RS-25-085, Subsequent License Renewal Application Affected Changes
| ML25118A280 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/28/2025 |
| From: | Constellation Energy Generation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML25118A278 | List: |
| References | |
| RS-25-085 | |
| Download: ML25118A280 (1) | |
Text
April 28, 2025 Enclosure A Page 1 of 30 Enclosure A Dresden Nuclear Power Station, Units 2 and 3, Subsequent License Renewal Application Affected Changes Introduction This enclosure contains five changes that are being made to the Subsequent License Renewal Application (SLRA) that were identified after its initial submission. For each item, a detailed description of the change is provided along with the affected page number(s) and specific portion(s) of the SLRA.
For clarity, entire sentences or paragraphs from the SLRA are provided with the deleted text highlighted by strikethroughs and the inserted text highlighted by bolded italics. In addition, any revisions to the SLRA tables are shown by providing excerpts from the affected tables.
April 28, 2025 Enclosure A Page 2 of 30 Table of Contents Change # 1 - Change to Isolation Condenser Fatigue TLAA..................................................... 3 Change # 2 - Removal of Core Spray Piping Leakages Assessment TLAA............................... 8 Change # 3 - Crack Growth Calculation Revision.....................................................................19 Change # 4 - AMP clarification for insulation jacketing.............................................................23 Change # 5 - AMR Updates for Primary Containment..............................................................27
April 28, 2025 Enclosure A Page 3 of 30 Change # 1 - Change to Isolation Condenser Fatigue TLAA Affected SLRA sections: Table 4.1-2, Section 4.3.7, and Section A.4.3.9.
Affected SLRA Page Numbers: 4.1-8, 4.3-28, and A-62.
Description of Change:
SLRA Section 4.3.7 and Appendix A.4.3.9 are revised to change the disposition of this TLAA from 10 CFR 54.21(c)(1)(i) to 10 CFR 54.21(c)(1)(iii).
Accordingly, SLRA Table 4.1-2, Section 4.3.7, and Section A.4.3.9 are revised as shown below.
April 28, 2025 Enclosure A Page 4 of 30 SLRA Table 4.1-2, Summary of Results - DNPS Time-Limited Aging Analyses, page 4.1-8 is revised as shown below:
Table 4.1-2 Summary of Results - DNPS Time-Limited Aging Analyses TLAA DESCRIPTION Disposition SLRA SECTION METAL FATIGUE ANALYSES 4.3 Dresden Transient Cycle and Cumulative Usage 80-Year Projections 4.3.1 ASME Section III, Class 1 Fatigue Analyses
§54.21(c)(1)(iii) 4.3.2 Environmental Fatigue Analyses for RPV and Class 1 Piping
§54.21(c)(1)(iii) 4.3.3 ASME Section III, Class 1 Fatigue Waivers
§54.21(c)(1)(ii) 4.3.4 ASME Section III, Class 2 & 3, and ANSI B31.1 Allowable Stress Analysis and Associated HELB Analyses
§54.21(c)(1)(i) 4.3.5 Reactor Vessel Internals 4.3.6 Core Shroud Support Fatigue Analysis
§54.21(c)(1)(iii) 4.3.6.1 Unit 2 Jet Pump Riser Repair/Mitigation Clamps Thermal Fatigue Analysis
§54.21(c)(1)(ii) 4.3.6.2 Fatigue Analysis of the Isolation Condensers
§54.21(c)(1)(i)
§54.21(c)(1)(iii) 4.3.7
April 28, 2025 Enclosure A Page 5 of 30 SLRA Section 4.3.7 Fatigue Analysis of the Isolation Condensers, page 4.3-28 is revised as shown below:
4.3.7 Fatigue Analysis of the Isolation Condensers TLAA
Description:
The DNPS isolation condensers provide core cooling when the RPV becomes isolated from the turbine and the main condenser. Fatigue evaluation of the isolation condensers was performed as part of original component design. The original fatigue evaluation analyses demonstrated that the 40-year cumulative usage factors (CUFs) for 13 subcomponents the critical components of the isolation condensers are less than the ASME Code Section III allowable value of 1.0.
The fatigue evaluations for the 13 subcomponents assumed either: 1) 240 Isolation Condenser actuations, 2) 240 Isolation Condenser actuations plus an additional 760 plant Heatup/Cooldown transient occurrences; or 3) 1000 occurrences which bound Isolation Condenser actuations or plant Heatup/Cooldown transients.
The specification for the isolation condensers and the supporting system piping and components requires analysis for 250 shutdown depressurization (i.e. isolation condenser operation) occurrences in 40 years.
