ML23181A154
| ML23181A154 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 02/28/2023 |
| From: | Constellation Energy Generation, TransWare Enterprises |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23181A149 | List: |
| References | |
| RS-23-079 LAS-FLU-001-R-002, Rev 1 | |
| Download: ML23181A154 (1) | |
Text
trans ware LAS-FLU-001-R-002 ENTER P RISES Revision 1 Page i of x
Topical Report r LASALLE COUNTY GENERATING
STATION UNIT 2 REACTOR PRESSURE VESSEL FLUENCE EVALUATION - END OF CYCLE 18
Document Number: LAS-FLU-001-R-002 Revision 1 February 2023
Prepared by: TransWare Enterprises Inc.
I Prepared for: Constellation Energy Corporation, LLC
LaSalle County Generating Station 2601 N 21st Rd Marseilles, IL 61341
Contract Number: 00808371
Project Manager: Ciara Miker
trans wa.re Controlled Copy Number: ____2____
ENTERPRISES
1565 Mediterranean Dr Sycamore, Illinois 60178-3141 815-895-4700, www.transware.net trans wa.re LAS-FLU-001-R-002 ENTER P RISES Revision 1 Page ii of x
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trans ware LAS -FLU -00 l-R-00 2 ENT(R P RISC S Revision I Page iii ofx
Topical Report
LASALLE COUNTY GENERATING STATION UNIT 2 REACTOR PRESSURE VESSEL FLUENCE EVALUATION - END OF CYCLE 18
Document Number: LAS-FLU-001-R - 002 Revision 1 February 2023
Prepared By: TransWare Enterprises Inc.
Project Team: M. E. Jewell, Project Enginee r H. J. Hepperrnann, Project Engineer E. A. Evans, Proj ect E ngi neer S. M. Wagstaff, Project Engineer Project Manager: K. E~re.:t Eng;"'"
D. B. Jor, Project Manager
Reviewed By: 2./,_ 3,/ '202-!;,
Date
~,~ I <..~<- !>
K. A.Jes, QA Spec ia li st Date
Approved By: 2.., 7 <-l,J < s;,,~
D B ?ii:l1/4::.i,n,g" Date
Prepared For: Constellation Energy Corporation, LLC LaSalle County Generating Station 2601 N 21st Rd Marseilles, IL 61341 Contract Number : 00808371
Project Manager: Ciara Miker
Tran s Ware Enterprises Inc.* 1565 Mediterranean Dr.
- Sycamo re, Illi nois 60 I 78 -3 14 1
+ 1-81 5-8 95 -4700
- www.transware.net trans wa.re LAS-FLU-001-R-002 ENTER P RISES Revision 1 Page iv of x
DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES This report was prepared by TransWare Enterprises Inc. (TransWare) as an account of work sponsored by Constellation Energy Corporation, LLC. This report contains information about the LaSalle County Generating Station Unit 2 (LaSalle Unit 2) that was approved for use by Constellation Energy Corporation, LLC. The information contained in this report is believed by TransWare to be an accurate and true representation of the facts known, obtained or provided to TransWare at the time this report was prepared. The use of this information by anyone other than Constellation Energy Corporation, LLC for any purpose other than that for which it is intended is not authorized. To the best of TransWares knowledge, this report does not contain any information that is known to be confidential or proprietary to reactor vendors, fuel supplier vendors, or other third-party vendors supplying materials and services to the LaSalle Unit 2 reactor. No person acting on behalf of TransWare Enterprises:
(a) makes any warranty or representation whatsoever, express or implied, (i) with respect to the use of any information, apparatus, method, process, or similar item disclosed in this report, including merchantability and fitness for a particular purpose, or (ii) that such use does not infringe on or interfere with privately owned rights, including any party's intellectual property, or (iii) that this report is suitable to any particular user's circumstance; or (b) assumes responsibility for any damages or other liability whatsoever, including any consequential damages, resulting from the selection or use of this report or any information, apparatus, method, process, or similar item disclosed in this report.
NOTICE OF CONFIDENTIALITY This report is intended for the sole use of Constellation Energy Corporation, LLC. If you have received this report in error, please notify TransWare immediately by email at Kathy.Jones@transware.net and proceed to its deletion. If you are not the intended recipient of this report, you are advised that the distribution, copying, or use of the information in this report for any purpose is prohibited by law.
QUALITY REQUIREMENTS This document has been prepared in accordance with the requirements of 10CFR50 Appendix B, 10CFR21, and TransWare Enterprises Inc.s 10CFR50 Appendix B quality assurance program.
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CONTENTS 1 Introduction...................................................................................................................... 1-1 1.1 Regulatory Requirements......................................................................................... 1-1 1.2 Limitations of the Fluence Evaluation....................................................................... 1-2 2 Summary of Results........................................................................................................ 2-1 3 Reactor Pressure Vessel Fast Neutron Fluence........................................................... 3-1 4 References....................................................................................................................... 4-1 4.1 References............................................................................................................... 4-1 4.2 Glossary................................................................................................................... 4-2
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LIST OF FIGURES Figure 3-1 LaSalle Unit 2 RPV Beltline Region at 54 EFPY.................................................... 3-2 Figure 3-2 Nozzle Fluence Edit Locations for N2 and N6 Nozzles.......................................... 3-3 Figure 3-3 Nozzle Fluence Edit Locations for N12 Nozzles.................................................... 3-3
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LIST OF TABLES Table 2-1 Maximum Fast Neutron Fluence for LaSalle Unit 2 RPV Beltline Welds, Nozzles, and Shell Plate Locations........................................................................ 2-2 Table 2-2 RPV Beltline Elevation Range for LaSalle Unit 2................................................... 2-2 Table 3-1 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Welds at EOC 18 (28.6 EFPY)............................................................................... 3-4 Table 3-2 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Welds at 32 EFPY.................................................................................................. 3-5 Table 3-3 Maximum Fast Neutron Fluence for LaSalle Unit 2 RPV Beltline Welds at 54 EFPY................................................................................................................. 3-5 Table 3-4 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Shell Plates............................................................................................................ 3-6 Table 3-5 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Nozzles at EOC 18 (28.6 EFPY)........................................................................... 3-6 Table 3-6 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Nozzles at 32 EFPY............................................................................................... 3-7 Table 3-7 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Nozzles at 54 EFPY............................................................................................... 3-8 Table 3-8 Reactor Beltline Elevation Range for LaSalle Unit 2.............................................. 3-8
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INTRODUCTION
This report presents the results of the reacto r pressure vessel (RPV) fast neutron fluence evaluation that was performed for the LaSalle County Generating Station Unit 2 (LaSalle Unit 2) reactor. The LaSalle Unit 2 reactor is owned and operated by Constellation Energy Corporation, LLC (Constellation). The information in this report was generated with the methods described in LaSalle County Generating Station Unit 2 Fluence Methodology Report [1].
In compliance with Regulatory Guide 1.190 [2], TransWare Enterprises Inc. (TransWare) has benchmarked the RAMA Fluence Methodology against industry standard benchmarks and plant-specific dosimetry measurements for boiling water reactors and pressurized water reactors. The results of the benchmarking show that the fluence methodology implemented by TransWare predicts specimen activities with no discernable bias in the computed fluence. The combined uncertainty for the LaSalle Unit 2 reactor pressure vessel is determined to be 8.6%. Based upon these results, there is no discernable bias in th e computed reactor pressure vessel fluence for the period of Cycle 1 through the end of Cycle 18 for the LaSalle Unit 2 reactor. The details of the uncertainty analysis can be found in Attachment 1 to the LaSalle County Generating Station Unit 2 Fluence Methodology Report [3].
The fluence provided in this report for the LaSalle Unit 2 reac tor pressure vessel and nozzles supersedes the fluence provided in the TransWare Report EXL-LSA-001-R-003 [4].
1.1 Regulatory Requirements Part 50 of Title 10 of the Code of Federal Regulations provides require ments for establishing irradiated material monitoring programs that serve to ensure the integrity of the reactor coolant pressure boundary of light water nuclear power reactors. Two appendices to Part 50 present the requirements that guide the fluence determina tions presented in this report: Appendix G, Fracture Toughness Requirements [5], and Appendix H, Reactor Vessel Material Surveillance Program Requirements [6].
Appendix G specifies fracture toughness requirements for the carbon and low-alloy ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary to ensure adequate margins of safety during any condition of normal operation, including anticipated conditions for system hydrostatic testing, to which the pressure boundary may be subjected over its service lifetime. These requir ements apply to base metal, welds, and weld heat-affected zones in the materials within the reactor pressure vessel beltline region.
