ML25080A136
| ML25080A136 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/14/2025 |
| From: | Mark Yoo NRC/NRR/DNRL/NLRP |
| To: | |
| References | |
| Download: ML25080A136 (1) | |
Text
BREAKOUT QUESTIONS Aging Management Audit Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application June 17, 2024 - March 14, 2025
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 2
SLRA Section B.2.1.1, Inservice Inspection Question Number LRA/SLRA Section LRA/SLRA Page Background / Issue (As applicable/needed)
Discussion Question /
Request 1
SLRA Section B.2.1.1 B-18 The SLRA section states that an AMP effectiveness review was performed in 2023 which assessed the AMP effectiveness from a license renewal perspective.
The staff understands that there are periodic ISI program self-assessments and/or peer-assessments based on the current licensing bases. Please provide the latest ISI program assessment report to the portal.
If any issues or deficiencies were identified in the assessment, briefly discuss how they have been addressed.
SLRA Section: 3.1, 3.2, A.2.1.2 and B.2.1.2 - Water Chemistry Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
Table 3.1.2-1 3.1-72 In Table 3.1.2-1 there is an AMR item for stainless steel Valve Body (Class 1) exposed to a treated water environment. This item cites Table 1 item 3.4.1-084 which refers to Stainless steel, nickel alloy piping, piping components exposed to steam.
Please clarify the use of item 3.4.1-084 given that the environment of the Table 2 item appears different than what is cited in that item.
2 Table 3.1.2-1 3.1-64 In Table 3.1.2-1 there is an AMR item for Carbon steel Piping, piping components: Class 1 greater than or equal to 4" NPS exposed to a steam internal environment. This item cites Table 1 item 3.1.1-079 which refers to Stainless steel; steel with nickel alloy or stainless steel cladding; and nickel alloy reactor coolant pressure boundary components exposed to reactor coolant.
Please clarify the use of item 3.1.1-079 given that the environment of the Table 2 item appears different than what is cited in that item.
3 Table 3.1.2-1 3.1-59 and 3.1-60 In Table 3.1.2-1 there is an AMR item for CASS Flow Device (Class 1) exposed to a steam internal environment. This item cites Table 1 item 3.1.1-079 which refers to Stainless steel; steel with nickel alloy or stainless steel cladding; and nickel alloy reactor Please clarify the use of item 3.1.1-079 given that the environment of the Table 2 item appears different than what is cited in that item.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 3
coolant pressure boundary components exposed to reactor coolant.
4 Table 3.2.2-3 3.2-79 and 3.2-84 In Table 3.2.2-3 there are AMR items for carbon steel piping, piping components and valve body exposed to a steam internal environment. This item cites Table 1 item 3.2.1-016 which refers to Steel piping, piping components exposed to treated water.
Please clarify the use of item 3.2.1-016 given that the environment of the Table 2 item appears different than what is cited in that item.
5 A.2.1.2 and B.2.1.2 A-11 and B-22 The UFSAR and the program description for the Water Chemistry program state, In addition, the Water Chemistry program provides for temperature and radioactivity monitoring of the isolation condenser shell side water. Shell side water temperature is monitored daily and is alarmed in the control room if shell side water temperature exceeds the alarm setpoint. Radioactivity monitoring is performed through monthly sampling of the shell side water for isotopic activity and tritium activity. However, the meaning of this statement is unclear to the staff, and how the Water Chemistry program is being used to manage aging for components exposed to this water is unclear to the staff.
Please clarify if the Water Chemistry program is being used to monitor those parameters or being controlled to manage aging effects.
If it is being controlled to manage aging effects, please clarify how the Water Chemistry program is being used to manage aging for components exposed to this environment.
6 Table 3.1.2-2 3.1-89 and 3.1-90 In Table 3.1.2-2 the BWR Vessel Internals program is substituted for the One-Time Inspection program for nickel alloy and stainless steel Reactor Vessel Penetrations: control rod drive stub tubes and housing; in core monitor housings exposed to Reactor coolant (GALL item IV.A1.RP-157 and Table 1 item 3.1.1-085). These items cite generic note E and plant-specific note 4, which states, The BWR Vessel Internals (B.2.1.7) program is substituted to manage the aging effects applicable to this component, material, and environment combination. However, the basis for substituting the BWR Vessel Internals program is unclear to the staff.
Please clarify the basis for substituting the BWR Vessel Internals program for the One-Time Inspection program to manage loss of material for these components.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 4
7 B.2.1.2 B-23 The justification for the exception states that a One-Time Inspection will be used to verify the effectiveness of the Water Chemistry program. It is unclear to the staff how the One-Time Inspection program will be used to verify the effectiveness of the Water Chemistry program for these components.
Please clarify the following:
How the One-Time Inspection program will be used to verify the effectiveness of the Water Chemistry Program for these components.
Do these components have any OE that would make One-Time Inspection inconsistent with the SLR guidance?
Are these components accessible so that a One-Time Inspection can be conducted?
8 N/A N/A Table 1-6 of the EPRI Stator Cooling Water guidelines lists electrochemical potential (ECP) as a key parameter for monitoring stator water chemistry.
However, according to the stator water specifications this parameter is not measured at Dresden.
Please confirm whether ECP is measured at Dresden for the stator water.
If it is not, please explain the basis for not measuring ECP as it is listed as a key parameter in the EPRI guidelines.
SLRA Section B.2.1.8, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)
Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question /
Request 1
The applicants Technical Basis Document in portal states that the applicant has chosen to use an enhanced visual examination and/or a qualified ultrasonic testing (UT) to monitor cracking of susceptible CASS components due to loss of fracture toughness from thermal aging embrittlement, and the examination methods that meet the criteria of ASME Code,Section XI, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems, is used.
