. 8. NRC Regulatory Issue Summary 2014-11, " Information on Licensing Applicat i ons for Fracture Toughness Requirements for Ferri tic Reactor Coolant Pressure Boundary Components," U.S. Nuclear Regulatory Co mmission , October 2014. [Ag e nc ywide Do cu m ent Management System (ADAMS) Accessi o n Numbe r ML14149Al65}
: 9. ASME Boiler and Pressure Vessel (B&PV) Code, Section III, Division 1 , Subsection NB, "C lass 1 Compo n ents." 10. NUREG-0800 , Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, C hapter 5 L WR Edition, Branch Technical Po sit ion 5-3, "Fracture Toughness Requirements
: 9. ASME Boiler and Pressure Vessel (B&PV) Code, Section III, Division 1 , Subsection NB, "C lass 1 Compo n ents." 10. NUREG-0800 , Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, C hapter 5 L WR Edition, Branch Technical Po sit ion 5-3, "Fracture Toughness Requirements
," Revision 2, U.S. Nuclear Regulatory Commission, Ma r ch 2007. 11. Combu s tion E ngineering Report A-PENG-ER-002 , Revision 0 , "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the ANO 2 Reactor Pressure Vessel Plates, Forgings, Welds a nd Cladding," October 1995. 12. Battelle -Columbus Report BMI-0584 , "F inal Report on Examination, Testing, and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Arkansas Nuclear One Unit 2 Generating Plant to Arkansas Power and Light Company ," May 1984. 13. Arkans a s Nuclear One -Unit 2 License Renewal Application, October 2003. [Ava ilable on th e N RC website} 14. AREV A NP, Inc. Report BAW-2399 , Revision 1 , "Ana lysis of Capsule W-104 Entergy Operations , Inc. Ar k ansas Nuclear One Unit 2 Power Plant Reactor Vessel Material Surveillance Program ," February 2005. 15. Code of Federal Regulations , 10 CFR 50.61 , " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register , Volume 60, No. 243, dated December 19 , 199 5, effective January 18 , 1996. 16. Combu s tion Engineering Owners Group Report CE NPSD-1039, Revision 2, "Best Estimate Copper and Nic k el Values in CE Fabricated Reactor Vessel Welds," June 1997. WCAP-18169-NP June2018 Revi s ion 1 r CALC-AN02-EP-17  
," Revision 2, U.S. Nuclear Regulatory Commission, Ma r ch 2007. 11. Combu s tion E ngineering Report A-PENG-ER-002 , Revision 0 , "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the ANO 2 Reactor Pressure Vessel Plates, Forgings, Welds a nd Cladding," October 1995. 12. Battelle -Columbus Report BMI-0584 , "F inal Report on Examination, Testing, and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Arkansas Nuclear One Unit 2 Generating Plant to Arkansas Power and Light Company ," May 1984. 13. Arkans a s Nuclear One -Unit 2 License Renewal Application, October 2003. [Ava ilable on th e N RC website} 14. AREV A NP, Inc. Report BAW-2399 , Revision 1 , "Ana lysis of Capsule W-104 Entergy Operations , Inc. Ar k ansas Nuclear One Unit 2 Power Plant Reactor Vessel Material Surveillance Program ," February 2005. 15. Code of Federal Regulations , 10 CFR 50.61 , " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register , Volume 60, No. 243, dated December 19 , 199 5, effective January 18 , 1996. 16. Combu s tion Engineering Owners Group Report CE NPSD-1039, Revision 2, "Best Estimate Copper and Nic k el Values in CE Fabricated Reactor Vessel Welds," June 1997. WCAP-18169-NP June2018 Revi s ion 1 r CALC-AN02-EP-17  
-00002 Rev O Page 62 of 88 Westinghouse Non-Proprietary Class 3 9-2 17. Combustion Engineering Report TR-MCD-002, "Arkansas Power & Light Arkansas Nuclear One -Unit 2 Eva luation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program," March 1976. 18. Westin g house Report WCAP-17501-NP, Revision 0, "Ana lysis of Capsule 104&deg; from the Calvert Cliffs Unit No. 2 Reactor Vessel Radiation Surveillance Program," February 2012. 19. Westin g house Report WCAP-16012, Revision 0, "Analysis of Capsule W-83 from the Dominion Nucle ar Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003. 20. Westin g house Report WCAP-16918-NP, Revision 1, "Analysis of Capsule V from the Southern Nuclea r Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program," April 2008. 21. Westin g house Report WCAP-18166-NP, Revision 0 , "Ana ly sis of Capsule 284&deg; from the Entergy Operations, Inc. Arkansas Nuclear One Unit 2 Reactor Vessel Radiation Surveillance Program," September 2016. 22. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Hando ut s, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. [ADAMS Accession Number MLJ 10070570]  
-00002 Rev O Page 62 of 88 Westinghouse Non-Proprietary Class 3 9-2 17. Combustion Engineering Report TR-MCD-002, "Arkansas Power & Light Arkansas Nuclear One -Unit 2 Eva luation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program," March 1976. 18. Westin g house Report WCAP-17501-NP, Revision 0, "Ana lysis of Capsule 104&deg; from the Calvert Cliffs Unit No. 2 Reactor Vessel Radiation Surveillance Program," February 2012. 19. Westin g house Report WCAP-16012, Revision 0, "Analysis of Capsule W-83 from the Dominion Nucle ar Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003. 20. Westin g house Report WCAP-16918-NP, Revision 1, "Analysis of Capsule V from the Southern Nuclea r Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program," April 2008. 21. Westin g house Report WCAP-18166-NP, Revision 0 , "Ana ly sis of Capsule 284&deg; from the Entergy Operations, Inc. Arkansas Nuclear One Unit 2 Reactor Vessel Radiation Surveillance Program," September 2016. 22. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Hando ut s, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. [ADAMS Accession Number MLJ 10070570]
: 23. RSICC Computer Code Collection CCC-650, "DOORS 3.2: One, Two-and Three Dimensional Discret e Ordinates Neutron/Photon Transport Code System," April 1998. WCAP-1 8169-NP June2018 Revision I I I l_ CALC-AN02-EP-17-00002 Rev O Page 63 of 88 Westinghouse Non-Proprietary Class 3 A-1 A P PENDIX A THERMAL STRESS INTENSITY FACTORS (Ku) Tables A-1 and A-2 contain the thermal stress intensity factors (K 11) for the maximum heatup and cooldown rates at 54 EFPY for Arkansas Nuclear One Unit 2. The reactor vessel cylindrical shell radii to the l/4T and 3/4T locations are as follows:
: 23. RSICC Computer Code Collection CCC-650, "DOORS 3.2: One, Two-and Three Dimensional Discret e Ordinates Neutron/Photon Transport Code System," April 1998. WCAP-1 8169-NP June2018 Revision I I I l_ CALC-AN02-EP-17-00002 Rev O Page 63 of 88 Westinghouse Non-Proprietary Class 3 A-1 A P PENDIX A THERMAL STRESS INTENSITY FACTORS (Ku) Tables A-1 and A-2 contain the thermal stress intensity factors (K 11) for the maximum heatup and cooldown rates at 54 EFPY for Arkansas Nuclear One Unit 2. The reactor vessel cylindrical shell radii to the l/4T and 3/4T locations are as follows:
* l/4T Radius= 81.688 inches
* l/4T Radius= 81.688 inches
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The cooldown transient is analyzed as it results in tensile stresses at the inside surface of t h e nozzle comer. A 1/4 T axial flaw is postulated at the inside s urface of the reactor vessel nozzle comer and stress intensity factors are determined based on the rounded curvature of the nozzle geometry.
The cooldown transient is analyzed as it results in tensile stresses at the inside surface of t h e nozzle comer. A 1/4 T axial flaw is postulated at the inside s urface of the reactor vessel nozzle comer and stress intensity factors are determined based on the rounded curvature of the nozzle geometry.
The allowable pressure is then calculated based on the fracture toughness of the nozzle material and the stress intensity factors for the l/4T flaw. B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The fracture toughness (Kie) used for the inlet and outlet nozzle material is defined in Appendix G of the Section XI ASME Code, as discussed in Section 6 of this report. The K,c fracture toughness curve is dependent on the Adjusted Reference Temperature (ART) value for irradiated materials.
The allowable pressure is then calculated based on the fracture toughness of the nozzle material and the stress intensity factors for the l/4T flaw. B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The fracture toughness (Kie) used for the inlet and outlet nozzle material is defined in Appendix G of the Section XI ASME Code, as discussed in Section 6 of this report. The K,c fracture toughness curve is dependent on the Adjusted Reference Temperature (ART) value for irradiated materials.
The ART values for the inle t and outlet nozzle materials are determined using the methodology contained in Regulatory Guide 1.99, Revision 2 [Ref. B-3], which is described in Section 7 of this report, and weight percent (wt. %) copper (Cu) and nickel (Ni), initial RT NDT value, and projected neutron fluence as inputs. The material p ro perties for each of the reactor vessel inlet and outlet nozzle forging materials are documented in Table B-1 and a summary of the limiting inlet and outlet nozzle ART values for Arkansas Nucle ar One Unit 2 is pr ese nted in Table B-2. Nozzle Material Properties The Arkansas Nuclear One Unit 2 nozzle material properties are provided in Tabl e B-1. Nickel (Ni), Manganese (Mn), and Phosphorus (P) weight percent (wt. %) values were obtained as the average of the material-specific analyses documented in Combustion Engineering reportA-PENG-ER-002  
The ART values for the inle t and outlet nozzle materials are determined using the methodology contained in Regulatory Guide 1.99, Revision 2 [Ref. B-3], which is described in Section 7 of this report, and weight percent (wt. %) copper (Cu) and nickel (Ni), initial RT NDT value, and projected neutron fluence as inputs. The material p ro perties for each of the reactor vessel inlet and outlet nozzle forging materials are documented in Table B-1 and a summary of the limiting inlet and outlet nozzle ART values for Arkansas Nucle ar One Unit 2 is pr ese nted in Table B-2. Nozzle Material Properties The Arkansas Nuclear One Unit 2 nozzle material properties are provided in Tabl e B-1. Nickel (Ni), Manganese (Mn), and Phosphorus (P) weight percent (wt. %) values were obtained as the average of the material-specific analyses documented in Combustion Engineering reportA-PENG-ER-002
[Ref. B-4] for WCAP-18169-NP June2018 Re v i si on l CALC-AN02-EP-17  
[Ref. B-4] for WCAP-18169-NP June2018 Re v i si on l CALC-AN02-EP-17  
-00002 Rev O Page 67 of 88 Westinghouse Non-Proprietary Class 3 B-2 each of the Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzles. Copper weight percent values for each of the Arkansas Nuclear One Unit 2 outlet nozzles were also taken to be the average of all available analyses contained in Combustion Engineering report A-PENG-ER-002  
-00002 Rev O Page 67 of 88 Westinghouse Non-Proprietary Class 3 B-2 each of the Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzles. Copper weight percent values for each of the Arkansas Nuclear One Unit 2 outlet nozzles were also taken to be the average of all available analyses contained in Combustion Engineering report A-PENG-ER-002
[Ref. B-4]. However , the A-PENG-ER-002 report did not contain copper weight percent values for the inlet nozzles, because at the time that the Arkansas Nuclear One Unit 2 nozzles were manufactured, these values were not required to be documented for SA-508, Class 2 low-alloy steel. Therefore, no material-specific copper weight percent value is available for the Arkansas Nuclear One Unit 2 inlet nozzles. Per NRC RIS 2014-11 [Ref. B-1], a copper weight percent value is not required for calculation of the Arkansas Nuclear One Unit 2 nozzle material ART values , because the nozzles have fluence values less than 1 x 10 1 7 n/cm 2* However, if a copper weight percent value is ever needed, a best-estimate copper weight percent value is available from Section 4 of the NRC-approved Boiling Water Reactor Vessel and Internals Project (BWRVlP [proprietary])
[Ref. B-4]. However , the A-PENG-ER-002 report did not contain copper weight percent values for the inlet nozzles, because at the time that the Arkansas Nuclear One Unit 2 nozzles were manufactured, these values were not required to be documented for SA-508, Class 2 low-alloy steel. Therefore, no material-specific copper weight percent value is available for the Arkansas Nuclear One Unit 2 inlet nozzles. Per NRC RIS 2014-11 [Ref. B-1], a copper weight percent value is not required for calculation of the Arkansas Nuclear One Unit 2 nozzle material ART values , because the nozzles have fluence values less than 1 x 10 1 7 n/cm 2* However, if a copper weight percent value is ever needed, a best-estimate copper weight percent value is available from Section 4 of the NRC-approved Boiling Water Reactor Vessel and Internals Project (BWRVlP [proprietary])
report, BWRVlP-173-A  
report, BWRVlP-173-A
[Ref. B-5], and this value could be utilized for the Arkansas Nuclear One inlet nozzles. A mean plus two standard deviations methodology was applied to the data in BWRVIP-173-A to determine a conservative copper weight percent value. The data in the BWRVIP report was tabulated from an industry-wide database of SA-508, Class 2 forging materials.
[Ref. B-5], and this value could be utilized for the Arkansas Nuclear One inlet nozzles. A mean plus two standard deviations methodology was applied to the data in BWRVIP-173-A to determine a conservative copper weight percent value. The data in the BWRVIP report was tabulated from an industry-wide database of SA-508, Class 2 forging materials.
The Charpy V-Notch forging specimen orientation for the inlet and outlet nozzles was not reported in PENG-ER-002
The Charpy V-Notch forging specimen orientation for the inlet and outlet nozzles was not reported in PENG-ER-002
; thus , it was conservatively assumed that the orientation was the "strong direction" for each nozzle forging. The initial RT NOT values were therefore determined for each of the Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzle forging materials using the Branch Technical Position (BTP) 5-3 , Position 1.1(3) methodology  
; thus , it was conservatively assumed that the orientation was the "strong direction" for each nozzle forging. The initial RT NOT values were therefore determined for each of the Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzle forging materials using the Branch Technical Position (BTP) 5-3 , Position 1.1(3) methodology
[Ref. B-6]. The initial RT N OT values for all of the nozzle materials were determined directly from the data or by using a CVGRAPH, Version 6.02 hyperbolic tangent curve fit through the minimum data points , in accordance with ASME Code Section III, Subarticle NB-2331, Paragraph (a)(4) [Ref. B-7]. The initial RTNoT values were determined using both BTP 5-3 Position 1.1(3)(a) and Position 1.1(3)(b), and the more limiting initial RT N o T value was chosen for each nozzle forging material.
[Ref. B-6]. The initial RT N OT values for all of the nozzle materials were determined directly from the data or by using a CVGRAPH, Version 6.02 hyperbolic tangent curve fit through the minimum data points , in accordance with ASME Code Section III, Subarticle NB-2331, Paragraph (a)(4) [Ref. B-7]. The initial RTNoT values were determined using both BTP 5-3 Position 1.1(3)(a) and Position 1.1(3)(b), and the more limiting initial RT N o T value was chosen for each nozzle forging material.
The Arkansas Nuclear One Unit 2 initial RT N O T values for the inlet and outlet nozzles materials are summarized in Table B-1. Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > 1 MeV) exposure of the Arkansas Nuclear One Unit 2 reactor vessel materials is discussed in Section 2 of this report. The fluence values used in the inlet and outlet nozzle ART calculations were calculated at the lowest extent of the nozzles (i.e., the nozzle to nozzle shell weld locations) and were chosen at an elevation lower than the actual elevation of the postulated flaw, which is at the inside comer of the nozzle , for conservatism.
The Arkansas Nuclear One Unit 2 initial RT N O T values for the inlet and outlet nozzles materials are summarized in Table B-1. Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > 1 MeV) exposure of the Arkansas Nuclear One Unit 2 reactor vessel materials is discussed in Section 2 of this report. The fluence values used in the inlet and outlet nozzle ART calculations were calculated at the lowest extent of the nozzles (i.e., the nozzle to nozzle shell weld locations) and were chosen at an elevation lower than the actual elevation of the postulated flaw, which is at the inside comer of the nozzle , for conservatism.
Line 257: Line 257:
WCAP-18169-NP June2018 Revision 1 CALC-AN02-EP-17  
WCAP-18169-NP June2018 Revision 1 CALC-AN02-EP-17  
-00002 Rev O Page 69 of 88 Westinghouse Non-Proprietary Class 3 B-4 B.2 NOZZLE COOLDOWN PRESSURE-TEMPERATURE LIMITS Allowable pressures are determined for a given temperature based on the fracture toughness of the limiting nozzle material along with the appropriate pressure and thermal stress intensity factors. The Arkansas Nuclear One Unit 2 nozzle fracture toughness used to determine the P-T limits is calculated using the limiting inlet and outlet nozzle ART values from Table B-2. The stress intensity factor correlations used for the nozzle comers are provided in ORNL study , ORNL/TM-2010
-00002 Rev O Page 69 of 88 Westinghouse Non-Proprietary Class 3 B-4 B.2 NOZZLE COOLDOWN PRESSURE-TEMPERATURE LIMITS Allowable pressures are determined for a given temperature based on the fracture toughness of the limiting nozzle material along with the appropriate pressure and thermal stress intensity factors. The Arkansas Nuclear One Unit 2 nozzle fracture toughness used to determine the P-T limits is calculated using the limiting inlet and outlet nozzle ART values from Table B-2. The stress intensity factor correlations used for the nozzle comers are provided in ORNL study , ORNL/TM-2010
/2 46 [Ref. B-8], and are consistent with ASME PVP2011-57015  
/2 46 [Ref. B-8], and are consistent with ASME PVP2011-57015
[Ref. B-9]. The methodology includes postulating an inside surface 1/4T nozzle comer flaw, and calculating through-wall nozzle comer stresses for a cooldown rate of 100&deg;F /hour. The through-wall stresses at the nozzle corner location were fitted based on a third-order polynomial of the form: where, CJ = through-wall stress distribution x = through-wall distance from inside surface Ao, A 1 , A 2 , A 3 = coefficients of polynomial fit for the third-order polynomial , used in the stress intensity factor expression discussed below The stress i ntensity factors generated for a rounded nozzle corner for the pressure and thermal gradient were calculated based on the methodology provided in ORNL/TM-2010
[Ref. B-9]. The methodology includes postulating an inside surface 1/4T nozzle comer flaw, and calculating through-wall nozzle comer stresses for a cooldown rate of 100&deg;F /hour. The through-wall stresses at the nozzle corner location were fitted based on a third-order polynomial of the form: where, CJ = through-wall stress distribution x = through-wall distance from inside surface Ao, A 1 , A 2 , A 3 = coefficients of polynomial fit for the third-order polynomial , used in the stress intensity factor expression discussed below The stress i ntensity factors generated for a rounded nozzle corner for the pressure and thermal gradient were calculated based on the methodology provided in ORNL/TM-2010
/246. The stress intensity factor expression for a rounded corner is: where, K 1 stress intensity factor for a circular corner crack on a nozzle with a rounded inner radius comer a crac k depth at the nozzle corner, for use with 1/4 T (25% of the wall thickness)
/246. The stress intensity factor expression for a rounded corner is: where, K 1 stress intensity factor for a circular corner crack on a nozzle with a rounded inner radius comer a crac k depth at the nozzle corner, for use with 1/4 T (25% of the wall thickness)
The Arkansas Nuclear One Unit 2 reactor vesse l inlet and outlet nozzle P-T limit curves are s hown in Figures B-1 and B-2, respectively, based on the stress intensity factor expression discussed above; also show n in these figures are the traditional beltline cooldown P-T limit curves from Figure 8-2. The nozzle P-T limit curves are provided for a cooldown rate of 100&deg;F/hr , along with a steady-state curve. An outside s urface flaw in the nozzle was not considered because the pressure stress is significantly lower at the outside surface than the inside surface. A heatup nozzle P-T limit curve is also not provided since it would be less limiting than the cooldown nozzle P-T limit curve in Figures B-1 and B-2 for an inside surface flaw. Additionally, the cooldown transient i s more limiting than the heatup transient since it results in tensile stresses at the inside surface of the nozzle corner. WCAP-18169-NP June2018 Revi s ion 1 Conclusion CALC-AN02-EP-17-00002 Rev O Page 70 of 88 Westinghouse Non-Proprietary Class 3 B-5 Based on the results shown in Figures B-1 and B-2 , it is concluded that the nozzle P-T limits are bounded by the traditional beltline curves. Therefore, the P-T limits provided in Section 8 for 54 EFPY remain limiting for the beltline and non-beltline reactor vessel components. W C AP-1816 9-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 71 of 88 Westin g h o u se Non-Propri e tary C la ss 3 -C) (/J Q. -2500 2250 2000 1750 1500 L I 1250 ----4~1 t-4----1 ::::s 1/) 1/) Q. E 1000 .2l 1/) >, (/J -r:: C'l:J 0 0 0 ... 0 750 I --+ I t -l-I I t, Cooldown Rate :E 500 -~~ -100&deg;F/Hr 0:: 250 Minimum Boltup Tern . = 60&deg;F Inlet Nozz l e Cooldown -lOO"F/Hr I r Inlet Nozzle Steady State Cooldown Rates steady-state  
The Arkansas Nuclear One Unit 2 reactor vesse l inlet and outlet nozzle P-T limit curves are s hown in Figures B-1 and B-2, respectively, based on the stress intensity factor expression discussed above; also show n in these figures are the traditional beltline cooldown P-T limit curves from Figure 8-2. The nozzle P-T limit curves are provided for a cooldown rate of 100&deg;F/hr , along with a steady-state curve. An outside s urface flaw in the nozzle was not considered because the pressure stress is significantly lower at the outside surface than the inside surface. A heatup nozzle P-T limit curve is also not provided since it would be less limiting than the cooldown nozzle P-T limit curve in Figures B-1 and B-2 for an inside surface flaw. Additionally, the cooldown transient i s more limiting than the heatup transient since it results in tensile stresses at the inside surface of the nozzle corner. WCAP-18169-NP June2018 Revi s ion 1 Conclusion CALC-AN02-EP-17-00002 Rev O Page 70 of 88 Westinghouse Non-Proprietary Class 3 B-5 Based on the results shown in Figures B-1 and B-2 , it is concluded that the nozzle P-T limits are bounded by the traditional beltline curves. Therefore, the P-T limits provided in Section 8 for 54 EFPY remain limiting for the beltline and non-beltline reactor vessel components. W C AP-1816 9-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 71 of 88 Westin g h o u se Non-Propri e tary C la ss 3 -C) (/J Q. -2500 2250 2000 1750 1500 L I 1250 ----4~1 t-4----1 ::::s 1/) 1/) Q. E 1000 .2l 1/) >, (/J -r:: C'l:J 0 0 0 ... 0 750 I --+ I t -l-I I t, Cooldown Rate :E 500 -~~ -100&deg;F/Hr 0:: 250 Minimum Boltup Tern . = 60&deg;F Inlet Nozz l e Cooldown -lOO"F/Hr I r Inlet Nozzle Steady State Cooldown Rates steady-state  
-25"F/Hr -60"F/Hr Acceptable 0 eration I r--0 50 100 150 200 250 300 350 400 450 500 550 Averaged Reactor Coolant System Temperature  
-25"F/Hr -60"F/Hr Acceptable 0 eration I r--0 50 100 150 200 250 300 350 400 450 500 550 Averaged Reactor Coolant System Temperature
(&deg;F) B-6 Figure B-1 Compari s on of Arkansas Nuclear One Unit 2 Beltline P-T Limit s to Inlet No z zle Limits WCA P-181 6 9-NP Jun e 2 018 R evisio n l CALC-AN02-EP-17-00002 Rev O Page 72 of 88 Westinghouse No n-Proprietary C l ass 3 2500 -,------,--,~-.,------,-----,-----
(&deg;F) B-6 Figure B-1 Compari s on of Arkansas Nuclear One Unit 2 Beltline P-T Limit s to Inlet No z zle Limits WCA P-181 6 9-NP Jun e 2 018 R evisio n l CALC-AN02-EP-17-00002 Rev O Page 72 of 88 Westinghouse No n-Proprietary C l ass 3 2500 -,------,--,~-.,------,-----,-----
,---,------, -C) en 0.. -2250 2000 1750 1500 I t 1250 -+----:::, (I) (I) 0.. E 1000 .! en -C: 750 0 0 0 ... 0 0 cu 500 Q) a::: 250 0 0 50 Outlet Nozzle M.l.------t----t Cooldown l -100&deg;F/Hr Outlet Nozzle ------t------f Steady State I t -+------------+---+ Cooldown Rate -100&deg;F/Hr Minimum Boltup Tern . = 60" F t Cooldown Rates Lowest Service Tern . = 150" F --+-Acceptable 0 eration ... 100 150 200 250 300 350 400 450 500 550 Averaged Reactor Coolant System Temperature  
,---,------, -C) en 0.. -2250 2000 1750 1500 I t 1250 -+----:::, (I) (I) 0.. E 1000 .! en -C: 750 0 0 0 ... 0 0 cu 500 Q) a::: 250 0 0 50 Outlet Nozzle M.l.------t----t Cooldown l -100&deg;F/Hr Outlet Nozzle ------t------f Steady State I t -+------------+---+ Cooldown Rate -100&deg;F/Hr Minimum Boltup Tern . = 60" F t Cooldown Rates Lowest Service Tern . = 150" F --+-Acceptable 0 eration ... 100 150 200 250 300 350 400 450 500 550 Averaged Reactor Coolant System Temperature
(&deg;F) B-7 F igure B-2 C ompari s on of Arkansas Nuclear One U nit 2 Beltline P-T Limit s to Outlet Nozzle Limits WCA P-181 6 9-NP Jun e 20 1 8 R evision 1 B.3 REFERENCES CALC-AN02-EP-17-00002 Rev O Page 7 3 of 88 Westinghouse Non-Proprietary Class 3 B-8 B-1 NRC Regu l atory Issue Summary 2014-11 , " Informat i on on Licensing Applications for Fracture To u ghness Requirements for Ferritic Reactor Coo l a nt Pressure Bo u ndary Components
(&deg;F) B-7 F igure B-2 C ompari s on of Arkansas Nuclear One U nit 2 Beltline P-T Limit s to Outlet Nozzle Limits WCA P-181 6 9-NP Jun e 20 1 8 R evision 1 B.3 REFERENCES CALC-AN02-EP-17-00002 Rev O Page 7 3 of 88 Westinghouse Non-Proprietary Class 3 B-8 B-1 NRC Regu l atory Issue Summary 2014-11 , " Informat i on on Licensing Applications for Fracture To u ghness Requirements for Ferritic Reactor Coo l a nt Pressure Bo u ndary Components
," U.S. Nuclear Reg ul atory Commission , Octo b er 2014. [A DAMS A c c es sion Numb e r ML 1 4 1 49AJ65] B-2 W e stinghouse Report WCAP-14040-A, Revision 4 , "Methodology Used to D eve l op Col d Ov e rpressure Mitigating System Setpoints and RCS Heatup and Coo l down Limit C u rves ," May 2004. B-3 U.S. Nuclear Regulatory Commission , Office of Nuclear Reg u latory Research , Reg ul atory Gui d e 1.99 , Revision 2 , " Rad i ation Embrittlement o f Reactor Vessel Materia l s ," May 1988. B-4 Co m bustion Engineering Report A-PENG-ER-002, R evision 0 , "The Reactor Vessel Group Re c ords Evaluation Program Phase II Fina l Report for th e ANO 2 Reactor Pressure Vessel Plates, Forgings , Welds and Cladding ," October 1995. B-5 BWR V IP-173-A: B W R V e ssel and Int e rnals Proj e ct: Evaluation of Ch e mi s try Data for BWR V esse l N o zz le Forging Mat e rials. EPRI, Pa l o Alto , CA: 2011. 1022835. B-6 NUREG-0800, Standard Review Plan for t h e Review of Safety Ana l ysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition , Branch Tec hni ca l Posi t ion 5-3, " Fracture Toughness Requir e ments ," Re v ision 2 , U.S. Nuclear Reg u latory Commission , March 2007. B-7 ASME Boiler and Pressure Vessel (B&PV) Code , Sect i on III , Division 1 , Subsection NB , " C l ass 1 Components." B-8 Oak Ridge N ational Laboratory Report , ORNL/TM-2 010/246 , " Stress and Frac tu re Mechanics An a lyses of Boiling Water Reactor an d Pressurized Water Reactor Pressure Vessel Nozzles -Revision 1 ," June 2012. [ADAMS Acc ession N umb e r ML 1 10060164]
," U.S. Nuclear Reg ul atory Commission , Octo b er 2014. [A DAMS A c c es sion Numb e r ML 1 4 1 49AJ65] B-2 W e stinghouse Report WCAP-14040-A, Revision 4 , "Methodology Used to D eve l op Col d Ov e rpressure Mitigating System Setpoints and RCS Heatup and Coo l down Limit C u rves ," May 2004. B-3 U.S. Nuclear Regulatory Commission , Office of Nuclear Reg u latory Research , Reg ul atory Gui d e 1.99 , Revision 2 , " Rad i ation Embrittlement o f Reactor Vessel Materia l s ," May 1988. B-4 Co m bustion Engineering Report A-PENG-ER-002, R evision 0 , "The Reactor Vessel Group Re c ords Evaluation Program Phase II Fina l Report for th e ANO 2 Reactor Pressure Vessel Plates, Forgings , Welds and Cladding ," October 1995. B-5 BWR V IP-173-A: B W R V e ssel and Int e rnals Proj e ct: Evaluation of Ch e mi s try Data for BWR V esse l N o zz le Forging Mat e rials. EPRI, Pa l o Alto , CA: 2011. 1022835. B-6 NUREG-0800, Standard Review Plan for t h e Review of Safety Ana l ysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition , Branch Tec hni ca l Posi t ion 5-3, " Fracture Toughness Requir e ments ," Re v ision 2 , U.S. Nuclear Reg u latory Commission , March 2007. B-7 ASME Boiler and Pressure Vessel (B&PV) Code , Sect i on III , Division 1 , Subsection NB , " C l ass 1 Components." B-8 Oak Ridge N ational Laboratory Report , ORNL/TM-2 010/246 , " Stress and Frac tu re Mechanics An a lyses of Boiling Water Reactor an d Pressurized Water Reactor Pressure Vessel Nozzles -Revision 1 ," June 2012. [ADAMS Acc ession N umb e r ML 1 10060164]
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The replacement closure head is considered in the cylindrical beltline P-T limit curves as described in Section 6 of this report. Furthermore , the replacement closure head has not undergone neutron embrittlement that would affect P-T limits. Therefore , no further consideration is necessary for this component with regards to P-T limits. The pressurizer was constructed to the 1968 Edition through 1970 Summer Addenda Section III ASME C ode and met all applicable requirements at the time of construction and is original to the plant. Furthermore , the pressurizer has not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for this component with regards to P-T limits. The replacement steam generators were constructed to the 1989 Edition Section III ASME Code and met a ll applicable requirements at the time of construction.