The fatigue analysis of the DNPS isolation condensers is a TLAA because it is part of the current licensing basis, is used to support a safety determination, and is based on the cycles predicted for a 40-year plant life.
TLAA Evaluation The fatigue evaluations of the thirteen subcomponents were reviewed and the six most bounding subcomponents were added to the SI:FatigueProTM software.
The SI:FatigueProTM calculation of usage for these six components includes adjustments for EPU operating conditions and environmentally assisted fatigue, if applicable. These are items number 5 through 10 on SLRA Table 4.3.1-3. The 80-year CUF or CUFen projections for these six components remain below 0.4.
The projected cycle count for 80 years is well below the number of assumed in the fatigue analysis. To ensure the projected CUF and CUFen values in Table 4.3.1-3 remain acceptable for the 80-year period of operation for Units 2 and 3, the Fatigue Monitoring program (B.3.1.1) will continue to monitor actual transient cycles and the associated CUF and CUFen values for all limiting locations and ensure corrective actions are taken, if necessary, prior to exceeding the ASME Section III acceptance criterion.
April 28, 2025 Enclosure A Page 6 of 30 TLAA Disposition: 10 CFR 54.21(c)(1)(iii) - The effects of fatigue aging on the intended function(s) of the Isolation Condensers will be managed by the Fatigue Monitoring Aging Management program (B.3.1.1) through the SPEO.
The isolation condenser fatigue analysis found that the limiting component (tube-to-tubesheet junction) only permitted a design life of 280 combined pressure, thermal, and thermal shock cycles. The limiting thermal shock event is caused by an isolation event that raises the fluid on the secondary side of the isolation condenser to boiling at atmospheric pressure, or 212F, followed by injection of makeup water as cold as 70F into the secondary side. The code analysis shows that each isolation condenser operating cycle includes a thermal shock event.
Based on the number of isolation condenser actuations through June 30, 2022 (26 for Unit 2, 22 for Unit 3), the projected 80-year cycle counts are 34 for Unit 2 and 30 for Unit 3. This projected cycle count is less than the 250 isolation condenser actuation design life, as well as the 280 actuations in the limiting fatigue analysis, therefore, the design analysis remains valid for the SPEO TLAA Disposition: 10 CFR 54.21(c)(1)(i) - The Fatigue Analysis of the isolation condensers has been demonstrated to remain valid through the SPEO.
April 28, 2025 Enclosure A Page 7 of 30 SLRA Section A.4.3.9 Fatigue Analysis of the Isolation Condensers, page A-62 is revised as shown below:
A.4.3.9 Fatigue Analysis of the Isolation Condensers The DNPS Isolation Condensers provide core cooling when the reactor pressure vessel becomes isolated from the turbine and the main condenser. A fatigue evaluation of the 13 isolation condensers subcomponents was performed as part of the original component design, demonstrating that the 40-year cumulative usage factors (CUFs) for the critical components of the Isolation Condensers are below the ASME Code Section III allowable value of 1.0. Since the fatigue analysis is based upon a number of transient cycles postulated to occur in 40 years of service, it has been identified as a TLAA that requires evaluation for the SPEO. The specification for the Isolation Condensers and the supporting system piping and components requires design and analysis for 250 shutdown depressurization (i.e., Isolation Condenser operation) transients in 40 years. Based on the number of Isolation Condenser actuations to date, the projected total cycle count for 80 years is well below the number of design cycles. The fatigue analysis of the Isolation Condensers has been demonstrated to remain valid through the SPEO in accordance with 10 CFR 54.21(c)(1)(i).
Based on the number of Isolation Condenser actuations to date, the projected total cycle count of all critical subcomponents for 80 years is well below the number of assumed in the fatigue analysis. In addition, the six most bounding subcomponents were added to the SI:FatigueProTM software. The Fatigue Monitoring program will continue to use SI:FatigueProTM to compute actual CUF and CUFen values for these bounding locations and ensure corrective action is taken prior to the CUF or CUFen values exceeding the acceptance criterion of 1.0, in accordance with 10 CFR 54.21(c)(1)(iii).
April 28, 2025 Enclosure A Page 8 of 30 Change # 2 - Removal of Core Spray Piping Leakages Assessment TLAA Affected SLRA sections: Table of Contents, Section 3.1.2.3, Table 4.1-1, Table 4.1-2, Section 4.7.5, Section A.4.0, Section A.1.4, and Section A.4.7.5.
Affected SLRA Page Numbers: xii, 3.1-30, 4.1-6, 4.1-9, 4.7-1, 4.7-8, 4.7-9, A-4, A-9, and A-69.
