ML18085A160
| ML18085A160 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/30/2015 |
| From: | Wilmart V Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| WCAP-18028-NP, Rev 01 | |
| Download: ML18085A160 (22) | |
Text
Serial Number 18-098 Docket Nos. 50-280/281 Attachment 2 WCAP-18028-NP, REV. 0 1 EXTENDED BELTLINE PRESSURE VESSEL FLUENCE EVALUATIONS APPLICABLE TO SURRY UNITS 1 & 2 SEPTEMBER 2015 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
SURRY POWER STATION UNITS 1 AND 2
Westinghouse Non-Proprietary Class 3 WCAP-18028-N P September 2015 Revision 0 Extended Beltline Pressure Vessel Fluence Evaluations Applicable to Surry Units 1 & 2
@ Westinghouse
Westinghouse Non-Proprietary Class 3 WCAP-18028-NP Revision 0 Extended Beltline Pressure Vessel Fluence Evaluations Applicable to Surry Units 1 & 2 Valerie Wilmart*
Safety & Systems Engineering September 2015 Reviewer: Benjamin W. Amiri*
Nuclear Operations & Radiation Analysis Approved: Laurent P. Houssay*, Manager Nuclear Operations & Radiation Analysis
- Electronically approved records are authenticated in the electronic document management system.
Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA
© 2015 Westinghouse Electric Company LLC All Rights Reserved
Westinghouse Non-Proprietary Class 3 II Record of Revisions Rev. Revision Descri1 tion Com leted 0-A Original Draft Issue June 2015 0-B This revision includes the calculated fluence at the V,i T flaw in the nozzles.
0 Original Official Issue WCAP-18028-NP September 20 15 Revision 0
Westinghouse Non-Proprietary Class 3 Ill TABLE OF CONTENTS LIST OF TABLES .... ........... .. .. ........... .. ................. .. ..... ........................... .. ..... .. .... ......... .. ............. .. ............. iv LIST OF FIGURES ....................... ...... ..................... ********************************************************************** ************** V 1 INTRODUCTION ................. .. ................ ...................................... ....... ...................... ... .. .. .... ....... 1-1 2 CALCULATED NEUTRON FLUENCE .. .......... ............................. .. .. ......... .. .................... ......... 2-l
2.1 INTRODUCTION
.. ......... .... ....................................................................................... .. ... 2-l 2.2 DISCRETE ORDINATES ANALYSIS .................................. .. .... ................... .. ........ ... ... 2-2 2.2.1 Method Discussion ........ .. .. .. .......................................................... ......... ......... 2-2 2.2.2 Reactor Geometry ............ ............ .... ........... ........ ....... .... .. ......................... ....... 2-2 2.2.3 Cycle-Specific Information ... .. .. ... .................................................................. . 2-6 2.2.4 Results ... ....................... ...................... ................................ ............................. 2-7 2.2.5 Recommendations ........ .... ... ....... ....... ................................................... .........2-13 3 REFERENCES ................ .. ................................ ..... .. ........... ....... .. ........ .... ..................... .. ............. 3-1 WCAP-1 8028-NP September 201 5 Revision 0
Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 2-1 Surry Units 1 & 2 - Pressure Vessel Material Locations ..... ....... .. .......... ............... ......... .. 2-8 Table 2-2 Surry Unit 1 - Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by Pressure Vessel Materials in the Extended Beltline .... .. .......................... ...... .. .... ................. ........... 2-9 Table 2-3 Surry Unit 2 - Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by Pressure Vessel Materials in the Extended Beltline... ... .. ................... ............ ............ ................... 2-10 WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 V LIST OF FIGURES Figure 2-1 Surry [r,0] Reactor Geometry at Core Midplane ........................... .. ...... .......................... 2-4 Figure 2-2 Surry [r,z] Reactor Geometry ...................................................................... ........ .. .......... 2-5 Figure 2-3 Surry Unit 1 - Axial Boundary of the l .OE+ 17 n/cm2 Fluence Threshold in the +Z Direction (at 54 and 72 EFPY) ............................. ........................ .. .. ............................ 2-11 Figure 2-4 Surry Unit 2 - Axial Boundary of the 1.0E+ 17 n/cm2 Fluence Threshold in the +Z Direction (at 54 and 72 EFPY) ..................................................................................... 2-12 WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION In the assessment of the state of embrittlement of light water reactor (LWR) pressure vessels, an accurate evaluation of the neutron exposure of each of the materials comprising the beltline region of the vessel is required. In Section II F of 10 CFR 50 Appendix G [Ref. 2], the beltline region is defined as:
"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 reactor core and adjacent regions 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".