The replacement closure head is considered in the cylindrical beltline P-T limit curves as described in Section 6 of this report. Furthermore , the replacement closure head has not undergone neutron embrittlement that would affect P-T limits. Therefore , no further consideration is necessary for this component with regards to P-T limits. The pressurizer was constructed to the 1968 Edition through 1970 Summer Addenda Section III ASME C ode and met all applicable requirements at the time of construction and is original to the plant. Furthermore , the pressurizer has not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for this component with regards to P-T limits. The replacement steam generators were constructed to the 1989 Edition Section III ASME Code and met a ll applicable requirements at the time of construction.
Furthermore , the replacement steam generators have not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for these components with regards to P-T limits. C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission , Federal R e gister, Volume 60, No. 243 , December 19, 1995. C-2 ASME Boiler and Pressure Vessel (B&PV) Code , Section III , Division 1, Subsection NB , "Class 1 Components
Furthermore , the replacement steam generators have not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for these components with regards to P-T limits. C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission , Federal R e gister, Volume 60, No. 243 , December 19, 1995. C-2 ASME Boiler and Pressure Vessel (B&PV) Code , Section III , Division 1, Subsection NB , "Class 1 Components
." W C AP-18169-NP June 2018 Re v ision l CALC-AN02-EP-17-00002 Rev O Page 75 of 88 Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D CREDIBILITY EVALUATION OF THE WELD HEAT# 10137 SURVEILLANCE DATA D.1 INTRODUCTION Regulatory Guide 1. 99 , Revision 2 [Ref. D-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position 2.1 of Regulatory Guide 1.99 , Revision 2, describes the method fo r calculating the adjusted reference temperature of reactor vessel beltline materials using surveillance capsule data. The methods of Position 2.1 can only be applied when two or more credible surveillance data sets become available from the reactor in question. The credibility of all surveillance program data applicable to the Arkansas Nuclear One Unit 2 beltline was assessed in WCAP-18166-NP  
." W C AP-18169-NP June 2018 Re v ision l CALC-AN02-EP-17-00002 Rev O Page 75 of 88 Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D CREDIBILITY EVALUATION OF THE WELD HEAT# 10137 SURVEILLANCE DATA D.1 INTRODUCTION Regulatory Guide 1. 99 , Revision 2 [Ref. D-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position 2.1 of Regulatory Guide 1.99 , Revision 2, describes the method fo r calculating the adjusted reference temperature of reactor vessel beltline materials using surveillance capsule data. The methods of Position 2.1 can only be applied when two or more credible surveillance data sets become available from the reactor in question. The credibility of all surveillance program data applicable to the Arkansas Nuclear One Unit 2 beltline was assessed in WCAP-18166-NP
[Ref. D-2]. However, the Arkansas Nuclear One Unit 2 extended beltline contains two welds with sister plant data , the Upper Shell Longitudinal Welds l-203A , B , & C (Heat# BOLA) and the Upper to Intermediate Shell Girth Weld 8-203 (Heat# 1013 7). Note that no surveillance data is available for the other two Heats (# 6329637 and # FAGA) which were also used to make the Upper to Intermediate Shell Girth Weld 8-203. The weld Heat # BOLA sister plant data is available from the J.M. Farley Unit 2 surveillance program. Since this s urveillance data is analyzed by itself, the credibility conclusion documented in Appendix D of WCAP-16918, Revision 1 [Ref. D-3] is applicable to Arkansas Nuclear One Unit 2; thus, the credibility conclusion of the Heat# BOLA data need not be upd ated. The J.M. Farley Unit 2 surveillance weld data (Heat# BOLA) is non-credible in regard to the Arkansas Nuclear One Unit 2 reactor vessel materials. The weld Heat# 10137 sister plant data is available from both the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs.
[Ref. D-2]. However, the Arkansas Nuclear One Unit 2 extended beltline contains two welds with sister plant data , the Upper Shell Longitudinal Welds l-203A , B , & C (Heat# BOLA) and the Upper to Intermediate Shell Girth Weld 8-203 (Heat# 1013 7). Note that no surveillance data is available for the other two Heats (# 6329637 and # FAGA) which were also used to make the Upper to Intermediate Shell Girth Weld 8-203. The weld Heat # BOLA sister plant data is available from the J.M. Farley Unit 2 surveillance program. Since this s urveillance data is analyzed by itself, the credibility conclusion documented in Appendix D of WCAP-16918, Revision 1 [Ref. D-3] is applicable to Arkansas Nuclear One Unit 2; thus, the credibility conclusion of the Heat# BOLA data need not be upd ated. The J.M. Farley Unit 2 surveillance weld data (Heat# BOLA) is non-credible in regard to the Arkansas Nuclear One Unit 2 reactor vessel materials. The weld Heat# 10137 sister plant data is available from both the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs.
The Millstone Unit 2 surveillance program includes two distinct welds, Heat# 10137 an d Heat# 90136. In previous analyses, this weld surveillance data was treated as one combined we ld and subsequently analyzed together.
The Millstone Unit 2 surveillance program includes two distinct welds, Heat# 10137 an d Heat# 90136. In previous analyses, this weld surveillance data was treated as one combined we ld and subsequently analyzed together.
However, these two weld metal heats were not melted tog e ther into a tandem weld; they were individually deposited.
However, these two weld metal heats were not melted tog e ther into a tandem weld; they were individually deposited.
It cannot be determined with full confidence how much of the overall surveillance weld is which weld metal heat and, furthermore, exactly which weld heat specimens are contained in which surveillance capsules in the Millstone Unit 2 program. The Millstone Unit 2 (combined) surveillance weld data met the second and third credibility criteria of Regulatory Guide 1.99 , Revision 2 [Ref. D-1]. Additionally, Table D-2 of WCAP-16012  
It cannot be determined with full confidence how much of the overall surveillance weld is which weld metal heat and, furthermore, exactly which weld heat specimens are contained in which surveillance capsules in the Millstone Unit 2 program. The Millstone Unit 2 (combined) surveillance weld data met the second and third credibility criteria of Regulatory Guide 1.99 , Revision 2 [Ref. D-1]. Additionally, Table D-2 of WCAP-16012
[Ref. D-4] indicates that all of the measured weld L'iRT NDT values were within the 1-sigma scatter band; therefore , suggesting that there is good agreement between the measured capsule data and the embrittlement correlations.
[Ref. D-4] indicates that all of the measured weld L'iRT NDT values were within the 1-sigma scatter band; therefore , suggesting that there is good agreement between the measured capsule data and the embrittlement correlations.
If the two heats of weld material were evaluated individually, one would expect that the scatter in the data would decrease since the irradiated material would em brittle differently for the two separate welds with differ ent, as-measured, copper and nicke l contents. However , since the ( combined) weld mater ia l already passes the Regulatory Guide 1.99 , Revision 2 credibility analysis, a re-evaluation of the material ( as two separate heats) is not expected to significantly change the overall results of the subsequent reactor vessel integrity analyses.
If the two heats of weld material were evaluated individually, one would expect that the scatter in the data would decrease since the irradiated material would em brittle differently for the two separate welds with differ ent, as-measured, copper and nicke l contents. However , since the ( combined) weld mater ia l already passes the Regulatory Guide 1.99 , Revision 2 credibility analysis, a re-evaluation of the material ( as two separate heats) is not expected to significantly change the overall results of the subsequent reactor vessel integrity analyses.
Thus, the surveillance weld metal will be considered to be only Heat# 10137 for the evaluations contained herein. All currently determined input data for Position 2.1 chemistry factor determination (See Section 5) and surveillance data credibility assessment WCAP-18169
Thus, the surveillance weld metal will be considered to be only Heat# 10137 for the evaluations contained herein. All currently determined input data for Position 2.1 chemistry factor determination (See Section 5) and surveillance data credibility assessment WCAP-18169
-NP June 2018 Re vis ion I CALC-AN02-EP-17-00002 Rev O Page 76 of 88 Westinghouse Non-Proprietary Class 3 D-2 documented in this Appendix will be used "as-is," as documented in the Millstone Unit 2 surveillance capsule an a lyses of record. For conservatism, no reduction in the margin term of Regulatory Guide 1.99, Revision 2 [Ref. D-1] was taken to account for the additional uncertainties , despite the data remaining credible (see Section D.2). Note that this approach should also be followed when completing analyses per 10 CFR 50.61 [Ref. D-5]. Despite thi s additional conservatism , the Arkansas Nuclear One Unit 2 Upper to Intermediate Shell Girth Weld 8-20 3 (Heat# 10137) was not the limiting material for the Arkansas Nuclear One Unit 2 P-T limit curves. D.2 EVALUATION Per Appen d ix D of WCAP-17501-NP  
-NP June 2018 Re vis ion I CALC-AN02-EP-17-00002 Rev O Page 76 of 88 Westinghouse Non-Proprietary Class 3 D-2 documented in this Appendix will be used "as-is," as documented in the Millstone Unit 2 surveillance capsule an a lyses of record. For conservatism, no reduction in the margin term of Regulatory Guide 1.99, Revision 2 [Ref. D-1] was taken to account for the additional uncertainties , despite the data remaining credible (see Section D.2). Note that this approach should also be followed when completing analyses per 10 CFR 50.61 [Ref. D-5]. Despite thi s additional conservatism , the Arkansas Nuclear One Unit 2 Upper to Intermediate Shell Girth Weld 8-20 3 (Heat# 10137) was not the limiting material for the Arkansas Nuclear One Unit 2 P-T limit curves. D.2 EVALUATION Per Appen d ix D of WCAP-17501-NP
[Ref. D-6], the Calvert Cliffs Unit 2 surveillance weld data (Heat# 10137) was deemed credible , and per Appendix D of WCAP-16012  
[Ref. D-6], the Calvert Cliffs Unit 2 surveillance weld data (Heat# 10137) was deemed credible , and per Appendix D of WCAP-16012
[Ref. D-4], the Millstone Unit 2 surveillance weld data (Heat# 10137) was also deemed credible.
[Ref. D-4], the Millstone Unit 2 surveillance weld data (Heat# 10137) was also deemed credible.
Thus , when analyzed individually, these surve illance welds pass all five of the Regulatory Guide 1.99, Revision 2 [Ref. D-1] credibility criterion.
Thus , when analyzed individually, these surve illance welds pass all five of the Regulatory Guide 1.99, Revision 2 [Ref. D-1] credibility criterion.
The only c r edibility criterion that must be updated as a result of analyzing the two surveillance welds together is Criterion  
The only c r edibility criterion that must be updated as a result of analyzing the two surveillance welds together is Criterion
: 3. This evaluation is documented herein. Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of LlRT N OT values about a best-fit line drawn as described in Regulatory Guide 1.99 , Revision 2 [Ref. D-1] normally should be less than 28&deg;F for welds and l 7&deg;F for base metal. Eve n if the fluence range is large (two or more orders of magn i tude), the scatter should not exceed twice those val ues. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upp er-s helf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. D-7]. The functional form of the least-squares method as described in Regulatory Guide 1.99, Revision 2 will be utilized to determine a be st-fit line for this data and to determine if the scatter of these LlRT NDT values about this line is less than 28&deg;F for the weld. Following i s the calculation of the best-fit line as described in Reference D-1. In addition, the recommended NRC method s for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13 , 1998 [Ref. D-8]. At this meeting the NRC presented five cases. Of the five cases, Case 5 ("Surveillance Data from Other Sources Only") most closely repre sents the situation for the Arkansas Nuclear One Unit 2 reactor vessel Upper to Intermediate Shell Girth Weld 8-203 (Heat# 10137) as described below: Heat# 10137 {Case 5) -This weld heat p ertai ns to the Upper to Intermediate Shell Girth Weld 8-203 in the Arkansas Nuclear One Unit 2 reactor vessel. This weld heat is not contained in the Arkansas Nuclear One Unit 2 s urveillance program. Howeve r, it is contained in the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs.
: 3. This evaluation is documented herein. Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of LlRT N OT values about a best-fit line drawn as described in Regulatory Guide 1.99 , Revision 2 [Ref. D-1] normally should be less than 28&deg;F for welds and l 7&deg;F for base metal. Eve n if the fluence range is large (two or more orders of magn i tude), the scatter should not exceed twice those val ues. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upp er-s helf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. D-7]. The functional form of the least-squares method as described in Regulatory Guide 1.99, Revision 2 will be utilized to determine a be st-fit line for this data and to determine if the scatter of these LlRT NDT values about this line is less than 28&deg;F for the weld. Following i s the calculation of the best-fit line as described in Reference D-1. In addition, the recommended NRC method s for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13 , 1998 [Ref. D-8]. At this meeting the NRC presented five cases. Of the five cases, Case 5 ("Surveillance Data from Other Sources Only") most closely repre sents the situation for the Arkansas Nuclear One Unit 2 reactor vessel Upper to Intermediate Shell Girth Weld 8-203 (Heat# 10137) as described below: Heat# 10137 {Case 5) -This weld heat p ertai ns to the Upper to Intermediate Shell Girth Weld 8-203 in the Arkansas Nuclear One Unit 2 reactor vessel. This weld heat is not contained in the Arkansas Nuclear One Unit 2 s urveillance program. Howeve r, it is contained in the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs.
NRC Case 5 per Reference D-8 is entitled "Surve illance Data from Other Sources Onl y" and most closely represents the situation for Arkansas Nuclear One Unit 2 weld Heat# 10137. WCAP-18169-NP June 2018 Revi s ion I CALC-AN02-EP-17  
NRC Case 5 per Reference D-8 is entitled "Surve illance Data from Other Sources Onl y" and most closely represents the situation for Arkansas Nuclear One Unit 2 weld Heat# 10137. WCAP-18169-NP June 2018 Revi s ion I CALC-AN02-EP-17  
Line 290: Line 290:
Therefore, in accordance with Case 5 , the combined data from both Calvert Cliffs Unit 2 and Millstone Unit 2 will now be assessed to determine the credibility conclusion for all applicable data for weld Heat# 10137. W C AP-1816 9-NP June 2018 Revi s ion l CALC-AN02-EP 00002 Rev O Page 78 of 88 Westinghouse Non-Proprietary Class 3 Credibility Assessment Case 5: Weld Heat# 10137 (All data) D-4 In acc orda n ce wi th the NRC Case 5 guidelines, th e data from Calvert Cliffs Unit 2 and Millstone Unit 2 wi ll now b e ana l yze d together.
Therefore, in accordance with Case 5 , the combined data from both Calvert Cliffs Unit 2 and Millstone Unit 2 will now be assessed to determine the credibility conclusion for all applicable data for weld Heat# 10137. W C AP-1816 9-NP June 2018 Revi s ion l CALC-AN02-EP 00002 Rev O Page 78 of 88 Westinghouse Non-Proprietary Class 3 Credibility Assessment Case 5: Weld Heat# 10137 (All data) D-4 In acc orda n ce wi th the NRC Case 5 guidelines, th e data from Calvert Cliffs Unit 2 and Millstone Unit 2 wi ll now b e ana l yze d together.
Dat a is adjusted to the m ea n chemical composition and operating temperatur e of the s urveillanc e capsules.
Dat a is adjusted to the m ea n chemical composition and operating temperatur e of the s urveillanc e capsules.
This is performed in Table D-1. Table D-1 Mean Chemical Composition and Operating Temperature for Calvert Cliffs Unit 2 and Millstone Unit 2 Material Capsule 263&deg; Weld Metal Heat# 10137 (Ca lv ert Cliffs Unit 2 Data) 97&deg; 104&deg; 97&deg; Weld Meta l Heat# 10137 (Millstone Un it 2 Data) 104&deg; 83&deg; MEAN Note: (a) C hemistry data obtained from Table 3-1. (b) Temperatu r e data obtained from Table 4-2. Cu Ni In l et Temperature during Wt.%<*> Wt.%<*> Period of Irradiation (0 F)(b) 550 0.21 0.06 549 548 544.3 0.21 0.0 6 547.6 548.0 0.21 0.06 547.8 Since the mean che mic a l compositio n of the s urveill ance capsule data is identical to the actual chemical compos ition data for eac h caps ule , no chemistry adjustment is nec essary. However, since the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance capsule operating temperatures are not identical to the mean operating temperature, the s ur veillance capsule data will be ad ju ste d to the mean operating tempera tu re. The capsule-specific temperature a dju s tment s are as s hown in Table D-2. Table D-2 Operating Te mperature Adjustments for the Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data Material Caps ule Inlet Temperature d u ring Mean Operating Temperature Period of Irradiation  
This is performed in Table D-1. Table D-1 Mean Chemical Composition and Operating Temperature for Calvert Cliffs Unit 2 and Millstone Unit 2 Material Capsule 263&deg; Weld Metal Heat# 10137 (Ca lv ert Cliffs Unit 2 Data) 97&deg; 104&deg; 97&deg; Weld Meta l Heat# 10137 (Millstone Un it 2 Data) 104&deg; 83&deg; MEAN Note: (a) C hemistry data obtained from Table 3-1. (b) Temperatu r e data obtained from Table 4-2. Cu Ni In l et Temperature during Wt.%<*> Wt.%<*> Period of Irradiation (0 F)(b) 550 0.21 0.06 549 548 544.3 0.21 0.0 6 547.6 548.0 0.21 0.06 547.8 Since the mean che mic a l compositio n of the s urveill ance capsule data is identical to the actual chemical compos ition data for eac h caps ule , no chemistry adjustment is nec essary. However, since the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance capsule operating temperatures are not identical to the mean operating temperature, the s ur veillance capsule data will be ad ju ste d to the mean operating tempera tu re. The capsule-specific temperature a dju s tment s are as s hown in Table D-2. Table D-2 Operating Te mperature Adjustments for the Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data Material Caps ule Inlet Temperature d u ring Mean Operating Temperature Period of Irradiation
(&deg;F) Temperature  
(&deg;F) Temperature
(&deg;F) A djustment  
(&deg;F) A djustment
(&deg;F) 263&deg; 550 2.2 Weld Me t a l Heat# 10137 (Calvert C liff s Unit 2 Data) 97&deg; 549 1.2 104&deg; 548 0.2 547.8 97&deg; 544.3 -3.5 Weld Meta l Heat# 10137 (Mi ll stone Unit 2 Data) 104&deg; 547.6 -0.2 830 548.0 0.2 Using the chemical composition and operating temperature a djustm ents described an d calculated above, an interim c hemistry factor is calc ulat ed for wel d Heat # 10137 using the Ca l vert Cliffs Unit 2 and Millstone Unit 2 data. This calculation is shown on the following page in Table D-3. WCAP-18 1 69-NP June 2018 Revision I Table D-3 Material Weld Metal Hea t# 10137 (Calve rt Cliffs Unit 2 Da t a) Weld Metal Hea t# 1013 7 (Mi llsto ne Unit 2 Data) Notes: CALC-AN02-EP-17-00002 Rev O Page 79 of 88 Westinghouse Non-Proprietary Class 3 D-5 Calculation of Weld Heat # 10137 Interim Chemistry Factor for the Credibility Evaluation Using Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data Capsule t<*l .1RTNDT (c) FF*.1RTNDT Capsule FF<bl (x 10 19 n/cm2, E > 1.0 MeV) (OF) {&deg;F) 263&deg; 0.825 0.9460 74.9 (72.7) 70.86 97&deg; 1.95 1.1825 84.1 (82.9) 99.45 104&deg; 2.44 1.2401 69.9 (69.7) 86.68 97&deg; 0.324 0.6902 62.4 (65.93) 43.09 104&deg; 0.949 0.9853 51.9 (52.12) 51.16 830 1.740 1.1523 56.3 (56.09) 64.86 SUM: 416.10 CF H e at# 10137 = L(FF * ~RT NDT) .;-L(FF 2) = (416.10) .;-(6.606) = 63.0&deg;F FF 2 0.895 1.398 1.538 0.476 0.971 1.328 6.606 (a) f= fluence; (b) FF= fluence factor= t<0*28 -O.IO'l o g I). (c) t>RT N OT values are the measured 30 ft-lb shift values. Each t>RT N OT value ha s been adjusted according to the temperature adjustments summar i zed in Table 0-2. The t>RT N OT va lu es for each surveillance weld data point are not adjusted by the ratio procedure, because the mean chemical composition is id e ntical to eac h capsule chemical composition (pre-adjusted values are listed in parentheses and we r e taken from Table 4-2). The scatter of ~RT NDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99 , Revision 2 , Position 2.1 [Ref. D-1] is presented in Table D-4. Table D-4 Material Weld Metal Heat# 10137 (Calvert Cliffs Unit 2 Data) Weld Meta l Heat# 10137 (Mi ll sto ne Unit 2 Data) Best-Fit Evaluation for Surveillance Weld Metal Heat# 10137 Using Calvert Cliffs Unit 2 and Millstone Unit 2 Data CF Capsule f Measured Predicted Residual Capsule (Slopebest-fit) (x 10 19 n/cm2, FF .1RTNDT .1RTNDT .1RTNDT <28&deg;F (Weld) (OF) E > 1.0 MeV) (OF) (OF) (OF) 263&deg; 63.0 0.825 0.9460 74.9 59.6 15.3 Yes 97&deg; 63.0 1.95 1.1825 84.l 74.5 9.6 Yes 104&deg; 63.0 2.44 1.240 1 69.9 78.1 8.2 Yes 97&deg; 63.0 0.324 0.6902 62.4 43.5 18.9 Yes 104&deg; 63.0 0.949 0.9853 51.9 62.1 10.2 Yes 830 63.0 1.740 1.1523 56.3 72.6 16.3 Yes The scatter of ~RT N DT values a bout the best-fit line , drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [R ef. D-1 ], should be less than 28&deg;F for weld metal. Table D-4 indicates that 100% (six out of s ix) of the survei llanc e data points fall within the +/-la of 28&deg;F scatter band for survei llanc e weld materials.
(&deg;F) 263&deg; 550 2.2 Weld Me t a l Heat# 10137 (Calvert C liff s Unit 2 Data) 97&deg; 549 1.2 104&deg; 548 0.2 547.8 97&deg; 544.3 -3.5 Weld Meta l Heat# 10137 (Mi ll stone Unit 2 Data) 104&deg; 547.6 -0.2 830 548.0 0.2 Using the chemical composition and operating temperature a djustm ents described an d calculated above, an interim c hemistry factor is calc ulat ed for wel d Heat # 10137 using the Ca l vert Cliffs Unit 2 and Millstone Unit 2 data. This calculation is shown on the following page in Table D-3. WCAP-18 1 69-NP June 2018 Revision I Table D-3 Material Weld Metal Hea t# 10137 (Calve rt Cliffs Unit 2 Da t a) Weld Metal Hea t# 1013 7 (Mi llsto ne Unit 2 Data) Notes: CALC-AN02-EP-17-00002 Rev O Page 79 of 88 Westinghouse Non-Proprietary Class 3 D-5 Calculation of Weld Heat # 10137 Interim Chemistry Factor for the Credibility Evaluation Using Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data Capsule t<*l .1RTNDT (c) FF*.1RTNDT Capsule FF<bl (x 10 19 n/cm2, E > 1.0 MeV) (OF) {&deg;F) 263&deg; 0.825 0.9460 74.9 (72.7) 70.86 97&deg; 1.95 1.1825 84.1 (82.9) 99.45 104&deg; 2.44 1.2401 69.9 (69.7) 86.68 97&deg; 0.324 0.6902 62.4 (65.93) 43.09 104&deg; 0.949 0.9853 51.9 (52.12) 51.16 830 1.740 1.1523 56.3 (56.09) 64.86 SUM: 416.10 CF H e at# 10137 = L(FF * ~RT NDT) .;-L(FF 2) = (416.10) .;-(6.606) = 63.0&deg;F FF 2 0.895 1.398 1.538 0.476 0.971 1.328 6.606 (a) f= fluence; (b) FF= fluence factor= t<0*28 -O.IO'l o g I). (c) t>RT N OT values are the measured 30 ft-lb shift values. Each t>RT N OT value ha s been adjusted according to the temperature adjustments summar i zed in Table 0-2. The t>RT N OT va lu es for each surveillance weld data point are not adjusted by the ratio procedure, because the mean chemical composition is id e ntical to eac h capsule chemical composition (pre-adjusted values are listed in parentheses and we r e taken from Table 4-2). The scatter of ~RT NDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99 , Revision 2 , Position 2.1 [Ref. D-1] is presented in Table D-4. Table D-4 Material Weld Metal Heat# 10137 (Calvert Cliffs Unit 2 Data) Weld Meta l Heat# 10137 (Mi ll sto ne Unit 2 Data) Best-Fit Evaluation for Surveillance Weld Metal Heat# 10137 Using Calvert Cliffs Unit 2 and Millstone Unit 2 Data CF Capsule f Measured Predicted Residual Capsule (Slopebest-fit) (x 10 19 n/cm2, FF .1RTNDT .1RTNDT .1RTNDT <28&deg;F (Weld) (OF) E > 1.0 MeV) (OF) (OF) (OF) 263&deg; 63.0 0.825 0.9460 74.9 59.6 15.3 Yes 97&deg; 63.0 1.95 1.1825 84.l 74.5 9.6 Yes 104&deg; 63.0 2.44 1.240 1 69.9 78.1 8.2 Yes 97&deg; 63.0 0.324 0.6902 62.4 43.5 18.9 Yes 104&deg; 63.0 0.949 0.9853 51.9 62.1 10.2 Yes 830 63.0 1.740 1.1523 56.3 72.6 16.3 Yes The scatter of ~RT N DT values a bout the best-fit line , drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [R ef. D-1 ], should be less than 28&deg;F for weld metal. Table D-4 indicates that 100% (six out of s ix) of the survei llanc e data points fall within the +/-la of 28&deg;F scatter band for survei llanc e weld materials.
Therefore, the s ur veillance weld material (Heat # 1013 7) is deemed "credi ble" per the third criterion when all available data is cons idered. WCAP-1 8169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 80 of 88 Westinghouse Non-Proprietary C l ass 3 D-6 In conclus i on, the combined survei ll ance data from Calvert C li ffs Unit 2 and Millstone Unit 2 for weld Heat# 10137 may be applied to the Arkansas Nuclear One Unit 2 reactor vessel weld. The Po sition 2.1 chemistry factor calc ul ation, as applica bl e to the Arkansas Nuclear One Unit 2 reactor vessel wel d , is con t ained in Section 5. This Position 2.1 CF va lu e could be u se d with a reduced margin term in the ART calc ulat ions contained in Section 7. However, consistent w ith the discussion in Section D.1 of this Appendix, the ART va lu es calculate d with the Position 2.1 CF value for weld Heat# 10137 utili ze a full margin term for conservatism. D.3 REFERENCES D-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Ra diation Embrittlement of Reactor Vessel Materials," May 1988. D-2 Westinghouse Report WCAP-1 8166-NP, Revision 0, "Ana l ysis of Ca p s ul e 284&deg; from th e Entergy Operations , Inc. Arkansas Nuclear One Unit 2 Reactor Vessel Radiation Surveillance Program ," September 2016. D-3 W e stinghouse Report WCAP-16918-NP, Revision 1 , "Analysis of Capsule V from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Radiation Surveillance Program ," April 2008. D-4 Westinghouse Report WCAP-16012, Revision 0 , " Analysis of Caps ule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003. D-5 Code of Federal Regulations , 10 CFR 50.61, "Frac ture Toughness Requirements for Protection Ag a inst Pressurized Thermal Shock Events," Federa l Register , Volume 60 , No. 243, dated December 19, 1995 , effective January 18 , 1996. D-6 Westinghouse R e port WCAP-17501-NP, Revision 0, "Analysis of Caps ul e 104&deg; from the Calvert Cli ff s Unit No. 2 Reactor Vessel Radiation Surveillance Program ," Fe bru ary 2012. D-7 ASTM El85-82 , "Stan dard Practice for Conduct i ng Surveillance Tests for Light-Water Coo l ed Nuclear Power Reactor Vessels," ASTM, Jul y 1982. D-8 K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, N RC/lndustry Workshop on RPV Int egrity Issu es, February 12 , 1998. [A DAMS Accession Numbe r MLI 100 7057 0] WCAP-1 816 9-NP June 2018 Revision I}}
Therefore, the s ur veillance weld material (Heat # 1013 7) is deemed "credi ble" per the third criterion when all available data is cons idered. WCAP-1 8169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 80 of 88 Westinghouse Non-Proprietary C l ass 3 D-6 In conclus i on, the combined survei ll ance data from Calvert C li ffs Unit 2 and Millstone Unit 2 for weld Heat# 10137 may be applied to the Arkansas Nuclear One Unit 2 reactor vessel weld. The Po sition 2.1 chemistry factor calc ul ation, as applica bl e to the Arkansas Nuclear One Unit 2 reactor vessel wel d , is con t ained in Section 5. This Position 2.1 CF va lu e could be u se d with a reduced margin term in the ART calc ulat ions contained in Section 7. However, consistent w ith the discussion in Section D.1 of this Appendix, the ART va lu es calculate d with the Position 2.1 CF value for weld Heat# 10137 utili ze a full margin term for conservatism. D.3 REFERENCES D-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Ra diation Embrittlement of Reactor Vessel Materials," May 1988. D-2 Westinghouse Report WCAP-1 8166-NP, Revision 0, "Ana l ysis of Ca p s ul e 284&deg; from th e Entergy Operations , Inc. Arkansas Nuclear One Unit 2 Reactor Vessel Radiation Surveillance Program ," September 2016. D-3 W e stinghouse Report WCAP-16918-NP, Revision 1 , "Analysis of Capsule V from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Radiation Surveillance Program ," April 2008. D-4 Westinghouse Report WCAP-16012, Revision 0 , " Analysis of Caps ule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003. D-5 Code of Federal Regulations , 10 CFR 50.61, "Frac ture Toughness Requirements for Protection Ag a inst Pressurized Thermal Shock Events," Federa l Register , Volume 60 , No. 243, dated December 19, 1995 , effective January 18 , 1996. D-6 Westinghouse R e port WCAP-17501-NP, Revision 0, "Analysis of Caps ul e 104&deg; from the Calvert Cli ff s Unit No. 2 Reactor Vessel Radiation Surveillance Program ," Fe bru ary 2012. D-7 ASTM El85-82 , "Stan dard Practice for Conduct i ng Surveillance Tests for Light-Water Coo l ed Nuclear Power Reactor Vessels," ASTM, Jul y 1982. D-8 K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, N RC/lndustry Workshop on RPV Int egrity Issu es, February 12 , 1998. [A DAMS Accession Numbe r MLI 100 7057 0] WCAP-1 816 9-NP June 2018 Revision I}}