Description of Change:
SLRA Section 4.7.5, Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment, and Appendix A Section A.4.7.5, Dresden Nuclear Power Station Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment, are revised to remove the leakage assessment as a TLAA.
In June of 2023, GEH issued revision 0 of a proprietary report which forms the basis for the leakage assessment of the Core Spray Replacement Piping described in SLRA Section 4.7.5.
The Dresden SLRA was submitted in April 2024. In March 2025, GEH issued revision 1 of this proprietary report. The updated March 2025 report shows that the original leakage assessment did not consider a corrosion aging effect of the mating surfaces of the mechanical fittings.
Instead, the leakage assessment was based on conservatively assumed clearances (gaps) between the entire mating surfaces of the mechanical fittings. The postulated gaps were based on the worst-case localized machining tolerances of each mating surface. High points based on the worst-case tolerances for each surface were then assumed be in contact, thus creating complete separation (a gap) between the two mating surfaces. This methodology results in the conservative assumption that the assumed gap is open along the entire diameter of the fitting.
This assumption bounds the worst case expected condition over the life of the new configuration.
The mating surfaces of these fittings were designed with refined surface finishes and are held together with bolting with specified torque values. Mating surfaces in complete contact, fabricated from the same materials, are not expected to experience general corrosion. Any potential volumes between the mating surfaces created by surface imperfections would be extremely small. Potential corrosion in these small volumes would be insignificant and would result in the formation of small tenacious oxide layers with densities less than the densities of the base material. The less dense oxides would fill the small volumes and the remaining mating surfaces, which are in contact, would not separate. Additional corrosion would require the removal of the tenacious oxides from the small volumes to expose new bare metal. This would require significant disruptions, such as high flow rates, steam impingement, cavitation, or mechanical perturbations. No such mechanisms are present in this application. Therefore, it was reasonable not to consider corrosion effects on the mating surfaces in the design of these mechanical fittings.
Based on this new information in the updated GEH analysis, it is concluded that the leakage assessment described in SLRA Section 4.7.5 and Appendix A Section A.4.7.5 does not meet the criteria of 10 CFR 54.3(2), and is therefore, not a TLAA.
Accordingly, the SLRA Table of Contents, Section 3.1.2.3, Table 4.1-1, Table 4.1-2, Section 4.7, Section 4.7.5, Section A-4.0, Section A.1.4, and Section A.4.7.5 are revised as shown below.
April 28, 2025 Enclosure A Page 9 of 30 SLRA Table of Contents on page xii is revised as shown below:
4.7 Other Plant-Specific Time-Limited Aging Analyses......................................... 4.7-1 4.7.1 Reactor Building Overheard Crane Load Cycles................................................... 4.7-2 4.7.2 Crack Growth Calculation of a Postulated Flaw in The Heat Affected Zone of an Arc Strike in The Torus Shell............................................................................ 4.7-4 4.7.3 Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam........ 4.7-5 4.7.4 Generic Letter 81-11 Crack Growth Analysis to Demonstrate Conformance to The Intent of NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking............................................................................... 4.7-6 4.7.5 Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment.......... 4.7-8 4.7.6 Unit 2 Reactor Pressure Vessel Closure Flange Flaw......................................... 4.7-10 4.7.7 Isolation Condenser Weld Flaw TLAA................................................................. 4.7-11 4.7.8 Protective Coatings............................................................................................ 4.7-12
April 28, 2025 Enclosure A Page 10 of 30 The third bullet from the bottom of the list contained in SLRA Section 3.1.2.3 Time-Limiting Aging Analysis, page 3.1-30 is revised as shown below:
Section 4.7.5, Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment
April 28, 2025 Enclosure A Page 11 of 30 SLRA Table 4.1-1 Review of Generic TLAAs Listed in NUREG-2192, Tables 4.1-2 and 4.7-1 page 4.1-6 is revised as shown below:
Table 4.1-1 Review of Generic TLAAs Listed in NUREG-2192, Tables 4.1-2 and 4.7-1 NUREG-2192 Examples of Potential Generic and Plant-Specific TLAAs Applies for DNPS?