In Section II A of l O CFR 50 Appendix H [Ref. 2], the lower limit of neutron exposure for consideration of radiation induced material damage is specified by a neutron fluence (E > 1.0 MeV) threshold of LOE+ 17 n/cm2 . Each of the materials that is anticipated to experience a neutron exposure that exceeds this fluence threshold must be considered in the overall embrittlement assessments for the pressure vessel.
The existing fluence analysis of the Surry Units l and 2 pressure vessels [Refs. 3 and 4] was limited to an axial range that extended approximately 1.5 foot above and 1 foot below the active fuel stack. This model did not include all the pressure vessel materials that could potentially exceed the I .OE+ 17 n/cm2 (E > 1.0 MeV) fluence threshold defined in 10 CFR 50 Appendix H [Ref. 2]. The purpose of this extended beltline fluence evaluation is to define which materials in the Surry Units 1 and 2 pressure vessels are projected to exceed the LOE+ 17 n/cm2 threshold neutron fluence before the End of License Extension (EOLE); and, to project the neutron fluence for each of these specific materials. This will help Dominion to fulfill its commitment with respect to the USNRC Request for Additional Information (RAI)
Docket Nos.: 50-280/281 [Ref. 8] in determining whether the neutron fluence exposure (E > 1.0 MeV) of the inlet and outlet nozzle materials would be greater than l.OE+l7 n/cm2
- In subsequent sections of this report, the methodologies used to perform neutron transport calculations are described in some detail and the results of the plant-specific transport calculations are given for each of the materials located in the traditional and extended beltline regions of the Surry Units 1 and 2 pressure vessels.
WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-1 2 CALCULATED NEUTRON FLUENCE
2.1 INTRODUCTION
A discrete ordinates SN transport analysis was performed for the Surry Units 1 and 2 reactors to determine the neutron radiation environment within the extended beltline of the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant- and fuel-cycle-specific basis.
All of the calculations described in this report were based on nuclear cross-section data derived from ENDF/B-VI. Furthermore, the neutron transport evaluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 5). Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-14040-A, Revision 4
[Ref. 1).
Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [Ref. 5], describes state-of-the-art calculation and measurement procedures that are acceptable to the USNRC staff for determining pressure vessel fluence. Also included in Regulatory Guide 1.190 is a discussion of the steps required to qualify and validate the methodology used to determine the neutron exposure of the pressure vessel wall. One important step in the validation process is the comparison of plant-specific neutron calculations with available measurements. An evaluation of the dosimetry sensor sets from three of the first four surveillance capsules withdrawn from Surry Unit 1 is provided in Reference 3 while the evaluation for the first five surveillance capsules withdrawn from Surry Unit 2 is provided in Reference 4. The dosimetry analyses documented in References 3 and 4 showed that the
+/-20% (lcr) acceptance criteria specified in Regulatory Guide 1.190 [Ref. 5] is met.
The results of the present extended beltline analysis are consistent with those of References 3 and 4.
Therefore, the Regulatory Guide 1.190 acceptance criteria continue to be met. The validated calculations form the basis for providing future projections of the neutron exposure of the reactor pressure vessel. In line with References 3 and 4, projections up to 54 EFPY are provided in Section 2.2.4. Extended projections up to 72 EFPY are also provided. In addition, as per Dominion's request, projections corresponding to 80 years of life are provided based on 18-month cycles with an average outage time of 25 days. This was determined to correspond to 68 EFPY.
WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-2 2.2 DISCRETE ORDINATES ANALYSIS 2.2.1 Method Discussion In performing the fast neutron exposure evaluations for the Surry Units 1 and 2 reactor vessels, a series of fuel-cycle-specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:
<p(r,9,z) = <p(r,9)x <p(r,z)
<p(r) where <p(r, 9, z) is the synthesized three-dimensional neutron flux distribution, <p(r, 9) is the transport solution in r,0 geometry, <p(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 same source per unit height as that used in the [r,0] two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at Surry Units 1 and 2.
All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code [Ref. 6] 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 specifically for light-water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 Legendre expansion and angular discretization was modeled with an S 16 order of angular quadrature.
Energy- and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle-specific basis.
2.2.2 Reactor Geometry The analyses documented in References 3 and 4 formed the basis for the current extended beltline evaluation. In completing the current analysis, the [r,9] and [r] models from Reference 3 and 4 were retained as is while the [r,z] model was expanded to encompass all axial elevations that were anticipated to experience a neutron fluence greater than I.OE+ 17 n/cm2
- The [r,z] model was expanded by about 2.5 feet in the +Z direction (relative to the core midplane) to encompass these axial elevations.
For the Surry Units 1 and 2 transport calculations, the [r,0] model depicted in Figure 2-1 was utilized since the reactor is octant symmetric. This [r,0] model includes the core, the reactor internals, the thermal shield - including explicit representations of the surveillance capsules at 15°, 25°, 35° and 45° - the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the water-filled shield tank. The symmetric [r,0] model was utilized to perform both the surveillance capsule dosimetry evaluations with subsequent comparisons with calculated results [Refs. 3 and 4], and to generate the maximum fluence levels at the pressure vessel wall. In developing this analytical model, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core, bypass and downcomer regions of the WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-3 reactor were taken to be representative of full-power operating conditions. The coolant densities were treated on a fuel-cycle-specific basis (see Section 2.2.3). The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the [r,8] reactor model consisted of 156 radial by 83 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 flux convergence criterion utilized in the [r,8] calculations was set at a value of0.001.
The [r,z] model used for the Surry Units 1 and 2 calculations is shown in Figure 2-2. The model extends radially from the centerline of the reactor core out to a location interior to the water filled shield tank and over an axial span from an elevation approximately 1 foot below to 4 feet above the active fuel. As in the case of the [r,8] models, nominal design dimensions and full-power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baflle and core barrel regions were also explicitly included in the model. The [r,z] geometric mesh description of these reactor models consisted of 148 radial by 106 axial intervals. As in the case of the
[r,8] calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the [r,z]
calculations was also set at a value of 0.001.
The one-dimensional radial model used in the synthesis procedure consisted of the same 148 radial mesh intervals included in the [r,z] model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.
WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-4 0
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Figure 2-1 Surry [r,0] Reactor Geometry at Core Midplane WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-5
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Westinghouse Non-Proprietary Class 3 2-6 2.2.3 Cycle-Specific Information Since the analyses presented in References 3 and 4 represent the basis for the current evaluation, most of the core design data and operating parameters were taken directly from those analyses. In particular, Reference 3 used Unit 1 cycle-specific core design information for Cycles 1 through 21 while Reference 4 used Unit 2 cycle-specific core design information for Cycles 1 through 20. Projections were used beyond that point. The input data of References 3 and 4 were obtained from Reference 9 and included a power uprate. Since then, Cycles 22 through 26 were implemented at Unit 1 and Cycles 21 through 25 at Unit 2.
The cycle-specific core design data for these latter cycles were taken from Reference 10. Note that the Measurement Uncertainty Recapture (MUR) uprate occurred at the onset of Unit 1 Cycle 24 on December 1, 2010 and at the onset of Unit 2 Cycle 24 on June 15, 2011.
The future projections were based on the assumption that the core power distribution and associated plant operating characteristics from the latest implemented cycle were representative of future plant operation.
Therefore, for Unit 1, projections for Cycles 27 and beyond were based on Cycle 26 while for Unit 2, projections for Cycles 26 and beyond were based on Cycle 25.