The staff notes that the ASME Code,Section XI, Appendix VIII, Supplement 9, Qualification Requirements for Cast Austenitic Piping Welds, has been under development. However, the NRC has Clarify whether the applicant will use ASME Code Case N-824 for qualification of its UT for CASS examination. If not, provide explanation.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 5
approved ASME Code Case N-824, Ultrasonic Examination of Cast Austenitic Piping Welds from the Outside Surface,Section XI, by incorporation by reference into 10 CFR 50.55a via inclusion in Regulatory Guide (RG) 1.147, Revision 20, Inservice Inspection Code Case Acceptability, ASME Section XI, with conditions.
The applicant did not mention the use of ASME Code Case N-824 in its SLRA or its technical basis document.
SLRA AMP B.2.1.27, Buried and Underground Piping and Tanks Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.27 B-146 B-147 SLRA Section B.2.1.27, Buried and Underground Piping and Tanks, states [t]he cathodic protection survey results are evaluated using the -850mV relative to CSE (copper/copper sulfate reference electrode) instant off criterion specified in NACE SP0169 to determine cathodic protection system effectiveness for steel piping and tanks.
Exception No. 1 to SLRA Section B.2.1.27 states this [buried aluminum] piping is provided with cathodic protection which will mitigate any potential corrosion.
GALL-SLR Report AMP XI.M41, Buried and Underground Piping and Tanks, recommends that cathodic protection is provided for buried aluminum piping, with a different acceptance criterion when compared to steel piping. The SLRA addresses the acceptance criterion for steel but not aluminum. The staff requests a discussion on this topic.
2 B.2.1.27 B-147 Exception No. 1 to SLRA Section B.2.1.27 states additional assurance of the long-term reliability of the piping is provided by a cold applied tape coating system that is installed on the pipe as well as a secondary protective encasement of polyethylene wrap, which was installed in accordance with AWWA C105.
An ePortal request for any additional information with respect to the tape coating and polyethylene wrap used on the external surfaces of buried aluminum piping was received, but the staff is leaving this topic as a placeholder for additional discussion (if needed).
The staff reviewed WO 00899125-09 and WO 00899125-10 and noted that
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 6
the piping was externally wrapped with (a) Tapecoat M50 RC, which has thickness of 50 +/- 2 mils; and (b) a polyethylene film, which has a nominal thickness of 8 mils. Is this correct?
3 B.2.1.27 B-147 Exception No. 1 to SLRA Section B.2.1.27 states
[e]lectrical resistance corrosion monitoring is also performed, and a corrosion probe is installed in the CLSM directly adjacent to the [buried aluminum]
HPCI line.
The staff requests a discussion on the following:
- 1. Is the corrosion probe constructed from steel or aluminum?
- 2. If steel, how is this representative of aluminum piping?
- 3. If aluminum, ER probes are intended to indicate metal loss by general corrosion (i.e., not suited for aluminum where pitting and crevice corrosion are the aging mechanisms in a soil environment). More information on this can be found in NUREG-2221, Supplement 1, pages 2-94 and 2-95 (ML23180A208). Based on the above rationale, it is unclear to the staff how aluminum ER probes would provide an accurate external corrosion rate for the buried aluminum HPCI line.
4 B.2.1.27 B-148 Exception No. 1 to SLRA Section B.2.1.27 states [i]f examination results indicate active corrosion is occurring (guided wave indications classified as The staff requests a discussion on the use of guided wave given that the current position in GALL-SLR Report
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 7
greater than minor), then direct examination of suspect areas will be performed.
AMP XI.M41 is that this technique may not be substituted for direct inspections (i.e., external visual or internal UT).
Specifically, the staff would like to discuss the following:
- 1) The ability of guided wave to identify localized corrosion, such as pitting, that may occur in buried aluminum piping.
- 2) The impact of the coating system and CLSM backfill on the inspectable range.
- 3) The definition of a minor guided wave indication.
5 B.2.1.27 B-148 Exception No. 1 to SLRA Section B.2.1.27 states
[f]urther, routine level monitoring in the contaminated condensate storage tanks ensures that system intended functions are maintained and would provide for prompt identification of significant leakage that could threaten system intended functions. In addition, existing monitoring of the groundwater wells in the area of the piping would detect minor leakage from the condensate tanks or piping prior to the leak progressing to a size that could potentially impact intended functions.
The staff requests a discussion on why these activities (i.e., monitoring of tank levels and groundwater wells), which appear to be credited for managing the effects of aging on buried aluminum piping, are not described in the enhancements or the UFSAR supplement description for this program.
6 B.2.1.27 B-149 Enhancement No. 1 to SLRA Section B.2.1.27 states
[p]erform direct visual inspections of one 10-linear foot section of buried stainless steel piping during each 10-year period beginning 10 years prior to the subsequent period of extended operation.
For a two-unit site, GALL-SLR Report AMP XI.M41 recommends two inspections of buried stainless steel piping in each 10-year interval, not one inspection. The staff requests a discussion on this topic.
7 B.2.1.27 B-151 Enhancement No. 8 to SLRA Section B.2.1.27 states
[p]erform direct visual inspection of one 10-linear foot section of underground steel pipe located in the condensate piping vault during each 10-year period For a two-unit site, GALL-SLR Report AMP XI.M41 recommends the smaller of 2 percent of the piping length or 3 inspections for underground steel
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 8
beginning 10 years prior to the subsequent period of extended operation.
piping. Without knowledge of the amount of underground steel piping in linear feet, it is unclear if the 2 percent recommendation from GALL-SLR Report AMP XI.M41 is met. The staff requests a discussion on this topic.
8 B.2.1.27 B-150 Enhancement No. 5 to SLRA Section B.2.1.27 states
[e]xpansion of sample size may be limited by the extent of piping subject to the observed degradation mechanism or if the piping system or portion of the system is replaced or otherwise mitigated
[emphasis added by staff] within the same 10-year inspection interval in which the original degradation was identified or within four years after the end of the 10-year interval, if the degradation was identified in the latter half of the 10-year interval.