SLRA Section NUREG-2192, Table 4.7 Examples of Potential Plant-Specific TLAA Topics (BWRs)
Re-flood Thermal Shock of the Reactor Pressure Vessel Yes 4.2.7 Re-flood Thermal Shock of the Core Shroud and Other Reactor Vessel Internals Yes 4.2.8 Loss of Preload for Core Plate Rim Hold-Down Bolts No N/A Erosion of the Main Steam Line Flow Restrictors No N/A Susceptibility to Irradiation-Assisted Stress Corrosion Cracking No N/A Fatigue of Cranes (Crane Cycle Limits)
Yes 4.7.1 Fatigue of the Spent Fuel Pool Liner No N/A Corrosion Allowance Calculations Yes No 4.7.5 N/A Flaw Growth due to Stress Corrosion Cracking No N/A Predicted Lower Limit No N/A
April 28, 2025 Enclosure A Page 12 of 30 SLRA Table 4.1-2 Summary of Results - DNPS Time-Limited Aging Analyses, page 4.1-9 is revised as shown below:
Table 4.1-2 Summary of Results - DNPS Time-Limited Aging Analyses TLAA DESCRIPTION Disposition SLRA SECTION OTHER PLANT-SPECIFIC ANALYSES 4.7 Reactor Building Overhead Crane Load Cycles
§54.21(c)(1)(i) 4.7.1 Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Torus Shell
§54.21(c)(1)(i) 4.7.2 Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam
§54.21(c)(1)(i) 4.7.3 Generic Letter 81-11 Crack Growth Analysis to Demonstrate Conformance to the Intent of NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking
§54.21(c)(1)(iii) 4.7.4 Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment
§54.21(c)(1)(i) 4.7.5 Unit 2 Reactor Pressure Vessel Closure Flange Flaw
§54.21(c)(1)(i) 4.7.6 Isolation Condenser Weld Flaw
§54.21(c)(1)(i) 4.7.7 Protective Coatings
§54.21(c)(1)(i) 4.7.8
April 28, 2025 Enclosure A Page 13 of 30 SLRA Section 4.7 Other Plant-Specific Time-Limiting Aging Analyses, page 4.7-1 is revised as shown below:
4.7 OTHER PLANT-SPECIFIC TIME-LIMITED AGING ANALYSES This section evaluates:
Reactor Building Overhead Crane Load Cycles (Section 4.7.1)
Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Torus Shell (Section 4.7.2)
Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam (Section 4.7.3)
Generic Letter 81-11 Crack Growth Analysis to Demonstrate Conformance to the Intent of NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (Section 4.7.4)
Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment (Section 4.7.5)
Unit 2 Reactor Pressure Vessel Closure Flange Flaw (Section 4.7.6)
Isolation Condenser Weld Flaw (Section 4.7.7)
Protective Coatings (Section 4.7.8)
April 28, 2025 Enclosure A Page 14 of 30 SLRA Section 4.7.5 Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment, page 4.7-8 and 4.7-9 is revised as shown below:
4.7.5 UNIT 2 CORE SPRAY REPLACEMENT PIPING FATIGUE AND LEAKAGE ASSESSMENT TLAA
Description:
In November 2009, all four lower sectionals on the Core Spray System downcomers were replaced, removing all known piping flaws in the DNPS Unit 2 RPV. DNPS Unit 3 had performed a similar piping replacement in 2004, however the DNPS Unit 3 repair is not included in this TLAA because it was analyzed for a 60-year service life, which extends beyond the 80-year SPEO.
Although the lower sectional replacement removed all the known flaws in the DNPS Unit 2 Core Spray piping, the design of the repair introduced a number of small openings at mechanical connections in the core spray piping and in its attachment to the shroud wall, thus providing leakage paths for the core spray coolant. Therefore, a leakage assessment was performed to demonstrate that the piping would perform its intended function, assuming the maximum corrosion tolerance based on the 40-year design life.
In addition, a A fatigue analysis of the new piping system and associated bolting concluded a maximum design CUF value of 0.0785.
Since the leakage assessment and fatigue analysis assumed a 40-year service life, until November 2049, they have it has been identified as a TLAA that requires evaluation for the SPEO, which ends in December 2049 for DNPS Unit 2.
TLAA Evaluation:
Leakage Assessment The leakage assessment was re-evaluated by GEH for an additional five (5) years, for a total of 45-years, until 2054. The assessment concluded that an additional five (5) years of corrosion to the austenitic stainless steel, alloy 718, and alloy X-750 materials that make up the leakage paths results in no changes to the calculated leakage.
The re-evaluation concluded that the original leakage assessment remains valid for an additional five (5) years, for a total of 45-years, until 2054.
Fatigue Evaluation Fatigue evaluations were performed in the hardware stress report. The highest cumulative usage factor (CUF) reported in the stress report was 0.0785 for the pipe flange. The stress report also identified fatigue usages of 0.0021 for the flange bolts, 0.021 for the bolts, and 0.0266 for the lower anchors.
April 28, 2025 Enclosure A Page 15 of 30 The associated fatigue analysis was re-evaluated by GEH for a 45-year service life, until 2054. The re-evaluation concluded that an additional five (5) years will not significantly increase the CUFs, and there is no effect on the stress report conclusions because of increased fatigue usage.