The data utilized for the core power distributions in plant-specific transport analyses included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions. This information was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle for use in the [r,8], [r,z], and [r] discrete ordinates transport calculations. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle-averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. The cycle length was taken from References 3 and 4 up until Unit 1 Cycle 21 and Unit 2 Cycle 20. Beyond that point, the data were taken from Reference 10.
In constructing these core source distributions, the Westinghouse generic approach was used. In this approach, the source term originates from the fission of six nuclides: 235 U, 238 U, 23 9I>u, 240Pu, 241 Pu and 242 Pu. Generic values are used including the fission spectra, fission sharing, energy released per fission and average number of neutrons per fission. The relative pin power distributions are taken from the Westinghouse Core Radiation Source Data (CRSD).
Water densities in the core, bypass, and downcomer regions as well as in the upper and lower core plena regions were determined on a fuel cycle-specific basis consistent with the average temperature rise in the 36 fuel assemblies located on the periphery of the reactor core. Since the neutron fluence at the pressure vessel is dominated by leakage from the peripheral fuel assemblies, the use of the peripheral water density in the analytical models is justified. The normal operating condition temperatures were taken from References 3 and 4 up until Unit l Cycle 21 and Unit 2 Cycle 20. Beyond that point, the data were taken from Reference 10.
WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-7 2.2.4 Results In Table 2-1, locations of the Surry Units 1 and 2 vessel welds and plates are provided. The axial position of each material is indexed to z = 0.0 cm, which corresponds to the mid-plane of the active fuel stack.
Selected results from the neutron transport analyses are provided in Tables 2-2 and 2-3 for Units 1 and 2 respectively. Calculated fast neutron (E > 1.0 MeV) fluence for reactor vessel materials, on the pressure vessel clad/base metal interface, is provided for the nominal end of Cycle 26 for Unit 1 (32.5 EFPY) and nominal end of Cycle 25 for Unit 2 (31.3 EFPY). In line with References 3 and 4, projections up to 54 EFPY are provided. Extended projections up to 72 EFPY are also provided. In addition, as per Dominion's request, projections corresponding to 80 years of life are provided based on 18-month cycles with an average outage time of 25 days. Surry Unit 1 will reach its 80-years EOLE on May 25, 2052 while Unit 2 will reach it on January 29, 2053 . Assuming that Unit 1 Cycle 27 started on May 2015 and that Unit 2 Cycle 26 started on May 2014, it was determined that 80-years of life correspond to 68 EFPY for both units.
Figures 2-3 and 2-4 show the relevant weld locations. For the regions beyond the upper circumferential weld, Figure 2-3 shows the axial boundary of the LOE+ 17 n/cm2 fluence threshold (at 54 and 72 EFPY) as a function of azimuthal position (Z versus 0) for Unit 1 while Figure 2-4 shows the information for Unit 2. It is noted that the nozzle materials located above the nozzle centerline remain below l .OE+ 17 n/cm2 through EOLE. Likewise, the lower shell to lower head circumferential weld remains out of the beltline region through EOLE.
WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-8 Table 2-1 Surry Units 1 & 2 - Pressure Vessel Material Locations Material Axial Location
. Azimuthal Location
[cm] [de2reesl 1/4 T Flaw in Outlet Nozzle Nozzle 1 276.98 25 Nozzle 2 276.98 145 Nozzle 3 276.98 265 1/4 T Flaw in Inlet Nozzle Nozzle 1 272.69 95 Nozzle 2 272.69 215 Nozzle 3 272.69 335 Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle I 261.94 25 Nozzle 2 261.94 145 Nozzle 3 261.94 265 Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1 256.22 95 Nozzle 2 256.22 215 Nozzle 3 256.22 335 Nozzle Shell to Intermediate Shell 203.6 1 to 205.61 0 to 360 Circumferential Weld Intermediate Shell Plate 1 -50.19 to 203 .61 45 to 225 Plate 2 -50.19 to 203 .61 225 to 45 Intermediate Shell Longitudinal Welds Weld 1 -50.19 to 203.61 45 Weld2 -50.19 to 203.61 225 Intermediate Shell to Lower Shell
-52.19 to -50.19 0 to 360 Circumferential Weld Lower Shell Plate 1 -305.69 to -52.19 135 to 3 15 Plate 2 -305.69 to -52.19 315 to 135 Lower Shell Longitudinal Welds Weld 1 -305 .69 to -52.19 135 Weld2 -305.69 to -52.19 315 Lower Shell to Lower Vessel Head
-305 .69 0 to 360 Circumferential Weld
- Axial elevations are indexed to Z = 0.0 at the rnidplane of the active fuel stack.
WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-9 Table 2-2 Surry Unit 1 - Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by Pressure Vessel Materials in the Extended Beltline Neutron Fluence [n/cm'l Material 32.5 EFPY 54 EFPY 68 EFPY* 72 EFPY 1/4 T Flaw in Outlet Nozzle l.53E+ l6 2.69E+ l6 3.45E+ l6 3.67E+l6 Nozzle 1 Nozzle 2 l .08E+ l6 l.93E+ l6 2.49E+ l6 2.65E+ l6 Nozzle 3*** 4.48E+ l6 7.59E+ l6 9.62E+ l6 1.02E+ 17 1/4 T Flaw in Inlet Nozzle 5.80E+ l6 9.82E+ l6 l.24E+ 17 l.32E+ 17 Nozzle I***
Nozzle 2 1.40E+ l6 2.50E+l6 3.22E+ l6 3.42E+ l6 Nozzle 3 l.98E+ l6 3.48E+l6 4.46E+l6 4.74E+l6 Outlet Nozzle Forging to Vessel Shell Welds -
Lowest Extent 3.62E+ l6 6.35E+ l6 8.13E+ l6 8.63E+ l6 Nozzle 1 Nozzle 2 2.55E+ l6 4.55E+ l6 5.86E+ l6 6.23E+l6 Nozzle 3** 1.06E+ l 7 l.79E+ l7 2.27E+ l7 2.40E+ l 7 Inlet Nozzle Forging to Vessel Shell Welds -
Lowest Extent l.42E+ l7 2.40E+ l 7 3.04E+l 7 3.22E+ l7 Nozzle 1**
Nozzle 2 3.43E+ l6 6.IOE+l6 7.84E+ l6 8.34E+l6 Nozzle 3** 4.85E+ l6 8.51E+ I6 1.09E+ l7 l.16E+ l7 Nozzle Shell 3.64E+ I8 6.00E+ l8 7.54E+ l8 7.98E+ l8 Nozzle Shell to Intermediate Shell Circumferential Weld 3.64E+ I8 6.00E+ l8 7.54E+ I8 7.98E+l8 Intermediate Shell Plate 1 3.17E+ l9 5.06E+ I9 6.29E+ l9 6.65E+l9 Plate 2 3.17E+ l9 5.06E+ l9 6.29E+ l9 6.65E+ I9 Intermediate Shell Longitudinal Welds Weld 1 5.75E+ l8 9.85E+ l8 l.25E+I9 l.33E+l9 Weld2 5.75E+l8 9.85E+ 18 1.25E+ I9 l.33E+l9 Intermediate Shell to Lower Shell Circumferential Weld 3.18E+l9 5.08E+ l9 6.3IE+ l9 6.67E+l9 Lower Shell Plate 1 3.20E+ l9 5.11E+ l9 6.35E+ l9 6.70E+l9 Plate 2 3.20E+ l9 5.11E+ l9 6.35E+ l9 6.70E+l9 Lower Shell Lo ngitudinal Welds Weld 1 5.80E+ l8 9.94E+l8 l.26E+ I9 l.34E+l9 Weld2 5.80E+ l8 9.94E+ l8 l .26E+ l9 1.34E+ l9 Lower Shell to Lower Vessel Head
< 1E+ l7 < 1E+ l7 < 1E+l7 < IE+ l7 Circumferential Weld
- Corresponds to 80 years of life
- l/4T flaw in Outlet Nozzle 3 is projected to reach I .OE+ 17 n/crn2 at approximately 70.7 EFPY, which corresponds to June 10, 2055 l /4T flaw in Inlet Nozzle 1 is projected to reach I .OE+ 17 n/cm2 at approximately 55 .0 EFPY, which corresponds to December 21 , 2038 WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2- 10 Table 2-3 Surry Unit 2 - Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by Pressure Vessel Materials in the Extended Beltline Neutron Fluence [n/cmi]
Material 31.3 EFPY 54 EFPY 68 EFPY* 72 EFPY 1/4 T Flaw in Outlet Nozzle 1.49E+l6 2.66E+16 3.38E+ l 6 3.58E+ 16 Nozzle 1 Nozzle 2 1.09E+ l 6 1.95E+ 16 2.48E+l6 2.63E+16 Nozzle 3*** 4.29E+l6 8.28E+ l6 l .07E+l 7 l.15E+ l7 1/4 T Flaw in Inlet Nozzle Nozzle 1*** 5.55E+ l6 1.07E+ l7 l.39E+ l7 1.48E+ l7 Nozzle 2 l.41E+ l6 2.52E+l6 3.21E+ l 6 3.40E+16 Nozzle 3 1.93E+l6 3.44E+ l6 4.37E+ l6 4 .63E+ 16 Outlet Nozzle Forging to Vessel Shell Welds -
Lowest Extent 3.52E+ l6 6.27E+l6 7.96E+ 16 8.45E+ l6 Nozzle 1 Nozzle 2 2.57E+ l6 4.60E+ l6 5.85E+ l6 6.20E+l6 Nozzle 3** 1.0IE+ l7 l. 95E+ l7 2.53E+ 17 2.70E+l7 Inlet Nozzle Forging to Vessel Shell Welds -