The statement or otherwise mitigated is not included in GALL-SLR Report AMP XI.M41. The staff requests a discussion on why the following language is included in the enhancement. A similar question was asked during the Clinton LRA breakout.
9 N/A N/A N/A The staff has the following clarification question. Are the external coatings for buried cast iron piping the same as buried steel piping as noted in the program basis document? If not, what are the external coating system(s) used for buried cast iron piping?
10 B.2.1.27 B-146 SLRA Section B.2.1.27 states the program addresses pipingcomposed of carbon fiber reinforced polymer (CFRP)
The staff requests a discussion with respect to whether the CFRP is (a) structural or non-structural; and (b) applied to the inner or outer diameter of the piping.
11 B.2.1.27 B-146 SLRA Section B.2.1.27 states [a]ging management of the buried Fire Protection System piping will be accomplished through monitoring the activity of the diesel fire pump to detect system leakage.
DR-PBD-AMP-XI.M41, Program Basis Document Buried and Underground Piping and Tanks, GALL-SLR Program XI.M41 - Buried and Underground The staff requests a discussion on the following:
- 1. What is the leakage rate (approximate) during normal operation?
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 9
Piping and Tanks, states [a]t DNPS, the fire protection system does not utilize a jockey pump.
The header pressure is maintained between 90 to 115 psig using a cross connection with the plant service water system. If fire header pressure drops to 80 psig, then the 2/3 Diesel Fire Pump (DFP) will automatically start, and an alarm is actuated. This cross connect provides a flow between 60 and 90 gallons per minute (GPM) prior to actuation of the 2/3 DFP.
- 2. Can the cross connect flow rate be measured? If so, why would this not be monitored as well?
12 N/A N/A GALL-SLR Report AMP XI.M41 states [w]hen using the option of monitoring the activity of a jockey pump instead of inspecting buried fire water system piping, a flow test or system leak rate test is conducted by the end of the next refueling outage [emphasis added by staff] or as directed by the current licensing basis, whichever is shorter, when unexplained changes in jockey pump activity (or equivalent equipment or parameter) are observed.
The recommendation that a flow test or system leak rate test is conducted by the end of the next refueling outage does not appear to be included in the LRA or the program basis document.
The staff requests a discussion on this topic.
13 B.2.1.27 B-149 Enhancement No. 2 to SLRA Section B.2.1.27 states
[p]erform direct visual inspections of two 10-linear foot sections of buried polymeric piping during each 10-year period...
The staff requests a discussion on the following:
- 1. It is the staffs understanding that there are three types of buried polymeric piping (i.e.,
PVC, CFRP, and some other type of polymer). Is this correct?
- 2. What type of polymer is the buried polymeric piping in SLRA Table 3.3.2-8?
- 3. The overall sampling approach given there are three types of buried polymers and two inspections in each 10-year interval.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 10 14 Tables 3.3.2-8 and 3.4.2-1 3.3-192 3.4-60 SRP-SLR Report items 3.3.1-263 and 3.4.1-135 recommend that cracking or blistering due to exposure to ultraviolet light, ozone, radiation, or chemical attack is managed for polymeric piping exposed to soil.
The staff requests a discussion with respect to why this aging effect will not be managed for (a) polymeric piping and piping components exposed to soil in the fire protection system; and (b) carbon fiber reinforced polymer piping and piping components exposed to soil in the condensate system.
15 N/A N/A The staff noted the following during its review of plant-specific operating experience.
AR 04182528 discusses an underground fire header leak.
AR 00928304 notes [t]he root cause of the two through wall leaks was determined to be the degradation of the protective moisture barrier wrap, which allowed moisture to come in contact with the piping resulting in external corrosion.
ARs 01252357 and 01253628 discuss pin hole leaks in asbestos coated buried piping.
What was the cause of the leak referenced in AR 04182528?
The staff requests a discussion on AR 00928304, including the extent of use of the protective moisture barrier wrap on in-scope piping.
The staff requests a discission on ARs 01252357 and 01253628, including (a) whether they involved in-scope piping; (b) whether the cause was internal or external corrosion; and (c) the extent of use of the asbestos coating on in-scope piping.
SLRA Section/AMP: B.2.1.28 Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.28 B-157 to B-158 The Program Basis Document DR-PBD-AMP-XI.M42, Rev. 1 on the applicant Portal states that components in the scope of the AMP come from the following systems:
Closed Cycle Cooling Water System Fire Protection System Isolation Condenser System Discuss operating experience, from the following in-scope systems, if available, that provides objective evidence that SLRA Section B.2.1.28 AMP inspections will be effective in identifying and managing aging
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 11 Low Pressure Coolant Injection System Station Blackout Diesel Generator System SLRA Section B.2.1.28, Operating Experience, describes operating experience examples from two of the above systems (the Isolation Condenser System and the Low Pressure Coolant Injection System) and states that these examples provide objective evidence that internal coatings are inspected for degradation, and that adverse conditions are identified and addressed in the corrective action program to repair or replace degraded conditions.
effects and loss of coating integrity for the following in-scope systems:
Closed Cycle Cooling Water System Fire Protection System Station Blackout Diesel Generator System 2
Table 3.3.2-15 3.3-249,250,255 SLRA Table 3.3.2-15 lists two heat exchanger components (drywell equipment drain sump and reactor building equipment drain tank) which are internally coated carbon steel and for which the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program is proposed to manage for the aging effect of loss of material. Plant Specific Note 1 justifies the use of this program by claiming that these two components meet the six criteria in GALL-SLR AMP XI.M42, Element 4, to use the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program in lieu of the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks.
Of the six GALL-SLR criteria referenced in Plant Specific Note 1, NRC staff were only able to verify that the components only current licensing basis intended function is leakage boundary.