Therefore, both the leakage assessment and the fatigue analysis is valid through the SPEO.
TLAA Disposition: 10 CFR 54.21(c)(1)(i) - The DNPS Unit 2 replacement core spray piping leakage assessment and the fatigue analysis remain valid through the SPEO.
April 28, 2025 Enclosure A Page 16 of 30 SLRA Appendix A, Table of contents for Section A.4.0 Time-Limited Aging Analyses, page A-4 is revised as shown below:
A.4.7 Other Plant Specific Time-Limited Aging Analyses.............................................. A-66 A.4.7.1 Dresden Nuclear Power Station Reactor Building Overhead Crane Load Cycles................................................................................................................. A-66 A.4.7.2 Crack Growth Calculation Of A Postulated Flaw In The Heat Affected Zone Of An Arc Strike In The Suppression Chamber Shell.......................................... A-67 A.4.7.3 Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam...... A-67 A.4.7.4 Generic Letter 81-11 Crack Growth Analysis to Demonstrate Conformance to the Intent of NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking............................................................................. A-68 A.4.7.5 Dresden Nuclear Power Station Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment...................................................................... A-69 A.4.7.6 Dresden Nuclear Power Station Unit 2 Reactor Vessel Closure Flange Flaw TLAA.................................................................................................................. A-69 A.4.7.7 Dresden Nuclear Power Station Isolation Condenser Weld Flaw TLAA.............. A-70 A.4.7.8 Dresden Nuclear Power Station Protective Coatings.......................................... A-70
April 28, 2025 Enclosure A Page 17 of 30 SLRA Appendix A, Section A.1.4, Time Limited Aging Analyses, page A-9 is revised as shown below:
- 35. Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam (Section A.4.7.3)
- 36. Generic Letter 81-11 Crack Growth Analysis to Demonstrate Conformance to the Intent of NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (Section A.4.7.4)
- 37. Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment (Section A.4.7.5)
- 38. Unit 2 Reactor Pressure Vessel Closure Flange Flaw (Section A.4.7.6)
April 28, 2025 Enclosure A Page 18 of 30 SLRA Appendix A, Section A.4.7.5, Dresden Nuclear Power Station Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment, page A-69 is revised as shown below:
A.4.7.5 Dresden Nuclear Power Station Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment In November 2009, all four lower sections on the Core Spray System downcomers were replaced in the Unit 2 reactor vessel. Unit 3 had performed a similar piping replacement in 2004, however it is not included in this TLAA because it was analyzed for a 60-year service life which extends beyond the 80-year subsequent period of extended operation.
Although the lower section replacement removed all of the known flaws in the Unit 2 Core Spray piping, the design of the repair introduced a number of small openings at mechanical connections in the Core Spray Line and its attachment to the shroud wall, thus providing leakage paths for the Core Spray coolant. Therefore, a leakage assessment was performed to demonstrate that the piping would perform its intended function, assuming the maximum corrosion tolerance based on the 40-year design life.
In addition, a A fatigue analysis of the new piping system and its associated bolting concluded a maximum design CUF value of 0.0785. Since the leakage assessment and fatigue analysis assumed a 40-year service life, until November 2049, they have it has been identified as a TLAAs that requires evaluation for the subsequent period of extended operation, which ends in December 2049 for Unit 2.
The leakage and fatigue assessments was were re-evaluated by GEH for an additional five years, for a total of 45 years, until 2054. The assessment concluded that an additional five years of corrosion to the austenitic stainless steel, alloy 718, and alloy X-750 materials that make up the leakage paths results in no changes to the calculated leakage. The GEH re-evaluation concluded that an additional five years will not significantly increase the CUFs, and there is no effect on the existing stress report conclusions because of increased fatigue usage. Therefore, both the Unit 2 replacement core spray piping leakage assessment and the fatigue analysis remains valid through the subsequent period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).
April 28, 2025 Enclosure A Page 19 of 30 Change # 3 - Crack Growth Calculation Revision Affected SLRA sections: Section 4.7.2 and Appendix A Section A.4.7.2.
Affected SLRA Page Numbers: 4.7-4 and A-67.