Revision as of 04:51, 25 April 2019

WCAP-18169-NP, Rev 1, Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation.
ML18215A178
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Issue date: 06/30/2018
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CALC-ANO2-EP-17-00002, Rev 0 WCAP-18169-NP, Rev 1
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WCAP-18169-NP Revision 1 CALC-AN02-EP-17-00002 Rev O Page 3 of 88 Westinghouse Non-Proprietary Class 3 June 2018 Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation

@ Westinghouse CALC-AN02-EP-17-00002 Rev O Page 4 of 88 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCA P-1 8 1 6 9-N P Rev i s io n 1 Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation Reviewers:

Benjam in E. May s* Structural Design & Ana l ysis III J un e 2 0 18 D. Brett Lynch* Approved:

Structural Design & Analysis III Eugene T. Hayes* Radiation Engineering

& Analysis Lynn A. Patterson*, Manager Structural Design & Analysis III Laurent P. Houssay*, Manager Radiation Engineering

& Analysis *Electro n ically approved records are authenticated in the electronic document management system. Westinghouse Electric Company LLC 1000 Westinghouse Dr. Cranberry Township , PA 16066 © 2018 Westingho u se Electric Company LLC All Rights Reserved Revision 0: Revision 1: CALC-AN02-EP-17-00002 Rev O Page 5 of 88 Westinghouse Non-Proprietary C l ass 3 RECORD OF REVISION Original Issue II Revised issue. The purpose of this rev1s1on is to remove utilization of the RES/DE/CIB-2013-01 report methodology.

Therefore, calculated LlRT NDT values less than or equal to 25°F will not be reduced to zero. The pressure-temperature (P-T) limit curves are not affected by the changes. Changes are indicated with change bars. WCAP-18 1 69-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 6 of 88 Westin g hou s e Non-Propri e t a ry Clas s 3 TABLE OF CONTENTS lll LIST OF TABLES ...............................................

........................................................................................ iv LIST OF FIGURES ...................

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vi EXECUTIVE

SUMMARY

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.................................................................................... vii INTRODUCTION

........................................................................................................................ 1-1 2 CALCULATED NEUTRON FLUENCE ...............................

...................................................... 2-1

2.1 INTRODUCTION

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......................................... 2-l 2.2 DISCRETE ORDINATES ANALYSIS ...................................

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................ 2-1 2.3 CALCULATIONAL UNC E RTAINTIES

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............................... 2-3 3 FRACTURE TOUGHNESS PROPERTIES

................................................................................. 3-1 4 SURVEILLANCE DATA .............

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........................................ 4-1 5 CHEMISTRY FACTORS ....................................................................................................

......... 5-1 6 CRITERIAFORALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS

................ 6-1 6.1 OVERALLAPPROACH

................................................................................................. 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPM E NT ............................................................................................................ 6-1 6.3 CLOSURE H E ADNESS E L FLANGE REQUIREMENTS

........................................... 6-5 6.4 LOWEST SERVICE TEMPERATURE REQUIREMENTS

........................................... 6-5 6.5 BOLTUP T E MPERATURE REQUIREMENTS

............................................................. 6-5 7 CALCULATION OF ADWSTED REFERENCE TEMPERATURE

.......................................... 7-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ....................... 8-1 9 REF E RENCES .................

............................................................................................................ 9-1 APPENDIXA APPENDIXB APPENDIXC APPENDIXD WCAP-18169-NP TH E RMAL STRESS INTENSITY FACTORS (K 11) ************************************

          • A-1 REACTOR VESS E L INLET AND OUTLET NOZZLES ................................ B-1 NON-REACTOR V E SSEL FERRlTIC COMPONENTS

................................. C-1 CREDIBILITY EVALUATION OF THE WELD HEAT# 10137 SURVEILLANCE DATA .................................................................................. D-1 June2018 Re v ision 1 Table 2-1 Table 2-2 Table 2-3 Table 2-4 Table 2-5 Table 2-6 Table 2-7 Table 2-8 Table 3-1 Table 3-2 Table 4-1 Table 4-2 Table 4-3 Table 5-1 Table 5-2 Table 5-3 Table 5-4 Table 5-5 Table 7-1 CALC-AN02-EP-17-00002 Rev O Page 7 of 88 Westinghouse Non-Proprietary Class 3 LIST OF TABLES IV Pressure Vessel Material Locations

.................................................................................. 2-5 Calculated Maximum Fast Neutron (E > 1.0 MeV) Fluence of the Pressure Vessel Clad/Base Metal Interface

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.................... 2-6 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface

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2-7 Calculated Azimuthal Variation of the Maximum Fast Neutron (E > 1.0 MeV) Fluence Rate at the Reactor Vessel Clad/Base Metal Interface

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............ 2-8 Calculated Azimuthal Variation of the Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Reactor Vessel Clad/Base Metal Interface

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....... 2-9 Calculated Iron Atom Di splace ment Rate at the Pressure Vessel Clad/Base Metal Interface

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............................................................................. 2-10 Calculated Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface

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................... 2-l 1 Calculational Uncertainties

............................................................................................ 2-12 Summary of the Best-Est imate Chemistry and Initial RT N DT Values for the Arkansas Nuclear One Unit 2 Reactor Vessel Materials

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............................. 3-2 Summary of Arkansas Nuclear One Unit 2 Reactor Vessel Closure Heads, Vessel Flange and Balance of RCS Initial RTNDT Values .......................

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.. 3-4 Arkansas Nuclear One Unit 2 Surveillance Capsule Data ......................

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.4-2 Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data for Weld Heat # 10137 .........................................

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........... 4-3 J.M. Farley Unit 2 Surveillance Capsule Data for Weld Heat# BOLA. ..............

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.4-3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Intermediate Shell Plate C-8009-3 Using Surveillance Capsule Data ...........................................

....... 5-2 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat # 83650 Using Surveillance Capsule Data .................................

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....................... 5-2 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat # 10137 Using Surveillance Capsule Data .............................................

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....... 5-3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat # BOLA Using Surveillance Capsule Data .....................

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5-4 Summary of Arkansas Nuclear One Unit 2 Positions 1.1 and 2.1 Chemistry Factors ..... 5-5 Fluence Values and Fluence Factors for the Vessel Surface , l/4T and 3/4T Locations for the Arkansas Nuclear One Unit 2 Reactor Vessel Materials at 54 EFPY ........................

7-3 WCAP-18169-NP June 2018 Revision I Table 7-2 Table 7-3 Table 7-4 Table 8-1 Table 8-2 Table 8-3 Table A-1 Table A-2 Table B-1 Table B-2 Table D-1 Table D-2 T a ble D-3 Table D-4 CALC-AN02-EP-17-00002 Rev O Page 8 of 88 Westinghou s e Non-Proprietary Class 3 V Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the l/4T Location ...................... 7-4 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location ...................... 7-6 Summary of the Increased Limiting ART Values Used in the Generation of the Arkansas Nuclear One Unit 2 Heatup and Cooldown Curves at 54 EFPY ................

..................... 7-8 Arkansas Nuclear One Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors) .............................................................

............ 8-6 Arkansas Nuclear One Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements , and w/o Margins for Instrumentation Errors) .......................................... 8-8 Arkansas Nuclear One Unit 2 54 EFPY Inservice Hydrostatic and Leak Test Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c , w/ Flange and LST Requirements , and w/o Margins for Instrumentation Errors) .............

.............. 8-9 K1t Values for Arkansas Nuclear One Unit 2 at 54 EFPY 80°F/hr Heatup Curves (w/ Flange and LST Requirements, and w/o Margins for Instrument Errors) ....................... A-2 K 11 Values for Arkansas Nuclear One Unit 2 at 54 EFPY 100°F/hr Cooldown Curves (w/ Flange and LST Requirements, and w/o Margins for Instrument Errors) ....................... A-3 Summary of the Arkansas Nuclear One Unit 2 Reactor Vessel Nozzle Material Initial RT N D T , Chemistry , and Fluence Values at 54 EFPY ................................

....................... B-3 Summary o f the Limiting ART Values for the Arkansas Nuclear One Unit 2 Inlet and Outlet Nozzle Materials

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................... B-3 Mean Chemical Composition and Operating Temperature for Calvert Cliffs Unit 2 and Millstone Unit 2 .: ............................................................................................................ D-4 Operating Temperature Adjustments for the Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data .............................................................................................. D-4 Calculation of Weld Heat # 10137 Interim Chemistry Factor for the Credibility Evaluation Using Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data ... ..................................

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................. D-5 Best-Fit Evaluation for Surveillance Weld Metal Heat# 10137 Using Calvert Cliffs Unit 2 and Millstone Unit 2 Data .....................................

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D-5 WCAP-181 6 9-NP June 2018 Revision 1 F igure 2-1 Figure 2-2 Figure 2-3 Figure 8-1 Figure 8-2 Figure 8-3 F igure B-1 F igure B-2 CALC-AN02-EP-17

-00002 Rev O Page 9 of 88 Westin g house Non-Propriet a ry Class 3 LIST OF FIGURES VI Arkansas Nuclear One Unit 2 r , e Reactor Geometry Plan View at the Core Midplane with Surveillance Capsule s .................................................................................................... 2-13 Arkansas Nuclear One Unit 2 r,e Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules ....................................................................................... 2-14 Arkansas Nuclear One Unit 2 r,z Reactor Geometry Section View ............................... 2-15 Arkansas Nuclear One Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 50 , 60 , 70 , and 80°F/hr) Applicable for 54 EFPY (with Flange and LST Requirements a nd without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c) .......................................................................... 8-3 Arkansas Nuclear One Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0 , 25 , 60, and 100°F/hr) Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c) ............................................................ 8-4 Arkansas Nuclear One Unit 2 Reactor Coolant Sy s tem Inservice Hydrostatic and Leak Test Limitations Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K 1c) ............................................................................

.............. 8-5 Comparison of Arkansas Nuclear One Unit 2 Beltline P-T Limits to Inlet Nozzle Limits .....................................................................................

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................................. B-6 Comparison of Arkansa s Nuclear One Unit 2 Beltline P-T Limits to Outlet Nozzle Limits ........................................................................................

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..................... B-7 W C AP-181 6 9-NP June 2018 Re v ision 1 CALC-AN02-EP-17-00002 Rev O Page 10 of 88 Westinghouse Non-Proprietary Class 3 EXECUTIVE

SUMMARY

vii This report provides the methodology and results of the generation of heatup and cooldown temperature (P-T) limit curves for normal operation of the Arkansas Nuclear One Unit 2 reactor vessel. The heatup and cooldown P-T limit curves were generated using the limiting Adjusted Reference Temperature (ART) values for Arkansas Nuclear One Unit 2. The limiting ART values were those of Lower Shell Plate C-8010-1 (Position 1.1) at both 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations. The P-T l i mit curves were generated using the K,c methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code ,Section XI, Appendix G. The P-T limit curves were generated for 54 effective full-power years (EFPY) using heatup rates of 50, 60, 70, and 80°F/hr, and cooldown rates of O (steady-state), -25, -60, and -100°F/hr. The curves were developed with the flange and lowest service temperature (LST) requirements and without margins for instrumentation errors. They can be found in Figures 8-1 and 8-2. Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 54EFPY. Appendix B contains a P-T limit evaluation of the reactor vessel inlet and outlet nozzles based on a 1/4 T flaw postulated at the inside surface of the reactor vessel nozzle comer , where T is the thickness of the vessel at the nozzle comer region. As discussed in Appendix B, the P-T limit curves generated based on the limiting cylindrical beltline material (Lower Shell Plate C-8010-1) bound the P-T limit curves for the reactor ve s sel inlet and outlet nozzles for Arkansas Nuclear One Unit 2 at 54 EFPY. Appendix C contains discussion of the other non-reactor vessel ferritic Reactor Coolant Pressure Boundary (RCPB) components relative to P-T limits. As discussed in Appendix C, all of the non-reactor vessel ferritic RCPB components meet the applicable requirements of Section III of the ASME Code. Appendix D contains a credibility evaluation for weld Heat# 10137 considering all applicable sister plant surveillance program data. WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 11 of 88 Westinghouse Non-Proprietary Class 3 1 INTRODUCTION 1-1 Heatup and cooldown P-T limit curves are calculated using the adjusted RT NDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RT NDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced dRT NDT, and adding a margin. The unirradiated RT NDT is designated as the higher of either the drop weight nil-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. For the purposes of this report, heatup is defined as the process of heating the reactor coolant system (RCS) from ambient temperature to operating temperature.

Cooldown is defined as the process of cooling th e RCS from operating temperature to ambient temperature. Steady-state is defined as a 0°F/hr cooldown or heatup rate. Under steady-state , the thermal stress intensity factor is considered to be zero. The stead y-state curve is necessary from an engineering perspective for comparison with heatup and cooldown.

RT NDT increases as the material is exposed to fast-neutron radiation.

Therefore, to find the most limiting RT NDT at any time period in the reactor's life, dRT NDT due to the radiation exposure associated with that time period must be added to the unirradiated RT NDT (RT NDT(U)). The extent of the shift in RT NDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The U.S. Nu cl ear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 [Ref. l]. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values (RT NDT(U) + dRT NDT + margins for uncertainties) at the l/4T and 3/4T locations, where Tis the thickness of the vessel at the beltline region measured fr om the clad/base metal interface.

The heatu p and cooldown P-T limit curves documented in this report were generated using the most limiting ART values (plus an additional margin to account for future perturbations such as an uprate or surveillance capsule results) and the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [Ref. 2]. Specifically, the Kie methodology of the 1998 through the 2000 Addenda Edition of ASME Code,Section XI, Appendix G [Ref. 3] was used. The Kie curve is a lower bound static fracture toughness curve obtained from test data gathered from several different heats of pressure vessel steel. The limiting material is indexed to the Kic curve so that allowable stress intensity factors can be obtained for the material as a function of temperature.

Allowable operating limits are then determined using the allowable s tress intensity factors. The purpo s e of this report is to present the calculations and the development of the Arkansas Nuclear One Unit 2 heatup and cooldown P-T limit curves for 54 EFPY. This report documents the calculated ART values and the development of the P-T limit curves for normal operation.

The calculated ART values for 54 EFPY are documented in Section 7 of this report. The fluence projections used in calculation of the ART values are provided in Section 2 of this report. The P-T l i mit curves herein were generated without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [Ref. 4] have been incorporated in the P-T limit curves, along with the lowest service temperature (LST) requirements of ASME Code,Section III [Ref. 9]. As discussed i n Appendix B, the P-T limit curves generated in Section 8 bound the P-T limit curves for the reactor ve s sel inlet and outlet nozzles for Arkansas Nuclear One Unit 2 at 54 EFPY. Discussion of the other non-reactor vessel ferritic RCPB components relative to P-T limits is contained in Appendix C. WCAP-18169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 12 of 88 Westin g hou s e Non-Proprietary Class 3 2 CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

2-1 A discrete o rdinates (S N) transport analysis was performed for the Arkansas Nuclear One Unit 2 reactor to determine t he neutron radiation environment within the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant-and fuel-cycle-specific basis. An evaluation of the dosimetry sensor sets from the 284° and 97° surveillance capsules is provided in WCAP-18166-NP

[Ref. 21]. The dosimetry analysis documented in WCAP-18166-NP showed that the +/-20% (lcr) acceptance criterion specified in Regulatory Guide 1.190 [Ref. 5] is met. The validated calculations form the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 EFPY. All of the c alculations described in this section were based on nuclear cross-section data derived from the Evaluated Nuclear Data F i le (ENDF) database (specifically, ENDF/B-Vl).

Furthermore, the neutron transport e v aluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 5]. Additionally, the metho d s used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodolo g y described in WCAP-14040-A, Revision 4 [Ref. 2]. 2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the Arkansas Nuclear One Unit 2 reactor vessel , a s eries of fuel-cycle

-specific forward transport calculations were performed using the following dimensional fluence rate synthesis technique:

cp(r,0,z)

= cp(r,0) x cp(r,z) cp(r) where cp(r , 0,z) is the synthesized three-dimensional neutron fluence rate distribution , cp(r,0) is the transport solution in r , 0 geometry , cp(r,z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and cp(r) is the one-dimensional solution for a cylindrical reactor model using the sa me source per unit height as that used in the r,0 two-dimensional calculation. This synth e sis procedure was carried out for each operating cycle at Arkansas Nuclear One Unit 2. F or the Arkansas Nuclear One Unit 2 transport calculations , the r,0 model depicted in Figure 2-1 and Figure 2-2 were utilized since, with the exception of the capsules , the reactor is octant symmetric. These r , 8 models included the core , the reactor internals , octants with surveillance capsules at 7° and 14° and octants without surveillance capsules, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations.

Specificall y, the r , 8 model with surveillance capsules was utilized to perform capsule dosimetry evaluation s and subsequent comparisons with calculated results , while the r , 0 model without surveillance capsules w a s used to generate the maximum fluence levels at the pressure vessel wall. In developing these analytical models , nominal design dimensions were generally employed for the various structural c omponents.

Note that for the pressure vessel inner radiu s, however , the average of the as-built inner radii WCAP-181 6 9-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 13 of 88 Westinghouse Non-Proprietary Class 3 2-2 was used. In addition, water temperatures and, hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions.

The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, g uide tubes, et cetera. The geometric mesh description of the r , 0 reactor model in Figure 2-1 consisted of 166 radial by 123 azimuthal intervals. The geometric mesh description of the r,0 reactor model in Figure 2-2 consisted of 166 radial by 116 azimuthal intervals.

Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r , 0 calculations was set at a value of0.001. The r,z model used for the Arkansas Nuclear One Unit 2 calculations is shown in Figure 2-3. The model extends radially from the centerline of the reactor core out to the primary biological shield and axially from an elevation approximately 4.9 feet below to 6 feet a bove the active core. As in the case of the r,0 models, nominal design dimensions, with the exception of the pressure vessel inner radius, and full-power coolant densities were employed in the calculations.

In the r,z model, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel girth ribs located between the core shroud and core barrel regions were also explicitly included in the model. The geometric mesh description of the r,z reactor model in Figure 2-3 consisted of 161 radial by 222 axial i ntervals.

As in the case of the r,0 calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r,z calculations was set at a value of0.001. The one-dimensional radial model used in the synthesis procedure consisted of the same 161 radial mesh intervals included in the r,z model. Thus , radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The core power distributions used in the plant-specific transport analysis for each of the first 25 fuel cycles at Arkansas Nuclear One Unit 2 included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions (note that Cycles 1-24 have been completed; Cycle 25 is based on the expected core design for this cycle and an assumed cycle length of 1.37 EFPY). This information was used to develop spatial-and energy-dependent core source di stributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle-averaged neutron fluence rate, which, when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies.

From these assem bly-d ependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined. All of the transport calculations supporting this analysis were performed using the DORT discrete ordinates code [Ref. 23) and the BUGLE-96 cross-section library [Ref. 7). The BUGLE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifica lly for water reactor (LWR) applications. In these ana ly ses, anisotropic scattering was treated with a P 5 Legendre expansion and angular discretization was modeled with an S 1 6 order of angular quadrature.

Energy-and space-d ependent core power distributions , as well as system operating temperatures, were treated on a fuel-cycle

-specific basis. WCAP-1 8169-NP June 20 1 8 Revision I CALC-AN02-EP-17

-00002 Rev O Page 14 of 88 Westinghouse Non-Proprietary Class 3 2-3 The locations of the Arkansas Nuclear One Unit 2 vessel welds and plates are provided in Table 2-1. The axial posit i on of each material is indexed to z = 0.0 cm, which corresponds to the rnidplane of the active fuel stack. These data tabulations include both plant-and fuel-cycle-specific calculated neutron exposures at the end of Cycle 24 , at the end of projected Cycle 25, and at further projections to 54 EFPY. The calculations account for the uprate from 2815 MWt to 3026 MWt that occurred at the beginning of Cycle 16. The projection s are based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 23 , Cycle 24 , and the design of Cycle 25 are representative of future plant operation.

The future projections are based on the current reactor power level of 3026 MWt. Selected r e sults from the neutron transport analyses are provided in Table 2-2 through Table 2-7. In Table 2-2 , the calculated maximum fast neutron (E > 1.0 MeV) fluence values for the reactor pressure vessel materials are provided at future projections to 32, 36, 40 , 48 and 54 EFPY. The projections are based on the assumption that the core power distributions and associated plant operating characteristics from Cycles 23-25 are representative of future plant operation.

In Table 2-3 , the calculated maximum iron atom displ a cement values for the reactor pressure vessel materials are provided at future projections to 32, 36, 40 , 48 and 54 EFPY. The calcu l ated fast neutron (E > 1.0 MeV) fluence rate , fast neutron (E > 1.0 MeV) fluence , iron atom displacem e nt rate , and iron atom displacements are provided in Table 2-4 through Table 2-7, respectively, for the reactor pressure vessel inner radius at four azimuthal locations , as well as the maximum exposure observed w ithin the octant. The vessel data given in Table 2-4 through Table 2-7 were taken at the clad/base metal interface and represent maximum calculated exposure levels on the vessel. 2.3 CALCULATIONAL UNCERTAINTIES T he uncertainty associated with the calculated neutron exposure of the Arkansas Nuclear One Unit 2 reactor pres s ure vessel materials is based on the recommended approach provided in Regulatory Guide 1.190. In particular , the qualification of the methodology was carried out in the following four s tages: 1. Comparison of calculations with benchmark measurements from the Pool Critical As s embly (PCA) s imulator at the Oak Ridge National Laboratory (ORNL). 2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment.

3. An a nalytical sensitivity study addressing the uncertainty components resulting from important in p ut parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments. 4. Comparisons of the plant-specific calculations with all available dosimetry results from the Ar kansas Nuclear One Unit 2 surveillance program. WCAP-18 1 6 9-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 15 of 88 Westinghouse Non-Proprietary Class 3 2-4 The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operationa l or geometric variables that impact power reactor calculations. The second phase of the qualification (H.B. Robinson comparisons) addressed uncertainties in these additional a reas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations.

The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Arkansas Nuclear One Unit 2 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Arkansas Nuclear One Unit 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule or pressure vessel neutron exposures.

Table 2-8 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference

6. The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons given in Appendix A of Reference 21 support th e se uncertainty assessments for Arkansas Nuclear One Unit 2. WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 16 of 88 Westinghouse Non-Proprietary C la ss 3 Table 2-1 Pressure Vessel Material Locations Axial Location Ma t e rial R ela ti ve to Core Midplane at O cm (cm) Inl et Nozzle to Upper Shell Welds -Lowest Extent Nozzle I 301.625 (*) Nozzle 2 30 1.625 (*) Nozzle 3 30 1.625 (*) Nozzle 4 301.625 (*) Outlet Nozz l e to Upper Shell We ld s -Lowest Extent Nozzle I 301.625 (*) Nozzle 2 301.625<*> U pp er Shell to Intermediate She ll Ci r cumferent i a l Weld 8-203 248.722 Intermediate S h ell Plates C-8009-1 , -2 , 33.973 to 248.722 Int ermediate S h ell Longitudina l Welds 2-203 A -33.973 to 248.722 2-203 B -33.973 to 248.722 2-203 C -33.973 to 248.722 Intermediate S h ell to Lower Shell C i rcumferentia l Weld 9-203 -33.973 Lower Shell P l ates C-8 010-1 , -2, 3 1 5.892 to -33.973 Lower Shell Longitudinal Welds 3-203 A -3 I 5.892 to -33.973 3-2 03 B -315.892 to -33.973 3-203 C -3 I 5.892 to -33.973 Lower Shell to Bottom Head Circumferential Weld 10-203 -315.892 No te s: Azimut h al Location (degrees) 60 120 240 300 0 180 0 to 360 0 to 360 (b) 90 210 330 0 to 360 0 to 360(c) 90 210 330 0 to 360 (a) This axia l l ocation correspo nd s to the bottom of the vesse l s upport pad of t h e inl et nozzle , instead of the nozzle to upper shell we ld. T hi s provides a bounding fluence for the nozzle to upp er shell we ld. (b) Intermediate shell plates C-8009-1 , -2 , and -3 extend from azimut h al angles of330° to 90°, 90° to 210°, and 210° to 330°, r e spectively. (c) Lower she ll plates C-8 0 1 0-1 , -2, and -3 extend from azimuthal angles of210° to 330°, 330° to 90°, and 90° to 210°, respectively.