Appendix H specifies the requirements for a material surveillance program that serves to monitor changes in the fracture toughness properties of the ferritic materials in the reactor pressure vessel beltline region. The changes in fracture toughness properties of ferritic materials are attributed to the exposure of the materials to neutron irradiation and the thermal environment.Section III of trans wa.re LAS-FLU-001-R-002 ENTER P RISES Revision 1 Page 1-2 of 1-4
Appendix H specifies that a material surveillance program is required for light water nuclear power reactors if the peak fast neutron fluence with energy greater than 1 MeV (E > 1.0 MeV) at the end of the design life of the vessel is expected to exceed 1.0E+17 n/cm 2.
In compliance with the Appendix H requirements, fracture toughness test data is obtained from material specimens that are exposed to neutron irradiation in surv eillance capsules installed at or near the inner surface of the reactor pressure ve ssel. These capsules are withdrawn periodically from the reactor for measurement and analysis. Fast neutron fluence is not a measurable quantity and must be determined using analytical methods. It must be demonstrated that the analytical method used to determine the fast neutron fluence provides a conservative prediction over the beltline region of the pressure boundary when compared to the measurement data with allowances for all uncertainties in the measurements.Section III of Appendix H also allows for an Integrated Surveillance Program (ISP) in which representative materials for the reactor are irradiated in one or more other reactors of sufficiently similar design and operating features to permit accurate comparisons of the predicted amount of radiation damage.
Implementing guidelines addressing the requirements of Appendices G and H are provided in U. S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2 (Reg Guide 1.99R2), Radiation Embrittlement of Reactor Vessel Materials [7], and Regulatory Guide 1.190 (Reg Guide 1.190), Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [2]. Reg Guide 1.99R2 addresses the requirements of Appendix G for determining the damage fluence that is used in the evaluation of fracture toughness in light water nuclear reactor pressure vessel ferritic materials. Reg Guide 1.190 addresses the requirements for determining the fast neutron fluence and uncertainty in the fluence predictions that are used in fracture toughness evaluations.
The fast neutron fluence evaluations described in this report meet the requirements of Appendices G and H of Part 50 of Title 10 of the Code of Federal Regulations and U. S. NRC Regulatory Guides 1.190 and 1.99 Revision 2.
1.2 Limitations of the Fluence Evaluation The fast neutron fluence presented in this report is based on hist orical and projected operating conditions of the reactor. The RPV fast neutron fluence that is based on historical operating conditions is determined to meet the requirements of Reg Guide 1.190 with no discernable bias in the results. It is determined, therefore, that the RPV fast neutron fluence presented in this report is suitable for use in evaluating material embrittlement conditions of reactor pressure vessel materials in accordance with Reg Guide 1.99R2. Use of the results for other purposes is not demonstrated.
Fluence projections are determined using a projection cycle of predicted fuel loading. This cycle is assumed to be an equilibrium cycle representative of how the reactor will operate until the end of the reactors operating license. Continued use of the projected fluence presented in this report must be demonstrated as applicable as new operating history data from the reactor becomes available.
It is cited in Regulatory Position 1.2 of Reg Guide 1.190 that a best-estimate power distribution may be used for reactor vessel neutron fluence calculations. The best-estimate fluence presented trans wa.re LAS-FLU-001-R-002 ENTER P RISES Revision 1 Page 1-3 of 1-4
in this report meets the requirements of Regulatory Position 1. 2. Regulatory Position 1.2 further states that best-estimate power distribution must be updated if changes in core loadings, surveillance measurements, or ot her information indicate a significant change in projected fluence values. Under this requirement, the other information that can necessitate an update of the fluence model can include: implementation of power uprates/derates, introduction of new fuel designs, changes in projected cycle lengths, changes in core loading and/or operational strategies, changes in reactor flow, or other changes that could alter the power/flux profiles used in the fluence projections and uncertainty analysis.
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SUMMARY
OF RESULTS
This section provides a summary of the fast neutron fluence determined for the LaSalle Unit 2 RPV. Details of the RPV fluence evaluation are presented in Section 3, Reactor Pressure Vessel Fast Neutron Fluence.
LaSalle Unit 2 is a BWR/5 class reactor with a core loading of 764 fuel assemblies. The fluence evaluation for this reactor is based on historical and projected operati ng data through Cycle 18 (28.6 EFPY). Fluence evaluations are also performed at 32 EFPY and 54 EFPY.
Table 2-1 presents a summary of the maximum fast neutron fluence de termined for the RPV shell plates, welds and nozzles at EOC 18 (28.6 EFPY), 32 EFPY, and 54 EFPY. The significant fluence for the evaluated areas of the RPV occur at the inside surface of the RPV base metal, which is denoted as the 0T depth in the pressure vessel wall. The N2 nozzles remain under the fluence threshold of 1.0E+17 n/cm2. All other evaluated areas of the RPV have exceeded the fluence threshold of 1.0E+17 n/cm2. It is shown in Table 2-1 that the maximum fluence is determined to occur at the 0T location of the Shell Ring 2 with a value of 1.07E+18 n/cm2 at 54 EFPY. Note in Table 2-1 that all fluence that has exceeded the fluence threshold of 1.0E+17 n/cm2 are shown in red font and that the maximum fluences in the RPV are additionally shown in bold font.
Table 2-2 shows the axial span of the RPV beltline region that was determined for LaSalle Unit 2 at EOC 18 (28.6 EFPY), 32 EFPY, and 54 EFPY. The reactor beltline region is defined in Appendices G [5] and H [6] of 10CFR50 to include those regions that directly surround the effective height of the reactor core, as well as those adjacent areas of the RPV that are predicted to experience sufficient neutron irradiation damage. This definition of the RPV beltline is considered to include all materials that exceed a fast neutron fluence of 1.0E+17 n/cm2. At 54 EFPY the RPV beltline covers 421.2 cm, or approximately 13.8 ft of the reactor vessel. The scope of the fluence model was developed to provide an evaluation of the reactor pressure vessel over the full height of the RPV extended beltline region.
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Table 2-1 Maximum Fast Neutron Fluence for LaSalle Unit 2 RPV Beltline Welds, Nozzles, and Shell Plate Locations
Component Maximum Fast Neutron Fluence (n/cm2)
EOC 18 (28.6 EFPY) 32 EFPY 54 EFPY
RPV Beltline Welds AB 5.52E+17 5.98E+17 9.11E+17 BA 4.00E+17 4.35E+17 6.46E+17 BB 3.72E+17 4.04E+17 6.27E+17 BC 4.82E+17 5.24E+17 8.08E+17 BD 3.72E+17 4.07E+17 6.44E+17 BE 5.90E+17 6.45E+17 1.01E+18 BF 4.71E+17 5.15E+17 7.92E+17 Nozzle Forging-to-Base-Metal Welds Nozzle Weld N2 1.35E+16 1.47E+16 2.27E+16 Nozzle Weld N6 2.43E+17 2.72E+17 4.56E+17 Nozzle Weld N12 1.68E+17 1.90E+17 3.27E+17 Shell Plates Shell Ring 1 5.52E+17 5.98E+17 9.11E+17 Shell Ring 2 6.31E+17 6.90E+17 1.07E+18
Table 2-2 RPV Beltline Elevation Range for LaSalle Unit 2 Reactor Lifetime Lower Elevation Upper Elevation Axial Span of the RPV
[in (cm)] [in (cm)] Beltline [in (cm)]
EOC 18 (28.6 EFPY) 214.95 (545.96) 369.29 (938.00) 154.34 (392.04) 32 EFPY 213.95 (543.44) 370.40 (940.81) 156.45 (397.37) 54 EFPY 209.51 (532.14) 375.32 (953.31) 165.81 (421.17)
Section 3 provides detailed results for the RPV fast neutron fluence evaluation. RPV damage fluence is reported at the 0T, 1/4T, and 3/4T depths of the RPV wall for each horizontal (circumferential) weld, vertical (axial) wel d, shell plate, and nozzle in the RPV beltline.
Figure 3-1 illustrates the location of the welds, shell plates, and nozzles in the RPV. Fluence damage through the thickness of the RPV wall is determined using the displacements-per-atom (dpa) attenuation method prescribed in Regulatory Guide 1.99 [7]. to the LaSalle County Generating Station Unit 2 Fluence Methodology Report
[3] presents the results for the calculated-to-measurement (C/M) activities determined for the surveillance capsule and flux wire dosimetry removed LaSalle Unit 2. The total average C/M for LaSalle Unit 2 is determined to be 0.93 with a standard deviation of 0.07.