Description of Change:
SLRA Section 4.7.2 and Appendix A Section A.4.7.2 are being revised to address all normal operating transients that result in alternating stresses to the Torus shell in accordance with the existing arc strike crack growth evaluation. The existing arc strike crack growth evaluation determined that the identified flaw would remain acceptable following 850 stress cycles assuming conservative loading conditions as compared to normal operation transients. The DNPS Plant Specific Unique Analysis Report (PUAR) identifies the normal operation transients that result in alternating stress to the Torus shell as: SRV actuations, plant startup pressure transients, and plant-startup temperature transients. The updated evaluation in SLRA Section 4.7.2 and Appendix A.4.7.2, provided below, compares the allowable number of stress cycles from the crack growth evaluation to the number of occurrences of the applicable normal operation transients projected to occur from 1991, when the evaluation was performed, through the end of the Unit 3 subsequent period of extended operation in 2051.
Accordingly, SLRA Section 4.7.2 and Appendix A Section A.4.7.2 are revised as shown below.
April 28, 2025 Enclosure A Page 20 of 30 SLRA Section 4.7.2 Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Torus Shell, page 4.7-4 is revised as shown below:
4.7.2 CRACK GROWTH CALCULATION OF A POSTULATED FLAW IN THE HEAT AFFECTED ZONE OF AN ARC STRIKE IN THE TORUS SHELL TLAA
Description:
In 1991, DNPS Unit 3 evaluated the effect of an arc strike flaw on the Torus shell wall.
The flaw was evaluated for crack growth, assuming 850 SRV actuations. over 40-years and found to be acceptable The flaw was evaluated for crack growth in accordance with ASME Section XI 1986, Appendix A. The evaluation concluded that the flaw remains acceptable following 850 normal operation stress cycles.
The DNPS Plant Unique Analysis Report (PUAR), Volume 2 (Reference 4.8.4) identifies the three normal operation transients that result in alternating stress to the Torus shell as: SRV actuations, plant startup temperature transients, and plant startup pressure transients. A further evaluation was performed in 1997, and it was determined that the flaw depth of the arc strike was not of sufficient depth to warrant permanent repairs. The area of the arc strike, though not permanently repaired, is subject to the ASME Section XI, Subsection IWE Aging Management Program, (B.2.1.29) which will continue to ensure the acceptability of the condition.
Since the crack growth evaluations the number of SRV actuations provide the basis for crack growth acceptability for the arc strike flaw on the Torus wall through 850 normal operation stress cycles, these evaluations are considered TLAAs that must be evaluated for the SPEO.
TLAA Evaluation:
SLRA Table 4.3.1-2 documents that the number of SRV actuations through the end of the SPEO, based on the total SRV actuations experienced as of 2022, is projected to be 75 SRV actuations for the Unit 3 SRVs. This value is well below the 850 actuations assumed in the original flaw analysis.
The 1991 and 1997 crack growth evaluations were performed to demonstrate that the arc strike flaw was acceptable for continued service.
The DNPS PUAR identified three normal operation transients that result in alternating stress of varying magnitude on the Torus wall surrounding the flaw.
The applicable transients in the DNPS PUAR are SRV actuations, pressure transients during a plant startup in which Torus pressure increases from ambient to the operating pressure, and temperature transients during a plant startup in which Torus temperature increases from ambient to the operating temperature.
April 28, 2025 Enclosure A Page 21 of 30 SLRA Table 4.3.1-2 item 3 documents that Dresden Unit 3 is projected to experience 262 plant startups in 80 years. A review of the Fatigue Monitoring program shows that Unit 3 experienced 142 plant startups as of 1991. Therefore, the total number of startups projected to occur from 1991 to 2051 is 120. As discussed above, two stress cycles are conservatively assumed per startup resulting in 240 projected stress cycles due to startup pressure and temperature transients from 1991 through the end of the SPEO.
SLRA Table 4.3.1-2 items 35 through 39 documents that Dresden Unit 3 is projected to experience 375 SRV actuations in 80 years. Operating experience indicates that the majority of SRV actuations occurred early in the life of the unit. However, for the purposes of this evaluation, the entire 80-year projection of 375 SRV actuations is conservatively assumed to occur in the 60 years between 1991 and 2051. As such, the total number of normal operation stress cycles is conservatively projected as 615, which is significantly less than the 850 stress cycles that are allowable in accordance with the crack growth evaluation. Therefore, the crack growth evaluation remains valid through the subsequent period of extended operation TLAA Disposition: 10 CFR 54.21(c)(1)(i) - The crack growth calculation of a postulated flaw in the heat affected zone of an arc strike in the torus shell has been demonstrated to remain valid through this SPEO.