2-5 WCAP-18 169-NP June2018 Revision I CALC-AN02-EP-17

-00002 Rev O Page 17 of 88 Westinghouse Non-P roprietary C l ass 3 2-6 Tab le 2-2 Calculated Maximum Fast Neutron (E > 1.0 MeV) Fluence of the Pressure Vessel Clad/Base Meta l Interface Material Flue nce (b) (n/cm 2) 32 EFPY 36 EFPY 40 EFPY 48 EFPY 54 EFPY Inlet No zz le to Upper She ll Welds -Lowest Extent Nozzle 1<*) 4.78E+l6 5.36E+l6 5.93E+16 7.09E+l6 7.96E+l 6 Nozzle 2<*l 4.78E+l6 5.36E+l6 5.93E+l6 7.0 9E+l6 7.96E+l6 Nozzle 3(a) 4.78E+16 5.36E+16 5.93E+16 7.09E+l6 7.96E+l6 Nozzle 4 (*) 4.78E+l6 5.36E+16 5.93E+l6 7.09E+l6 7.96E+16 Out l et Nozzle to Uppe r Shell Welds -Lowest Extent Nozzle 1<*) 6.05E+16 6.73E+16 7.41E+1 6 8.77E+l6 9.80E+16 Nozzle 2<*) 6.05E+16 6.73E+16 7.4 1 E+l6 8.77E+l 6 9.80E+l6 Uppe r Shell t o Intermed i a t e Shell Circumferential Weld 8-203 3.53E+l7 3.96E+l7 4.38E+17 5.24E+l7 5.89E+l7 Intermediate Shell Plates C-8009-1 , -2 , -3 3.02E+19 3.36E+19 3.70E+19 4.39E+l9 4.91E+19 Intermed i ate Shell Longitudina l Welds 2-203 A 2.89E+l9 3.21E+19 3.53E+1 9 4.16E+l9 4.64E+l9 2-203 B 2.22E+19 2.48E+19 2.75E+1 9 3.28E+l9 3.68E+19 2-203 C 2.22E+l9 2.48E+19 2.75E+1 9 3.28E+\9 3.68E+l9 Intermediate Shell to Lower Shell Circ um ferential Weld 9-203 3.00E+19 3.35E+l9 3.69E+l 9 4.38E+l9 4.89E+l 9 Lowe r She ll Plates C-8 010-1 , -2, -3 3.02E+l9 3.37E+19 3.73E+l 9 4.44E+19 4.98E+1 9 Lowe r S h ell Longitudinal Welds 3-203 A 2.89E+I9 3.22E+19 3.55E+l9 4.2IE+I9 4.71E+19 3-203 B 2.22E+l9 2.49E+19 2.77E+19 3.32E+l9 3.73E+l9 3-203 C 2.22E+19 2.49E+l9 2.77E+l 9 3.32E+l 9 3.73E+1 9 Lower She ll to Bottom Head Circumferentia l Weld 10-203 5.41E+l6 6.06E+16 6.71E+l6 8.0IE+l 6 8.98E+l 6 N otes: (a) The axial location used corresponds to the bottom of the vessel s upport pad of the inlet no zz le , instead of the no zz le to upper shell weld. This provides a bo undin g fluence for the no zzle to upper shell weld. (b) Va lu es are based on the average power distribut ions and core operating conditions of Cyc l es 23-25. WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 18 of 88 Westinghouse Non-Proprietary Class 3 2-7 Table 2-3 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Material Iron Atom Displacements

<bl (dpa) 32 EFPY 36 EFPY 40 EFPY 48 EFPY 54 EFPY Inlet Nozzle to Upper She ll Welds -Lowest Extent Nozzle 1<*l 2.83E-04 3.1 7E-04 3.51E-04 4.1 9E-04 4.7 0E-04 Nozzle i*l 2.83E-04 3.17E-04 3.51 E-04 4.19E-04 4.70E-04 Nozzle 3<*l 2.83E-04 3.17E-04 3.51E-04 4.19E-04 4.70E-04 Nozz le 4<*l 2.83E-04 3.1 7E-04 3.51E-04 4.19E-04 4.70E-04 Outlet Nozzle to Upper Shell Welds -Lowest Extent Nozz l e 1<*l 3.37E-04 3.75E-04 4.14 E-04 4.91E-04 5.48E-04 Nozzle 2<*l 3.37E-04 3.75E-04 4.14E-04 4.91E-04 5.48E-04 Upper Shell to Intermediate Shell Circumfe r ential Weld 8-203 6.41E-04 7.1 8E-04 7.95E-04 9.50E-04 1.07E-03 I ntermediate Shell Plates C-800 9-1, -2, -3 4.59E-02 5. l lE-02 5.64E-02 6.69E-02 7.47E-02 Intermediate Shell Long itudin al Welds 2-203 A 4.42E-02 4.90E-02 5.39E-02 6.36E-02 7.09E-02 2-203 B 3.39E-02 3.80E-02 4.20E-02 5.02E-02 5.63E-02 2-203 C 3.39E-02 3.80E-02 4.2 0E-02 5.02E-02 5.63E-02 Intermediate Shell to Lower Shell Circumferential Weld 9-203 4.57E-02 5.IO E-02 5.62E-02 6.67E-02 7.45E-02 Lower Shell Plates C-8010-1 , -2 , -3 4.59E-02 5. I 3E-02 5.67E-02 6.76E-02 7.58E-02 Lower She ll Longitud in al Welds 3-203 A 4.41E-02 4.92E-0 2 5.42E-02 6.43E-02 7.19E-02 3-203 B 3.39E-0 2 3.81E-02 4.23E-02 5.08E-02 5.71E-02 3-203 C 3.39E-02 3.8 1 E-02 4.23E-02 5.08E-02 5.71E-02 Lower She ll to Bottom Head Circumferentia l Weld 10-203 2.85E-04 3.19E-04 3.53E-04 4.22E-04 4.73E-04 Notes: (a) The axial location u sed co rresponds to the bottom of the vessel support pad of the inlet no zz le , instead of the nozzle to upper s hell weld. This provides a bound i ng fluence for th e nozzle to upp er s h ell weld. (b) Values are based on the average power di st ributions a nd core operating conditions of Cyc l es 23-25. WCAP-18169-NP June2018 Revision 1 Table 2-4 Cycle 1 2 3 4 5 6 7 8 9 10 II 12 13 14 15 16 17 18 19 20 21 22 23 24 25<*) Note: CALC-AN02-EP-17-00002 Rev O Page 19 of 88 Westinghouse Non-Proprietary Class 3 2-8 Calculated Azimuthal Variation of the Maximum Fast Neutron (E > 1.0 MeV) Fluence Rate at the Reactor Vessel Clad/Base Metal Interface Cycle Cumulative Fluence Rate (n/cm 2-s) Operating Length Time o o 15° 30° 45° Maximum (EFPY) (EFPY) 0.89 0.89 3.62E+l0 3.60E+IO 2.71E+IO 2.63E+IO 3.81E+IO 0.80 1.69 4.59E+l0 4.51E+l0 3.38E+l0 3.22E+10 4.80E+l0 0.6 4 2.33 4.34E+l0 4.21E+l0 3.18E+l0 3.12E+l0 4.52E+IO 0.9 7 3.31 4.36E+l0 4.12E+l0 3.34E+l0 3.33E+10 4.46E+l0 0.85 4.16 4.68E+l0 4.52E+l0 3.34E+l0 3.30E+10 4.8 6E+IO 1.22 5.38 3.40E+IO 3.40E+l0 2.54E+l0 2.34E+10 3.64E+l0 1.13 6.51 3.l 7E+l0 3.03E+IO 2.39E+l0 2.20E+l0 3.30E+l0 1.15 7.66 3.25E+IO 3.20E+IO 2.42E+l0 2.17E+l0 3.44E+l0 1.18 8.84 3.20E+l0 3.15E+IO 2.38E+IO 2.09E+IO 3.38E+l0 1.32 10.16 2.98E+l0 2.98E+l0 2.35E+l0 2.26E+l0 3.19E+IO 1.33 11.49 2.36E+IO 2.06E+IO l.73E+IO 1.83E+IO 2.36E+IO 1.31 12.8 1 2.41E+l0 2.30E+l0 l.77E+l0 l.75E+l0 2.48E+l0 1.47 14.27 2.29E+l0 l.99E+l0 l.63E+l0 l.78E+l0 2.29E+l0 1.41 15.69 2.35E+l0 2.28E+ 10 l.79E+l0 l.78E+l0 2.42E+IO 1.29 16.98 2.31E+10 2.23E+10 l.90E+ 10 l.85E+l0 2.34E+l0 1.35 18.33 2.73E+l0 2.49E+ 10 1.90E+l0 1.92E+l0 2.75E+l0 1.36 19.69 2.56E+l0 2.47E+IO l.96E+l0 l.89E+l0 2.63E+10 1.43 21.12 2.77E+l0 2.83E+l0 2.21E+l0 2.13E+l0 2.95E+l0 1.34 22.46 2.71E+l0 2.72E+l0 2.09E+l0 2.06E+l0 2.85E+l0 1.36 23.82 2.72E+l0 2.72E+l0 2.lOE+lO 2.1 lE+lO 2.85E+l0 1.35 25.17 2.72E+l0 2.72E+l0 2.12 E+IO 2.12E+l0 2.86E+IO 1.45 26.61 2.59E+IO 2.59E+IO 2.03E+IO 2.05E+IO 2.72E+IO 1.36 27.98 2.50E+l0 2.68E+l0 2.13E+IO 2.16E+l0 2.76E+l0 1.26 29.24 2.40E+IO 2.57E+10 2.14E+l0 2.20E+l0 2.64E+l0 1.37 30.60 2.9 7E+IO 2.96E+l0 2.29E+l0 2.32E+l0 3.1 lE+lO (a) Cy cle 25 i s t he c urrent operating cycle. Values li ste d for this cycle a r e projections based on the Cycle 25 design. WCAP-1 8169-NP June2018 Revision I Table 2-5 Cy cle 1 2 3 4 5 6 7 8 9 1 0 1 1 12 13 14 15 16 17 18 19 20 2 1 22 23 24 25<*) Futu r e<b) Future (b) Future<b) Futu r e(b) Futu r e(b) Notes: CALC-AN02-EP-17-00002 Rev O Page 20 of 88 We stingho u se Non-Propri etary C l ass 3 2-9 Calculated Az imuthal Variation of the Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Reactor Vessel Clad/Base Metal Interface Cycle C umu la t ive Flu e n ce (n/c m 2 Op e r ating Lengt h Time oo 1 5° 30° 45° Maxi mum (EFPY) (E FP Y) 0.89 0.89 1.0 2E+l8 1.01 E+1 8 7.61E+17 7.37 E+l7 l.07E+1 8 0.80 1.69 2.1 6E+l8 2.14 E+1 8 l.6 1 E+18 1.54E+1 8 2.27E+1 8 0.64 2.33 3.04E+18 2.99E+1 8 2.25E+18 2.1 7E+1 8 3.18E+1 8 0.97 3.31 4.38E+18 4.26E+18 3.27E+1 8 3.1 9E+1 8 4.55E+1 8 0.85 4.1 6 5.6 4 E+18 5.4 8E+1 8 4.1 7E+18 4.08E+1 8 5.86E+l 8 1.22 5.38 6.95E+18 6.78E+18 5.1 5E+18 4.98E+1 8 7.26E+l 8 1.13 6.51 8.0 8E+18 7.86E+18 6.00E+18 5.77E+1 8 8.44E+1 8 1.15 7.66 9.26E+l 8 9.0 2E+18 6.88E+1 8 6.56E+l 8 9.68E+l 8 1.18 8.84 1.04 E+19 1.0 2E+19 7.76E+l 8 7.33E+1 8 l.09 E+19 1.32 10.16 l.1 7E+19 l.14 E+1 9 8.74E+1 8 8.27E+1 8 l.23E+19 1.33 1 1.49 l.27E+I9 l.23E+19 9.4 7E+1 8 9.04E+1 8 1.32E+1 9 1.31 1 2.81 l.37E+l9 1.3 2E+19 l.02E+1 9 9.76E+1 8 1.43E+1 9 1.47 14.27 l.47E+l9 1.4 2E+19 1.09E+1 9 l.06E+19 l.53E+19 1.41 1 5.69 1.5 7E+ 19 l.51E+l9 l.17E+1 9 1.14E+1 9 l.63E+l 9 1.29 1 6.98 1.6 7 E+l9 l.6 0 E+l 9 l.25E+1 9 l.21E+l 9 l.73E+1 9 1.35 1 8.33 l.78E+19 l.7IE+19 1.33E+l 9 l.29E+l 9 l.84E+l9 1.36 1 9.69 l.89E+19 l.81E+l9 l.41E+l 9 1.37E+I 9 l.95E+l 9 1.43 21.12 2.01E+l9 1.9 4 E+l9 l.51E+l 9 l.46E+l 9 2.08E+19 1.34 22.46 2.12E+l9 2.0 5E+l 9 1.59E+l 9 l.55E+l9 2.20E+l 9 1.36 23.8 2 2.23E+19 2.1 6E+l 9 1.68E+l 9 l.6 4 E+l9 2.32E+19 1.35 25.17 2.35E+l9 2.27E+l 9 l.77E+l 9 l.72E+l 9 2.43E+1 9 1.45 26.6 1 2.46E+l9 2.39E+l9 l.86E+l 9 l.81E+l 9 2.55E+I 9 1.36 27.98 2.57E+l9 2.50E+l9 l.95E+l 9 l.91E+l 9 2.67E+l 9 1.26 29.24 2.66E+l9 2.6 0E+l 9 2.03E+l 9 l.99E+l 9 2.77E+l9 1.37 30.60 2.7 8E+l9 2.72E+l 9 2.1 2E+l9 2.0 8E+l 9 2.9 0 E+l 9 32.00 2.89E+l9 2.84E+l9 2.22E+l 9 2.1 8 E+l 9 3.02E+l 9 36.00 3.22E+l9 3.1 8E+l9 2.49E+l 9 2.46E+l 9 3.37E+l 9 40.00 3.55E+ 19 3.53E+19 2.77E+l9 2.74E+l 9 3.73E+l 9 48.00 4.2 1 E+l9 4.22E+l 9 3.32E+l9 3.30E+l 9 4.44E+l 9 54.00 4.71E+l9 4.74E+l 9 3.73E+l 9 3.72 E+l 9 4.98E+l9 (a) Cyc l e 25 i s t h e cu rr e nt o p erating cyc l e. Va lu es listed for thi s cycle a r e p roj ect i o n s b ased o n t h e Cyc l e 25 des i gn. (b) Va l ues b e y o n d Cy cl e 25 a r e based o n t h e average powe r di s t ri bu tions a nd co r e operating co n d i t i ons of Cyc l es 23-25. WCA P-1 8 1 69-NP Ju n e2 01 8 R evisi on 1 T able 2-6 Cy cl e l 2 3 4 5 6 7 8 9 10 1 1 12 13 14 15 16 17 18 19 20 21 22 23 24 25<*) Note: CALC-AN02-EP-17-00002 Rev O Page 21 of 88 Wes ti ngho u se No n-P roprie t ary C l ass 3 2-10 C a lculated Iron A tom Di s placement Rate a t the Pr e ssure V esse l Clad/Ba se Metal I nt e rface Cycle Cumulative Iron Ato m Di s pl ace m e nt R a t e (dpa/s) O perating Length Time o o 15° 30° 45° Max imum (EFPY) (EFPY) 0.89 0.89 5.52E-11 5.4 7E-11 4.1 4E-ll 4.0 l E-11 5.79E-11 0.80 1.69 6.99E-l l 6.86E-11 5.1 6E-11 4.91E-1 l 7.29E-l l 0.64 2.33 6.62E-1 1 6.40E-1 1 4.85E-11 4.76E-l l 6.87E-1 l 0.97 3.31 6.65E-l l 6.27E-1 l 5.l OE-11 5.08E-11 6.78E-ll 0.85 4.16 7.13E-11 6.87E-l l 5.l OE-1 1 5.04 E-l l 7.39E-l l 1.22 5.38 5.20E-11 5.1 7E-1 1 3.88E-l l 3.58E-l 1 5.53E-l l 1.13 6.51 4.84E-1 l 4.61E-1 1 3.66E-l l 3.37E-11 5.02E-l l 1.15 7.66 4.97E-l l 4.86E-l l 3.69E-l 1 3.33E-11 5.24E-l l 1.18 8.84 4.89E-l 1 4.79E-1 l 3.64E-1 1 3.21E-l 1 5.1 5E-l l 1.32 1 0.16 4.56E-l l 4.54E-l l 3.59E-l l 3.45E-11 4.84E-l l 1.33 1 1.49 3.60E-1 1 3.1 4E-l l 2.65E-11 2.80E-l l 3.60E-11 1.31 12.81 3.68E-11 3.51E-ll 2.71E-l l 2.68E-l l 3.78E-l l 1.47 14.27 3.50E-l l 3.03E-l l 2.50E-1 1 2.73E-ll 3.50E-11 1.41 15.69 3.59E-1 l 3.48E-l l 2.7 3 E-l l 2.73E-ll 3.69E-1 l 1.29 1 6.98 3.53E-l l 3.41E-l l 2.91E-l 1 2.83E-l l 3.57E-11 1.35 18.33 4. l 7E-l l 3.80E-l l 2.91E-ll 2.95E-l l 4.19E-11 1.36 19.69 3.91E-l l 3.76E-1 1 3.0lE-1 1 2.89E-l l 4.0lE-11 1.43 21.12 4.23E-1 l 4.31E-l 1 3.38E-1 l 3.26E-1 1 4.49E-1 l 1.34 22.46 4.13E-11 4.14E-l l 3.20E-ll 3.16E-11 4.3 4 E-11 1.36 23.82 4.15E-ll 4.l3E-11 3.21E-11 3.23E-l 1 4.3 4 E-11 1.35 25.17 4.16E-ll 4.14E-ll 3.24E-11 3.25E-l l 4.35E-1 l 1.45 26.61 3.95 E-l l 3.95E-l l 3.lOE-1 1 3.13E-l l 4.14E-l l 1.36 27.98 3.82E-l l 4.08E-l 1 3.26E-l l 3.29E-11 4.20E-l 1 1.26 29.24 3.67E-1 l 3.91E-l l 3.27E-1I 3.36E-11 4.02E-11 1.37 30.60 4.53E-1 1 4.51E-1 1 3.50E-11 3.55E-l 1 4.74E-11 (a) C yc l e 25 i s the c urre n t operating c y cle. Va l ue s li s t e d for t h is c y cle a re projec t ions based o n t h e Cycle 25 de s ign. WCA P-1 8 1 69-NP June 2 01 8 R evisio n 1 CALC-AN02-EP-17-00002 Rev O Page 22 of 88 Westinghouse Non-Proprietary Class 3 2-11 Table 2-7 Calculated Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Cycle Cumulative Iron Atom Displacements (dpa) Operating Cycle Length Time o o 15° 30° 45° Maximum (EFPY) (EFPY) 1 0.89 0.89 l.55E-03 l.54E-03 1.16E-03 l.13E-03 l.63E-03 2 0.80 1.69 3.30E-03 3.25E-03 2.45E-03 2.36E-03 3.45E-03 3 0.64 2.33 4.63E-03 4.54E-03 3.43E-03 3.32E-03 4.83E-03 4 0.97 3.31 6.68E-03 6.47E-03 5.00E-03 4.88E-03 6.91E-03 5 0.85 4.16 8.60E-03 8.32E-03 6.37E-03 6.24E-03 8.90E-03 6 1.22 5.38 l.06E-02 l.03E-02 7.86E-03 7.61E-03 1.IOE-02 7 1.13 6.51 1.23E-02 l.20E-02 9.17E-03 8.82E-03 l.28E-02 8 1.15 7.66 l.41E-02 1.37E-02 l.05E-02 l.OOE-02 l.47E-02 9 1.18 8.84 1.59E-02 l.55E-02 l .19E-02 l.12E-02 l.66E-02 10 1.32 10.16 l.78E-02 l.74E-02 l.34E-02 1.27E-02 l.86E-02 11 1.33 11.49 l.94E-02 l.87E-02 l.45E-02 l.38E-02 2.0lE-02 12 1.31 12.81 2.09E-02 2.02E-02 l.56E-02 l.49E-02 2. l 7E-02 13 1.47 14.27 2.25E-02 2.15E-02 l.67E-02 l.62E-02 2.32E-02 14 1.41 15.69 2.40E-02 2.31E-02 l.79E-02 1.74E-02 2.48E-02 15 1.29 16.98 2.54E-02 2.44E-02 l.91E-02 l.85E-02 2.63E-02 16 1.35 18.33 2.72E-02 2.60E-02 2.03E-02 1.97E-02 2.80E-02 17 1.36 19.69 2.89E-02 2.76E-02 2.16E-02 2.lOE-02 2.97E-02 18 1.43 21.12 3.07E-02 2.95E-02 2.31E-02 2.24E-02 3.l 7E-02 19 1.34 22.46 3.24E-02 3.12E-02 2.44E-02 2.37E-02 3.35E-02 20 1.36 23.82 3.41E-02 3.29E-02 2.57E-02 2.50E-02 3.53E-02 21 1.35 25.17 3.58E-02 3.46E-02 2.70E-02 2.63E-02 3.70E-02 22 1.45 26.61 3.76E-02 3.64E-02 2.84E-02 2.77E-02 3.89E-02 23 1.36 27.98 3.92E-02 3.81E-02 2.98E-02 2.91E-02 4.06E-02 24 1.26 29.24 4.06E-02 3.96E-02 3.l lE-02 3.04E-02 4.22E-02 25<*) 1.37 30.60 4.25E-02 4.14E-02 3.25E-02 3.19E-02 4.41E-02 Future<bl 32.00 4.42E-02 4.32E-02 3.39E-02 3.33E-02 4.59E-02 Future(b) 36.00 4.92E-02 4.84E-02 3.81E-02 3.76E-02 5.13E-02 Future<b) 40.00 5.42E-02 5.37E-02 4.23E-02 4.19E-02 5.67E-02 Future<b) 48.00 6.43E-02 6.42E-02 5.08E-02 5.05E-02 6.76E-02 Future<b) 54.00 7.19E-02 7.21E-02 5.71E-02 5.69E-02 7.58E-02 Notes: (a) Cycle 25 is the current operating cycle. Values listed for this cycle are projections based on the Cycle 25 design. (b) Values b e yond Cycle 25 are based on the average power distributions and core operating conditions of Cycles 23-25. WCAP-18 1 69-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 23 of 88 Westinghouse Non-Proprietary Class 3 Table 2-8 Calculational Uncertainties Description Capsule and Vessel IR PCA Comparisons 3% H.B. Robinson Comparisons 3% Analytical Sensitivity Studies 11% Additional Uncertainty for Factors not Explicitly Evaluated 5% Net Calculational Uncertainty 13% WCAP-18169-NP 2-12 June 2018 Revision 1 Figure 2-1 CALC-AN02-EP-17-00002 Rev O Page 24 of 88 Westinghouse Non-Proprie tary Class 3 AN0-2 Vessel Model -OORT -r,t Geometry with Capsules Meshes: 166R,123 S -c.. _ .... _ .... _ -C.*-----*'-'°" --*--....,..,,,.

2-13 Arkansas Nuclear One Unit 2 r,0 Reactor Geometry Plan View at the Core Midplane with Surveillance Capsules WCAP-18169-NP June 2018 Revision 1 CALC-AN02-EP-17

-00002 Rev O Page 25 of 88 Westinghouse Non-Proprietary Class 3 A N0-2 Vessel M odel -DORT -r ,t Geometry without Capsules Meshes: 166R,116 9 _ ... -(,*--,._, *-°" -,._,, ..... -..... -... _ _ ... _ ----2-14 Figure 2-2 Arkansas Nuclear One Unit 2 r,0 Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules WCAP-18169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 R e v O Page 26 of 88 Westinghou s e Non-Proprietary C l ass 3 AN0-2 Vessel Model -DORT -r, z Geometry Mes h es: 161R , 222Z -r,,* -C<<t llrool -c.-.ir ..... -,,_, v .... 0:,6 -,-, , ..... >>"' -CH 0, '° "' e N "1 N I .., .., I;! I . ci !51 I "' .,; "' N I .... .; "' ',\:).O 9 U 182.9 2 7 0 3 65.8 R [cm l --Fig u re 2-3 Ar k a n s a s N ucl ear On e U nit 2 r,z React o r Geo m etry Se c t ion V ie w WCAP-18 1 69-NP 2-15 June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 27 of 88 Westinghouse Non-Proprietary Class 3 3 FRACTURE TOUGHNESS PROPERTIES 3-1 The requirements for P-T limit curve development are specified in 10 CFR 50 , Appendix G [Ref. 4]. The beltline region of the reactor vesse l is defined as the following in 10 CFR 50, Appendix G: "the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent r eg ions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage. " The Arkansas Nuclear One Unit 2 beltlin e materials traditionally included the intermediate and lower shell plate and weld materials; however, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. 8], any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 10 1 7 n/cm 2 (E > 1.0 MeV) at the end of the licensed operating period should be considered in the development of P-T limit curves. The materials that exceed this fluence threshold are referred to as extended beltline materials and are evaluated to ensure that the applicable acceptance criteria are met. As seen from Table 2-2 of this report, the extended beltline materials include the upper shell plates , upper shell longitudinal welds, and the upper to intermediate shell girth weld. Note that for reactor vessel welds, the terms " girth" and "circumferential" are used interchangeably; herein, these welds sha ll be referred to as girth welds. Similarly, for reactor vesse l welds , the terms "axial" and "longitudina l" are used intercha ng eably; herein, these we lds shall be referred to as longitudinal welds. Although the reactor vessel nozzles are not a part of the extended beltline, per NRC RIS 2014-11 , the nozzle materials must be evaluated for their potential effect on P-T limit curves regardless of exposure -See Appendix B for more details. As part of this P-T limit curve development effort, the methodology and eval uations used to determine the initial RT NOT values for the Arkansas Nuclear One Unit 2 reactor vessel beltline and extended beltline base metal materials were reviewed and updated , as appropriate.

The initial RT NDT value of Intermediate Shell Plate C-8009-3 was determined per ASME Code ,Section III , Subsection NB-2300 [R ef. 9]. The initial RT NOT values for each of the other eight reactor vessel plates were determined per BTP 5-3, Paragraph B l.1(3) [Ref. 10] in conj unction with ASME Code,Section III, Subsection NB-2300 [Ref. 9]. These initial RTNDT values were determined using both BTP 5-3 Position l.1(3)(a) and Position 1.1(3)(b), and the more limiting initial RT DT value was chosen for each material.