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REACTOR PRESSURE VESSEL FAST NEUTRON FLUENCE
This section presents the predicted best-estimate fast neutron fluence (energy > 1.0 MeV) for the LaSalle Unit 2 reactor pressure vessel (RPV) at EOC 18 (28.6 EFPY), 32 EFPY, and 54 EFPY.
It is reported in Reference 3 that the calculat ed fluence does not require a bias adjustment; therefore, the calculated fluence is the best-estimate fluence for LaSalle Unit 2.
The reactor pressure vessel fast neutron fluence is determined at the interface of the RPV base metal and cladding, which is denoted as the 0T location of the RPV wall. Damage fluence through the RPV wall is reported at the 1/4T and 3/4T depths in the wall. These values are determined based on the nominal RPV wall thickness of 15.7 cm (6.1875 in).
The fast neutron fluence that is used in material embrittlement evaluations should be determined using an appropriate damage function (such as displacements-per-atom of iron) rather than the computed fast neutron fluence obtained from transport calculations. Two acceptable methods for estimating the damage fluence are prescribed in Regulatory Guide 1.99 [7]. One method is based on a generic fluence attenuation formulation and the other is a plant-specific fluence attenuation.
The generic fluence attenuation formulation is used when computational fluence methods do not support plant-specific damage functions. The equation for generic attenuation is given in Regulatory Guide 1.99 as:
where x is the depth in the RPV wall, given in inches, and fsurf is the fast neutron fluence at the vessel wetted, 0T surface.
Plant-specific material damage assessments may be used to obtain a more accurate estimate of the damage fluence throughout the RPV wall in accordance with the following equation:
where dpax is the damage expressed as displacements-per-atom of iron (dpa) at depth x in the RPV wall, and dpasurf is the damage at vessel 0T. In these evaluations, 0T represents the inner surface of the RPV base metal. It has been demonstrated that the generic attenuation approach can become increasingly non-conservative at increasing axial distances from the reactor core mid-plane [8] elevation; therefore, plant-specif ic damage assessments are recommended for use in RPV material embrittlement evaluations.
Plant-specific fast neutron damage fluence for LaSalle Unit 2 is presented in this report for the RPV horizontal (circumferential) welds, vertical weld s, shell plates, and nozzles that reside in the RPV beltline region. The location and naming convention of the shell plates, welds, and nozzles trans wa.re LAS-FLU-001-R-002 ENTER P RISES Revision 1 Page 3-2 of 3-8
are shown in Figure 3-1. Also shown in Figure 3-1 is the calculated RPV beltline region for the LaSalle Unit 2 reactor at 54 EFPY.
180° 210
- 240
- 270
- 300
- 330
- o* 30
- 60 ° go
- 120* 150* 180 °
1 1 11 1 I I I 1 1 1 I I I 1 1 1 I I I 11 1 I I 1 1 1 1 I I I 111 1 11 1 1 1 I I I 111 I I I 1 11 1 I I 1 1 1 I I 1 111 I I I 111 1 1
0 '-' 0 0 '-' '-' 0 375.32 in (95 3.31 cm) - - -
N 12C N6C N12D N12A N6A N6B N 12 B Shell Ring 2 u..o 00 LI.Jo CD M CD If) CD;::: N C')
RPV BeltlineRegion*
AB 277.25 in (704.22 cm)
0 0
(.)~ ~~ co ~
co r--- co~ CD r---a, r---
~ C')
Shell Ring 1 209.51 in (532. 14 cm)
) 0 0 0 0 0 0 0 0 0 0 0 C N2 F N2G N2 H N2J N2 K N1A N2A N2 B N2C N2 D N2 E N1B
V No tes: This drawing is not to scale. Inside View Dimensions are given in inc hes (cm).
- RPV beltline region is shown for 54 EF PY.
Figure 3-1 LaSalle Unit 2 RPV Beltline Region at 54 EFPY
Table 3-1 through Table 3-7 report the maximum fast neutron damage fluence that is determined for each of the RPV welds, shell plates, and nozzles residing in the RPV beltline. Damage fluence is reported at the 0T, 1/4T, and 3/4T de pths in the RPV wall for the reporting periods of interest. In all tables, the damage fluence that exceeds the threshold fluence of 1.0E+17 n/cm 2 is shown in red font. The maximum damage fl uence determined for the weld s, shells and nozzles is also shown in bold font. It is observed in the tables th at all evaluated components other than the N2 Nozzles will have exceeded the fluence threshold of 1.0E+17 n/cm 2 before 54 EFPY.
Table 3-1 through Table 3-3 report the maximum damage fluence that is determined for the RPV horizontal (circumferential) and vertical welds at EOC 18 (28.6 EFPY), 32 EFPY, and 54 EFPY, respectively. The maximum damage fluence for each weld is de termined to occur at the 0T depth, with the maximum flue nce occurring in horizontal weld BE with a value of 1.01E+18 n/cm2 at 54 EFPY.
Table 3-4 reports the maximum damage fluence that is determined for each RPV shell plate at EOC 18 (28.6 EFPY), 32 EFPY, and 54 EFPY. Th e maximum damage fluence for each shell plate is determined to occur at the 0T depth, with the maximum fluence occurring in Shell Ring 2 with a value of 1.07E+18 n/cm 2 at 54 EFPY.
Table 3-5 through Table 3-7 report the maximum damage fluence that is determined for the RPV nozzles at EOC 18 (28.6 EFPY), 32 EFPY, and 54 EFPY, respectively. Damage fluence for the nozzles is presented along two paths: one along the nozzle forgi ng-to-base-metal weld and the other along an extraction path that extends from the inside corner of the forging to the outside surface of the RPV wall. Both paths are illustra ted for the N2 and N6 nozzles in Figure 3-2 and trans wa.re LAS-FLU-001-R-002 ENTER P RISES Revision 1 Page 3-3 of 3-8
for the N12 nozzles in Figure 3-3. It is noted that the extraction path is evaluated around the full circumference of the nozzle forging to determ ine the maximum fluence. The maximum damage fluence for the beltline nozzles is determined to occur at the 0T depth in the metal weld of the N6 nozzle with a value of 4.56E+17 n/cm 2 at 54 EFPY.
oi
.c Cf)
.9 :,
u Nozzle T Forging-to-Base -metal Ex traction Path Weld
Figure 3-2 Nozzle Fluence Edit Locations for N2 and N6 Nozzles
Nozzle Forging-to-Base-metal Ex traction Path Weld T~
Figure 3-3 Nozzle Fluence Edit Locations for N12 Nozzles
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Table 3-8 reports the elevations that define the RPV beltline at EOC 18 (28.6 EFPY), 32 EFPY, and 54 EFPY. It is shown that the RPV beltline at 54 EFPY covers 421.2 cm, or approximately 13.8 ft of the reactor vessel.
Table 3-1 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Welds at EOC 18 (28.6 EFPY)
Weld Azimuth(1) Elevation [in (cm)](1) Fast Neutron Fluence (n/cm2) 0T 1/4T 3/4T
Horizontal Welds
AB 25° 277.2 (704.2) 5.52E+17 3.82E+17 1.60E+17
Shell Ring 1 Vertical Welds
BA 317.5° 277.3 (704.2) 4.00E+17 2.80E+17 1.22E+17 BB 77.5° 277.3 (704.2) 3.72E+17 2.60E+17 1.12E+17 BC 197.5° 277.3 (704.2) 4.82E+17 3.36E+17 1.42E+17
Shell Ring 2 Vertical Welds
BD 350° 311.9 (792.2) 3.72E+17 2.61E+17 1.14E+17 BE 110° 316.8 (804.7) 5.90E+17 4.11E+17 1.72E+17 BF 230° 317.8 (807.3) 4.71E+17 3.29E+17 1.41E+17
(1) Azimuth and elevation values are listed for the 0T location only.
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Table 3-2 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Welds at 32 EFPY Weld Azimuth(1) Elevation [in (cm)](1) Fast Neutron Fluence (n/cm2) 0T 1/4T 3/4T
Horizontal Welds
AB 25° 277.2 (704.2) 5.98E+17 4.14E+17 1.74E+17
Shell Ring 1 Vertical Welds BA 317.5° 277.3 (704.2) 4.35E+17 3.04E+17 1.32E+17 BB 77.5° 277.3 (704.2) 4.04E+17 2.82E+17 1.22E+17 BC 197.5° 277.3 (704.2) 5.24E+17 3.64E+17 1.54E+17
Shell Ring 2 Vertical Welds BD 350° 311.9 (792.2) 4.07E+17 2.86E+17 1.24E+17 BE 110° 316.8 (804.7) 6.45E+17 4.49E+17 1.88E+17 BF 230° 317.8 (807.3) 5.15E+17 3.60E+17 1.55E+17
(1) Azimuth and elevation values are listed for the 0T location only.