Discuss how the GALL-SLR AMP XI.M42, Element 4, criteria related to downstream effects such as reduction in flow, drop in pressure, or reduction of heat
- transfer, accelerated corrosion of the base metal resulting from chemical contaminants in the internal environment, microbiologically influenced corrosion resulting from the internal environment, potential of galvanic couples resulting from the location of coated components in proximity to uncoated components, whether or not the component design credited the coating/lining,
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 12 were evaluated to justify the proposed alternative to GALL-SLR AMP XI.M42 for the drywell equipment drain sump and reactor building equipment drain tank.
Also, NRC staff notes that the basis for utilizing an alternative AMP should be docketed so that the staff can evaluate these generic note E items.
3 Table 3.4.2-5 3.4-98, 101 SLRA Table 3.4.2-5 includes the turbine oil reservoirs which are internally coated carbon steel and for which the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program is proposed to manage for the aging effect of loss of material. Plant Specific Note 1 justifies the use of this program by claiming that this component meets the six criteria in GALL-SLR AMP XI.M42, Element 4, to use the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program in lieu of the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks.
Of the six GALL-SLR criteria referenced in Plant Specific Note 1, NRC staff were only able to verify that the components only current licensing basis intended function is leakage boundary.
Discuss how the GALL-SLR AMP XI.M42, Element 4, criteria related to downstream effects such as reduction in flow, drop in pressure, or reduction of heat
- transfer, accelerated corrosion of the base metal resulting from chemical contaminants in the internal environment, microbiologically influenced corrosion resulting from the internal environment, potential of galvanic couples resulting from the location of coated components in proximity to uncoated components, whether or not the component design credited the coating/lining, were evaluated to justify the proposed alternative to GALL-SLR AMP XI.M42 for the turbine oil reservoirs.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 13 Also, NRC staff notes that the basis for utilizing an alternative AMP should be docketed so that the staff can evaluate these generic note E items.
4 B.2.1.28 B-156 The SLRA AMP B.2.1.28 claims consistency with the ten elements of the GALL-SLR AMP XI.M42.
The Program Basis Document DR-PBD-AMP-XI.M42, Rev. 1 on the applicant Portal states, in its discussion of consistency with GALL-SLR AMP XI.M42 Element 4 Detection of Aging Effects, that an operating experience review has not identified any significant MIC issues in the Plant Drainage System or Plant Radwaste System.
Discuss what is meant by significant MIC issues and if any MIC issues not considered significant have been identified in the Plant Drainage System or Plant Radwaste System.
SLRA Section B.2.1.9: Flow Accelerated Corrosion Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
N/A N/A BWRVIP-205 lists Dresden Unit 3 as a Category B for FAC wall thinning in the bottom head drain. Category B plants are considered to be only minimally affected by FAC. In contrast, Dresden Unit 2 has been categorized as Category C which is considered to have the highest amount of wall thinning.
The staff would like to know if any inspections were conducted on any portion of the Dresden Unit 2 or 3 bottom head drains.
Additionally, the staff requests the piping isometric drawing or equivalent documentation for the bottom head drain line of Dresden Units 2 and 3.
2 N/A N/A Susquehanna Unit 2 is designated as Category C. In 2019, Susquehanna Unit 2 experienced unexpected wall thinning in the bottom head drain line from flow accelerated corrosion.
Given both Susquehanna Unit 2 and Dresden Unit 2 are both Category C, how has Susquehanna operating experience been dispositioned by Dresden?
3 N/A N/A ER-AA-430-1004 Section 4.2 states that all components potentially susceptible to Erosion in Piping and Components (EPC) degradation will be Did a System Susceptibility Evaluation identify components potentially susceptible to EPC degradation at
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 14 identified through a System Susceptibility Evaluation.
It also states that Station piping specifications, line lists, and piping design tables can identify systems which contain components constructed of materials susceptible to EPC damage.
Dresden Units 2 and 3? If so, discuss the results of the evaluation. If not, discuss how potentially susceptible components were identified.
4 N/A N/A GALL-SLR AMP XI.17 Flow Accelerated Corrosion (FAC) states that this aging management program is informed and enhanced when necessary, through the systematic and ongoing review of both plant-specific and industry operating experience such that the effectiveness of the program is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report. Between 2017 and 2021, the reactor bottom head drain line at Susquehanna, Unit 2 experienced unexpected wall thinning due to FAC in the line downstream of the first 90-degree elbow. This resulted in the affected portion of the line being isolated in 2021 to prevent further wall loss and replaced in 2023. Subsequent root cause analysis directed by the licensee indicates that contributing factors to the FAC include (but are not necessarily limited to) the following factors:
Turbulent and high-volume flow rate in the line Use of expanders to increase the line diameter Base material and weld metal chemistry (primarily Cr content) of the pipe and weld segments.
It is not clear to the staff how Dresden has considered this industry operating experience in its FAC program.
a) Specifically discuss the following factors as they relate to FAC susceptibility in the Dresden, Unit 2 reactor bottom head drain line:
Flow rate in the line Geometry of the line (pipe diameters, number and degree of pipe bends, use of expanders/reducers, whether welds are butt welds or fillet welds)
Base material and weld metal chemistries of the pipe and weld segments in the line Discuss any other factors or lessons learned from the Susquehanna Unit 2 operating experience, if relevant to managing FAC at Dresden Unit 2.
5 N/A N/A Dresden Unit 2 Isometric Drawing ISI-501, Sheet 7, which was made available on the Portal, shows the general configuration of the reactor bottom head drain line from the bottom head to the pedestal. Note 1 of the Drawing states that the configuration is schematic a) While the staff understands the schematic nature of the drawing, the staff requests discussion of the following details in the
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 15 in nature and detailed only for the purpose of defining the components subject to system pressure testing.
breakout session: the relative location of piping elbows, orientation of piping between the elbows (i.e. horizontal, vertical),
relative location of welds that are known, relative location of piping reducers/expanders (if used),
and pipe diameter.