A summary of the b est-estimate copper (Cu), nickel (Ni), Manganese (Mn), and Phosphorus (P) contents, in units of weight percent (wt. %), as well as initial RT DT values for the reactor vessel beltline and extended beltline materials are provided in Table 3-1 for Arkansas Nuclear One Unit 2. Table 3-2 contains a summary of the initial RT N DT values of the reactor vessel flange, reactor vessel closure head, replacement reactor vessel closure head, and b alance of the reactor coolant system (RCS). These values serve as input to the P-T limit curves "flange-notch" and LST per Appendix G of 10 CFR 50 and ASME Code, Section Ill, respectively

-See Sections 6.3 and 6.4 for details. WCAP-1 8169-NP June2018 Revision l CALC-AN02-EP-17

-00002 Rev O Page 28 of 88 Westinghouse No n-Propri etary C la ss 3 3-2 Table 3-1 Summary of the Best-Estimate Chemistry and Initial RT N DT Values for the Arkansas Nuclear One Unit 2 Reactor Vessel Materials Reactor Vessel Material Chemical Composition<bl Heat Number<*>

and Identification Number<*>

Wt.% Wt.% Wt.% Cu Ni Mn Reactor Vessel Beltline Materials Intermediate Shell Plate C-8 009-1 C8161-3 0.098 0.605 1.35 Intermediate Shell Plate C-8009-2 C8161-l 0.085 0.600 1.37 Intermediate Shell Plate C-8 009-3 C8182-2 0.096 0.580 1.36 Lower Shell Pl ate C-8 010-1 C8161-2 0.085 0.585 1.33 Lower Shell Plate C-8 010-2 B 2545-l 0.083 0.668 1.36 Lower Shell Plate C-80 I 0-3 B2545-2 0.080 0.653 1.36 Intermediate Shell Longitudinal Welds Multiple (d) o.05<d> 1.oo<d> l.16 (d) 2-2 0 3A, B , & C Lower Shell Longitudinal Welds 10120 0.046 (e) 0.08zC 0> l.21 3-203A , B , & C Intermed i ate to Lower She ll Girth 83650 0.045<*) 0.08i 0> 1.24 Weld 9-203 Reactor Vessel Extended Beltline Materials Uooer Shell Plat e G> C-8 008-1 C8182-I 0.13 0.60 1.36 Uppe r Shell Plate C-80 08-2 C7605-l 0.13 0.55 1.36 Uooe r Shell Plate C-8008-3 C8571-2 0.08 0.55 l.29 Upper She ll Longitudinal Welds BOLA 0.02 0.93 1.02 l-203A , B , & C Upper to Intermediate She ll Girth 10137 0.22 0.02 0.94 6329637 0.21 0.11<1) 1.22 Weld 8-203 FAGA 0.03 0.95 1.00 Surveillance Weld Data<il Arka n sas N ucl ear One Unit 2 83650 0.045 0.083 1.33 Ca l vert Cliffs Unit 2 101 37 0.2 1 0.06 ---Millstone Unit 2 0.2 1 0.06 ---J.M. Farley Unit 2 BOLA 0.028 0.89 ---Notes on foll o wing page. WCA P-1 8 1 69-NP Wt.% p 0.010 0.010 0.012 0.009 0.00 8 0.008 o.014<d> 0.01 2 0.006 0.011 0.013 0.014 0.010 0.015 0.011 0.008 0.007 ---------Fracture Toughness Property Initial RT N D T (OF) -1.4 0.5 o.o<s) 1 2.0 -16.7 -22.6 5 6 .40<&) 1 2.2 60.5 27.3 -60 (g) 56 -24<hJ ------------June2018 Revision I (c)

Notes: CALC-AN02-EP-17-00002 R e v O Page 29 of 88 Westinghouse Non-Proprietary Class 3 3-3 (a) The reactor vessel plate and weld material identification and heat numbers were taken from the Arkansas Nuclear One Unit 2 Certified Materia l Test Reports (CMTRs) and/or A-PENG-ER-002

[Ref. 1 1], unless otherwise noted. (b) All chemistry va l ues obtained from A-PENG-ER-002 and/or the Arkansas Nuclear One Unit 2 CMTRs, u nl ess otherwise noted. Chemistry values for plates are t h e average of all available analyses. Chemistry values for welds are the average of all coated electrode deposit chemistry (CE D C) for the E-8018 stick electrodes or weld flux deposit chemistry (WFDC) for Linde 0091 welds, unless otherwise noted. Where avai l able, additional c h emistry analysis results from BMI-0584 [Ref. 1 2] were also included in the average. The chemistry va l ues for the beltline plates reported in this table are identical to those previously reported in the Arkansas Nuclear One Unit 2 License Renewal Application (LRA) [Ref. 13] and the previous capsule r eport, BAW-2399, Revision I [Ref. 14]. (c) The RT N OT (UJ values for the plates are based on drop-weight data , longitudinally-oriented Charpy V-notch test data and NUREG-0800 , BTP 5-3 Position 1.1(3)(a) and (b) [Ref. 1 0], with the more limiting RTNoT(UJ value being se l ected , un l ess otherwise noted. The RT NOT (UJ values for welds are t h e generic value for Linde 0091 flux type welds (-56°F) per IO CFR 50.61 [Ref. 15], unless otherwise noted. (d) The material Heat numbers for the Intermediate Shell Longitudinal Welds 2-203A, B, & Care unclear in the historical data. For conservatism, the material properties for the Intermediate Shell Longitudinal Welds 2-203A , B , & C reported are the most limiting values from welds relevant to the Intermediate Shell Longitudinal Welds 2-203A , B, & C per A-PENG-ER-002. These welds include Heat# 10120, Flux Type Linde 0091 , Lot# 3999 (sister plant weld), Heat# 10120 , 10120, Heat# AAGC, and the ana l ysis of the in-process weld depos i t chemistry. (e) The Cu and Ni wt. % values for the Lower Shell Longitudinal Welds 3-203A , B , & C and the Intermediate to Lower Shell Girth Weld 9-203 are consistent with BAW-2399 , Revision I [Ref. 14] and were originally taken from CE NPSD-1039, Revision 2 [Ref. 16]. (t) Weld Heat# 6329637 does not contain any WFDC Ni wt.% values , thus t h e bare wire chemical analysis (BWCA) value of 0.11 % from A-PENG-ER-002

[Ref. 11] was used. (g) Th i s RT N OT (U) value for the surveillance plate, Intermediate Shell Plate C-8009-3 , is based on drop-weig h t data, transverse orientation Charpy V-notch test data taken from the baseline capsule test report , TR-MCD-002

[Ref. 17] and ASME Code Section Ill Subarticle NB-2331 [Ref. 9]. The RT N OT (UJ value for the two weld materials, Heat numbers 83650 and BOLA, is based on drop-weight data , Charpy V-notch test data taken from A-PENG-ER-002 and ASME Code Section Ill Subarticle NB-233 1 [Ref. 9]. (h) Drop-weight test data is not available for this E-8018 weld heat. Therefore, to assign an RT N OT (UJ value to this E-8018 stick electrode weld, the data in Table 8 of A-PENG-ER-002 was analyzed. The average T NOT value for the 17 E-8018 weld heats is -57"F with a s tandard deviation of 16.5"F. This yields a bounding value of -24"F using a mean plus two sigma model; therefore, a value of -24"F is acceptable for the initial RT NOT value of this weld material , with consideration that its Charpy impact energy at IO"F, which i s less than TNoT + 60"F , was greater than 100 ft-lb. Furthermore, -24°F bounds all of the E-8018 stick electrode T NOT values present in A-PENG-ER-002

[Ref. 11). (i) Surveillance data exists for weld Heat# 83650, # IO 137 , and# BOLA from multiple sou r ces; see Section 4 for more detai l s. The data for Arkansas Nuclear One Unit 2 weld metal Heat# 83650 was taken as the average of the data available from MCD-002 , as well as the subsequent analyses completed during testing of the first capsule , BMl-0584 [Ref. 12]. The data for the Calvert Cliffs Unit 2 weld metal Heat# IO 137 was taken from Table 4-3 of WCAP-1750 I-NP [Ref. 18). The data for the Millstone Unit 2 weld metal Heat# 10137 was taken from Table 4-1 ofWCAP-16012

[Ref. 19]. The data for the J.M. Farley Unit 2 weld metal Heat# BOLA was taken from Table 4-1 of WCAP-16918, Revision I (Ref. 20]. U) Upper Shell Plate C-8008-1 shares the same material heat number as t h e Arkansas Nuclear One Unit 2 surveillance p l ate material , Intermediate Shell Plate C-8009-3; therefore, surveillance program test results apply to this material as well. WCAP-18169-NP June 2018 Revision I Table 3-2 CALC-AN02-EP-17-00002 Rev O Page 30 of 88 Westinghouse Non-Proprietary Class 3 3-4 Summary of Arkansas Nuclear One Unit 2 Reactor Vessel Closure Heads, Vessel Flange and Balance of RCS Initial RT NDT Values Reactor Vessel Material Initial RT NDT Methodology (OF) Current C lo sure Head 10 BTP 5-3, Paragraph B1.1(3)(a) and (b) [Ref. 10] (Heat# 125B404) Re p lacement Closure Head<*J -22 ASME Code,Section III , Subsection NB-2300 (Heat# R378 l/R3782) [Ref. 9] Vessel Flange 30 BTP 5-3, Paragraph B1.1(3)(a) and (b) [Ref. 10] (Heat# 122A440) Balance of Rcs Cb> 50 Note (b) N otes: (a) A replacement, si ngle forging, closure head has been fabricated; however , it has not yet been insta lled at Arkansas Nuclear One Unit 2. The vessel flange initial RT N oT value is higher than both the current and replacement clo sure head initial RT NDT values. Thus , the results contained herein are conservative for the current and replacement closure heads. (b) 50°F was conservatively assigned to all RCS material not specifically te s ted p e r Section 5.2.4.3 of the Arkansas Nuclear One Unit 2 updated Final Saf ety Analysis Report (FSAR). WCAP-18169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 31 of 88 Westinghouse Non-Proprietary C l ass 3 4 S U RVEILLANCE DATA 4-1 Per Regulatory Guide 1.99, Revision 2 [Ref. 1 ], calc ul ation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, data from survei llance programs at other plants which include a reactor vessel beltline or extended beltline material sho uld also be considered when calculating Position 2.1 chemistry factors. Data from a surveillance program a t another plant is often called 'sister plant' data. The survei ll ance caps ule plate material for Arkansas Nuclear One Unit 2 is from Intermediate Shell Plate C-80 09-3. Surveillance results from this plate also apply to Upper Shell Plate C-8008-1 , because the two plates wer e made from the same heat of material (Heat # C8182). The surveillance capsule weld material for Arkansas Nuclear One Unit 2 is Heat # 83650, which is applicable to the intermediate to lower shell girth weld. Table 4-1 summarizes the Arkansas Nuclear One Unit 2 surveillance data for the plate materia l an d we ld ma ter ia l (Heat # 83650) that will be used in the calc ulation of the Position 2.1 chemistry factor val ues for these materials.

The results of the last withdrawn and tested surveillance capsule , Caps ul e 28 4°, were documented in WCAP-18166-NP

[Ref. 21]. Appendix D of WCAP-18166-NP concluded that the surveillance plate and weld (Hea t # 83650) data are credi bl e; therefore , a reduced margin term will b e utilized in t he ART calculations contained in Section 7. The Arkansas Nuclear One Unit 2 reactor vessel upper to intermediate shell girth weld seam was fab r i cated using weld Heat# 10137. Weld Heat # 10137 is contained in the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs.

Thus, the Ca l vert C li ffs Unit 2 and Millstone Unit 2 data will be used in the calculation of the Position 2.1 chemistry factor val u e for Arkansas Nuclear One Unit 2 weld Heat # 10 13 7. Note that no s urveillance data is available for the other two Heats (# 6329637 and # FAGA) which were also used to make the Upper to Intermediate Shell Girth Weld 8-203. Table 4-2 su mmarize s the applica bl e surveillance capsule data pertaining to weld Heat # 10137. The combined surve illanc e data is deemed credi ble per Append i x D; however, as a result of the Mi ll stone Unit 2 surve illanc e data including both weld Heat # 10137 and 90136, the Position 2.1 chemistry factor calc ulation s for wel d Heat# 10137 will utilize a full margin term for conservatism.

See Appendix D for details. The Arkan s as Nuclear One Unit 2 reactor vesse l upper she ll lon gitudinal we ld seams were fabricated using weld Heat # BOLA. Weld Heat # BOLA is contained in the J.M. Farley Uni t 2 surveillance program. Thus, the J.M. Farley Unit 2 data will b e used in the calculation of the Position 2.1 chemistry facto r va lu e for Arkansas Nuclear One Unit 2 weld Heat# BOLA. Table 4-3 summarizes the applicab l e surve illanc e capsule data pertaining to weld Heat # BOLA. Per Appendix D of WCAP-1 6918-NP, Revision 1 [Ref. 20], the J.M. Farley Unit 2 survei ll ance weld data is deemed non-credible. Since the J.M. Farley Unit 2 survei llance weld is not analyze d with any a dditional surveillance capsule material herein, this credibility conclusion is applicable to the Arkansas N ucl ear One Unit 2 weld Heat # BOLA. Therefore, a full margin term will be utilized in the ART calc ulation s containe d in Section 7. WCAP-1816 9-NP June2018 Re visio n 1 Westinghouse Non-Proprietary C l ass 3 4-2 Table 4-1 Arkansas Nuclear One Unit 2 Surveillance Capsule Data Material Capsule(*> Intermediate Shell Plate C-8009-3 (Longitudina l) 97° 284° 97° Intermediate Shell Plate C-8009-3 (Transverse) 104° 284° 97° Surveillance Weld Material (Heat# 836 50) 104° 284° No t e: (a) Surveillance data was tak e n from Table 5-10 of WCAP-18166-NP

[Ref. 21]. WCAP-18169-NP Capsule Fluence<*> (x 10 1 9 n/cm2, E > 1.0 MeV) 0.303 3.67 0.303 2.15 3.67 0.303 2.15 3.67 Measured 30 ft-lb Transition Temperature Shift<*> (°F) 23.5 85.7 33.4 52.9 85.6 13.2 16.1 12.0 June 2018 Revi sio n 1 (") l> r (") )> z 0 "' I m I .... -.,j c:, 0 0 0 "' ::0 CD < 0 Ill IQ CD (,) "' 0 -co co Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 Ca l vert Cliffs Unit 2 and Millstone Unit 2 S u rveillance Capsule Data for We l d Heat# 10137 Ca p s ul e<*> C a p s ul e F lu e n ce<*> Meas ur ed 3 0 ft-l b Tra n siti on In l et Te mp eratu r e(bl Tem p e r at u re Material (x 10 19 n/cm2, E > 1.0 MeV) Temperat u re Shift<*> (°F) Adj u st m ent<<) (°F) {°F) 263° 0.825 72.7 550 -1.0 Calvert Cliffs Unit 2 Data 97° 1.95 82.9 549 -2.0 104° 2.44 69.7 548 -3.0 97° 0.324 65.93 544.3 -6.7 Millstone Unit 2 104° 0.949 52.12 547.6 -3.4 Data 83° 1.74 56.09 548.0 -3.0 Notes: (a) For surveillance weld Heat# 10137 , data pertaining to Calvert Cliffs Unit 2 were taken from Table 5-10 of WCAP-17501-NP [Ref. 18]. Data pertaining to Millstone Unit 2 were taken from Table 5-10 ofWCAP-16012

[Ref. 19). (b) Inlet temperatures were calculated as the average inlet temperature from all the previously completed cycles at the time of capsule withdrawal. (c) Temperature adjustment=

l .O*(T capsu t e -T p 1 a n1), where T p tan, = 55 l .0°F for Arkansas Nuclear One Unit 2 (applied to the weld ~T N DT data for each of the Calvert Cliffs Unit 2 and M i llstone Un i t 2 capsules in the Position 2.1 chemistry factor calculation

-See Section 5 for more details). T a bl e 4-3 J.M. Fa rl ey U nit 2 S ur ve ill a nc e Capsul e D a t a for We ld Heat# BO LA Ca p s ul e<*> Ca p s ul e F lu e n ce<*> Meas ur e d 3 0 ft-lb T r ans ition Inl et Te mp e r a tur e<hl Te mp e r a tur e Ma t e ri a l (x 10 19 n/c m 2, E > 1.0 MeV) Tem p e r a tur e S hi f t<*> (0 F) A dju s t ment<c) (°F) {°F) u 0.605 -28.4 544 -7.0 w 1.73 7.0 542 -9.0 J.M. Farley Unit X 2.98 -15.6 543 -8.0 2 z 4.92 1 0.2 543 -8.0 y 6.79 69.l 543 -8.0 V 8.73 56.5 542 -9.0 Notes: (a) For surveillance weld Heat# BOLA , data pertaining to J.M. Farley Unit 2 were taken from Table 5-10 ofWCAP-16918-NP, Revision I [Ref. 20]. (b) Inlet temperatures were calculated as the average inlet temperature from all the previously completed cycles at the time of capsule withdrawal. (c) Temperature adjustment=

J.O*(T caps ul e -Tp1an,), where Tplant = 551.0°F for Arkansas Nuclear One Unit 2 (applied to the weld li.RTNDT data for each of the J.M. Farley Unit 2 capsules in the Posi t ion 2.1 chem i stry factor calculation

-See Section 5 for more deta i ls). WCAP-18169-NP June 2018 Revision I 0 )> r 0 :i:,. z 0 N I m "'CJ I ..... ....., I 0 0 0 0 N ,, (D < 0 "'CJ Ill IQ (D w w 0 -co co CALC-AN02-EP-17-00002 Rev O Page 34 of 88 Westin g hou s e Non-Proprietary Class 3 5 CHEMISTRY FACTORS 5-1 The chemistry factors (CFs) were calculated using Regulatory Guide 1.99 , Revision 2 , Positions 1.1 and 2.1. Position 1.1 chemistry factors for each reactor vessel material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of Regulatory Guide 1.99 , Revision 2. The b est-estimate copper and nickel weight percent values for the Arkansas Nuclear One Unit 2 reactor vessel materials are provided in Table 3-1 of this report. The Position 2.1 chemistry factors are calculated for the materials that have available surveillance program result s. The calculation is performed using the method described in Regulatory Guide 1.99, Revision 2. The Arkansas Nuclear One Unit 2 surveillance data as well as the applicable sister plant data was summarized in Section 4 of this report , and will be utilized in the Position 2.1 chemistry factor calculations in this Section. The Position 2.1 chemistry factor calculations are presented in Tables 5-1 through 5-4 for Arkansas Nuclear One Unit 2 reactor vessel materials that have associated surveillance data. These value s were calculated using the surveillance data summarized in Section 4 of this report. All of the surveillance data i s adjusted for irradiation t e mperature and chemical composition differences in accordance with the guidance pres e nt e d at an industry meeting held by the NRC on February 12 and 13 , 1998 [Ref. 22]. Margin will be applied to th e ART calculations in Section 7 according to the conclusions of the credibility e v aluation for each of the surveillance materials , as documented in Section 4. The Position 1.1 chemistry fa ctors are summarized along with the Position 2.1 chemistry factors in Table 5-5 for Arkansa s Nuclear One Unit 2. W C AP-181 6 9-NP June 2018 Re v ision 1 CALC-AN02-EP-17

-00002 Rev O Page 35 of 88 Westinghouse Non-P roprietary C l ass 3 5-2 Table 5-1 Calculation of A rkansas N ucl e ar One U nit 2 Chemistr y Factor V alue for Intermediate S hell Plat e C-8009-3 U s ing Surveillance Capsule Data Int e rmediate Shell (IS) C a psul e J A RT N D T (d) FF*ART NDT C a p s ul e FF<cl FF 2 Pl a t e C-8009-3 D at a<*> (x 10 1 9 n/cm 2, E > 1.0 M eV) (O F) (O F) Longitu d i n a l Orientatio n 97° 0.3 0 3 0.6 728 23.5 1 5.8 1 0.453 284° 3.67 1.3373 85.7 1 1 4.60 1.788 97° 0.303 0.6 728 33.4 22.47 0.453 Transve r se Orientatio n 104° 2.15 1.2 0 79 52.9 63.90 1.459 284° 3.67 1.3373 85.6 114.47 1.788 SUM: 33 1.26 5.9 41 CF,s P l ate C-8 00 9-3 = L(FF

  • L'1RT NOT)-;-L(FF 2) = (33 1.26)-;-(5.94 1) = 55.8°F N ot es: (a) T h is surveilla n ce data applies to b ot h Intermediate S h e ll Plate C-8009-3 and Upper S h e ll Plate C-8008-1, since the two plates were made from the same heat of material (Heat# C8 l 82). (b) f= fluence. (c) FF= fluence factor= t<0*2 s-o.io*i o g ()_ (d) tiRT NDT values a r e the measured 30 ft-lb shift values. All values are taken from Table 4-1 of this report. Ta ble 5-2 C alculation of A rkan s a s N ucl e ar One U nit 2 Chemi s tr y F actor V alue for Weld He a t # 8 3 650 Using S urveillance Cap s ule Data We ld M e tal Ca p s ul e J<*l A RT N D T (c) FF*A RT NDT C a p s ul e FF (b l FF 2 H ea t# 8 3 650 (x 10 1 9 n/c m2, E > 1.0 MeV) (OF) (O F) 97° 0.303 0.6728 13.3 (13.2) 8.97 0.453 Arkansas N u clear 104° 2.15 1.2079 16.3 (16.1) 1 9.64 1.459 One Unit 2 D ata 284° 3.67 1.3373 12.1 (12.0) 1 6.21 1.788 SUM: 44.8 2 3.7 00 CFweldHeat
  1. 83 650= L(FF
  • L'1RT N oT)-;-L(FF 2) = (44.82)-;-(3.70)

= 12.1°F Notes: (a) f= fluence. (b) FF= fluence factor= t<0*2 s-o.i o*i osf)_ (c) tiRT N DT values are the meas u red 3 0 ft-lb shift va l ues. The tiRT NDT values are adjusted using t h e ratio p roced u re to ac c ount for diffe r ences in the s u rveillance weld c h emistry and t h e reacto r vesse l we l d chemistry (pre-adj u sted values a r e listed in parentheses and were taken from Table 4-1 of this report). Ratio applied to the Arkansas N u clear One Unit 2 surveillance data= CF v ess el W e ld/ Cf surv. Weld= 34.1 °F / 33.7°F = 1.0 I. WCA P-1 8 1 6 9-NP June 20 1 8 R evisio n 1 _j Table 5-3 Weld Metal Heat# 10137 Calvert Cliffs Unit 2 Data Millstone Unit 2 Data (dJ Notes: (a) f= tluence. CALC-AN02-EP-17

-00002 Rev O Page 36 of 88 Westinghouse Non-Proprietary Class 3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat# 10137 Using Surveillance Capsule Data Capsule t<*> L\.RTNDT (c) FF*L\.RTNDT Capsule FF<bl (x 10 19 n/cmZ, E > 1.0 MeV) (OF) (OF) 263° 0.8 25 0.9460 73.1 (72.7) 69.19 97° 1.95 1.1825 82.5 (82.9) 97.58 104° 2.44 1.2401 68.0 (69.7) 84.37 97° 0.324 0.6902 60.4 (65.93) 41.70 104° 0.949 0.9853 49.7 (52.12) 48.97 83° 1.740 1.1523 54.2 (56.09) 62.40 SUM: 404.20 CF we ld Heat# 10137 = }.:(FF

  • LlRT ND T).;.. l:(FF 2) = (404.20).;..

(6.606) = 61.2°F 5-3 FF 2 0.895 1.398 1.538 0.476 0.971 1.328 6.606 (b) FF= tluence factor= t<02 B-o.io*io g I)_ (c) ti.RT NDT val ues ar e the measured 30 ft-lb s hift va lues. The ti.RT N DT values are adju s ted first by the difference in operating temperature then using the ratio procedure to account for differen ces in the s urv eilla nce weld chemistry a nd the reactor vessel weld chemistry (pre-adjuste d values are listed in parenthe ses and were taken from T a ble 4-2 of this report). The temperature adjustments are listed i n Table 4-2. Ratio applied to the Calvert Cliffs Unit 2 surveilla nc e data = CF vesse l Weld / CFsurv. We l d = 98.5°F / 96.8°F = 1.02. Ratio applied to the Millstone Un it 2 surveillance data = CF vesse l We l d I CF s urv. We ld= 98.5°F / 96.8°F = 1.02. (d) Mi ll s tone Uni t 2 s urv eillance data contain s s pecimens from both weld H eat# 10137 and weld Heat# 90136. However , thi s inclusion of a n additio nal heat is not expected to negatively impact the s ub seq uent r eactor vessel integrity calculation results, as additional conservatisms are in place. See Appendix D for more det a il s. WCAP-18169-NP June 2018 Re visio n 1 Table 5-4 CALC-AN02-EP-17-00002 Rev O Page 37 of 88 Westinghouse Non-Proprietary Class 3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat# BOLA Using Surveillance Capsule Data {c) 5-4 Weld Metal Capsule f'3> ~RT NDT FF*~RTNDT Heat#BOLA Capsule (x 10 19 n/cm2, E > 1.0 MeV) FF{bl (OF) (OF) FF 2 u 0.605 0.8593 o.o<d> (-28.4) 0.00 0.738 w 1.73 1.1508 o.o<d) (7.0) 0.00 1.324 J.M. Farley X 2.98 1.2891 o.o<d) (-15.6) 0.00 1.662 Unit2 z 4.92 1.3992 2.2 (10.2) 3.08 1.958 y 6.79 1.4579 61.1 (69.1) 89.08 2.125 V 8.73 1.4960 47.5 (56.5) 71.06 2.238 SUM: 163.22 10.046 CF Heat# BOLA= I:(FF * ~RT N DT) I:(FF 2) = (163.22) -c-(10.046) = 16.2°F Notes: (a) f= fluence. (b) FF= fluence factor= t<0*2 s-O.IO'l ogfl_ (c) ~RT NDT va lues are the measured 30 ft-lb shift values. The ~RT NOT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the s urv eillance weld chemi s try and the reactor vessel weld chemistry (pre-adjusted va lu es are li s ted in parenthe ses and were taken from Table 4-3 of this report). The temperature adj u stments are listed in Table 4-3. A ratio of 1.00 was conservatively applied to the J.M. Farley Unit 2 surveillance data , s inc e CF vessel Weld< CFsurv. W e ld* (d) A negative ~RT NOT value was calculated afte r temperature adj u stment. Phy s ically , thi s s hould not occur; thus a conservative va lue of0.0°F was u sed. WCAP-18169-NP June2018 Revision 1 CA LC-AN02-EP-17-00002 Rev O Page 38 of 88 Westinghouse Non-Proprietary Class 3 5-5 Table 5-5 Summary of Arkansas Nuclear One Unit 2 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Chemistry Factor (°F) and Identification Number Heat Number Position 1.1 <*> Position 2.1 Reactor Vessel Beltline Materials Intermediate Shell Plate C-8009-1 C8161-3 63.6 ---Intermediate Shell Plate C-8009-2 C8161-l 54.5 ---Intermediate Shell Plate C-8009-3 C8182-2 62.2 55.8 (b) Lower Shell Plate C-8010-1 C8161-2 54.5 ---Lower Shell Plate C-8010-2 B2545-l 53.1 ---Lower Shell Plate C-8010-3 B2545-2 51.0 ---Inte rmedi a t e Shell Longitudinal Welds Multiple 68.0 ---2-203A, B, & C Lower Shell Longitudinal Welds 3-203A, B, & C 10120 34.0 ---Intermediate to Lower S hell Girth Weld 9-203 83650 34.1 12.1 (c) Reactor Vessel Extended Beltline Materials Upper Shell Plate C-8008-1 C8182-l 91.0 55.8 (b) Upper Shell Plate C-8008-2 C7605-l 89.5 ---Uppe r Shell Plate C-8008-3 C8571-2 51.0 ---Upper Shell Longitudinal Welds BOLA 27.0 16.2 (d) l-203A , B, & C 10137 98.5 6 l .2(e) Upper to Intermediate Shell Girth Weld 8-203 6329637 100.8 ---FAGA 41.0 ---Surveillance Weld Data Arkansas Nuclear One Unit 2 83650 33.7 ---Ca lv ert C liffs Unit 2 96.8 ---10137 Millstone Unit 2 96.8 ---J.M. Farley Unit 2 BOLA 38.2 ---Notes: (a) Position I.I ch e mistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of th i s report and Tables I and 2 of Regulatory Guide 1.99, Revi sion 2. (b) Position 2.1 chemistry factor was taken from Table 5-1 of this report. As discussed in Section 4, the surve ill a nc e plate data is credible a n d applicable to both Intermediate Shell Plate C-8009-3 and Upper Shell Plate C-8008-1. (c) Position 2.1 chemistry factor was taken from Table 5-2 of this report. As discussed in Section 4 , the surve illan ce weld data for Heat # 83650 is credi bl e. (d) Position 2.1 chemistry factor was taken from Table 5-4 of this report. As discussed in Section 4, the s urveillanc e weld data fo r Heat# BOLA is not credible. (e) Position 2.1 chemistry factor was taken from Table 5-3 of this report. As discussed in Section 4 , the survei llanc e weld data for Heat# l O 137 is credible; however no reduction in the margin term wi ll b e taken. WCAP-18169-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 3 9 of 88 Westinghouse Non-Proprietary Class 3 6 CRITE RI A F O R A L L O WA BL E P RE SSURE-TEM P ERAT URE RELATI O NSHIPS 6.1 O VERALL APPROACH 6-1 The ASME approach for calculating the allowa b le limit curves for vario u s heatup and coo l down rates specifies that the total stress intensity factor , K1, for the com b ined therma l and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress i n tensity factor, K, c , for the metal temperature at that time. K1c is o b tained from the refere n ce fracture to u ghness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Ref. 3]. The K1c curve is given by the following eq u ation: where , K 1 c (ksi-V in.) K l e =33 .2+ 20. 734 *e[0.02 (T-RT ND T )] (1) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ducti l ity temperature RT NDT This K1c curve is based on the l ower bound of static critica l K 1 val u es measured as a f unction of t emperature on s pecimens of SA-533 Grade B Class 1 , SA-508-1, SA-508-2, and SA-508-3 steel. 6.2 METHO DO LOGY F OR PRESSURE-TEM P E RAT URE LIMIT CU R VE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: w h e re, C C W C AP-1 81 6 9-NP stre s s intensity factor c a used by membrane (pres s ure) s tress stress intensity factor caused b y the therma l gradients (2) reference stress intensity factor as a function of the meta l temperature T and the metal r efe rence nil-ductility temperature RT N D T 2.0 for Leve l A and Level B service limits 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical June 2 018 Revi s ion I J CALC-AN02-EP-17-00002 Rev O Page 40 of 88 Westinghouse Non-Proprietary C l ass 3 For mem b rane tension, the corresponding K 1 for the postulated defect is: Kim= Mmx(pR/t) where, Mm for an inside axial surface flaw is given by: Mm 1.85 for Ji < 2 , Mm 0.926 Ji for 2::; Ji ::; 3.464, Mm 3.2 1 for Ji > 3.464 and, Mm for an outside axial surface flaw is given by: Mm 1. 77 for Ji < 2 , Mm 0.8 93 Ji for 2::; Ji ::; 3 .464, Mm 3.09 for Ji > 3.464 Similarly , Mm for an inside or an outside circumferential surface flaw is given b y: Mm 0.89 for Ji < 2, Mm 0.443 Ji for 2::; Ji ::; 3 .464 , Mm 1.53 for Ji > 3.464 Where: 6-2 (3) p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in). For bending stress, the corresponding K 1 for the postulated axial or circumferential defect is: K 1 b = Mb

  • Maximum Stress, where Mb is two-thirds of Mm (4) The maximum K 1 produced by radial thermal gradient for the postulated axial or circumferential inside surface defect ofG-2120 is: K11 = 0.953x10-3 x CR x t 2*5 (5) where CR i s the cooldown rate in °F/hr., or for a postulated axial or circumferential outside su r face defect K11 = 0.753x10-3 x HU x t 25 where HU is the heatup rate in °F /hr. WCAP-18169-NP (6) June2018 Revis i on 1 CALC-AN02-EP-17-00002 Rev O Page 41 of 88 Westinghouse Non-Propr ie tary Class 3 6-3 The throu g h-wall temperature difference associated with the maximum thermal K 1 can be determined fromASME Code,Section XI , Appendix G, Fig. G-2214-1.