Table 3-3 Maximum Fast Neutron Fluence for LaSalle Unit 2 RPV Beltlin e Welds at 54 EFPY Weld Azimuth(1) Elevation [in (cm)](1) Fast Neutron Fluence (n/cm2) 0T 1/4T 3/4T Horizontal Welds AB 25° 277.2 (704.2) 9.11E+17 6.31E+17 2.65E+17
Shell Ring 1 Vertical Welds BA 317.5° 277.3 (704.2) 6.46E+17 4.52E+17 1.97E+17 BB 77.5° 277.3 (704.2) 6.27E+17 4.38E+17 1.89E+17 BC 197.5° 277.3 (704.2) 8.08E+17 5.63E+17 2.39E+17
Shell Ring 2 Vertical Welds BD 350° 311.9 (792.2) 6.44E+17 4.52E+17 1.97E+17 BE 110° 317.8 (807.3) 1.01E+18 7.02E+17 2.94E+17 BF 230° 317.8 (807.3) 7.92E+17 5.54E+17 2.38E+17
(1) Azimuth and elevation values are listed for the 0T location only.
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Table 3-4 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Shell Plates Shell Plate Azimuth(1) Elevation [in (cm)](1) Fast Neutron Fluence Damage (n/cm2) 0T 1/4T 3/4T
Shell Ring 1 25° 277.2 (704.2) 5.52E+17 3.82E+17 1.60E+17 Shell Ring 2 25° 317.8 (807.3) 6.31E+17 4.35E+17 1.80E+17
32 EFPY
Shell Ring 1 25° 277.2 (704.2) 5.98E+17 4.14E+17 1.74E+17 Shell Ring 2 25° 317.8 (807.3) 6.90E+17 4.76E+17 1.97E+17
54 EFPY Shell Ring 1 25° 277.2 (704.2) 9.11E+17 6.31E+17 2.65E+17 Shell Ring 2 25° 317.8 (807.3) 1.07E+18 7.41E+17 3.08E+17
(1) Azimuth and elevation values are listed for the 0T location only.
Table 3-5 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Nozzles at EOC 18 (28.6 EFPY)
Location Azimuth Elevation [in (cm)](1) Fast Neutron Fluence at (n/cm2) 0T 1/4T 3/4T
N2 Nozzles Weld 1.35E+16 1.17E+16 8.40E+15 Extraction Path 2.03E+15 2.28E+15 4.22E+15 30°
Weld 60° 181.0 (459.7) 1.33E+16 1.16E+16 8.25E+15 Extraction Path 2.01E+15 2.25E+15 4.15E+15 Weld 7.02E+15 6.32E+15 5.20E+15 Extraction Path 1.16E+15 1.42E+15 2.95E+15 90°
N6 Nozzle Weld 45° 372.5 (946.2) 2.43E+17 1.73E+17 7.83E+16
Extraction Path 7.79E+16 7.04E+16 5.49E+16
N12 Nozzles Weld 20° 366.0 (929.6) 1.68E+17 1.19E+17 5.33E+16 Extraction Path 1.44E+17 1.17E+17 6.95E+16
- 1) Elevation values correspond to each nozzle centerline elevation.
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Table 3-6 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Nozzles at 32 EFPY Location Azimuth Elevation [in (cm)](1) Fast Neutron Fluence at (n/cm2) 0T 1/4T 3/4T
N2 Nozzles
Weld 30° 1.47E+16 1.28E+16 9.13E+15 Extraction Path 2.21E+15 2.48E+15 4.60E+15 Weld 60° 181.0 (459.7) 1.45E+16 1.26E+16 8.97E+15 Extraction Path 2.19E+15 2.45E+15 4.52E+15
Weld 90° 7.73E+15 6.95E+15 5.70E+15 Extraction Path 1.28E+15 1.56E+15 3.24E+15
N6 Nozzle Weld 2.72E+17 1.94E+17 8.76E+16 Extraction Path 8.82E+16 7.96E+16 6.17E+16 45° 372.5 (946.2)
N12 Nozzles Weld 20° 366.0 (929.6) 1.90E+17 1.34E+17 6.00E+16 Extraction Path 1.63E+17 1.32E+17 7.81E+16
(1) Elevation values correspond to each nozzle centerline elevation.
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Table 3-7 Maximum Fast Neutron Damage Fluence for LaSalle Unit 2 RPV Beltline Nozzles at 54 EFPY Location Azimuth Elevation [in (cm)](1) Fast Neutron Fluence at (n/cm2) 0T 1/4T 3/4T N2 Nozzles Weld 30° 2.27E+16 1.98E+16 1.41E+16 Extraction Path 3.46E+15 3.87E+15 7.10E+15
Weld 60° 181.0 (459.7) 2.24E+16 1.94E+16 1.38E+16 Extraction Path 3.41E+15 3.81E+15 6.97E+15 Weld 90° 1.26E+16 1.13E+16 9.07E+15 Extraction Path 2.10E+15 2.54E+15 5.16E+15 N6 Nozzle Weld 45° 372.5 (946.2) 4.56E+17 3.24E+17 1.45E+17 Extraction Path 1.52E+17 1.37E+17 1.04E+17 N12 Nozzles Weld 20° 366.0 (929.6) 3.27E+17 2.31E+17 1.02E+17 Extraction Path 2.82E+17 2.28E+17 1.32E+17
(1) Elevation values correspond to each nozzle centerline elevation.
Table 3-8 Reactor Beltline Elevation Range for LaSalle Unit 2 Reactor Lifetime Lower Elevation Upper Elevation Axial Span of the RPV
[in (cm)] [in (cm)] Beltline [in (cm)]
EOC 18 (28.6 EFPY) 214.95 (545.96) 369.29 (938.00) 154.34 (392.04)
32 EFPY 213.95 (543.44) 370.40 (940.81) 156.45 (397.37)
54 EFPY 209.51 (532.14) 375.32 (953.31) 165.81 (421.17)
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REFERENCES
4.1 References
- 1. LaSalle County Generating Station Unit 2 Fluence Methodology Report, TransWare Enterprises Document. LAS-FLU-001-R-001, Revision 0: 2021
- 2. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.190: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Washington, D.C.: Office of Nuclear Regulatory Research: 2001
- 3. Qualification of the LaSalle County Generating Station Unit 2 Fluence Model - Cycles 1 to 18 TransWare Enterprises Document LAS-FLU-001-R-001 Attachment 1, Revision 0:
2021
- 4. LaSalle County Generating Station Unit 2 Reactor Pressure Vessel Fluence Evaluation at End of Cycle 15 With Projections to 32 and 54 EFPY, TransWare Enterprises Inc., EXL-LSA-001-R-003, Revision 1, July 2014.
- 5. U.S. National Archives and Records Administration. Code of Federal Regulations.
Title 10, Appendix G to Part 50 - Fracture Toughness Requirements: 2013
- 6. U.S. National Archives and Records Administration. Code of Federal Regulations.
Title 10, Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements: 2008
- 7. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.99: Radiation Embrittlement of Reactor Vessel Materials. Revision 2. Washington, D.C.: Office of Nuclear Regulatory Research: 1988
- 8. E. N. Jones, Comparison of Regulatory Guide 1.99 Fluence Attenuation Methods, Journal of ASTM International. Vol. 9, No. 4, pp. 1-7 (2012).
trans wa.re LAS-FLU-001-R-002 ENTER P RISES Revision 1 Page 4-2 of 4-3 4.2 Glossary
AZIMUTHAL QUADRANT SYMMETRY - A type of core and pressure vessel azimuthal representation that represents a single quadrant of the reactor that can be rotated and mirrored to represent the entire 360-degree geometry. For example, the northeast quadrant can be mirrored to represent the northwest and southeast quadrants and can be rotated to represent the southwest quadrant.
BEST-ESTIMATE NEUTRON FLUENCE - See Neutron Fluence.
BOC - An acronym for beginning-of-cycle.
CALCULATED NEUTRON FLUENCE - See Neutron Fluence.
CALCULATIONAL BIAS - A calculational adjustment based on comparisons of calculations to measurements. If a bias is determined to exist, it may be applied as a multiplicative correction to the calculated fluence to produce the best-estimate neutron fluence.
CORE BELTLINE - The axial elevations corresponding to the active fuel height of the reactor core.
DAMAGE FLUENCE - See Neutron Fluence.
DPA - An acronym for displacements per atom which is typically used to characterize material damage in ferritic steels due to neutron exposure.