April 28, 2025 Enclosure A Page 22 of 30 SLRA Appendix A, Section A.4.7.2 Crack Growth Calculation of a Postulated Flaw in The Heat Affected Zone of an Arc Strike in The Torus Shell, page A-67 is revised as shown below:
A.4.7.2 Crack Growth Calculation Of A Postulated Flaw In The Heat Affected Zone Of An Arc Strike In The Torus Suppression Chamber Shell An existing arc strike flaw in the Dresden Station Unit 3 Torus suppression chamber wall was evaluated in 1991 and again in 1997 and was found to be acceptable. It was determined that the flaw depth of the arc strike was not of sufficient depth to warrant permanent repairs. The flaw evaluation for crack growth was based on the concluded that the flaw would remain acceptable following a total of 850 stress cycles.
Normal operation transients resulting in alternating stress cycles for the Torus shell include: SRV actuations, and plant startup temperature and pressure transients. over 40-years, which was assumed to be 850 SRV actuations. Since the number of SRV actuations provides the basis for crack growth acceptability for the arc strike flaw on the suppression chamber wall, this evaluation is considered a TLAA.Since the crack growth evaluations provide the basis for acceptability for the arc strike flaw on the Torus wall through 850 stress cycles experienced during normal operation transients, this evaluation is considered a TLAA. The total number of applicable stress cycles projected to occur during normal operation from 1991 to 2051 is 615. The number of SRV actuations through the end of the subsequent period of extended operation is projected to be 74 occurrences for each Unit 3 SRV, based on the total SRV actuations experienced as of 2022. This value is well below the 850 SRV actuations allowable stress cycles assumed in the original flaw analysis and therefore this TLAA evaluation remains valid through the subsequent period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).
The area of the arc strike is subject to the Containment ISI program, which will continue to ensure the acceptability of the condition.
April 28, 2025 Enclosure A Page 23 of 30 Change # 4 - AMP clarification for insulation jacketing Affected SLRA sections: Section 3.5.2.2.2.4, Table 3.5.1 Affected SLRA Page Number: 3.5-52, 3.5-102 Description of Change:
SLRA Section 3.5.2.2.2.4 is being updated to credit the One-Time Inspection program for management of cracking and loss of material for aluminum and stainless steel insulation jacketing exposed to air - outdoor environments. Additionally, Table 3.5.1-100 is being revised to remove reference to the External Surfaces Monitoring of Mechanical Components.
Accordingly, SLRA Section 3.5.2.2.2.4 and Table 3.5.1 are revised as shown below.
April 28, 2025 Enclosure A Page 24 of 30 The last paragraph in SLRA Section 3.5.2.2.2.4, Cracking Due to Stress Corrosion Cracking, and Loss of Material Due to Pitting and Crevice Corrosion, page 3.5-52 is revised as shown below:
The One-Time Inspection (B.2.1.20)External Surfaces Monitoring of Mechanical Components (B.2.1.23) program will be used to manage cracking and loss of material for aluminum and stainless steel insulation jacketing on piping exposed to air - outdoor environments.
April 28, 2025 Enclosure A Page 25 of 30 SLRA Table 3.5.1, Summary of Aging Management Evaluations for the Structures and Component Supports, page 3.5-102 is revised as shown below:
Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Item Number Component Aging Effect/Mechanism Aging Management Programs Further Evaluation Recommended Discussion 3.5.1-100 Aluminum, stainless steel support members; welds; bolted connections; support anchorage to building structure Loss of material due to pitting and crevice corrosion, cracking due to SCC AMP XI.M32, "One-Time Inspection," AMP XI.S6, "Structures Monitoring," or AMP XI.M36, "External Surfaces Monitoring of Mechanical Components" Yes Consistent with NUREG-2191. The External Surfaces Monitoring of Mechanical Components (B.2.1.23), One-Time Inspection (B.2.1.20), and Structures Monitoring (B.2.1.33) programs will be used to manage cracking and loss of material of aluminum and stainless steel in concrete elements (anchors), containment closure bolting, new fuel storage racks, hatches/plugs, thermal insulation jacketing (including clamps, bands, and fasteners), penetration sleeves, refueling bellows, miscellaneous structural applications (including catwalks, grating, handrails, kick plates, ladders, platforms, stairs, steel curbs, decking, sump covers, vents, permanent scaffolding, roof scuttles, fixed louvers, flood plates, siding, and closure plates), and supports associated with emergency diesel generator, HVAC system components, other miscellaneous mechanical equipment, platforms, pipe whip restraints, jet impingement shields, masonry walls, other miscellaneous structures, raceways, cable trays, conduit, HVAC ducts, tube track, instrument tubing, non-ASME piping and components, racks, panels, cabinets, and enclosures for electrical equipment and instrumentation (including support members, welds, bolted connections, and support anchorage to building structure) exposed to
April 28, 2025 Enclosure A Page 26 of 30 Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Item Number Component Aging Effect/Mechanism Aging Management Programs Further Evaluation Recommended Discussion air - indoor uncontrolled and air - outdoor environments.