Lowest Extent l.36E+l7 2.62E+ l7 3.40E+ l 7 3.62E+ l7 Nozzle 1**
Nozzle 2 3.45E+ l6 6.17E+ l6 7.84E+ l6 8.32E+ l6 Nozzle 3** 4.73E+ l 6 8.41E+ l6 l.07E+l7 l.13E+l7 Nozzle Shell 3.52E+ l8 6.70E+ 18 8.65E+l8 9.21E+ l8 Nozzle Shell to Intermediate Shell Circumferential Weld 3.52E+ l8 6.70E+ l8 8.65E+l8 9.21E+l8 Intermediate Shell Plate 1 3.10E+l9 5.64E+l9 7.20E+ l9 7.65E+ l9 Plate 2 3.10E+l9 5.64E+ l 9 7.20E+l9 7.65E+l9 Intermediate Shell Longitudinal Welds Weld 1 5.98E+ l8 1.03E+ l9 1.29E+ l9 I .36E+l 9 Weld2 5.98E+l8 l.03E+ l9 1.29E+ l9 1.36E+ l9 Intermediate Shell to Lower Shell Circumferential Weld 3.l 1E+l9 5.66E+ l 9 7.22E+l9 7.67E+l9 Lower Shell Plate I 3.12E+ l9 5.68E+l9 7.26E+ l9 7.71E+l9 Plate 2 3.12E+l9 5.68E+ l 9 7.26E+l9 7.71E+l9 Lower Shell Longitudinal Welds Weld I 6.03E+ l8 1.03E+l9 l.30E+ l9 l.37E+l9 Weld2 6.03E+ l8 1.03E+l9 l.30E+ l9 l.37E+l9 Lower Shell to Lower Vessel Head
< IE+ l7 < 1E+l7 < 1E+l7 < 1E+l7 Circumferential Weld
- Corresponds to 80 years of life
- l/4T flaw in Outlet Nozzle 3 is projected to reach 1.0E+ 17 n/cm2 at approximately 63.8 EFPY, which corresponds to July 6, 2048 l /4T flaw in Inlet Nozzle 1 is projected to reach I .OE+ 17 n/crn2 at approximately 50.9 EFPY, which corresponds to December 25, 2034 WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-11 400
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- -1E+17 n/cm2 threshold at 54 EFPY 326.7 (Nozzle Cen ertin'e) -
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-400 ..... 1/4 T Flaw Location in Inlet Nozzle Figure 2-3 Surry Unit 1 -Axial Boundary of the 1.0E+17 n/cm 2 Fluence Threshold in the +Z Direction (at 54 and 72 EFPY)
WCAP-18028-NP September 2015 Revision 0
Westinghouse Non-Proprietary Class 3 2-12 400
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- 1E+l7 n/cm2 threshold at 54 EFPY 326.7 (Nozzle Cen erllne) -
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-182.83 (bottomof rei - - - - - - - -
-200 45 90 135 180 225 270 315 360 Azimuthal location (degree)
Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent
- 1/4 T Flaw Location in Outlet Nozzle
-400 ***
- 1/4 T Flaw Locatlon in Inlet Nonie Figure 2-4 Surry Unit 2 -Axial Boundary of the 1.0E+17 n/cm 2 Fluence Threshold in the +Z Direction (at 54 and 72 EFPY)
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Westinghouse Non-Proprietary Class 3 2-13 2.2.5 Recommendations Tables 2-2 and 2-3 report the maximum fast neutron (E > 1.0 MeV) fluence at specific pressure vessel materials in the extended beltline for Surry Units 1 & 2, respectively. The nozzle shell, nozzle shell to intermediate shell circumferential weld, intermediate shell to lower shell circumferential weld, and the lower shell to lower vessel head circumferential weld have a single set of neutron fluence values for each unit. These neutron fluence values would be appropriate for use for P-T limit analyses for these materials.