EFFECTIVE FULL POWER YEARS (EFPY) - A unit of measurement representing one full year of operation at the reactors rated power level. For example, if a reactor operates for 12 months at full rated power, this represents 1.0 EFPY. If the reactor operates for 10 months at full rated power, then goes into a power uprate and continues operating for another 2 months at the new full rated power, this also represents 1.0 EFPY.
EOC - An acronym for end-of-cycle.
EXTENDED BELTLINE REGION - See RPV beltline.
FAST NEUTRON FLUENCE - Fluence accumulated by neutrons with energy greater than 1.0 MeV (E > 1.0 MeV).
NEUTRON FLUENCE - Time-integrated neutron flux reported in units of n/cm2. The term best-estimate fluence refers to the fast neutron fluence that is computed in accordance with the requirements of U.S. Nuclear Regulatory Commission Regulatory Guide 1.190. The term damage fluence, which is required for material embrittlement evaluations, refers to an adjusted fast neutron fluence that is determined using damage functions specified in U.S. Nuclear Regulatory Commission Regulatory Guide 1.99.
NOZZLE EXTRACTION PATH - The path, or trajectory through the nozzle blend radius along which fluence is determined for the nozzle.
NOZZLE FORGING-TO-BASE-METAL WELD - The weld between the nozzle forging and the RPV base metal materials. This is sometimes referred to as the Nozzle-to-Shell Weld OEM - An acronym for Original Equipment Manufacturer.
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RPV - An acronym for reactor pressure vessel. Unless otherwise noted, the reactor pressure vessel refers to the base metal material of the RPV wall (i.e., excluding clad/liner).
RPV BELTLINE - The RPV beltline is defined as that porti on of the RPV adjacent to the reactor core that attains sufficient neutron radiation damage that the integrity of the pressure vessel could be compromised. For purposes of this evaluation, the fa st neutron fluence threshold used to define the traditional RPV beltline is 1.0E+17 n/cm2. The axial span of the RPV that can exceed this threshold includes the RPV shells, welds, an d heat-affected zones. An extended beltline is also defined to include lower fluence regions of the pressure vessel but with higher stresses than the traditional beltline region, such as RPV nozzles. The combination of fluence and stress may result in a limiting location in the pressure vessel for determining pressure-temperature limits.
RPV ZERO ELEVATION - The RPV zero elevation is defined at the inside surface of the lowest point in the vessel bottom head, which is typically the bottom drain plug location. Axial elevations presented in this report are relative to RPV zero.
RVI - An acronym for reactor vessel internals.
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- E ",, ***, s, s on - rop n e ary LAS -FLU -00 t l -R -008 Attac hmen t 1, Revisi o n 1 Page i o f x
Topical Report
LASALLE COUNTY GENERATING STATION UNIT 2 FLUENCE METHODOLOGY REPORT-LICENSING REPORT
Attachment 1 Qualification of the LaSalle Unit 2 Reactor Fluence Model - Cycles 1 to 18
Document Number: LAS -F LU - 001 -R -008 Attachme n t 1, R evision 1 I Marc h 2023
Prepared b y : TransWare Enterprises Inc.
Prepared for: Constellation Energy Corporation, LLC LaSalle County Generating Station 2601 N 2 1st R d Ma rseilles, IL 61341
Contra c t Number: 0080837 1
Pro jec t Manager: Natalie McI n tosh
trans,
- Ja --
ENTERPRISES Controlled Copy Number: ______2
1565 Med iterranea n Dr Sycamo re, Illinois 60178 - 3141 815 - 895 - 4700
- www.t ransware. net trans wa.re Non-Proprietary LAS-FLU-001-R-008 ENTER P RISES Attachment 1, Revision 1 Page ii of x
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trans Wc..T '- N on -Pro prietary
[NTERPRISES LAS - FLU -00 l-R -00 8 Attach m ent I, Revision I Page ii i ofx
Topical Report
LASALLE COUNTY GENERATING S TAT ION UNIT 2 FLUENCE METHODOLOGY REPORT
Attachment 1 Qualification of the LaSa lle Unit 2 Reactor Flue n ce Model - Cycles 1 to 18
Document Number: LAS-FLU-001 - R-008 A ttachment 1, Revision 1 March 2023
Prepared By: TransWare Enterprises Inc.
Project Team: E. A. Eva n s, Proj ect Engi nee r H. J. H e pperm a nn, Project E ngineer M. E. J ewe ll, P roject E ng in eer S. M. W ags taff, Projec t E ng in eer K. E. Wa tk ins, Proj ect Eng in eer
Project Manager, LJ ~..,_ 3 /2/202 3 D. B. Jo ~ect Manage r Date
Reviewed By: 3/2 /2023 K. E. Watkin s, Projec t Engineer Date
3/2 /2023 K. A. fues, QA Sp ec ia li st Date Approved By: 3/2/2023 0. B.~~o n oge, Date
Prepared For: Constellation Energy Corporation, LLC LaSalle County Gene rating Sta tion 260 1 N 2 1st Rd Marseilles, IL 61 341 Contract Number: 008083 7 1
Project Manager : Natalie McInto sh
Tra ns W a re E nte rpri ses Inc.
- 1565 Medi terra nean Dr.
- Syca m o re, Illin o is 60 178 - 3 14 1
+ 1-81 5 - 895-4700
- www.tra nsware.ne t trans wa.re Non-Proprietary LAS-FLU-001-R-008 ENTER P RISES Attachment 1, Revision 1 Page iv of x
DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THE INFORMATION CONTAINED IN THIS REPORT IS BELIEVED BY TRANSWARE ENTERPRISES INC. TO BE AN ACCURATE AND TRUE REPRESENTATION OF THE FACTS KNOWN, OBTAINED OR PROVIDED TO TRANSWARE ENTERPRISES INC. AT THE TIME THIS REPORT WAS PREPARED. THE USE OF THIS INFORMATION BY ANYONE OTHER THAN THE CUSTOMER OR FOR ANY PURPOSE OTHER THAN THAT FOR WHICH IT IS INTENDED, IS NOT AUTHORIZED; AND WITH RESPECT TO ANY UNAUTHORIZED USE, TRANSWARE ENTERPRISES INC. MAKES NO REPRESENTATION OR WARRANTY AND ASSUMES NO LIABILITY AS TO THE COMPLETENESS, ACCURACY OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS DOCUMENT. IN NO EVENT SHALL TRANSWARE ENTERPRISES INC. BE LIABLE FOR ANY LOSS OF PROFIT OR ANY OTHER COMMERCIAL DAMAGE, INCLUDING BUT NOT LIMITED TO SPECIAL, CONSEQUENTIAL OR OTHER DAMAGES.
((
))
QUALITY REQUIREMENTS This document has been prepared in accordance with the requirements of 10CFR50 Appendix B, 10CFR21, and TransWare Enterprises Inc.s 10CFR50 Appendix B quality assurance program.
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CONTENTS
Title Page 1 Introduction...................................................................................................................... 1-1 2 Reactor Operating History.............................................................................................. 2-1 3 Reactor Statepoint Data.................................................................................................. 3-1 4 Surveillance Capsule Dosimetry Evaluation................................................................. 4-1 4.1 Summary of the Flux Wire Activation Analysis......................................................... 4-1 4.2 Comparison of Predicted Activation to Plant-specific Measurements...................... 4-3 4.2.1 Flux Wire Activation Analysis for the LaSalle Unit 2 30° Capsule............... 4-3 4.2.2 Cycle 6 Surveillance Capsule Activation Analysis...................................... 4-4 4.3 Reactor Pressure Vessel Lead Factors.................................................................... 4-5 5 Reactor Pressure Vessel Fluence Uncertainty Analysis............................................. 5-1 5.1 Comparison Uncertainty........................................................................................... 5-1 5.1.1 Operating Reactor Comparison Uncertainty............................................... 5-1 5.1.2 Benchmark Comparison Uncertainty.......................................................... 5-2 5.2 Analytic Uncertainty.................................................................................................. 5-2 5.3 Combined Uncertainty.............................................................................................. 5-3 6 References....................................................................................................................... 6-1 6.1 References............................................................................................................... 6-1 trans wa.re Non-Proprietary LAS-FLU-001-R-008 ENTER P RISES Attachment 1, Revision 1 Page vi of x
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LIST OF TABLES
Title Page Table 2-1 Summary of LaSalle Unit 2 Core Loading Inventory for Cycles 1 to 9.................... 2-1 Table 2-2 Summary of LaSalle Unit 2 Core Loading Inventory for Cycle 10 to 18.................. 2-2 Table 3-1 Statepoint Data for LaSalle Unit 2 per Cycle Basis................................................. 3-2 Table 4-1 Summary of the Fluence and Activity Comparisons for the LaSalle Unit 2 Dosimetry................................................................................................................ 4-3 Table 4-2 Summary of the Activity Comparisons for the Iron and Copper Flux Wires Removed From the LaSalle Unit 2 Reactor............................................................ 4-3 Table 4-3 Comparison of the Calculated-to-Measured Activities for the Iron Flux Wires Removed From the LaSalle Unit 2 30° Flux Wire Holder........................................ 4-4 Table 4-4 Comparison of the Calculated-to-Measured Activities for the Iron and Copper Flux Wires Removed From the LaSalle Unit 2 300° Surveillance Capsule at EOC 6..................................................................................................................... 4-4 Table 4-5 Best-Estimate Fluence Determined for the LaSalle Unit 2 Surveillance Capsules. 4-5 Table 4-6 Lead Factors Determined for the LaSalle Unit 2 Surveillance Capsules................. 4-5 Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements......... 5-2 Table 5-2 LaSalle Unit 2 RPV Combined Uncertainty for Energy > 1.0 MeV.......................... 5-3 trans wa.re Non-Proprietary LAS-FLU-001-R-008 ENTER P RISES Attachment 1, Revision 1 Page viii of x
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LIST OF FIGURES
Title Page Figure 4-1 Positioning of the Surveillance Capsules Installed in the LaSalle Unit 2 Reactor... 4-2 trans wa.re Non-Proprietary LAS-FLU-001-R-008 ENTER P RISES Attachment 1, Revision 1 Page x of x
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INTRODUCTION
This attachment provides the reactor operating history, comparisons to activation measurements, and uncertainty analysis that are essential for validating the LaSalle County Generating Station Unit 2 (LaSalle Unit 2) fluence methodology. The methodology that is used for determining the neutron fluence in the LaSalle Unit 2 reactor is detailed in the LaSalle County Generating Station Unit 2 Fluence Methodology Report [1].