The staff also requests that during the breakout session the applicant provide and discuss the isometric drawing showing the reactor bottom head drain line downstream of the reactor pedestal.
SLRA Section: B.2.1.10, Bolting Integrity Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.10 B.2.1.10 B-64 B-68 1-SLRA page, B-64: Second sentence in second paragraph states In addition, the program manages aging of submerged mechanical bolting for the traveling screens.
2-SLRA page B-68: second paragraph, enhancement Item No 5 states Revise engineering procedures to require volumetric examination in accordance with ASME Code Section XI, Table IWB-2500-1, Examination Category B-G-1, regardless of the code classification of the bolting, should high strength bolting greater than 2 inches in diameter be installed during the subsequent period of extended operation.
Is there any mechanical bolting greater than 2 inches used for the traveling screens?
If yes, how the required volumetric examinations will be performed in submerged conditions using the ASME Code Section XI, Table IWB-2500-1, Examination Category B-G-1?
2 B.2.1.10 B-65 1-SLRA page B-65: first paragraph states Aging management reviews have determined that high strength bolting material with actual yield strength It is not clear why enhancement Item 5 with volumetric examination in accordance with ASME Code Section
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 16 B.2.1.10 B-68 of 150 ksi or greater is not used for closure bolting with bolt diameters greater than 2 inches on pressure-retaining components within the scope of this program. Therefore, sample based volumetric inspection of closure bolting to detect indications of cracking is not applicable.
2-SLRA page B-68: second paragraph, enhancement Item No 5 states Revise engineering procedures to require volumetric examination in accordance with ASME Code Section XI, Table IWB-2500-1, Examination Category B-G-1, regardless of the code classification of the bolting, should high strength bolting greater than 2 inches in diameter be installed during the subsequent period of extended operation.
XI, Table IWB-2500-1, Examination Category B-G-1 is required.
Since the aging management reviews determined that high strength bolting material with actual yield strength of 150 ksi or greater is not used for closure bolting with bolt diameters greater than 2 inches on pressure-retaining components within the scope of this program.
3 B.2.1.10 All AMR items associated with TRP 18 except those that are not applicable.
B-64 The only exception presented in SLRA Section B.2.1.10 is: NUREG-2191 recommends that the scope of the Bolting Integrity program manages aging of closure bolting on pressure-retaining components. The scope of the Bolting Integrity aging management program will also include the aging management of mechanical bolting for the traveling screens.
Discussions for all AMR items associated with the Bolting Integrity AMP (i.e., AMR items 3.1.1-062, 063, and 067; 3.2.1-014, 015, 076, and 079; 3.3.1-012, 015, 142, and 145; 3.4.1-006, 009, and 073) state that these items are consistent with NUREG-2191 with exceptions and also an exception applies to the NUREG-2191 recommendations for Bolting Integrity (B.2.1.10) program implementation. All of the Table 2 items associated with these AMR items Exceptions to NUREG-2191
- 1. Clarify what exception is applicable to the Bolting Integrity AMR items and associated Table 2 items with generic note B. Revise, if applicable, the generic note from B to A for the Table 2 items associated with the Bolting Integrity AMR items.
- 2. Confirm whether a new component type line needs to be included as Traveling Screen Bolting in SLRA Table 3.3.2-12.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 17 cite generic note B, which implies that AMP takes some exceptions to NUREG-2191 AMP.
If the exception described in SLRA Section B.2.1.10 is the one referred to in these discussions, the generic note may need to be changed to A because those Table 2 items in the application are not related to the traveling screen.
SLRA Table 3.3.2-12 includes the component type of Bolting (Closure) using the Bolting Integrity AMP (B.2.1.10) and correctly note B was identified.
However, a new component type line item may need to be added for Traveling Screen Bolting in SLRA Table 3.3.2-12.
4 B.2.1.10 B-64 Enhancement 1 states the recommended guidance for proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting is a requirement at DNPS in accordance with the guidelines provided in EPRI TR-104213. The staff noted that the program basis document credits additional documents such as EPRI Report NP-5769, EPRI Report 1015336, EPRI Report 1015337, and NUREG-1339.
Enhancement 1 Include citations/references in the SLRA for the documents that provide guidance for proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting.
5 B.2.1.10 B-64 Justification for Exception states Although the submerged mechanical bolts for the traveling screens are not closure bolting on pressure-retaining components, the Bolting integrity aging management program has been determined to address the aging effects of loss of preload and loss of material as recommended in NUREG-2191.
In addition, Enhancement 2 states If the minimum sample size is not achieved during a 10-year period, then alternative inspections may be performed. For Enhancement 2 Confirm whether the treated water environment is applicable for this case.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 18 submerged bolting exposed to treated water, alternative inspections may include (a) diver inspections or (b) remote video inspections.
However, the staff noted that the Traveling Water Screens are not exposed to the treated water environment.
6 B.2.1.10 B-64 Enhancement 3 states Revise procedures governing the direct visual examination of bolted joints to include inspection parameters such as lighting, distance, and offset. This requirement is not clear regarding how the Detection of Aging Effects program element is being enhanced by inspection parameters such as lighting, distance, and offset.
The staff noted that the program basis document correctly states Existing repetitive tasks will be enhanced to specify adequate lighting and minimum inspection distances to adequately identify loss of material in submerged environments.
Enhancement 3 Confirm whether this enhancement needs to be revised to reflect the information provided in the basis document.
7 Item 3.4.1-007 3.4-26 Discussion for AMR item 3.4.1-007 states that this item is consistent with NUREG-2191 with exceptions. However, there are no Table 2 items associated with AMR item 3.4.1-007.