See Subsection 3.5.2.2.2.4.
The aluminum flood plates (in the Isolation Condenser Pump House) are constructed of 6061-T6 aluminum alloy, which is not susceptible to stress corrosion cracking.
April 28, 2025 Enclosure A Page 27 of 30 Change # 5 - AMR Updates for Primary Containment Affected SLRA sections: 3.5.1, 3.5.2-9 Affected SLRA Page Numbers: 3.5-100, 3.5-159, 3.5-164 Description of Change:
Reduction of strength; loss of mechanical properties is being added to SLRA Table 3.5.2-9 component type Concrete: Interior (accessible areas - drywell floor and reactor pedestal) and Concrete: Interior (inaccessible areas - drywell floor and reactor pedestal). The discussion in Table 3.5.1 Item Number 3.5.1-097 is being updated to reflect this revision.
The aging effect of loss of fracture toughness for the ring girder assemblies is being added to SLRA Table 3.5.2-9 component type Steel Components: ring girder assemblies (under reactor vessel skirt).
Accordingly, SLRA sections 3.5.1 and 3.5.2-9 are revised as shown below.
April 28, 2025 Enclosure A Page 28 of 30 SLRA Table 3.5.1, Summary of Aging Management Evaluations for the Structures and Component Supports, page 3.5-100 is revised as shown below:
Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Item Number Component Aging Effect/Mechanism Aging Management Programs Further Evaluation Recommended Discussion 3.5.1-097 Group 4: Concrete (reactor cavity area proximate to the reactor vessel): reactor (primary/biological) shield wall; sacrificial shield wall; reactor vessel support/pedestal structure Reduction of strength; loss of mechanical properties due to irradiation (i.e.,
radiation interactions with material and radiation-induced heating)
Plant-specific aging management program or plant-specific enhancements to selected AMPs Yes Consistent with NUREG-2191 as supplemented by SLR-ISG-2021-03-STRUCTURES. The fluence levels or irradiation dose received by any portion of the reactor shield wall, drywell floor, and reactor pedestal concrete during the SPEO from neutron or gamma radiation do not exceed the respective threshold level limits. Therefore, a plant-specific aging management program is not required to manage this aging effect. The Structures Monitoring (B.2.1.33) program will be used to assure integrity of reinforced concrete in the reactor shield wall, drywell floor, and reactor pedestal exposed to an air - indoor uncontrolled environment.
See Subsection 3.5.2.2.2.6.
April 28, 2025 Enclosure A Page 29 of 30 SLRA Section 3.5.2-9, Primary Containment, Summary of Aging Management Evaluation, page 3.5-159 is revised as shown below:
Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs NUREG-2191 Item NUREG-2192 Table 1 Item Notes Concrete: Interior (accessible areas -
drywell floor and reactor pedestal)
HELB/MELB Shielding Structural Support Reinforced concrete Air - Indoor, Uncontrolled Cracking, Loss of Bond, and Loss of Material (Spalling, Scaling)
Structures Monitoring (B.2.1.33)
III.A4.TP-26 3.5.1-066 A
Cracking Structures Monitoring (B.2.1.33)
III.A4.TP-25 3.5.1-054 A
Reduction of Strength; Loss of Mechanical Properties Structures Monitoring (B.2.1.33)
III.A4.T-35 3.5.1-097 A
Concrete: Interior (inaccessible areas
- drywell floor and reactor pedestal)
HELB/MELB Shielding Structural Support Reinforced concrete Air - Indoor, Uncontrolled Cracking, Loss of Bond, and Loss of Material (Spalling, Scaling)
Structures Monitoring (B.2.1.33)
III.A6.TP-104 3.5.1-065 A
Cracking Structures Monitoring (B.2.1.33)
III.A4.TP-204 3.5.1-043 A
Reduction of Strength; Loss of Mechanical Properties Structures Monitoring (B.2.1.33)
III.A4.T-35 3.5.1-097 A
April 28, 2025 Enclosure A Page 30 of 30 SLRA Section 3.5.2-9, Primary Containment, Summary of Aging Management Evaluation, page 3.5-164 is revised as shown below:
Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs NUREG-2191 Item NUREG-2192 Table 1 Item Notes Steel Components:
ring girder assemblies (under reactor vessel skirt)
Structural Support Carbon Steel Air - Indoor, Uncontrolled Loss of Material Structures Monitoring (B.2.1.33)
III.A4.TP-302 3.5.1-077 A
Loss of Fracture Toughness Structures Monitoring (B.2.1.33)
H, 5