The power history data presented in this report covers the time period from start of commercial operation (circa 1984) to the end of operating cycle 18. All surveillance dosimetry removed from the reactor over that time period, and which is available in the form of activation measurements, is evaluated. A combined uncertainty factor for the fluence model based on the modeling approach and measurement comparisons is determined which demonstrates that the computational fluence method used by TransWare Enterprises Inc. is qualified for use in determining neutron fluence for the LaSalle Unit 2 reactor pressure vessel in accordance with U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.190 [2].
In compliance with Regulatory Guide 1.190, it is shown in this report that the calculated-to-measured (C/M) ratio and standard deviation is 0.93 +/- 0.07 for all reactor dosimetry evaluated for the LaSalle Unit 2 reactor. The combined uncertainty for the LaSalle Unit 2 reactor is determined to be 8.57%. Based upon these results, there is no discernable bias in the computed reactor pressure vessel fluence for the period Cycle 1 through the end of Cycle 18 for the LaSalle Unit 2 reactor.
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REACTOR OPERATING HISTORY
R e a ctor op erat ing h i s t o1y i s th e m eas ure o f daily re act o r pow er l ev el s t h at ch arac terize th e rad i ati on exp osur e hi sto1y of a r eac t o r ove r its operat ing life. The daily powe r h i s t o1y dat a fo r th e La S all e U nit 2 re acto r w as p rov i de d by Co nste llati on in discr ete fo nn fo r Cycles 1 thro ugh 18.
An other impo r tan t e l em en t o f th e r eac t o r ope rat in g histo1y is the fue l d es igns t h at wer e loaded in th e r eac t o r fo r e a ch ope rat in g cy cle. E ach fu el des ign h as a d iffer en t powe r s ign ature in th e co re and, t h erefo r e, r esults in d ifferen t s p at i al powe r, exp os ure, an d fue l isot op i c distributi o n s th rou gh ou t the cor e r eg io n.
T ab le 2 -1 an d Table 2-2 prov i de a s ummai y o f the fuel des i gns th at we re l oaded t h e La S all e U nit 2 re acto r fo r eac h op erat ing cycle. T ab le 2 -1 l ists the fuel des igns th at wer e loaded in th e r eac t o r fo r Cycles 1 thro u gh 9. T ab le 2 -2 l ist s the fuel des i gns th at we re l oaded fo r Cycles 10 th rou gh 18. T h e do mina n t fue l des ign t h at was l oade d in t h e cor e fo r eac h cycle i s s h own in b old fo n t. T h e d o min ant fue l d es i gn t h at was l o a ded o n th e co re p eriphe1y is iden t ifi ed in blue fo n t.
T a bl e 2 -1 Summ ary o f L a Sa lle Uni t 2 C o re L oa ding Invent ory for Cy cles 1 to 9 Fuel De s ign s Cyc le 8x8 9x 9 GE5 GE7B GE8 B GE9 ATRIUM-9 1 764 2 540 224
3 304 220 240 4 88 224 240 212 5 88 240 436 6 96 668 7 764 8 508 256
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Table 2-2 Summary of LaSalle Unit 2 Core Loading Inventory for Cycle 10 to 18 Fuel Designs Cycle 9x9 10x10 ATRIUM-9 ATRIUM-10 GE14 GNF2 GNF3 10 382 382 11 96 382 286 12 478 286 13 625(1) 139 14 76411) 15A 460(1) 304 15B 460(1) 304 16 192 568 4(2) 17 760 4(2) 18 760 4(2) 19prj 488 276 20+(3 ) 764
- 1) In cycles 13, 14, and 15, t here are a few ATRIUM-10 MX Lead Test Assemb ly (LTA) bundles. These ar e trea ted as ATRIUM - 10 bundles.
- 2) In cycle 17, 17, and 18, there are 4 GNF3 LUA bund l es. These are treated the same as l ater GN F3 bundles.
- 3) For purposes of fl u ence projections, operation beyon d projec tion Cycle 19 uses an equilibrium p rojec tion cycle featuring a full core loading ofGNF3 (Cycle 20+). This cycle was provided by Cons tellation to pred i ct fluence at the end of the extended p lant license period.
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REACTOR STATEPOINT DATA
The reactor operating history is defined in discrete exposure steps referred to as statepoints. A statepoint is defined as a snapshot in time th at characterizes an equilibrium power-flow condition of a reactor core at a moment in time. The importance of statepoints is the affect that changes in reactor core power and core flow have on the granular radial and axial power distributions in the reactor core region. In particular, the changes in the granular power distributions have a proportional effect on the neutron flux that is calculated in the structural components adjacent to the core region, including the reactor pressure vessel. As power and flow do not vary in a proportional manner, several statepoints are needed to accurately characterize the operating states of the reactor core and the integral power that affects a fluence calculation.
Several statepoints are generally used to represent the different operating states of a reactor core over the course of an operating cycle. The core power distribution for a statepoint is generally determined using core simulator software. Because core simulator codes are used for a variety of core analysis functions, 10s to 100s of core calculations may be performed to track and monitor the operation of a reactor. Not all core calculations are suitable for use in fluence evaluations.
Therefore, each cycle of operating data is investigated to select the statepoints that are suitable for use in fluence calculations. When all reactor conditions are considered, the number of core simulator statepoints selected for a reactor fluence evaluation can vary from cycle to cycle.
Core simulator data was provi ded by Constellation to characterize the historical operating conditions for the LaSalle Unit 2 reactor for Cycles 1 through 18. Table 3-1 shows that a total of
(( )) statepoints were used to represent the operating states of the LaSalle Unit 2 reactor for the first 18 cycles of operating history. It is also shown in Table 3-1 that the number of statepoints used per cycle in the fluence calculation varied. ((
))
Table 3-1 also shows the rated thermal power of the reactor for a cycle and the accumulated effective full power years (EFPY) of exposure accumulated for that cycle.
A separate neutronics transport calculation is performed for each statepoint listed in Table 3-1.
The neutron fluxes calculated for each statepoint are then combined with daily thermal power information to provide an integral accounting of the neutron fluence for the reactor pressure vessel, reactor vessel internals, and surveillanc e capsules. The periods of reactor shutdown are also accounted for in this process, particularly to allow for an accurate calculation of irradiated surveillance capsule activities.
In addition to the 18 cycles of operating history data, two projection cycles, Cycle 19 and Cycle 20+, are also shown in Table 3-1. Cycle 19 was provided by Constellation as a transition cycle to Cycle 20+. Cycle 20+ was provided by Constellation and represents an equilibrium cycle that will be used to project fast neutron fluence to the end of 60 years of operation.