Clarify if AMR item 3.4.1-007 is applicable. If yes, provide Table 2 items associated with AMR item 3.4.1-007. If no, revise the discussion for AMR item 3.4.1-007 and provide justifications for its non-applicability.
SLRA Sections: 4.7.1 and B.2.1.13 - Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.13 B-82 SLRA page B-82: second paragraph, last sentence states Approximately 33 cranes and hoists, including the reactor building overhead cranes, turbine building overhead cranes, numerous equipment handling The DNPS SLRA identified approximately 33 cranes and hoists, trollies, and monorails that are managed by the AMP (XI.M18).
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 19 4.7.1 4.7-1 cranes, hoists, trollies, and monorails are managed by the program. Also, within the scope of the program are handling systems that lift light loads including equipment, tools, and fuel, over fuel and safety-related equipment within the spent fuel pool and reactor cavity.
SLRA page 4.7-1: first paragraph, first sentence states Section 4.7.1 of the Dresden Nuclear Power Station (DNPS) first License Renewal Application documented a TLAA associated with the Reactor Building Overhead Crane.
However, only the Reactor Building Overhead Crane was considered for the TLAA evaluation.
1-Justify why the other cranes, monorails, and hoists that are in scope for the subsequent license renewal AMP were not considered for the TLAA evaluations.
2-Provide the specific number of overhead heavy load and light load handling systems that are in the scope of the SLR.
Note: An approximation is not appropriate to use in this case, the total number of overhead heavy load and light load handling systems should be known and specifically stated in the SLRA.
3-Provide the list of overhead heavy load and light load handling systems that are in scope of the SLR AMP.
Revise the SLRA accordingly.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 20 2
AMR item 3.3.1-052 Table 3.3.2-5 3.3-68 3.3-155 SLRA AMR item 3.3.1-052 is associated with aging effects of loss of material, wear, deformation, and cracking, and the discussion for this item states that it is consistent with NUREG-2191. SLRA Table 2 items associated with AMR item 3.3.1-052 do not consider all the aging effect/mechanism as listed in SRP, AMR item 3.3-1,052.
In SRP, AMR item 3.3-1,052 lists four (4) aging effect/mechanism of Loss of material due to general corrosion, wear, deformation and cracking.
Revise SLRA line items associated with AMR item 3.3.1-052 by including all the aging effect/mechanism as listed in SRP.
SLRA Sections: 4.7.2 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
SLRA Section 4.7.2 4.7-4 TLAA Evaluation in SLRA Section 4.7.2 states 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.
However, the staff noted the followings:
- 1. A footnote in Section 4.7.2 of the SLRA basis document DR-TLAABD states Reference 4.9.36 (COM-27-217) clarifies that the 850 load cycles include 550 SRV actuations, 150 pressure cycles and 150 temperature cycles. A similar Dresden torus fatigue usage table (from PUAR volume 2, reference 4.84), assumes 300 SRV actuations, 150 pressure
- 1. Provide a basis for comparing the projected number of 75 SRV actuations with 850 actuations, considering the effects of pressure cycles and temperature cycles in the evaluation.
- 2. Clarify the basis of 220 Design Transient Occurrences in SLRA Table 4.3.1-2. Clarify how this number is related to the number of SRV actuations used in SLRA Section 4.7.2.
- 3. Correct the inconsistency in the projected number of actuations specified in SLRA Sections 4.7.2 and A.4.7.2
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 21 cycles and 150 temperature cycles, so the Quad Cities analysis bounds the Dresden analysis.
- 3. SLRA Section A.4.7.2 states 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.
Therefore, it is not clear whether the projected number of 75 SRV actuations can be compared with 850 actuations and how pressure cycles and temperature cycles would affect this evaluation. In addition, it is not clear what is the basis of 220 Design Transient Occurrences in SLRA Table 4.3.1-2 and how this number is used in Section 4.7.2. Also, the projected number of actuations specified in SLRA Sections 4.7.2 (75) and A.4.7.2 (74) are inconsistent.
SLRA Section: B.2.1.15 - Fire Protection Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
2.3.3.8 2.3-72 SLRA Section 2.3.3.8 states that the Fire Protection System includes fire rated enclosures.
However, SLRA Tables 2.3.3-8 and 3.3.2-8 do not include component type fire rated enclosures. In addition, DPR-PBD-AMP-XI.M26 (Fire Protection Program basis document) does not refer to component type fire rated enclosures.
The staff recognizes that one of the component types in SLRA Tables 2.3.3-8 Please discuss the following:
- 1. Where fire rated enclosures are addressed in the SLRA.
- 2. Whether concrete slabs refer to concrete ceilings and floors.
- 3. Whether the ventilation dampers in the Radwaste Structures have a fire barrier intended function.
- 4. Whether the component type fire barriers (damper assembly) includes ventilation dampers with a fire barrier intended function.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 22 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request and 3.3.2-8 may include fire rated enclosures, but the SLRA is not clear.
SLRA Section 2.3.3.8 states the Fire Protection System includes fire barrier walls and slabs. However, SLRA Tables 2.3.3-8 and 3.3.2-8 do not include component type fire barrier slabs. It appears that the SLRA and the Fire Protection Program basis document are referring to fire barrier concrete ceilings and floors when referring to slabs.
Therefore, the staff wants to ensure there understanding is correct.
SLRA Section 2.3.3.8 states that the Fire Protection System includes fire barrier dampers and ventilation dampers.
However, SLRA Tables 2.3.3-8 and 3.3.2-8 do not include component type ventilation damper. In addition, DPR-PBD-AMP-XI.M26 (Fire Protection Program basis document) does not refer to component type fire ventilation damper. The staff notes that SLRA Section 2.4.10, Radwaste Structures, states that ventilation dampers are evaluated with the Fire Protection System. It is unclear whether the ventilation dampers in the Radwaste Structures have a fire barrier intended function and whether the component type fire barriers (damper assembly)
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 23 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request includes ventilation dampers with a fire barrier intended function.