The temperature at any radial distance from the vessel surface can be determined from ASME Code,Section XI, Appendix G , Fig. G-2214-2 for the maximum thermal K 1* (a) The maximum thermal K 1 relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2). (b) Alternatively, the K 1 for radial thermal gradient can be calculated for any thermal stress di s tribution and at any specified time during cooldown for a 11,i-thickness axial or circumferential in s ide surface defect using the relationship

!01 = (1.0359Co

+ 0.6322Ci + 0.4753C 2 + 0.3855C 3) * .fi;. (7) or similarly, Ku during heatup for a 11,i-thickness outside axial or circumferential surface defect using the relationship:

K1t = (l .043Co + 0.630C, + 0.481C2 + 0.401C 3) * .f;; (8) where the coefficients C 0 , C 1 , C 2 and C 3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form: o-(x) = Co+ C,(x I a)+ C2(x I a)2 + C3(x I a)3 (9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front , and a is the maximum crack depth (in). Note that Equations 3, 7 , and 8 were implemented in the OPERLIM computer code, which is the program used to ge n erate the pressure-t em perature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4 , "Method ology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 2] Section 2.6 ( equations 2.6.2-4 a nd 2.6.3-1 ). Finally, the reactor vessel metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A, Revision 4 ( equation 2.6.1-1 ). This equation is solved utilizing values for thermal diffusivity of 0.518 ft2/hr at 70°F and 0.379 ft 2/hr at 550°F and a constant convective heat-transfer coefficient val ue of 7000 Btu/hr-ft 2-°F. At any time during the heatup or cooldown transient, K1c is determined by the metal temperature at the tip of a postu l ated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the s ection thickness per ASME Code,Section XI, Paragraph G-2120), the appropriate value for RT N OT, and the reference fracture toughness curve (Equation 1 ). The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress WCAP-18169-NP June 2018 Revision I --------------,

CALC-AN02-EP-1 7-0 00 02 Rev O Page 42 of 88 Westinghouse Non-Proprietary Class 3 6-4 intensity factors, K 1 i, for the 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 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because th e thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside s urface that would tend to open (propagate) the existing flaw. Allowable pressure-temperature curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.

The use of the composite curve 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 material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter.

This condition, of course, is not true for the steady-state situation. It follows that , at any given reactor coolant temperature, the ~T (temperature) across the vessel wall developed during cooldown results in a higher value of Kie at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1c exceeds K1t, 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 1/4T location an d , therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals along a c ooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure.

The metal temperature at the crack tip lags the coolant temperature; therefore , the K 1 e for the inside 1/4T flaw during heatup is lower than the K1c for the flaw during state conditions at the same coolant temperature.

During heatup , especially at the end of the transient , conditions may exist so that the effects of compressive thermal stresses and lower K1c values do not offset each other, and the pressure-te mperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases hav e 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 third portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the ca s e in which a l/4T flaw located at the 1/4T location from the outside surface 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 WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 43 of 88 Westinghouse Non-Proprietary Class 3 6-5 temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing h eatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature curves for the steady-state and finite heatup rate situations, t he final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein , over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

6.3 CLOSURE

HEADNESSEL FLANGE REQUIREMENTS 10 CFR Part 50 , Appendix G [Ref. 4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RT NDT by at least l 20°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure , which is calculated to be 622 psig. The initial RT N OT va lues of the reactor vessel closure head , replacement reactor vessel closure head , and vessel flange are documented in Table 3-2. The limiting unirradiated RT NOT of 30°F is associated with the vessel flange of the Arkansas Nuclear One Unit 2 vessel , so the minimum allowa ble temperature of this region is l 50°F at pressures greater than 622 psig (without margins for instrument uncertainties).

This limit is shown in Figures 8-1 and 8-2. 6.4 LOWEST SERVICE TEMPERATURE REQUIREMENTS The lowest service temperature (LST) is the minimum allowable temperature at which pressure can exceed 20% of the pre-service hydrostatic test pressure (3110 psig). This temperature is defined by Paragraphs NB-3211 and NB-2332 of ASME Code Section lil [Ref. 9] as the most limiting RT NOT for the balance of the RCS components plus 100°F. The balance of the reactor coolant system components includes consideration of the ferritic materials outside the reactor vessel cylindrical shell beltline , nozzle comer (see Appendix B), closure head, and vessel flange regions, but within the primary system. Per Table 3-2, the most limitin g RT NDT for the balance of RCS is 50°F. Therefore , without margins for instrument errors, the LST for Arkansas Nuclear One Unit 2 is 150°F. For Arkansas Nuclear One Unit 2 , this limit is identical to the vessel flange limit described in Section 6.3 and is shown in Figures 8-1 and 8-2. 6.5 BOLTUP TEMPERATURE REQUIREMENTS The minimum boltup temperature is the minimum allowable temperature at which the reactor vessel closure head bolts can be preloaded.

It is determined by the highest reference temperature , RT NOT, in the closure flange region. This requirement is esta blished in Appendix G to 10 CFR 50 [Ref. 4]. Per the NRC-approved methodology in WCAP-14040-A, Revision 4 [Ref. 2], the minimum boltup temperature should be 60°F or the limiting unirradiated RT NDT of the closure flange region , whichever is higher. Since the limiting unirradiated RT NDT of this region is below 60°F per Table 3-2 , the minimum boltup temperature for the Arkansas Nuclear One Unit 2 reactor vessel is 60°F (without margins for instrument uncertainties).

This limit is shown in Figures 8-1 and 8-2. WCAP-18169-NP June2018 Re v ision I CALC-AN02-EP-17-00002 Rev O Page 44 of 88 Westinghouse Non-Proprietary Class 3 7-l 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART= Initial RT N DT + Lill.T NDT + Margin (10) Initial RT NDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [Ref. 9]. If measured values of the initial RT N DT for the material in question are not available , generic mean values for that class of material may be used , provided if there are sufficient test results to establish a mean and standard de viati on for the class. Lill.T N DT i s the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows: ~RT NDT =CF* f (0.28-0.IO!ogf) (11) To calculate Lill.T N DT at any depth ( e.g., at 1/4T or 3/4T), the following formula must first be used to atten uate the fluence at the specific depth. £ = L s u-'ace

  • e (-O.Z 4 x) (dept h x) "* (12) where x inches (reactor vessel cylindrical shell beltline thickness is 7.875 inches) is the depth into the vesse l wall measured from the vessel clad/base metal interface.

The resultant fluence is then placed in Eq uation 1 1 to calculate the Lill.T NDT at the specific depth. The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-14040-A, Revision 4, "Me thodolo gy Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 2]. Table 7-1 c ontains the surface fluence values at 54 EFPY, which were used for the development of the T limit curves contained in this report. Table 7-1 also contains the 1/4T and 3/4T calculated fluence values and fluence factors (FFs), per Regulatory Guide 1.99, Revision 2 [Ref. l]. The values in thi s table will be us e d to calculate the 54 EFPY ART va lues for the Arkansas Nuclear One Unit 2 reactor vessel materials. Margin is c alculated as M = 2 cr; + cr ! . The standard deviation for the initial RT NDT margin term ( cri) is 0°F when t he initial RT NDT is a measured value, and l 7°F when a generic value is available.

The standard deviation for the ~RT NDT margin term , a ,.., is l 7°F for plates or forgings when surveillance data is not used or is non-credible , and 8.5°F (half the va lue) for plates or forgings when credible surveillance data is used. For w elds, a,.. is equal to 28°F when surveillance capsule data is not used or is non-credible , and is 14°F (half the value) when credible surveillance capsule data is used. The value for cr ,.. need not exceed 0.5 times the mean value of ~RT NDT* WCAP-181 6 9-NP June2018 Revision l CALC-AN02-EP-17-00002 Rev O Page 45 of 88 Westinghouse Non-Proprietary Class 3 7-2 Contained in Tables 7-2 and 7-3 are the 54 EFPY ART calcu lations at the 1/4T and 3/4T locations for generation of the Arkansas Nuclear One Unit 2 heatup and cooldown curves. The inlet and outlet nozzle forging materials for Arkansas Nuclear One Unit 2 have projected fluence val ue s that do not exceed the 1 x 10 1 7 n/cm 2 fluence threshold at 54 EFPY per Table 2-2 at the lowest extent of the nozzle; therefore , per NRC RIS 2014-11 [Ref. 8], neutron radiation embrittlement need not be consid e red herein for these materials.

Thus, ART calculations for the inlet and outlet nozzle forging materials u tilizing the 1/4T and 3/4T fluence values are excluded from Tables 7-2 and 7-3 , respectively. Limiting ART values for the nozzle materials are contained in Appendix B. The limiti n g ART va lue s for Arkansas Nuclear One Unit 2 to be used in the generation of the P-T limit curves are based on Lower Shell Plate C-8010-1 (Position 1.1). In order to provide an additional margin of conservatism , the limit ing calcu lated ART va lue s were rounded up and increased by 5°F. The increased limiting ART values, using the "Axial Flaw" methodology, for Lower Shell Plate C-8010-1 are s ummarized in Table 7-4. WCAP-18169-NP June 2018 Revision l CALC-AN02-EP-17-00002 Rev O Page 46 of 88 Westinghouse Non-Proprietary Class 3 7-3 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Arkansas Nuclear One Unit 2 Reactor Vessel Materials at 54 EFPY Surface Fluence, 1/4T f Reactor Vessel Region f <*> (n/cm2, (n/cm2, 1/4T E > 1.0 MeV) E > 1.0 MeV) FF Reactor Vessel Beltline Materials Intermediate Shell Plates 4.91 X 10 1 9 3.06 X 10 19 1.2955 Lower Shell Plates 4.98 X 10 1 9 3.10 X 10 1 9 1.2988 Intermediate Shell Longitudinal 4.64 X 10 1 9 2.89 X 10 19 1.2820 Welds Lower Shell Longitudinal 4.71 X 10 1 9 2.94 X 10 1 9 l.2856 Welds Intermediate to Lower Shell 4.89 X 10 1 9 Girth Weld 3.05 X 10 19 1.2945 Reactor Vessel Extended Beltline Materials Upper Shell Plates 5.89 X 10 1 7 3.67 X 10 1 7 0.2467 Upper Shell Longitudinal Welds 5.89 X 10 17 3.67 X 10 17 0.2467 Upper to Intermediate Shell 5.89 X 10 1 7 3.67 X 10 1 7 0.2467 Girth Weld Note: (a) 5 4 EFPY fluence value s are documented in Table 2-2. WCAP-18169-NP 3/4T f (n/cm2, E > 1.0 MeV) 1.19 X 10 19 1.21 X 10 19 1.12 X 10 1 9 1.14 X 10 1 9 1.18 X 10 1 9 1.43 X 10 1 7 1.43 X 10 1 7 1.43 X 10 1 7 3/4T FF 1.0485 1.0524 1.0327 1.0369 l.0474 0.1389 0.1389 0.1389 June 2018 Re v ision 1 Westinghouse Non-Proprietary Class 3 7-4 Table 7-2 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location Reactor Vessel Material and ID Heat CF l/4T Fluence l/4T RT NDT(U) (a) Number Number {°F) (n/cm2, E > 1.0 MeV) FF (OF) Reactor Vessel Beltline Materials Intermediate Shell Plate C-8009-1 C8161-3 63.6 3.06 X 10 1 9 1.2955 -1.4 Intermediate Shell Plate C-8009-2 C8161-l 54.5 3.06 X 10 1 9 1.2955 0.5 Intermediate Shell Plate C-8009-3 C8182-2 62.2 3.06 X 10 1 9 1.2955 0.0 U s ing c r e dible Arkan s as Nucl e ar C8182-2 55.8 3.06 X 10 1 9 1.2955 0.0 On e Unit 2 surv e illan ce data Lower Shell Plate C-8010-1 C8161-2 54.5 3.10 X 10 1 9 1.2988 12.0 Lower Shell Plate C-8010-2 B2545-l 53. l 3.10 X 10 1 9 1.2988 -16.7 Lower Shell Plate C-80 l 0-3 B2545-2 51.0 3.10 X 10 1 9 1.2988 -22.6 Intermediate Shell Longitudinal Welds 2-203A, B , & C Multiple 68.0 2.89 X 10 1 9 1.2820 -56 Lower Shell Longitudinal Welds 3-203A, B, & C 10120 34.0 2.94 X 10 1 9 1.2856 -56 Intermediate to Lower Shell Girth Weld 9-203 83650 34.1 3.05 X 10 1 9 1.2945 -40 Using cr e dible Arkansas Nuclear 83650 12.1 3.05 X 10 1 9 1.2945 -40 One Unit 2 surveillance data Reactor Vessel Extended Beltline Materials Upper Shell Plate C-8008-1 C8182-1 91.0 0.0367 X 10 1 9 0.2467 12.2 Using cr e dibl e Arkansas Nuclear On e Unit 2 surveillan ce data C8182-l 55.8 0.0367 X 10 1 9 0.2467 12.2 Upper Shell Plate C-8008-2 C7605-l 89.5 0.0367 X 10 1 9 0.2467 60.5 Upper Shell Plate C-8008-3 C8571-2 51.0 0.0367 X 10 19 0.2467 27.3 Upper Shell Longitudinal Welds BOLA 27.0 0.0367 X ] 0 1 9 0.2467 -60 1-203A, B , & C Using non-credibl e J.M Farley Unit 2 surveillan c e data BOLA 16.2 0.0367 X 10 1 9 0.2467 -60 WCAP-18169-NP ARTNDT a/*> (J A (b) (OF) (OF) (OF) 82.4 0.0 17.0 70.6 0.0 17.0 80.6 0.0 17.0 72.3 0.0 8.5 70.8 0.0 17.0 69.0 0.0 17.0 66.2 0.0 17.0 87.2 17.0 28.0 43.7 17.0 21.9 44.1 0.0 22.1 15.7 0.0 7.8 22.5 0.0 11.2 13.8 0.0 6.9 22.1 0.0 11.0 12.6 0.0 6.3 6.7 0.0 3.3 4.0 0.0 2.0 Margin ART<c> (OF) 34.0 34.0 34.0 17.0 34.0 34.0 34.0 65.5 55.4 44.1 15.7 22.5 13.8 22.1 12.6 6.7 4.0 (OF) 115.0 l 05.1 114.6 89.3 116.8 86.3 77.6 96.7 43.1 48.3 -8.7 57.1 39.7 104.7 52.5 -46.7 -52.0 June 2018 Revision 1 C') )> r C') )> z 0 "' ri, ,, I ..... "'-I 6 0 0 0 "' :::c (!) < 0 ,, Cl (C (!) "'-I 0 .... co co Westinghouse Non-Proprietary Class 3 7-5 Table 7-2 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the l/4T Location Reactor Vessel Material and ID Heat CF I/4T Fluence 1/4T RTNDT(U) (a) ARTNDT e1/*l G ,1(b) Margin ART<c> Number Number (OF) (n/cm2, E > 1.0 MeV) FF (OF) {°F) {°F) (OF) {°F) (OF) 10137 98.5 0.0367 x 10 19 0.2467 -56 24.3 17.0 12.2 41.8 I 0.1 Using credible Upper to Intermediate Shell Girth Calvert Cliffe 61.2 0.0367 X 10 19 0.2467 -5 6 I 5. I I 7.0 28.0 65.5 24.6 Unit 2 and Weld 8-203 Millstone Unit 2 data 6329637 100.8 0.0367 X 10 1 9 0.2467 -56 24.9 17.0 12.4 42.1 I 1.0 FAGA 41.0 0.0367 X 10 19 0.2467 -24 IO.I 0.0 5.1 10.l -3.8 Notes: (a) The plate material initial RT NDT values are measured values. For weld materials with generic initial RT N OT values, cr 1 = l 7°F. For weld materials with measured initial RT NOT values , OJ = 0°F. (b) As discussed in Section 4 , the surveillance plate and weld Heat # 83650 data were deemed credible, while the weld Heat # BOLA data we r e deemed non-credible.

The surveillance weld data for Heat# IO 137 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. I], the base metal cr 6 = l 7°F for Position 1.1 and cr 6 = 8.5°F for Position 2.1 with credible surveillance data. The weld metal cr 6 = 28°F for Position 1.1 and 2.1 with non-credible surveillance data (Heat# BOLA), and the weld metal cr 6 = 14°F for Position 2.1 with credible surveillance data (Heat# 83650). Since a full margin term will be used for Heat# IO 137 , cr 6 = 28°F with credible surveillance data for Position 2.1 for this weld heat. However, cr 6 need not exceed 0.5*ll.RT NOT* (c) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART= RT NDT (U) + Li.RT NOT+ Margin. WCAP-1 8169-NP June 2018 Revision I n )> r n )> z 0 "' ,;, " I ...... -...I I 0 0 0 0 "' ;:o (I) < 0 " II) cc (I) "" co 0 -co co Westinghouse Non-Proprietary Class 3 7-6 Table 7-3 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location Reactor Vessel Material and ID Heat CF 3/4T Fluence RTNDT(U) 3/4T FF Number Number (OF) (n/cm2, E > 1.0 MeV) (OF) Reactor Vessel Beltline Materials Intermedi a te Shell Plate C-8009-1 C8161-3 63.6 1.19 X 10 1 9 1.04 8 5 -1.4 Intermediate Shell Plate C-8009-2 C8161-1 54.5 l.19x 10 1 9 1.0485 0.5 Intermedi a te Shell Pl a te C-8009-3 C8182-2 62.2 1.19 X 10 1 9 1.0485 0.0 Usi n g cred ibl e Arkansas Nuclea r C8182-2 55.8 1.19 X 10 1 9 1.0485 0.0 On e U nit 2 s urv e illan ce data Lower Shell Plate C-80 I 0-1 C8161-2 54.5 1.21 X 10 1 9 1.0524 12.0 Lower Shell Plate C-8010-2 B2545-l 53.1 1.21 X 10 1 9 1.0524 -16.7 Lower Shell Pl a te C-80 l 0-3 B 2 545-2 51.0 1.21 X 10 1 9 1.0524 -22.6 Intermediate Shell Longitudinal Multiple 68.0 1.12 X 10 19 1.0327 -5 6 Weld s 2-203A , B , & C Lower Shell Longitudinal Welds 3-203A , B, & C 10120 34.0 1.14 X 10 1 9 1.0369 -56 Intermedi a te to Lower Shell Girth 83650 34.1 l.18xl0 1 9 1.0474 -40 Weld 9-203 Usi n g credi bl e Arkansas N ucl e ar 83650 12.1 1.(8 X 10 19 1.0474 -40 On e Un it 2 surveillance data Reactor Vessel Extended Beltline Materials Upper Shell Plate C-8008-1 C8 1 8 2-1 91.0 0.0143 X 10 1 9 0.1389 12.2 Using credi bl e Arkansas N ucl ea r On e Un it 2 s urv e illan ce data C8182-1 55.8 0.0143 X 10 1 9 0.1389 12.2 Upper Shell Plate C-8008-2 C7605-1 89.5 0.0143 X 10 1 9 0.1389 60.5 Upper Shell Plate C-8008-3 C8571-2 51.0 0.0143 X I 0 1 9 0.1389 27.3 Upper Shell Longitudinal Welds 1-203A, B, & C BOLA 27.0 0.0143 X 10 1 9 0.1389 -60 Usi n g non-credible J.M Farley BOLA 16.2 0.0143 X 10 1 9 0.1389 -60 Unit 2 surveillance data WCAP-1 8 169-NP (*) iiRTNDT o}*> G a (b) (OF) (OF) (OF) 66.7 0.0 17.0 57.1 0.0 17.0 65.2 0.0 17.0 58.5 0.0 8.5 57.4 0.0 17.0 55.9 0.0 17.0 53.7 0.0 17.0 70.2 17.0 28.0 35.3 17.0 17.6 35.7 0.0 17.9 12.7 0.0 6.3 12.6 0.0 6.3 7.8 0.0 3.9 12.4 0.0 6.2 7.1 0.0 3.5 3.8 0.0 1.9 2.3 0.0 1.1 Margin (OF) 34.0 34.0 34.0 17.0 34.0 34.0 34.0 65.5 49.0 35.7 12.7 12.6 7.8 12.4 7.1 3.8 2.3 ART<c> (OF) 99.3 91.6 99.2 75.5 103.4 73.2 65.1 79.7 28.2 31.4 -14.7 37.5 27.7 85.4 41.5 -52.5 -55.5 June 2018 Revision 1 0 )> r-0 )> z 0 "' I m '"ti I ..... -..I 6 0 0 0 "' ;:c (!) < 0 '"ti (C (!) <D 0 -00 00 Westinghouse Non-Proprietary Class 3 7-7 Table 7-3 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location Reactor Vessel Material and ID Heat CF 3/4T Fluence RTNDT(U) (a) ARTNDT (J?) <J ,1 (b) Margin ART<c> 3/4T FF Number Number (OF) (n/cm2, E > 1.0 MeV) (OF) (OF) (OF) (OF) (OF) {°F) 10137 98.5 0.0143 X 10 19 0.1389 -56 13.7 17.0 6.8 36.6 -5.7 Using credible Upper to Intermediate Shell Girth Calvert Cliffs 61.2 0.0143 X 10 19 0.1389 -56 8.5 17.0 28.0 65.5 18.0 Unit 2 and Weld 8-203 Millstone Unit 2 data 6329637 100.8 0.0143 X 10 1 9 0.1389 -56 14.0 17.0 7.0 36.8 -5.2 FAGA 41.0 0.0143 X 10 1 9 0.1389 -24 5.7 0.0 2.8 5.7 -12.6 Notes: (a) The plate material initial RT NOT values are measured values. For weld materials with generic initial RT NOT values, cr 1 = l 7°F. For weld materials with measured initial RT NDT values , cr 1 = 0°F. (b) As discussed in Section 4, the surveillance plate and we ld Heat # 83650 data were deemed credible, w hile the weld Heat # BOLA data were deemed non-credible. The surveillance weld data for Heat# 10137 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. I], the base metal cr 6 = l 7°F for Position 1.1 and cr 6 = 8.5°F for Position 2.1 with credib le surveillance data. The weld metal cr 6 = 28°F for Position 1.1 and 2.1 with non-credible surveillance data (Heat# BOLA), and the weld metal cr 6 = 14°F for Position 2.1 with credible surveillance data (Heat# 83650). Since a full margin term will be used for Heat# IO 137, cr 6 = 28°F with credible surveillance data for Position 2.1 for this weld heat. However, cr 6 need not exceed 0.5*~RT NOT* (c) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART= RTNoT(U) + ~RTNoT + Margin. WCAP-18169-NP June 2018 Revision 1 0 )> r-0 > z 0 "' rn "ti I .... ...... 6 0 0 0 "' ::tJ CD < 0 "ti II) (C CD 0, 0 0 .... 00 00 Table 7-4 Notes: CALC-AN02-EP-17

-00002 Rev O Page 51 of 88 Westinghouse Non-Proprietary Class 3 Summary of the Increased Limiting ART Values Used in the Generation of the Arkansas Nuclear One Unit 2 Heatup and Cooldown Curves at 54 EFPY l/4T Limiting ART<*> 3/4T Limiting ART<*> 122°F 109°F Lower Shell Plate C-8010-1 (Position 1.1) 7-8 (a) The ART values used for P-T limit curve development in this report are the limiting ART values calcu l ated in Tables 7-2 and 7-3 rounded up and increased by 5°F to add additional margin; this approach is conservative. WCAP-181 6 9-NP June2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 52 of 88 Westinghouse Non-Proprietary Class 3 8-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel cylindrical beltline region using the methods discussed in Sections 6 and 7 of this report. This approved methodology is also presented in WCAP-14040-A , Revision 4 [Ref. 2]. Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 50, 60, 70 , and 80°F/hr applicable for 54 EFPY, with the flange and lowest service temperature requirements and using the "Axial Flaw" methodology. Figure 8-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 25 , 60 , and 100°F/hr applicable for 54 EFPY, with the flange and lowest service temperature requirements and using the " Axial Flaw" methodology.

The heatup and cooldown curves were generated using the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 8-1 and 8-2. This is in addition to other criteria, which must be met before the reactor is made critical , as discussed in the following paragraphs. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality l imit line shown in Figure 8-1 (heatup curve only). The first straight-line portion of the c riticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50 (see Figure 8-3 and Table 8-3). The governing equation for the hydrostatic test is defined in the 1998 through the 2000 Addenda ASME Code Section XI , Appendix G as follows: where, K 1 m is the stress intensity factor covered by membrane (pressure) stress [see page 6-2, Equation (3)], K, c = 33.2 + 20.734 e [O.OZ(T-R T N oT l l [s ee page 6-1 , Equation (l)], T is the minimum permissible metal temperature, and RT N D T is the metal reference nil-ductility temperature.

The criticality limit curve s pecifies pressure-temperature limits for core operation in order to provide additional margin during actual power production.