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Table 3-1 Statepoint Data for LaSalle Unit 2 per Cycle Basis
Cycle Number Number of Reactor Rated Thermal Power Statepoints (MWt) Accumulated EFPY
1 II 3323 1.36 2 3323 2.47 3 3323 3.36 4 3323 4.76 5 3323 5.96 6 3323 6.97 7 3323 8.06 8(1) 3323 9.60
9A 3489 11.37 9B 3489 11.56 10 3489 13.47 11 3489 15.40 12 3489 17.23 13 3489 19.2 1 14<2> 3546 21.12
15 3546 22.92 16 3546 24.76 17 3546 26.66 18 3546 28.59 19<3> 3546 30.42 20+<4 ) 11 3546 54.0
( 1) A power uprat e was implemen ted midway through cycle 8 from 3323 MW th to 3489 MW th.
(2) A power upra te was implemented at the beginning of cycle 14 from 3489 MW th to 3546 MW th.
(3) Cycle 19 consists of p rojection data for tran sitioning to a full GNF3 core loading.
( 4) Cycle 20+ is a projec tion cycle using a full core of GNF3 fuel. This cycle will be used to proj ect reacto r fluence at the end of the exten ded license period.
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SURVEILLANCE CAPSULE DOSIMETRY EVALUATION
This section presents the results of the activation analysis and the determination of fast neutron fluence for the LaSalle Unit 2 surveillance capsule dosimetry. Lead factors associating the peak RPV fluence with the capsule fluence are also reported. The results presented in this section form the basis for the validation and qualification of the fluence methodology as applied to the LaSalle Unit 2 reactor in accordance with the Regulatory Guide 1.190 [1].
Regulatory Guide 1.190 requires that fluence calculational methods to be validated by comparisons with activation measurements from operating reactor dosimetry. It is preferred that the activation data be taken from the reactor be ing evaluated. However, comparative data from plants of similar design may be used in cases where insufficient dosimetry measurements exist.
In the case for the LaSalle Unit 2 reactor, th ere is sufficient plant-specific measurements available to qualify the calculational method without reference to other plants of similar design.
In order to report computed fluence as the best-estimate fluence, Regulatory Guide 1.190 requires that the standard deviation resulting from the comparison of calculated to measurement data should be 20%. It is determined that overall calculated-to-measured (C/M) comparison ratio and standard deviation for the LaSalle Unit 2 reactor is 0.93 +/- 0.07. Therefore, the computational fluence model for the LaSalle Unit 2 reactor meets the Regulatory Guide 1.190 criteria and, as such, no bias adjustment is required to be applied to the computed RPV fluence.
4.1 Summary of the Flux Wire Activation Analysis The LaSalle Unit 2 reactor hosts three (3) surveillance capsules that are mounted near the reactor core mid-plane elevation. The capsules are positioned near the inner surface of the reactor pressure vessel wall at the 30, 120, and 300 azimuths around the circum ference of the reactor pressure vessel. Figure 4-1 illustrates the positioning of the surveillance capsules in the LaSalle Unit 2 reactor.
The 300° surveillance capsule was removed from the LaSalle Unit 2 reactor for testing of radiation effects on the Charpy specimens and activation analysis of the flux wires that were contained in the capsules. In addition, one (1) set of flux wires were extracted from the flux wire holder attached to the 30 surveillance capsule for activation analysis.
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Reactor North Site West o*
330 ° 30 ° Surveillance Capsule
Jet Pump Assembly
300 °
+ + Shroud
+ + Core Reflector
+ + + +
+ + + + +
+ + + + + +
+ + + + + i{;f{~;l{W,W,;l{W, 270° + + + + -tf 4-'¥¥¥¥¥¥ 90
- Reactor East Site North
+ + + + + + *'+ ++ + ++ +
+ + + + + + + + + Reactor Pressure
+ ++ t+ + + + + Vessel and Clad
+ + + ++ + + ++
+ + + + + Vessel Insula tion
+ + and Clad
240°
Cavity Reg ions
Biological Shield and Clad 210
- 150°
Notes: This drawing is not to scale. 180 ° F = Fuel bundle loc ations.
(Locations shown only for the northeast quadrant.)
+ = Control rod locations.
Figure 4-1 Positioning of the Surveillance Capsules Installed in the LaSalle Unit 2 Reactor Table 4-1 provides a summary of th e activation comparisons for each set of flux wire that were irradiated in LaSalle Unit 2 reactor. The table shows that the overall calculated-to-measured (C/M) ratio and associated standa rd deviation for 9 irradiated sp ecimens was determined to be 0.93 +/- 0.07.
Table 4-1 also provides the irradiation period in terms of cycle exposure, the accumulated exposure in terms of Effective Full Power Years (EFPY) of reactor operation, the average fast neutron fluence (E > 1.0 MeV) for each set of fl ux wires, and the number of specimens (viz., flux wires) evaluated for each surveillance capsule and set of flux wires removed from the LaSalle Unit 2 reactor.
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Table 4-1 Summary of the Fluence and Activity Comparisons for the LaSalle Unit 2 Dosimetry
Dosimeter Exposure Fluence Measured Deviation Cycles of Accumulated Fast Neutron Number of Calculated vs. Standard Exposure (EFPY) ( > 1 MeV, n / cm 2 ) Spe c imens (C /M) (a)
30° Flux W ire 1 - 1 1.36 2.42 E+16 6 0.92 0.05
300° Capsu le 1 - 6 6.97 1.29 E+17 9 0.94 0.08
Overall Average C/ M and Standard Deviation 15 0.93 0.07
Table 4 -2 p rovides t he overa ll calc u lated-to-meas ured ratios and standard dev i ati o ns fo r t he copp er, iron, an d nickel fl ux wires t h at wer e inadiated in th e L aSalle U ni t 2 reactor.
Table 4-2 Summary of the Activity Comparisons for the Iron and Copper Flux Wires Removed From the LaSalle Unit 2 Reactor
Flux Wire Number of Specimens C/ M (7
Copper 6 0.85 0.02 Iron 6 0.97 0.01 Nickel 3 1.00 0 Overall Average C/ M 15 0.93 0.07 and Standard Deviation
4.2 Comparison of Predicted Activation to Plant-specific Measurements The com p ar iso n of calcu l ated activatio ns to meas ureme n ts for t h e dos im etry are prese n te d in t his s u bsectio n. Flu ence and l ead factors fo r each caps u le ar e reported in Su bsectio n 4.3, Reactor Pressure Vessel Lead Factors.
4.2.1 Flux Wire Activation Analysis for the LaSalle Unit 2 30 ° Capsule Table 4 -3 shows t h e act ivat i on measurements, per iods of in adi ati on, comp u ted activatio ns, and th e comp u ted-to -meas ured (C/M) rat i os for th e flux wires inadiated in t h e 30° flux wire h older.
Activatio n calcu l ati on s were p er fo1med fo r t h e follow ing r eact i ons: 54Fe (n,p) 54Mn an d 63Cu (n, a,) 6°Co.
Th e average C/M ratio and associated standa r d deviation dete1min ed for all of the flux w ires inadiated in t h e LaSa ll e Unit 2 30° fl u x wire h older i s 0.92 +/- 0.05.
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Table 4-3 Comparison of the Calculated-to-Measured Activities for the Iron Flux Wires Removed From the LaSalle Unit 2 30° Flux Wire Holder
Flux Wire Dosimeter Measured Calculated C/ M (dps / g) (3) (dps /g) C1
Iron Fe-1 3.06E+ 04 2.97 E+04 0.97 -
Fe-2 3.11 E+ 04 2.97 E+04 0.96 -
Fe-3 3.07E+ 04 2.97 E+04 0.97 -
Iron Average 0.97 0.01 Copper Cu - 1 1.89E+ 03 1.65 E+03 0.87 -
Cu -2 1.89E+ 03 1.65 E+03 0.87 -
Cu -3 1.90E+ 03 1.65 E+03 0.87 -
Copper Average 0.87 0.00 Average C/ M and Standard Deviation 0.92 0.05
4.2.2 Cycle 6 Surveillance Capsule Activation Analysis Iron, ni cke l, an d coppe r fl u x wir es wer e inadiated in t h e L aSa ll e U nit 2 300° survei ll ance caps u les during th e firs t six (6) cycles ofreacto r op eratio n. At t he tim e of th eir rem ova l, t h e fl ux wires had been inadiated for a total of 6.97 EFPY.
Activatio n calcu l ati on s were p er fo1med fo r t h e follow ing r eacti ons: 54Fe (n,p) 54Mn, 58Ni (n,p) 58Co, and 63 Cu (n,a) 6°Co. The preci se locat i on of th e indivi dua l w ires w i thin th e caps u le i s not known; t h erefo r e, t he activatio n calc ul atio ns wer e perfo1m ed at t h e cen ter of t h e caps u le con tainer.