2 2.4.14, 2.4.16 2.4-63, 2.4-71 SLRA Section 2.4.14, Switchyard Structures, states that there is a firewall between the Unit 1 and Unit 2 reserve auxiliary transformers and that it is evaluated in Yard Structures. However, SLRA Section 2.4.16, Yard Structures, refers to the concrete block firewall that separates the Unit 1 and Unit 2 reserve auxiliary transformers, and goes on to state that fire protection components are evaluated separately in the Fire Protection System. SLRA Tables 2.4-16 and 3.5.2-16 do not have concrete block components, nor do they have components with a fire barrier intended function.
Please confirm the staffs understanding that the component type fire barriers (masonry walls) in SLRA Tables 2.3.3-8 and 3.3.2-8 include the firewall between the Unit 1 and Unit 2 reserve auxiliary transformers.
3 3.3 3.3-185 SLR-ISG-2021-02-Mechanical (ML20181A434) added AMR item VII.G.A-807 to Table VII.G in Volume 1 of NUREG-2191 and Table 3.3-1 in NUREG-2192. The aging effects for silicates used as fireproofing/fire barriers exposed to air are loss of material, change in material properties, cracking, delamination, and separation. These aging effects are consistent with Section 6, Fire Barriers, of EPRI 3002013084, Long-Term Operations: Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools), November 2018.
Please provide the technical basis for why change in material properties, delamination, and separation were not cited as applicable aging effects for gypsum.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 24 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request LRA Table 3.3.2-8 identifies cracking and loss of material as the only aging effects for gypsum fire barriers (penetration seals and fire stops) exposed to indoor uncontrolled air.
It is unclear why change in material properties, delamination, and separation were not cited as applicable aging effects for gypsum given that it is similar to cementitious fireproofing/fire barriers.
The staff notes that Section 2.0 of the Fire Protection Program basis document that it appears to indicate that delamination is applicable to gypsum.
4 A.2.1.15 A-19 Table XI.01 in NUREG-2191, Volume 2 (ML17187A204), states, The program also includes periodic inspection and testing of halon/carbon dioxide or clean agent fire suppression systems.
The first sentence in SLRA Section A.2.1.15 states, The Fire Protection aging management program is an existing condition monitoring and performance monitoring program that includes fire barrier visual inspections and low-pressure carbon dioxide system visual inspections and functional testing.
This sentence does not refer to the halon fire suppression system, however, the Please discuss whether both the first and last sentences should refer to the halon fire suppression system.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 25 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request last sentence does, which states, The program also includes visual inspections and periodic functional tests of the low-pressure carbon dioxide and halon fire suppression systems using the National Fire Protection Association Codes and Standards for guidance.
5 N/A N/A Section 3.10 of the Fire Protection Program basis document includes discussion of AR02638773 and WO1737330 related to securing a metal clamp around ceramic blanket and pipe.
Section 6.1 of EPRI 3002013084, Long-Term Operations: Subsequent License Renewal Aging Effects for Structures and Structural Components (Structural Tools), November 2018, states, in part, wire and other appurtenances used to secure fire wrap to the item being protected - is considered to be part of the fire wrap itself.
Please identify the materials used for securing fire wraps and where they are addressed in the SLRA, including AMR items for managing applicable aging effects.
6 N/A N/A Section 2.1 of the Fire Protection Program basis document appears to indicate that cracking and delamination are applicable aging effects for reinforced concrete, concrete block, grout, and gypsum. However, SLRA Table 3.3.2-8 does not identify delamination as an applicable aging effect for these materials.
The staff notes that Section 3.1 of the Fire Protection Program basis document Please discuss whether it was the intent to identify delamination as an applicable aging effect for reinforced concrete, concrete block, grout, and gypsum (see question 3 for gypsum).
Please discuss whether the intent was to identify loss of bond as an applicable aging effect for fire barrier materials.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 26 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request refers to loss of bond. However, loss of bond is not identified as an applicable aging effect in SLRA Table 3.3.2-8.
SLRA Section: B.2.1.11 - Open-Cycle Cooling Water System Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.11
Background:
As shown below in B.2.1.11, the OCCW System manages carbon fiber reinforced polymer for loss of material, cracking, hardening or loss of strength, reduction of heat transfer, and flow blockage.
B.2.1.11 Open-Cycle Cooling Water System Program Description The Open-Cycle Cooling Water System aging management program is an existing preventive, mitigative, condition monitoring and performance monitoring program based on the implementation of NRC Generic Letter (GL) 89-13 and includes nonsafety-related portions of the open cycle cooling water system. The program includes: (a) surveillance and control to significantly reduce the incidence of flow blockage problems as a result of biofouling, (b) tests to verify heat transfer of heat exchangers, and (c) periodic inspection and maintenance so that corrosion, erosion, cracking, fouling, and biofouling cannot degrade the performance of systems serviced by the open cycle cooling water systems. The program also includes evaluation, repair, or replacement of components that do not meet minimum wall thickness requirements. The program applies to components constructed of steel, cast iron, stainless A word search of the SLRA did not reveal any Table 2 locations where flow blockage or cracking were being managed for CFRP.
Please provide the technical basis for not including flow blockage and cracking as an AERM.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 27 steel, copper alloys, and carbon fiber reinforced polymer.
There is no cement or cementitious piping within the scope of the program. This program includes guidance beyond the requirements contained in NRC GL 89-13, such as inputs from industry reports and documents (e.g., EPRI documents) that address operating experience such that aging effects are adequately managed.
The program manages piping, piping components, and heat exchanger components in safety-related and nonsafety-related raw water systems that are exposed to a raw water environment for loss of material, cracking, hardening or loss of strength, reduction of heat transfer, and flow blockage.