The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test , and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding temperature curve for heatup and cooldown calculated a s described in Section 6 of this report. For the heatup and cooldown curves without margins for instrumentation errors , the minimum temperature for the inservice hydrostatic test at 2485 psig for the Arkansas Nuclear One Unit 2 reactor vessel at 54 EFPY is WCAP-181 6 9-NP June20I8 Re v ision I CALC-AN02-EP-17-00002 Rev O Page 53 of 88 Westin g house Non-Proprietary Class 3 8-2 l 79°F. The 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 ve s sel. Figures 8-1, 8-2, and 8-3 define all of the above limits for ensuring prevention of non-ductile failure for the Arkansas Nuclear One Unit 2 reactor vessel for 54 EFPY with the flange and lowest service temperature requirements and without instrumentation uncertainties. The data points used for developing the heatup and cooldown P-T limit curves shown in Figures 8-1 and 8-2 are presented in Tables 8-1 and 8-2. The data points used for developing the inservice hydrostatic and leak test P-T limit curve s hown in Figure 8-3 are presented in Table 8-3. The P-T limit curves shown in Figures 8-1 , 8-2 , and 8-3 were generated based on the limiting ART values for the cylindrical beltline and extended beltline reactor vessel ma t erials rounded up and increased by 5°F to add additional margin; this approach is conservative. As discus s ed in Appendix B , the P-T limits developed for the cylindrical beltline region bound the P-T limits for t he reactor vessel inlet and outlet nozzles for Arkansas Nuclear One Unit 2 at 54 EFPY. WCAP-18169-NP June 2018 Re v ision 1 CALC-AN02-EP-17-00002 Rev O Page 54 of 88 Westinghouse Non-Proprietary Class 3 MATERIAL PROPERTY BASIS 8-3 LIMITING MATERIAL:

Lower Shell Plate C-8010-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 54 EFPY: l/4T, 122°F (Axial Flaw) 3/4T , 109°F (Axial Flaw) 2500 2250 2000 1750 -(!) en 1500 a. -Q) I,. :::, t/j t/j Q) 1250 I,. ll. -0 Q) ... 1000 :::, cu (.) 750 500 250 0 ' loper l imAna l ysis Versi o n:5.4 R u n:24301 Oper l im.x l sm Version: 5.4.1 I I I ' ,_ ~I Unacceptable I Ooeration I I ' IHeatup Rater 50°F/Hr I L.---"' f "' IHeatup Rater !Crit i cal Limit! 60°F/Hr u 50°F/Hr I I I I !Cr i tica l Limitj I IHeatup Rate j' "' 70°F/Hr I 60°F/Hr ~IHeatup RateV,...

_!Critical Limitl i'... 70°F/Hr 80°F/Hr ,---I I !Critical Limitl 80°F/Hr 1 1 Lowest Service Temp.= 150°F I Acceptab l e I Ope r ation .. Minimum +-Boltup Cr i ticality Lim i t based on Temo. = 60°F I+--inservice hydrostatic test temperature (179°F) for the service per i od up to 54 EFPY I I I I I I 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 8-1 Arkansas Nuclear One Unit 2 Reactor Coolant System Heatup Limitations (Heatup WCAP-18169-NP Rates of 50, 60, 70, and 80°F/hr) Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c) June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 55 of 88 Westinghouse Non-Proprietary Class 3 8-4 MATERJAL PROPERTY BASIS LIMITING MATERJAL:

Lower Shell Plate C-8010-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 54 EFPY: l/4T, 122°F (Axial Flaw) 3/4T, 109°F (Axial Flaw) -(!) en c... -Q) ... :, Ill Ill Q) ... c... "C Q) .... :, (.) cu (.) 2500 2250 2000 1750 1500 1250 1000 750 500 250 0 ' loperlimAnalysis Version: 5.4 R u n:24301 Operlim.x l sm Versio n: 5.4.1 I , f-1 Unacceptable I Operation I 0 J I ~-Cooldo w n " ~Rates I Lowest Service steady-state I Acceptable I Temp.= 150" F -25°F/Hr Operation

-60°F/H r I r. ~I Cooldown Ratel 1 1-100°F/Hr I ' I I I Minimum ! I I +--Boltup ! I Temp.= 60'F I ' ! 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 8-2 Arkansas Nuclear One Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of O, 25 , 60, and 100°F/hr)

Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c) WCAP-181 69-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 56 of 88 We s tin g hou s e Non-P ro priet a ry C la ss 3 8-5 MA TERIA L PROPERTY B A SIS LI MITING MATE RIAL: Lower Sh e ll Pl a te C-8010-1 u si ng Regulatory Guide 1.9 9 Position 1.1 d a ta LIMITING AR T VALUES A T 54 EFPY: 1/4T , 122°F (Axial Fla w) 3/4T , 109°F (Axial F law) 250 0 225 0 200 0 175 0 -C) en 150 0 0... -Q) .... ::, II) II) Q) 125 0 .... 0... i:, Q) -ns ::, u 100 0 cu 0 75 0 50 0 25 0 0 ' loperli mAn a l ysis Ve r sio n: 5.4 R un: 24301 Operli m.x l sm Version: 5.4.1 I ' lln service Hyd r ostatic! ._j Unacceptable I . Ooeration I I and Leak Test Limit 0 I Lowest Service Tem:J. = 150'F I Acceptab l e I Operation I Minimum I+-Boltup Temo. = 60'F I ! I l I I l I I 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 8-3 Arkansas N uclear One Unit 2 Reactor Coolant S y stem Inservice Hydrostatic and Leak Test Limitations Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 A dd e nda App. G M e thodolog y (w/ K1c) WCA P-1 8 1 69-NP J un e2 01 8 R evisio n I Ta bl e 8-1 5 0°F/h r Heat up T (OF) p (p sig) 60 0 60 622 65 622 70 622 75 622 80 622 85 622 90 622 95 622 JOO 622 105 622 110 622 115 622 120 622 125 622 1 30 622 135 622 140 622 145 622 150 622 150 1176 1 55 1 238 WCAP-18169-NP Westinghouse Non-Proprietary C l ass 3 8-6 A rkan sas N ucl ea r On e Un it 2 54 EFPY He atup C u rve D at a Po in ts u s i ng th e 199 8 t h ro u g h th e 2000 A dd e nd a A pp. G Me thod o log y (w/ Kic, w/ F l a ng e a nd L ST Re qui re m e nt s, and w/o M ar g ins for Ins trum e ntation E rrors) 50°F/hr 6 0°F/hr C riti ca li ty 60°F/h r He atup Cr iti ca li ty T (°F) p T (°F) p T (OF) p (ps i g) (p s i g) (p s i g) 179 0 60 0 179 0 179 622 60 622 179 622 180 622 65 622 180 622 185 622 70 622 1 85 622 1 90 622 75 622 190 622 190 1176 80 622 190 1114 195 1238 85 622 195 11 70 200 1308 90 622 200 1234 205 1 384 95 622 205 1303 210 1468 100 622 2 1 0 1380 2 1 5 1 56 1 105 622 215 1 466 220 1664 110 622 220 1 560 225 1 778 115 622 225 1664 230 1903 1 20 622 230 1 778 235 2042 125 622 235 1 905 240 2 1 95 1 3 0 622 240 2045 245 2363 135 622 245 2200 140 622 250 2370 1 4 5 622 1 50 622 150 1 1 1 4 155 1 170 70°F/hr 70°F/h r Hea tup Cr i tica li ty T (°F) p T (°F) p (p s i g) (psi g) 60 0 179 0 60 622 179 622 65 622 180 622 70 622 185 622 75 622 190 622 80 622 190 1057 85 622 195 1 109 90 622 200 1 166 95 622 205 12 3 0 100 622 2 1 0 1300 1 05 622 2 1 5 1378 110 622 22 0 1 464 115 622 225 1559 120 622 23 0 1664 125 622 23 5 1 780 130 622 24 0 19 0 8 135 622 245 2050 140 622 25 0 2206 145 622 2 55 2378 150 622 150 1057 155 1 1 09 80°F/hr H ea tup T (°F) p (p s i g) 60 0 60 622 65 622 70 622 75 622 80 622 85 622 90 622 95 622 100 622 1 05 622 11 0 622 115 622 1 20 622 1 25 622 13 0 622 1 35 622 1 40 622 1 45 622 1 50 622 1 50 1007 1 55 1054 80°F/hr Cr itic a li ty T (°F) 179 179 180 185 190 190 1 95 200 205 210 2 1 5 220 225 230 235 2 40 245 250 255 260 p (psi g) 0 622 622 622 622 1007 1 054 1106 11 6 4 1228 1 299 1 37 8 1 46 5 1 56 1 1 667 1 784 1 9 1 3 2056 22 1 4 2388 June 20 1 8 Revision l C') )> r C') )> z 0 N m "'C ' .... ..... b 0 0 0 N ::0 (I) < 0 "'C Ill (C (I) U1 ..... 0 -CIO CIO Table 8-1 50°F/hr Heatup T (°F) p (psig) 160 1308 165 1384 170 1468 175 1561 180 1664 185 1778 190 1903 195 2042 200 2195 205 2363 WCAP-18169-NP Westinghouse Non-Proprietary Class 3 8-7 Arkansas Nuclear One Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors) 50°F/hr 60°F/hr Criticality 60°F/hr Heatup Criticality 70°F/hr Heatup T (°F) p T (°F) p T (°F) p T (°F) p (psig) (psig) (psig) (psig) 160 1234 160 1166 165 1303 165 1230 170 13 80 170 1300 175 1466 175 1378 180 1560 180 1464 185 1664 185 1559 190 1778 190 1664 195 1905 195 1780 200 2045 200 1908 205 2200 205 2050 210 2370 210 2206 215 2378 70°F/hr Criticality 80°F/hr Heatup T (°F) p T (°F) p (psig) (psig) 160 1106 165 1164 170 1228 175 1299 180 1378 185 1465 190 1561 195 1667 200 1784 205 1913 210 2056 215 2214 220 2388 80°F/hr Criticality T (°F) p (psig) June 2018 Revision 1 0 )> r 0 )> z 0 N m "ti I ..... .....i I 0 0 0 0 N :;o l'D < 0 "ti Q) u:i l'D UI 00 0 -00 00 CALC-AN02-EP-17

-00002 Rev O Page 59 of 88 W e st i n g hou se Non-Pro pr i e t a ry Cl ass 3 Table 8-2 Arkansas Nuclear One Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors) St e ad y-State -25°F/hr. -60°F/hr.

-100°F/hr. T (OF) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 0 6 0 0 6 0 0 6 0 0 60 622 6 0 62 2 6 0 622 6 0 5 76 65 6 22 65 622 65 6 22 65 5 94 70 622 7 0 6 2 2 7 0 622 7 0 613 75 622 75 622 75 622 75 622 8 0 622 8 0 62 2 8 0 6 22 8 0 622 85 622 85 622 85 622 8 5 6 22 90 6 22 9 0 6 22 9 0 622 90 622 95 622 95 622 9 5 622 95 622 1 00 6 22 100 622 1 0 0 622 100 6 22 1 05 6 22 10 5 622 105 622 10 5 622 1 10 622 110 6 2 2 110 622 llO 6 22 115 622 11 5 622 115 622 11 5 622 1 20 6 22 1 2 0 622 1 20 62 2 1 2 0 622 125 622 1 25 622 1 25 622 1 2 5 622 130 6 22 1 3 0 622 13 0 62 2 1 3 0 62 2 135 622 1 35 622 135 622 1 35 622 1 40 622 140 6 22 1 4 0 622 140 622 145 622 14 5 622 145 622 145 622 150 622 1 5 0 622 1 50 622 1 5 0 622 150 1 321 1 5 0 1 32 1 150 1 32 1 1 5 0 1 3 21 1 55 1 393 1 55 1 393 1 55 1 393 15 5 13 93 160 14 74 1 6 0 14 7 4 160 147 4 1 60 14 7 4 1 65 1 562 1 65 1 5 62 165 1 562 1 6 5 15 6 2 1 70 1 660 1 7 0 1 66 0 1 70 1 66 0 1 7 0 1 66 0 1 75 1 768 17 5 1768 175 1 768 1 75 1 768 1 80 1 888 1 8 0 1888 1 80 1 888 1 8 0 1 888 1 85 2020 1 8 5 2 0 2 0 1 85 202 0 1 8 5 2 0 2 0 1 90 2 1 66 1 9 0 2 1 6 6 190 2 1 66 1 9 0 2 1 6 6 1 95 2 3 28 1 95 2 3 28 195 2328 1 95 2328 8-8 WCA P-1 8 1 69-NP Jun e 2 01 8 R evisio n 1 T a ble 8-3 WCAP-18 1 6 9-NP CALC-AN02-EP-17-00002 Rev O Page 60 of 88 Wes t inghouse No n-Proprie t ary C l ass 3 A rkansas Nucl e ar One Unit 2 54 EFPY lns e rvice H y dro s tatic and Le ak T e st Cur ve Data Point s u s ing the 1998 through the 2000 A ddenda A pp. G Methodolog y (w/ K 1c, w/ Flange and LST Requirements , a nd w/o M argins for Instrumentation E rrors) T (°F) P (psi g) 60 0 60 622 65 622 70 622 75 622 80 622 85 622 90 622 95 622 100 622 105 622 110 622 115 622 120 622 125 622 130 622 135 622 140 622 145 622 150 622 150 176 1 155 1858 160 1965 162 2000 165 2083 170 22 1 4 175 2358 179 2 485 8-9 J une 20 1 8 Revision l CA L C-A N02-EP-17

-0 0002 Rev O Page 61 of 88 Westinghouse Non-Proprietary Class 3 9 REFERENCES 9-1 1. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research , Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials ," May 1988. 2. Westin g house Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpr e ssure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004. 3. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure." 4. Code o f Federal Regulations, 10 CFR Part 50, Appendix G , "Frac ture Toughness Requirements," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19 , 1995. 5. Regula t ory Guide 1.190 , "Ca lculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, March 2001. 6. Westin g hou se Report WCAP-16083-NP, Revision 1, " Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Do s imetry," April 2013. 7. RSICC Data Library Collection DLC-1 8 5 , "BUGLE-96:

Coupled 47 Neutron, 20 Gamma-Ray Group Cross-S ection Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applic a tions ," March 1996. 8. NRC Regulatory Issue Summary 2014-11, " Information on Licensing Applicat i ons for Fracture Toughness Requirements for Ferri tic Reactor Coolant Pressure Boundary Components," U.S. Nuclear Regulatory Co mmission , October 2014. [Ag e nc ywide Do cu m ent Management System (ADAMS) Accessi o n Numbe r ML14149Al65}

9. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1 , Subsection NB, "C lass 1 Compo n ents." 10. NUREG-0800 , Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, C hapter 5 L WR Edition, Branch Technical Po sit ion 5-3, "Fracture Toughness Requirements

," Revision 2, U.S. Nuclear Regulatory Commission, Ma r ch 2007. 11. Combu s tion E ngineering Report A-PENG-ER-002 , Revision 0 , "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the ANO 2 Reactor Pressure Vessel Plates, Forgings, Welds a nd Cladding," October 1995. 12. Battelle -Columbus Report BMI-0584 , "F inal Report on Examination, Testing, and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Arkansas Nuclear One Unit 2 Generating Plant to Arkansas Power and Light Company ," May 1984. 13. Arkans a s Nuclear One -Unit 2 License Renewal Application, October 2003. [Ava ilable on th e N RC website} 14. AREV A NP, Inc. Report BAW-2399 , Revision 1 , "Ana lysis of Capsule W-104 Entergy Operations , Inc. Ar k ansas Nuclear One Unit 2 Power Plant Reactor Vessel Material Surveillance Program ," February 2005. 15. Code of Federal Regulations , 10 CFR 50.61 , " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register , Volume 60, No. 243, dated December 19 , 199 5, effective January 18 , 1996. 16. Combu s tion Engineering Owners Group Report CE NPSD-1039, Revision 2, "Best Estimate Copper and Nic k el Values in CE Fabricated Reactor Vessel Welds," June 1997. WCAP-18169-NP June2018 Revi s ion 1 r CALC-AN02-EP-17

-00002 Rev O Page 62 of 88 Westinghouse Non-Proprietary Class 3 9-2 17. Combustion Engineering Report TR-MCD-002, "Arkansas Power & Light Arkansas Nuclear One -Unit 2 Eva luation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program," March 1976. 18. Westin g house Report WCAP-17501-NP, Revision 0, "Ana lysis of Capsule 104° from the Calvert Cliffs Unit No. 2 Reactor Vessel Radiation Surveillance Program," February 2012. 19. Westin g house Report WCAP-16012, Revision 0, "Analysis of Capsule W-83 from the Dominion Nucle ar Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003. 20. Westin g house Report WCAP-16918-NP, Revision 1, "Analysis of Capsule V from the Southern Nuclea r Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program," April 2008. 21. Westin g house Report WCAP-18166-NP, Revision 0 , "Ana ly sis of Capsule 284° from the Entergy Operations, Inc. Arkansas Nuclear One Unit 2 Reactor Vessel Radiation Surveillance Program," September 2016. 22. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Hando ut s, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. [ADAMS Accession Number MLJ 10070570]

23. RSICC Computer Code Collection CCC-650, "DOORS 3.2: One, Two-and Three Dimensional Discret e Ordinates Neutron/Photon Transport Code System," April 1998. WCAP-1 8169-NP June2018 Revision I I I l_ CALC-AN02-EP-17-00002 Rev O Page 63 of 88 Westinghouse Non-Proprietary Class 3 A-1 A P PENDIX A THERMAL STRESS INTENSITY FACTORS (Ku) Tables A-1 and A-2 contain the thermal stress intensity factors (K 11) for the maximum heatup and cooldown rates at 54 EFPY for Arkansas Nuclear One Unit 2. The reactor vessel cylindrical shell radii to the l/4T and 3/4T locations are as follows:
  • l/4T Radius= 81.688 inches
  • 3/4T Radius= 85.625 inches W C AP-1816 9-NP June 2018 Revision 1 TableA-1 Water Temp. (OF) 6 0 65 70 75 8 0 85 9 0 95 100 10 5 11 0 11 5 1 2 0 1 25 13 0 1 35 1 40 145 1 5 0 1 55 1 6 0 1 65 1 7 0 1 75 18 0 1 85 1 90 1 95 2 00 2 0 5 2 10 CALC-AN02-EP-17-00002 Rev O Page 64 of 88 W estingho u se No n-P roprieta ry C l ass 3 A-2 Ku Values for Arkansas Nuclear One Unit 2 at 54 EFPY 80°F/hr Heatup Curves (w/ Flang e and LST Requirements, and w/o Margins for Instrument Errors) V e ssel Temp e rature 1/4T Therma l Stress Vessel Temperature 3/4T Thermal Stres s at 1/4T Location for Int e nsity Factor at 3/4T Location for Intensity Factor 80°F/hr H e a t up (°F) (ksi -V in.) 80°F/hr Heatup (°F) (ksi -V in.) 56.321 -1.028 55.105 0.545 59.387 -2.397 55.641 1.5 4 0 62.751 -3.44 9 56.862 2.407 66.385 -4.400 58.707 3.14 9 7 0.275 -5.1 55 61.052 3.764 74.297 -5.8 1 6 63.832 4.283 7 8.513 -6.35 1 66.967 4.7 1 6 82.818 -6.82 0 70.400 5.0 8 3 87.260 -7.203 74.080 5.3 90 91.763 -7.540 77.960 5.654 96.358 -7.818 82.009 5.878 10 0.997 -8.0 66 86.196 6.070 1 05.699 -8.271 90.500 6.236 11 0.432 -8.456 94.900 6.3 79 11 5.207 -8.6 11 99.381 6.5 0 3 1 2 0.006 -8.753 1 03.929 6.6 1 3 1 24.832 -8.872 1 08.533 6.7 0 9 1 29.675 -8.984 113.183 6.794 134.537 -9.080 1 17.872 6.8 7 1 1 39.412 -9.170 122.594 6.9 4 0 1 44.298 -9.249 1 27.343 7.00 2 1 49.195 -9.324 1 32.1 1 4 7.060 1 54.099 -9.39 1 1 36.904 7.11 3 159.011 -9.457 1 41.710 7.1 62 163.927 -9.5 1 6 1 46.529 7.2 0 9 168.849 -9.574 1 51.358 7.253 1 73.774 -9.627 1 56.198 7.294 1 78.703 -9.68 1 1 61.044 7.3 34 1 83.634 -9.730 1 65.897 7.373 1 88.568 -9.78 0 1 70.756 7.410 1 93.503 -9.826 1 75.6 1 9 7.44 6 WCAP-1 8 1 69-NP Ju ne2 01 8 R evis ion 1 TableA-2 CALC-AN02-EP-17-00002 Rev O Page 65 of 88 Westinghouse Non-Proprietary Class 3 A-3 Ktt Values for Arkansas Nuclear One Unit 2 at 54 EFPY 100°F/hr Cooldown Curves (w/ Flange and LST Requirements, and w/o Margins for Instrument Errors) Water Vessel Temperature at l/4T 100°F/hr Cooldown Temp. Location for 100°F/hr l/4T Thermal Stress (OF) Cooldown (°F) Intensity Factor (ksi --./in.) 210 232.413 13.504 205 227.339 13.448 200 222.265 13.392 195 217.192 13.336 190 212.118 13.281 185 207.044 13.225 180 201.970 13.169 175 196.897 13.113 170 191.823 13.058 165 186.749 13.002 160 181.676 12.947 155 176.602 12.891 150 171.529 12.836 145 166.456 12.781 140 161.3 8 3 12.726 135 156.310 12.671 130 151.237 12.616 125 146.164 12.561 120 141.091 12.507 115 136.018 12.452 110 130.946 12.398 105 125.874 12.343 100 120.801 12.289 95 115.729 12.235 90 110.657 12.182 85 105.585 12.128 80 100.514 12.074 75 95.442 12.020 70 90.371 11.967 65 85.299 11.914 60 80.229 11.860 WCAP-18169-NP June 2018 Revi s ion 1 CALC-AN02-EP-17

-0 0 002 Rev O Page 66 of 88 Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B REACTOR VESSEL INLET AND OUTLET NOZZLES As describ ed in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. B-1], reactor vessel non-beltline materials may define pressure-temperature (P-T) limit curves that are more limiting than those calculated for the reactor vesse l cylindrical s hell beltline materials. Reactor vessel nozzles , penetrations , and other discontinuities have complex geometries that can exhibit significantly higher stresses than those for the reactor vessel beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even if the reference temperatures (RT NOT) for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries.

The methodology contained in WCAP-14040-A, Revision 4 [Ref. B-2] was used in the main body of this repo rt to d evelop P-T limit curves for the limiting Arkansas Nuclear One Unit 2 cylindrical shell beltline material; however , WCAP-14040

-A, Re vision 4 does not consider ferritic materials in the area adjacent to the beltlin e, specifically the stressed inlet and outlet nozzles. Due to the geometric discontinuity , the inside corn er regions of these nozzles are the most highly stressed ferritic component outside the beltline region of the reactor vessel; therefore, these components are analyzed in this Appendix.

P-T limit curves are determined for the reactor vessel nozzle comer region for Arkansas Nuclear One Unit 2 and compared to the P-T limit curves for the reactor vessel traditional beltline region in order to determine if the nozzles can be more limiting than the reactor vessel beltline as the plant ages and the vessel accumulates more neutron flu e nce. The increase in neutron fluence as the plant ages causes a concern for embrittlement of the reactor vessel above the beltline region. Therefore, the P-T limit curves are developed for the nozzle inside comer region since the geometric d isco ntinuity results in high stresses due to internal pressure and the cooldown transient.

The cooldown transient is analyzed as it results in tensile stresses at the inside surface of t h e nozzle comer. A 1/4 T axial flaw is postulated at the inside s urface of the reactor vessel nozzle comer and stress intensity factors are determined based on the rounded curvature of the nozzle geometry.

The allowable pressure is then calculated based on the fracture toughness of the nozzle material and the stress intensity factors for the l/4T flaw. B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The fracture toughness (Kie) used for the inlet and outlet nozzle material is defined in Appendix G of the Section XI ASME Code, as discussed in Section 6 of this report. The K,c fracture toughness curve is dependent on the Adjusted Reference Temperature (ART) value for irradiated materials.

The ART values for the inle t and outlet nozzle materials are determined using the methodology contained in Regulatory Guide 1.99, Revision 2 [Ref. B-3], which is described in Section 7 of this report, and weight percent (wt. %) copper (Cu) and nickel (Ni), initial RT NDT value, and projected neutron fluence as inputs. The material p ro perties for each of the reactor vessel inlet and outlet nozzle forging materials are documented in Table B-1 and a summary of the limiting inlet and outlet nozzle ART values for Arkansas Nucle ar One Unit 2 is pr ese nted in Table B-2. Nozzle Material Properties The Arkansas Nuclear One Unit 2 nozzle material properties are provided in Tabl e B-1. Nickel (Ni), Manganese (Mn), and Phosphorus (P) weight percent (wt. %) values were obtained as the average of the material-specific analyses documented in Combustion Engineering reportA-PENG-ER-002

[Ref. B-4] for WCAP-18169-NP June2018 Re v i si on l CALC-AN02-EP-17

-00002 Rev O Page 67 of 88 Westinghouse Non-Proprietary Class 3 B-2 each of the Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzles. Copper weight percent values for each of the Arkansas Nuclear One Unit 2 outlet nozzles were also taken to be the average of all available analyses contained in Combustion Engineering report A-PENG-ER-002

[Ref. B-4]. However , the A-PENG-ER-002 report did not contain copper weight percent values for the inlet nozzles, because at the time that the Arkansas Nuclear One Unit 2 nozzles were manufactured, these values were not required to be documented for SA-508, Class 2 low-alloy steel. Therefore, no material-specific copper weight percent value is available for the Arkansas Nuclear One Unit 2 inlet nozzles. Per NRC RIS 2014-11 [Ref. B-1], a copper weight percent value is not required for calculation of the Arkansas Nuclear One Unit 2 nozzle material ART values , because the nozzles have fluence values less than 1 x 10 1 7 n/cm 2* However, if a copper weight percent value is ever needed, a best-estimate copper weight percent value is available from Section 4 of the NRC-approved Boiling Water Reactor Vessel and Internals Project (BWRVlP [proprietary])

report, BWRVlP-173-A

[Ref. B-5], and this value could be utilized for the Arkansas Nuclear One inlet nozzles. A mean plus two standard deviations methodology was applied to the data in BWRVIP-173-A to determine a conservative copper weight percent value. The data in the BWRVIP report was tabulated from an industry-wide database of SA-508, Class 2 forging materials.

The Charpy V-Notch forging specimen orientation for the inlet and outlet nozzles was not reported in PENG-ER-002

thus , it was conservatively assumed that the orientation was the "strong direction" for each nozzle forging. The initial RT NOT values were therefore determined for each of the Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzle forging materials using the Branch Technical Position (BTP) 5-3 , Position 1.1(3) methodology

[Ref. B-6]. The initial RT N OT values for all of the nozzle materials were determined directly from the data or by using a CVGRAPH, Version 6.02 hyperbolic tangent curve fit through the minimum data points , in accordance with ASME Code Section III, Subarticle NB-2331, Paragraph (a)(4) [Ref. B-7]. The initial RTNoT values were determined using both BTP 5-3 Position 1.1(3)(a) and Position 1.1(3)(b), and the more limiting initial RT N o T value was chosen for each nozzle forging material.

The Arkansas Nuclear One Unit 2 initial RT N O T values for the inlet and outlet nozzles materials are summarized in Table B-1. Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > 1 MeV) exposure of the Arkansas Nuclear One Unit 2 reactor vessel materials is discussed in Section 2 of this report. The fluence values used in the inlet and outlet nozzle ART calculations were calculated at the lowest extent of the nozzles (i.e., the nozzle to nozzle shell weld locations) and were chosen at an elevation lower than the actual elevation of the postulated flaw, which is at the inside comer of the nozzle , for conservatism.