Table 4 -4 p rovide com par i so ns of t he calculated - to -m easured speci fi c activit i es fo r each iron,
ni ck el, an d coppe r fl u x wir e rem oved from t h e LaSa ll e U nit 2 reactor at en d o f Cy cle 6. The average of t h e calc u lated-to - meas ur ed ratio fo r each fl u x wir e i s rep o1ie d in the tab l es below. It is noted that only one measurement is provided for each flux wire. Therefore, a comp u ted standard deviation for each flux wire is not deten nined.
Table 4-4 Comparison of the Calculated-to-Measured Activities for the Iron and Copper Flux Wires Removed From the LaSalle Unit 2 300° Surveillance Capsule at EOC 6
Flux Wire Measured Calculated C/M (dps /g) (4) (dps /g) C1
Iron 5.10 E+04 5.00 E+04 0.98 -
Nickel 8.29 E+05 8.31 E+05 1.00 -
Copper 6.68 E+03 5.52 E+03 0.83 -
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4.3 Reactor Pressure Vessel Lead Factors T able 4 -5 an d Tab l e 4-6 prov i de the best -estim ate fas t neu tron fl uen ce detennined fo r th e reacto r press ur e vesse l an d dosi m eters and th e associated lead fac tors for eac h evaluate d tim e p erio d for th e L aSalle U ni t 2 r eactor.
Table 4-5 Best-Estimate Fluence Determined for the LaSalle Unit 2 Surveillance Capsules Dosimeter Fluence Peak RPV Fluence Evaluated Time (n / cm 2 ) (n / cm 2 Period 30 ° 300 ° OT 1/4T
EOC6 -- 1.29 E+17 1.44 E+17 9.95 E+ 16 32 EFPY 6.63 E+17 -- 6.90 E+17 4. 76 E+ 17 54 EFPY 1.00 E+18 -- 1.07 E+18 7.41 E+ 17
Table 4-6 Lead Factors Determined for the LaSalle Unit 2 Surveillance Capsules
Lead Factor Evaluated Time 30° 300 ° Period OT 1/4T OT 1/4T
EOC6 -- - 0.89 1.29 32 EFPY 0.96 1.39 - --
54 EFPY 0.94 1.35 - --
- 1) The le ad fac t or i s defined as the ratio of the fa st neu tron fluence a t the center of the surv eillance cap sule t o the peak fast neutron fluence at the base metal inner swface (01) of the RPV. A s econd lead fac tor is also pro v ided assuming the peak damage fluence at the l/4T depth of the RPV wall.
Notab le is t he fact t hat n o lead facto rs ar e prov ided fo r t he 120° cap sul e. This caps u le was broke n fro m its attachme n t h ardware an d was transfe 1Ted to th e sp ent fuel pool with ou t meas ur eme n t.
The calcu l ated -to -meas ured activation com parisons for t he survei ll ance capsu l es presented in t he previo us sect i on s sh ow n o discemable bi as in th e com putatio n al flu en ce m eth od. T h erefo r e, the best -estimate fl u ence r eported fo r each capsu l e in T able 4-5 is t h e fast n eutro n fl u ence com pute d by th e flue n ce methodo l ogy.
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REACTOR PRESSURE VESSEL FLUENCE UNCERTAINTY ANALYSIS
This section presents the combined uncertainty analysis and the determination of bias for the LaSalle Unit 2 reactor pressure vessel (RPV) fluence evaluation. The combined uncertainty is comprised of two components. One component is the uncertainty factors developed from plant-specific measurements and the other is an analytic uncertainty factor. When combined, these components provide a basis for determining the combined uncertainty (1) and bias in the computed RPV fluence.
The requirements for determining the combined uncertainty and bias for light water reactor pressure vessel fluence evaluations are provided in Regulatory Guide 1.190 [2]. The approach for determining combined uncertainty and bias for reactor pressure vessel fluence is demonstrated in Reference 5.
For pressure vessel fluence evaluations, two uncertainty factors are considered: comparison factors and uncertainty introduced by the measuremen t process. After analysis of these factors, it is determined that the combined uncertainty for LaSalle Unit 2 RPV fluence is 8.57%, and that no adjustment for bias is required for the RPV fast neutron fluence determined for the period Cycle 1 through the end of Cycle 18.
5.1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activity measurements. For pressure vessel fluence evaluations, two comparison uncertainty factors are considered: operating reactor comparison factors and benchmark comparison factors.
5.1.1 Operating Reactor Comparison Uncertainty TransWare has evaluated activation measurem ents for several BWR plants ranging from BWR/2-class plants to BWR/6-class plants. Each class of BWRs can have one or several variations of reactor core configurations, each having different radial diameters for the core shroud, reactor pressure vessel, and biological shield components. In addition, each can have different placements of the jet pumps and surveillance capsules in the reactor vessels.
The LaSalle Unit 2 reactor is a BWR/5 class design. ((
)) The overall comparison ratio for all BWR class plants evaluated as of the date of this evaluation is 1.01 +/- 0.10. ((
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5.1.2 Benchmark Comparison Uncertainty Th e benchmark compar i so n unce rtainty i s base d on a set of in dustiy stan dard simulatio n ben ch mark compa r ison s. In accordance with t he gu ide lin es provided in Regu l ato1y G uide 1.190, i t is appropr i ate to incl u de compar isons of vessel sim u latio n benc hm ark meas ur eme n ts in th e overa ll fl u ence u n ce1iainty eval uation. Two vessel sim u latio n benc hm arks are eval uated : th e Pool Cri tical Assembly (PCA) an d VENUS-3 expe r ime n tal ben chma rks.
Th e PCA expe rime n tal benc hm ark include s (( )) act ivat i on mea surement s at the mid -p l an e elevation in va r ious s imu l ated r eacto r compo n en ts. The VENUS -3 exper imental ben ch mark include s (( )) act ivat i on mea surement s at a ran ge of el evatio ns in var i ou s sim u lated reacto r compo n ents. Table 5 - 1 summar i zes the calculated - to -mea surement (C/M) result s detenn ined for th ese vessel sim u latio n benc hm arks.
Table 5-1 Summary of Comparisons to Vessel Simulation Ben c hmark Measurements
Ben c hmark Measurements Cal c ulated-to-St. Dev. (1a) Number of Average Measured (C/ M)
Po o l Critic al Assembl y ((
VENUS-3 To ta l Simul a ted Vessel ))
Comparisons
5.2 Analytic Uncertainty Th e calcu l ati on al models used for flue n ce analyses are compr i sed of n umero u s analytical parameters that h ave associated u n ce1iainties in the ir values. Th e u n ce1iainty in th ese parameter s needs to be tested for i ts contrib u tion to t h e overa ll fluence u n ce1iainty.
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5.3 Combined Uncertainty Th e comb ined unce 1tainty for th e r eactor pre ssur e v e ss el flue n ce ev aluation i s deten nin ed with a we i gh ting function ((
)) Table 5-2 sh ows th at th e c ombined unce1tainty ( lcr) determ ined for th e L a S alle U ni t 2 r eactor pre ss ur e v e ssel fl uen ce i s 8.57 % for neu tro n ener gy exceed ing 1. 0 MeV.
It is show n in T able 5 -2 t h at t h e c ombined unce1tainty is w ell bel ow t h e 20% unce 1tainty limi t specified in Regu l ato1y G uide 1.190. In ac c or dan ce w i th Regu l ato1y G uide 1.190, t h ere is n o d i scem ab l e bia s in t h e c om p u ted RPV flue n ce. Th erefo r e, n o adju stme n t t o t h e RPV fast neu tro n flu en ce fo r t h e per i o d cone spon ding to Cycle 1 th rou gh th e en d o f Cy cle 18 is r equired.
Table 5-2 LaSalle Unit 2 RPV Combined Uncertaint y for Energy > 1.0 MeV
Uncertainty Term Value
Combined Uncerta inty (1cr) 8.57%
B ias None !1>
- 1) The bias tenni s l ess than its con stitue nt uncertainty v alues, c oncl udin g that no s ta tis t ic a lly s igni fic ant b ias exists.
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REFERENCES
6.1 References
- 1. LaSalle County Generating Station Unit 2 Fluence Methodology Report, TransWare Enterprises Inc. Document Number LAS-FLU-001-R-001, Revision 0: 2021
- 2. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.190: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Washington, D.C.: Office of Nuclear Regulatory Research: 2001
- 3. Flux Wire Dosimetry Evaluation for LaSalle Nuclear Power Station, Unit 2, General Electric Company, SASR 87-59, DRF A00-02764, Revision 0, October 1987.
- 4. LaSalle Unit 2 RPV Surveillance Materials Testing and Analysis, General Electric Company, GE-NE-B1301786-01, Revision 2, April 2010.