The intermediate shell, intermediate shell longitudinal weld, lower shell, and lower shell longitudinal weld report two sets of fluence values for each unit due to each of these locations having two separate plates or welds. However, the results for the two sets are equivalent for each location. Therefore, either set of neutron fluence values would be appropriate for use for P-T limit analyses for these materials.
Regarding the inlet and outlet nozzles, two separate fast neutron fluence values are given at two locations for each nozzle. One location represents the lowest extent of the nozzle forging to vessel shell welds, whereas the second location conservatively represents the 1/4T flaw in the nozzles. Although the fluence results at the nozzle forging to vessel shell welds are more limiting, the l/4T flaw location is more representative of the fluence for the nozzle at the peak stress location. Therefore, fluence at either location (lowest extent of the nozzle forging to vessel shell welds or at the l /4T flaw) can be used for the P-T limit analyses for the inlet and outlet nozzles for each unit.
A full three-dimensional discrete ordinates model provides a more geometrically detailed analysis. This more detailed representation can be utilized as a next step for further analysis of the maximum fast neutron fluence analyses for each unit if needed.
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Westinghouse Non-Proprietary Class 3 3-1 3 REFERENCES
- 1. Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.
- 2. Code of Federal Regulations, l O CFR Part 50, Appendix G, "Fracture Toughness Requirements," and Appendix H, "Reactor Vessel Materials Surveillance Program Requirements".
- 3. Westinghouse Calculation Note CN-REA-08-74, "Pressure Vessel Neutron Fluence Evaluation to Support the MUR for Surry Unit l," July 2009.
- 4. Westinghouse Calculation Note CN-REA-08-75, "Pressure Vessel Neutron Fluence Evaluation to Support the MUR for Surry Unit 2," July 2009.
- 5. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, March 2001.
- 6. RSICC Computer Code Collection CCC-650, "DOORS 3.2: One-, Two-, and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System," April 1998.
- 7. RSICC Data Library Collection DLC-185, "BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.
Proposed License Amendment Request - Clarification of Reactor Coolant System Heatup and Cooldown Limitations Technical Specifications Figures - Response to Request for Additional Information," February 4, 2015 (http://pbadupws.nrc.gov/docs/ML1504/ML15041A720.pdf).
- 9. Dominion Letter DRS-SPS-MUR-08-008, "Surry MUR Uprate Project - Dominion Resources Services Transmittal of Data for Fluence Analysis," Steve Pietryk, March 20, 2008.
- 10. Dominion Letter MEMO-NCD-20150012, "Dominion Purchase Order 70288223 and Transmittal of Core Power Information for Surry Power Station," May 22, 2015.
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