Per Table 2-2, the inlet nozzles are determined to receive a projected maximum fluence of 7.96 x 10 1 6 n/cm 2 (E > 1 MeV) at the lowest extent of the nozzles at 54 EFPY. Similarly, the outlet nozzles are projected to achieve a maximum fluence value of 9.80 x 10 1 6 n/cm 2 (E > 1 MeV) at the lowest extent of the nozzles at 54 EFPY. Thus , the maximum neutron fluence values for the nozzle materials are not projected to exceed a fluence of 1 x 10 1 7 n/cm 2 at 54 EFPY. Per NRC RIS 2014-11 [Ref. B-1], embrittlement of reactor vessel materials, with projected fluence values less than 1 x 10 1 7 n/cm 2 , does not need to be considered.

Therefore, the initial RT N OT values documented in Table B-1 are identical to the nozzle ART values. The neutron fluence values used in the ART calculations for the Arkansas Nuclear One Unit 2 inlet and outlet nozzle forging materials are summarized in Table B-1. WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17

-00002 Rev O Page 68 of 88 Westinghouse Non-Proprietary Class 3 B-3 Table B-1 Summary of the Arkansas Nuclear One Unit 2 Reactor Vessel Nozzle Material Initial RT NDT, Chemistry, and Fluence Values at 54 EFPY Chemical Composition<*>

Fluence at Lowest RTNDT(U) (c) Extent of Nozzle

1.0 MeV) Wt.% Wt.% Wt.% Wt.% (OF) Cu Ni Mn p In l et Nozzle C-8015-1 Note (b) 0.69 0.65 0.007 30 7.96 X 10 1 6 In let Nozzle C-8015-2 Note (b) 0.6 0 0.70 0.010 10 7.96 X 10 1 6 In let Nozzle C-8015-3 Note (b) 0.69 0.65 0.006 10 7.96 X 10 1 6 In let Nozzle C-8015-4 Note (b) 0.65 0.7 0 0.007 30 7.96 X 10 1 6 Outlet Nozzle C-8016-1 0.12 0.63 0.64 0.006 0 9.80 X 10 1 6 Outlet Nozzle C-8016-2 0.17 0.76 0.83 0.017 13.5 9.80 X 10 16 Notes: (a) Chemistry values are the average of all available material-specific chemical analyses , unless otherwise not ed. (b) T h e Arkansas Nuclear One Unit 2 copper we i ght percent val ues are not documented in the historical r ecords. This value is not needed for the c urr ent analysis per NRC RIS 20 14-11 [R ef. 8-1], s in ce the fluence va lu es for these materials are below 1.0 x 10 1 7 n/cm 2. If a copper we i ght p e rcent va lu e i s needed in the future , a best-estimate copper weight perce nt value is available from Section 4 of the NRC-approved 8WRVIP (proprietary) report , 8WRVIP-l73-A [Ref. 8-5]. (c) RT NDT(U) va lues wer e determined using NUREG-0800 , 8TP 5-3 Position 1.1(3)(a) and (b) [Ref. 8-6] methodology with the more limiting RT N DT(UJ value being se l ected for eac h nozzle material. (d) F l uence values conservatively correspond to 54 EFPY flu e nce va lu es at the lowest extent of the nozzle weld. Table B-2 Summary of the Limiting ART Values for the Arkansas Nuclear One Unit 2 Inlet and Outlet Nozzle Materials Nozzle Material and ID Limiting ART Value EFPY N umber (OF) Inlet Nozzle C-8015-1 and C-8015-4 30 54 Outlet Nozz l e C-8016-2 13.5 The u se of the embrittlement conclusion of NRC RlS 2014-11 [R ef. B-1], and thus the limitin g ART values summarized in Table B-2, will remain unchanged as long as the fluence values assigned to the inlet and outlet nozzles remain below 1.0 x 10 1 7 n/cm 2 (E > 1.0 MeV). If these fluence values are reached, the Arkansas Nuclear One Unit 2 nozzle mat erial ART va lue s shou ld be re-evaluated. WCAP-18169-NP June2018 Revision 1 CALC-AN02-EP-17 -00002 Rev O Page 69 of 88 Westinghouse Non-Proprietary Class 3 B-4 B.2 NOZZLE COOLDOWN PRESSURE-TEMPERATURE LIMITS Allowable pressures are determined for a given temperature based on the fracture toughness of the limiting nozzle material along with the appropriate pressure and thermal stress intensity factors. The Arkansas Nuclear One Unit 2 nozzle fracture toughness used to determine the P-T limits is calculated using the limiting inlet and outlet nozzle ART values from Table B-2. The stress intensity factor correlations used for the nozzle comers are provided in ORNL study , ORNL/TM-2010 /2 46 [Ref. B-8], and are consistent with ASME PVP2011-57015 [Ref. B-9]. The methodology includes postulating an inside surface 1/4T nozzle comer flaw, and calculating through-wall nozzle comer stresses for a cooldown rate of 100°F /hour. The through-wall stresses at the nozzle corner location were fitted based on a third-order polynomial of the form: where, CJ = through-wall stress distribution x = through-wall distance from inside surface Ao, A 1 , A 2 , A 3 = coefficients of polynomial fit for the third-order polynomial , used in the stress intensity factor expression discussed below The stress i ntensity factors generated for a rounded nozzle corner for the pressure and thermal gradient were calculated based on the methodology provided in ORNL/TM-2010 /246. The stress intensity factor expression for a rounded corner is: where, K 1 stress intensity factor for a circular corner crack on a nozzle with a rounded inner radius comer a crac k depth at the nozzle corner, for use with 1/4 T (25% of the wall thickness) The Arkansas Nuclear One Unit 2 reactor vesse l inlet and outlet nozzle P-T limit curves are s hown in Figures B-1 and B-2, respectively, based on the stress intensity factor expression discussed above; also show n in these figures are the traditional beltline cooldown P-T limit curves from Figure 8-2. The nozzle P-T limit curves are provided for a cooldown rate of 100°F/hr , along with a steady-state curve. An outside s urface flaw in the nozzle was not considered because the pressure stress is significantly lower at the outside surface than the inside surface. A heatup nozzle P-T limit curve is also not provided since it would be less limiting than the cooldown nozzle P-T limit curve in Figures B-1 and B-2 for an inside surface flaw. Additionally, the cooldown transient i s more limiting than the heatup transient since it results in tensile stresses at the inside surface of the nozzle corner. WCAP-18169-NP June2018 Revi s ion 1 Conclusion CALC-AN02-EP-17-00002 Rev O Page 70 of 88 Westinghouse Non-Proprietary Class 3 B-5 Based on the results shown in Figures B-1 and B-2 , it is concluded that the nozzle P-T limits are bounded by the traditional beltline curves. Therefore, the P-T limits provided in Section 8 for 54 EFPY remain limiting for the beltline and non-beltline reactor vessel components. W C AP-1816 9-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 71 of 88 Westin g h o u se Non-Propri e tary C la ss 3 -C) (/J Q. -2500 2250 2000 1750 1500 L I 1250 ----4~1 t-4----1 ::::s 1/) 1/) Q. E 1000 .2l 1/) >, (/J -r:: C'l:J 0 0 0 ... 0 750 I --+ I t -l-I I t, Cooldown Rate :E 500 -~~ -100°F/Hr 0:: 250 Minimum Boltup Tern . = 60°F Inlet Nozz l e Cooldown -lOO"F/Hr I r Inlet Nozzle Steady State Cooldown Rates steady-state -25"F/Hr -60"F/Hr Acceptable 0 eration I r--0 50 100 150 200 250 300 350 400 450 500 550 Averaged Reactor Coolant System Temperature (°F) B-6 Figure B-1 Compari s on of Arkansas Nuclear One Unit 2 Beltline P-T Limit s to Inlet No z zle Limits WCA P-181 6 9-NP Jun e 2 018 R evisio n l CALC-AN02-EP-17-00002 Rev O Page 72 of 88 Westinghouse No n-Proprietary C l ass 3 2500 -,------,--,~-.,------,-----,----- ,---,------, -C) en 0.. -2250 2000 1750 1500 I t 1250 -+----:::, (I) (I) 0.. E 1000 .! en -C: 750 0 0 0 ... 0 0 cu 500 Q) a::: 250 0 0 50 Outlet Nozzle M.l.------t----t Cooldown l -100°F/Hr Outlet Nozzle ------t------f Steady State I t -+------------+---+ Cooldown Rate -100°F/Hr Minimum Boltup Tern . = 60" F t Cooldown Rates Lowest Service Tern . = 150" F --+-Acceptable 0 eration ... 100 150 200 250 300 350 400 450 500 550 Averaged Reactor Coolant System Temperature (°F) B-7 F igure B-2 C ompari s on of Arkansas Nuclear One U nit 2 Beltline P-T Limit s to Outlet Nozzle Limits WCA P-181 6 9-NP Jun e 20 1 8 R evision 1 B.3 REFERENCES CALC-AN02-EP-17-00002 Rev O Page 7 3 of 88 Westinghouse Non-Proprietary Class 3 B-8 B-1 NRC Regu l atory Issue Summary 2014-11 , " Informat i on on Licensing Applications for Fracture To u ghness Requirements for Ferritic Reactor Coo l a nt Pressure Bo u ndary Components ," U.S. Nuclear Reg ul atory Commission , Octo b er 2014. [A DAMS A c c es sion Numb e r ML 1 4 1 49AJ65] B-2 W e stinghouse Report WCAP-14040-A, Revision 4 , "Methodology Used to D eve l op Col d Ov e rpressure Mitigating System Setpoints and RCS Heatup and Coo l down Limit C u rves ," May 2004. B-3 U.S. Nuclear Regulatory Commission , Office of Nuclear Reg u latory Research , Reg ul atory Gui d e 1.99 , Revision 2 , " Rad i ation Embrittlement o f Reactor Vessel Materia l s ," May 1988. B-4 Co m bustion Engineering Report A-PENG-ER-002, R evision 0 , "The Reactor Vessel Group Re c ords Evaluation Program Phase II Fina l Report for th e ANO 2 Reactor Pressure Vessel Plates, Forgings , Welds and Cladding ," October 1995. B-5 BWR V IP-173-A: B W R V e ssel and Int e rnals Proj e ct: Evaluation of Ch e mi s try Data for BWR V esse l N o zz le Forging Mat e rials. EPRI, Pa l o Alto , CA: 2011. 1022835. B-6 NUREG-0800, Standard Review Plan for t h e Review of Safety Ana l ysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition , Branch Tec hni ca l Posi t ion 5-3, " Fracture Toughness Requir e ments ," Re v ision 2 , U.S. Nuclear Reg u latory Commission , March 2007. B-7 ASME Boiler and Pressure Vessel (B&PV) Code , Sect i on III , Division 1 , Subsection NB , " C l ass 1 Components." B-8 Oak Ridge N ational Laboratory Report , ORNL/TM-2 010/246 , " Stress and Frac tu re Mechanics An a lyses of Boiling Water Reactor an d Pressurized Water Reactor Pressure Vessel Nozzles -Revision 1 ," June 2012. [ADAMS Acc ession N umb e r ML 1 10060164] B-9 ASME PVP2011-57015 , "Additional Improvements to Appendix G of ASME Section X1 Code for No z zles," G. Ste v ens, H. Mehta , T. Gries b ach , D. Sommerville , Ju l y 2011. W C AP-18169-NP June 2018 Revi s ion I CALC-AN02-EP-17-00002 Rev O Page 74 of 88 Westinghouse Non-Proprietary Clas s 3 C-1 APPENDIX C NON-REACTOR VESSEL FERRITIC COMPONENTS 10 CFR Part 50 , Appendix G [Ref. C-1], requires that all Reactor Coolant Pressure Boundary (RCPB) componen t s meet the requirements of Section III of the ASME Code. The lowest service temperature requirement for all RCPB components, which is specified in NB-3211 and NB-2332 of the Section III ASME Code, is the relevant requirement that would affect the pressure-temperature (P-T) limits. The lowest service temperature (LST) requirement of NB-3211 and NB-2332 of the Section III ASME Code is applicable to material for ferritic piping, pumps and valves with a nominal wall thickness greater than 2 Yi inches [Ref. C-2]. Arkansas Nuclear One Unit 2 reactor coolant system piping contains ferritic materials i n the Class 1 piping; the pumps and valves do not contain ferritic material. The LST requirements of NB-3211 and NB-2332 are considered in Section 6.4 of this report. The other ferritic RCPB components that are not part of the reactor vessel consist of the replacement closure head , the pressurizer and the replacement steam generators. The replacement closure head is considered in the cylindrical beltline P-T limit curves as described in Section 6 of this report. Furthermore , the replacement closure head has not undergone neutron embrittlement that would affect P-T limits. Therefore , no further consideration is necessary for this component with regards to P-T limits. The pressurizer was constructed to the 1968 Edition through 1970 Summer Addenda Section III ASME C ode and met all applicable requirements at the time of construction and is original to the plant. Furthermore , the pressurizer has not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for this component with regards to P-T limits. The replacement steam generators were constructed to the 1989 Edition Section III ASME Code and met a ll applicable requirements at the time of construction. Furthermore , the replacement steam generators have not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for these components with regards to P-T limits. C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission , Federal R e gister, Volume 60, No. 243 , December 19, 1995. C-2 ASME Boiler and Pressure Vessel (B&PV) Code ,Section III , Division 1, Subsection NB , "Class 1 Components ." W C AP-18169-NP June 2018 Re v ision l CALC-AN02-EP-17-00002 Rev O Page 75 of 88 Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D CREDIBILITY EVALUATION OF THE WELD HEAT# 10137 SURVEILLANCE DATA D.1 INTRODUCTION Regulatory Guide 1. 99 , Revision 2 [Ref. D-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position 2.1 of Regulatory Guide 1.99 , Revision 2, describes the method fo r calculating the adjusted reference temperature of reactor vessel beltline materials using surveillance capsule data. The methods of Position 2.1 can only be applied when two or more credible surveillance data sets become available from the reactor in question. The credibility of all surveillance program data applicable to the Arkansas Nuclear One Unit 2 beltline was assessed in WCAP-18166-NP [Ref. D-2]. However, the Arkansas Nuclear One Unit 2 extended beltline contains two welds with sister plant data , the Upper Shell Longitudinal Welds l-203A , B , & C (Heat# BOLA) and the Upper to Intermediate Shell Girth Weld 8-203 (Heat# 1013 7). Note that no surveillance data is available for the other two Heats (# 6329637 and # FAGA) which were also used to make the Upper to Intermediate Shell Girth Weld 8-203. The weld Heat # BOLA sister plant data is available from the J.M. Farley Unit 2 surveillance program. Since this s urveillance data is analyzed by itself, the credibility conclusion documented in Appendix D of WCAP-16918, Revision 1 [Ref. D-3] is applicable to Arkansas Nuclear One Unit 2; thus, the credibility conclusion of the Heat# BOLA data need not be upd ated. The J.M. Farley Unit 2 surveillance weld data (Heat# BOLA) is non-credible in regard to the Arkansas Nuclear One Unit 2 reactor vessel materials. The weld Heat# 10137 sister plant data is available from both the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs. The Millstone Unit 2 surveillance program includes two distinct welds, Heat# 10137 an d Heat# 90136. In previous analyses, this weld surveillance data was treated as one combined we ld and subsequently analyzed together. However, these two weld metal heats were not melted tog e ther into a tandem weld; they were individually deposited. It cannot be determined with full confidence how much of the overall surveillance weld is which weld metal heat and, furthermore, exactly which weld heat specimens are contained in which surveillance capsules in the Millstone Unit 2 program. The Millstone Unit 2 (combined) surveillance weld data met the second and third credibility criteria of Regulatory Guide 1.99 , Revision 2 [Ref. D-1]. Additionally, Table D-2 of WCAP-16012 [Ref. D-4] indicates that all of the measured weld L'iRT NDT values were within the 1-sigma scatter band; therefore , suggesting that there is good agreement between the measured capsule data and the embrittlement correlations. If the two heats of weld material were evaluated individually, one would expect that the scatter in the data would decrease since the irradiated material would em brittle differently for the two separate welds with differ ent, as-measured, copper and nicke l contents. However , since the ( combined) weld mater ia l already passes the Regulatory Guide 1.99 , Revision 2 credibility analysis, a re-evaluation of the material ( as two separate heats) is not expected to significantly change the overall results of the subsequent reactor vessel integrity analyses. Thus, the surveillance weld metal will be considered to be only Heat# 10137 for the evaluations contained herein. All currently determined input data for Position 2.1 chemistry factor determination (See Section 5) and surveillance data credibility assessment WCAP-18169 -NP June 2018 Re vis ion I CALC-AN02-EP-17-00002 Rev O Page 76 of 88 Westinghouse Non-Proprietary Class 3 D-2 documented in this Appendix will be used "as-is," as documented in the Millstone Unit 2 surveillance capsule an a lyses of record. For conservatism, no reduction in the margin term of Regulatory Guide 1.99, Revision 2 [Ref. D-1] was taken to account for the additional uncertainties , despite the data remaining credible (see Section D.2). Note that this approach should also be followed when completing analyses per 10 CFR 50.61 [Ref. D-5]. Despite thi s additional conservatism , the Arkansas Nuclear One Unit 2 Upper to Intermediate Shell Girth Weld 8-20 3 (Heat# 10137) was not the limiting material for the Arkansas Nuclear One Unit 2 P-T limit curves. D.2 EVALUATION Per Appen d ix D of WCAP-17501-NP [Ref. D-6], the Calvert Cliffs Unit 2 surveillance weld data (Heat# 10137) was deemed credible , and per Appendix D of WCAP-16012 [Ref. D-4], the Millstone Unit 2 surveillance weld data (Heat# 10137) was also deemed credible. Thus , when analyzed individually, these surve illance welds pass all five of the Regulatory Guide 1.99, Revision 2 [Ref. D-1] credibility criterion. The only c r edibility criterion that must be updated as a result of analyzing the two surveillance welds together is Criterion
3. This evaluation is documented herein. Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of LlRT N OT values about a best-fit line drawn as described in Regulatory Guide 1.99 , Revision 2 [Ref. D-1] normally should be less than 28°F for welds and l 7°F for base metal. Eve n if the fluence range is large (two or more orders of magn i tude), the scatter should not exceed twice those val ues. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upp er-s helf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. D-7]. The functional form of the least-squares method as described in Regulatory Guide 1.99, Revision 2 will be utilized to determine a be st-fit line for this data and to determine if the scatter of these LlRT NDT values about this line is less than 28°F for the weld. Following i s the calculation of the best-fit line as described in Reference D-1. In addition, the recommended NRC method s for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13 , 1998 [Ref. D-8]. At this meeting the NRC presented five cases. Of the five cases, Case 5 ("Surveillance Data from Other Sources Only") most closely repre sents the situation for the Arkansas Nuclear One Unit 2 reactor vessel Upper to Intermediate Shell Girth Weld 8-203 (Heat# 10137) as described below: Heat# 10137 {Case 5) -This weld heat p ertai ns to the Upper to Intermediate Shell Girth Weld 8-203 in the Arkansas Nuclear One Unit 2 reactor vessel. This weld heat is not contained in the Arkansas Nuclear One Unit 2 s urveillance program. Howeve r, it is contained in the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs.
NRC Case 5 per Reference D-8 is entitled "Surve illance Data from Other Sources Onl y" and most closely represents the situation for Arkansas Nuclear One Unit 2 weld Heat# 10137. WCAP-18169-NP June 2018 Revi s ion I CALC-AN02-EP-17 -00002 Rev O Page 77 of 88 Westinghouse Non-Proprietary Class 3 D-3 Credibility Assessment Case 5: Weld Heat# 10137 (Calvert Cliffs Unit 2 Data Only) Following the NRC Case 5 guidelines, the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance weld metal (Heat# 10137) will be evaluated for credibility. Weld Heat# 10137 pertains to Arkansas Nuclear One Unit 2 reactor vessel Upper to Intermediate Shell Girth Weld 8-203 , but is not contained in the Arkansas Nuclear One Unit 2 surveillance program. In accordance with the NRC Case 5 guidelines , the data from only Calvert Cliffs Unit 2 will be analyzed first, since the irradiation environment for Calvert Cliffs Unit 2 is judged closer to that of Arkansas Nuclear One Unit 2 as evidenced by the temperature adjustments documented in Table 4-2. This assessment was performed in Appendix D of WCAP-17501 -NP [Ref. D-6] and concluded that the s urveillance data for Heat# 10137 from Calvert Cliffs Unit 2 only was credible. Therefore, in accordance with Case 5 , the combined data from both Calvert Cliffs Unit 2 and Millstone Unit 2 will now be assessed to determine the credibility conclusion for all applicable data for weld Heat# 10137. W C AP-1816 9-NP June 2018 Revi s ion l CALC-AN02-EP 00002 Rev O Page 78 of 88 Westinghouse Non-Proprietary Class 3 Credibility Assessment Case 5: Weld Heat# 10137 (All data) D-4 In acc orda n ce wi th the NRC Case 5 guidelines, th e data from Calvert Cliffs Unit 2 and Millstone Unit 2 wi ll now b e ana l yze d together. Dat a is adjusted to the m ea n chemical composition and operating temperatur e of the s urveillanc e capsules. This is performed in Table D-1. Table D-1 Mean Chemical Composition and Operating Temperature for Calvert Cliffs Unit 2 and Millstone Unit 2 Material Capsule 263° Weld Metal Heat# 10137 (Ca lv ert Cliffs Unit 2 Data) 97° 104° 97° Weld Meta l Heat# 10137 (Millstone Un it 2 Data) 104° 83° MEAN Note: (a) C hemistry data obtained from Table 3-1. (b) Temperatu r e data obtained from Table 4-2. Cu Ni In l et Temperature during Wt.%<*> Wt.%<*> Period of Irradiation (0 F)(b) 550 0.21 0.06 549 548 544.3 0.21 0.0 6 547.6 548.0 0.21 0.06 547.8 Since the mean che mic a l compositio n of the s urveill ance capsule data is identical to the actual chemical compos ition data for eac h caps ule , no chemistry adjustment is nec essary. However, since the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance capsule operating temperatures are not identical to the mean operating temperature, the s ur veillance capsule data will be ad ju ste d to the mean operating tempera tu re. The capsule-specific temperature a dju s tment s are as s hown in Table D-2. Table D-2 Operating Te mperature Adjustments for the Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data Material Caps ule Inlet Temperature d u ring Mean Operating Temperature Period of Irradiation (°F) Temperature (°F) A djustment (°F) 263° 550 2.2 Weld Me t a l Heat# 10137 (Calvert C liff s Unit 2 Data) 97° 549 1.2 104° 548 0.2 547.8 97° 544.3 -3.5 Weld Meta l Heat# 10137 (Mi ll stone Unit 2 Data) 104° 547.6 -0.2 830 548.0 0.2 Using the chemical composition and operating temperature a djustm ents described an d calculated above, an interim c hemistry factor is calc ulat ed for wel d Heat # 10137 using the Ca l vert Cliffs Unit 2 and Millstone Unit 2 data. This calculation is shown on the following page in Table D-3. WCAP-18 1 69-NP June 2018 Revision I Table D-3 Material Weld Metal Hea t# 10137 (Calve rt Cliffs Unit 2 Da t a) Weld Metal Hea t# 1013 7 (Mi llsto ne Unit 2 Data) Notes: CALC-AN02-EP-17-00002 Rev O Page 79 of 88 Westinghouse Non-Proprietary Class 3 D-5 Calculation of Weld Heat # 10137 Interim Chemistry Factor for the Credibility Evaluation Using Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data Capsule t<*l .1RTNDT (c) FF*.1RTNDT Capsule FF<bl (x 10 19 n/cm2, E > 1.0 MeV) (OF) {°F) 263° 0.825 0.9460 74.9 (72.7) 70.86 97° 1.95 1.1825 84.1 (82.9) 99.45 104° 2.44 1.2401 69.9 (69.7) 86.68 97° 0.324 0.6902 62.4 (65.93) 43.09 104° 0.949 0.9853 51.9 (52.12) 51.16 830 1.740 1.1523 56.3 (56.09) 64.86 SUM: 416.10 CF H e at# 10137 = L(FF * ~RT NDT) .;-L(FF 2) = (416.10) .;-(6.606) = 63.0°F FF 2 0.895 1.398 1.538 0.476 0.971 1.328 6.606 (a) f= fluence; (b) FF= fluence factor= t<0*28 -O.IO'l o g I). (c) t>RT N OT values are the measured 30 ft-lb shift values. Each t>RT N OT value ha s been adjusted according to the temperature adjustments summar i zed in Table 0-2. The t>RT N OT va lu es for each surveillance weld data point are not adjusted by the ratio procedure, because the mean chemical composition is id e ntical to eac h capsule chemical composition (pre-adjusted values are listed in parentheses and we r e taken from Table 4-2). The scatter of ~RT NDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99 , Revision 2 , Position 2.1 [Ref. D-1] is presented in Table D-4. Table D-4 Material Weld Metal Heat# 10137 (Calvert Cliffs Unit 2 Data) Weld Meta l Heat# 10137 (Mi ll sto ne Unit 2 Data) Best-Fit Evaluation for Surveillance Weld Metal Heat# 10137 Using Calvert Cliffs Unit 2 and Millstone Unit 2 Data CF Capsule f Measured Predicted Residual Capsule (Slopebest-fit) (x 10 19 n/cm2, FF .1RTNDT .1RTNDT .1RTNDT <28°F (Weld) (OF) E > 1.0 MeV) (OF) (OF) (OF) 263° 63.0 0.825 0.9460 74.9 59.6 15.3 Yes 97° 63.0 1.95 1.1825 84.l 74.5 9.6 Yes 104° 63.0 2.44 1.240 1 69.9 78.1 8.2 Yes 97° 63.0 0.324 0.6902 62.4 43.5 18.9 Yes 104° 63.0 0.949 0.9853 51.9 62.1 10.2 Yes 830 63.0 1.740 1.1523 56.3 72.6 16.3 Yes The scatter of ~RT N DT values a bout the best-fit line , drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [R ef. D-1 ], should be less than 28°F for weld metal. Table D-4 indicates that 100% (six out of s ix) of the survei llanc e data points fall within the +/-la of 28°F scatter band for survei llanc e weld materials. Therefore, the s ur veillance weld material (Heat # 1013 7) is deemed "credi ble" per the third criterion when all available data is cons idered. WCAP-1 8169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 80 of 88 Westinghouse Non-Proprietary C l ass 3 D-6 In conclus i on, the combined survei ll ance data from Calvert C li ffs Unit 2 and Millstone Unit 2 for weld Heat# 10137 may be applied to the Arkansas Nuclear One Unit 2 reactor vessel weld. The Po sition 2.1 chemistry factor calc ul ation, as applica bl e to the Arkansas Nuclear One Unit 2 reactor vessel wel d , is con t ained in Section 5. This Position 2.1 CF va lu e could be u se d with a reduced margin term in the ART calc ulat ions contained in Section 7. However, consistent w ith the discussion in Section D.1 of this Appendix, the ART va lu es calculate d with the Position 2.1 CF value for weld Heat# 10137 utili ze a full margin term for conservatism. D.3 REFERENCES D-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Ra diation Embrittlement of Reactor Vessel Materials," May 1988. D-2 Westinghouse Report WCAP-1 8166-NP, Revision 0, "Ana l ysis of Ca p s ul e 284° from th e Entergy Operations , Inc. Arkansas Nuclear One Unit 2 Reactor Vessel Radiation Surveillance Program ," September 2016. D-3 W e stinghouse Report WCAP-16918-NP, Revision 1 , "Analysis of Capsule V from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Radiation Surveillance Program ," April 2008. D-4 Westinghouse Report WCAP-16012, Revision 0 , " Analysis of Caps ule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003. D-5 Code of Federal Regulations , 10 CFR 50.61, "Frac ture Toughness Requirements for Protection Ag a inst Pressurized Thermal Shock Events," Federa l Register , Volume 60 , No. 243, dated December 19, 1995 , effective January 18 , 1996. D-6 Westinghouse R e port WCAP-17501-NP, Revision 0, "Analysis of Caps ul e 104° from the Calvert Cli ff s Unit No. 2 Reactor Vessel Radiation Surveillance Program ," Fe bru ary 2012. D-7 ASTM El85-82 , "Stan dard Practice for Conduct i ng Surveillance Tests for Light-Water Coo l ed Nuclear Power Reactor Vessels," ASTM, Jul y 1982. D-8 K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, N RC/lndustry Workshop on RPV Int egrity Issu es, February 12 , 1998. [A DAMS Accession Numbe r MLI 100 7057 0] WCAP-1 816 9-NP June 2018 Revision I
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