In Table 3.3.1, the OCCW System AMP is called out to manage hardening or loss of strength, and loss of material for CFRP but flow blockage and cracking are not mentioned, despite being listed as aging effects requiring management (AERMs).
In Table 3.3.2-12, hardening or loss of strength, and loss of material for CFRP are listed but flow blockage and cracking are not.
SLRA Section: B.2.1.16 - Fire Water System Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
3.3 3.3-196 Volume 1 of NUREG-2191 includes the following AMR items that address flow blockage of stainless steel components exposed to condensation in the Fire Protection System:
Please provide a technical basis for why flow blockage was not identified as an applicable aging effect for the stainless steel spray nozzles exposed
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 28 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 3.3.1-130, flow blockage of metallic sprinklers 3.3.1-131, flow blockage of stainless steel piping and piping components SLRA Table 3.3.2-8 does not cite flow blockage as an applicable aging effect for the stainless steel spray nozzles exposed to condensation nor does it provide a technical basis for why flow blockage is not applicable.
The staff notes that SLRA Table 3.3.2-8 cites AMR item 3.3.1-130 to manage flow blockage of the copper alloy spray nozzles exposed to condensation.
to condensation in the Fire Protection System.
2 2.3.3.8, 3.3, A.2.1.16, B.2.1.16 2.3-72, 2.3-75, 3.3-187, A-20, B-92 SLRA Sections 2.3.3.8, A.2.1.16, and B.2.1.16 state that the Fire Protection System includes standpipes and hose stations. However, SLRA Tables 2.3.3-8 and 3.3.2-8 only include hose stations.
The staff notes that Section 3.1 of Revision 1 of DR-PBD-AMP-XI.M27 (program basis document) states, Fire hose stations and standpipes are included as piping components.
Given that hose stations are explicitly listed in SLRA Tables 2.3.3-8 and 3.3.2-8 and not included as piping components, please discuss where standpipes are addressed in the SLRA.
3 A.2.1.16, B.2.1.16 A-20, B-92 SLRA Section A.2.1.16 states The program manages aging of fire protection system components exposed to raw water, while SLRA Section B.2.1.16 refers to raw water and condensation environments.
Please discuss whether SLRA Section A.2.1.16 should also refer to condensation.
4 A.2.1.16, B.2.1.16 A-20, B-92 SLRA Section A.2.1.16 states, Testing or replacement of sprinklers that have been in place for 50 years and 75 years is performed using the guidance of NFPA 25, 2011 Edition.
Please discuss whether dry sprinklers are utilized at DNPS.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 29 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request SLRA Section B.2.1.16 states, The Fire Water System program includes replacement or testing of a representative sample of sprinklers before they reach 50 years and 75 years of service.
The staff notes that Section 3.4 of the program basis document states that fast response sprinkler elements and solder type sprinklers with a temperature classification of extra high (325°F) or greater are not used at DNPS.
Section 5.3.1.1.1.6 of the 2011 Edition of NFPA 25 states, Dry sprinklers that have been in service for 10 years shall be replaced or representative samples shall be tested and then retested at 10-year intervals. The SLRA and program basis document dont appear to address dry sprinklers.
5 B.2.1.16 B-93 SLRA Section B.2.1.16 includes an exception to Section 13.2.5 of the 2011 Edition of NFPA 25 that specifies A main drain test shall be conducted annually at each water-based fire protection system riser to determine whether there has been a change in the condition of the water supply piping and control valves. The exception states, The DNPS program performs main drain testing of forty-seven (47) of the seventy-two (72) in scope water-based fire protection systems on an 18 month frequency. Systems omitted from main drain testing share a common supply with tested systems such that the supply for non-tested systems is verified. In addition, DNPS performs alternative testing and inspections to ensure that the intent of the recommended main drain testing is met.
Please address the following:
- 1. Will 47 or 53 main drain tests be performed in accordance with NFPA 25?
- 2. Confirm that the exception only applies to 25 (72 - 47) or 19 (72 -
- 53) in scope water-based fire protection systems.
- 3. What condition(s) noted in NEIL is being met?
- 4. Identify all alternative inspections and tests, including the procedures, that are credited in this exception.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 30 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request The justification for the exception states, The NEIL LCM, however, includes an allowance to omit main drain testing, and The allowance requires the performance of various other tests and inspection (e.g., valve position verification and cycling, flow tests, flushes, and suppression system inspection) which, in place of the main drain test, ensures that no major obstructions exist in the fire suppression system piping. The justification further states, Dresden Nuclear Power Station manages the fire water supply and performs testing and inspections in accordance with its NRC approved fire protection program as supplemented by additional NEIL requirements.
With regards to the allowance for main drain testing, NEIL identifies conditions that need to exist and surveillances that should be performed if certain conditions exist. The justification is not explicit with regards to the NEIL requirements.
The justification states, Therefore, given the rigorous regime of alternative tests, inspections, and OE, reasonable assurance is provided that major obstructions to the supply of the in-scope systems would be identified during the 53 main drain tests that are currently performed. It is unclear whether 47 or 53 main drain tests will be performed.
6 B.2.1.16 B-95 Enhancement 1 in SLRA Section B.2.1.16 is related to a one-time volumetric wall thickness inspection on a representative sample that is periodically subject to flow during testing.
Please discuss clarifying in the SLRA that DCNPP does not have portions of fire protection system components that are normally dry but periodically
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 31 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request The fifth item under Operating Experience in SLRA Section B.2.1.16 notes that three systems out of six were identified as requiring correction to either the slope of piping or a reconfiguration of drainage.
However, it doesnt state whether the corrections were performed, like the program basis document appears to.
The SLRA does not clearly state, like the program basis document does, that DCNPS does not have sections that collect water or dont drain properly.
Without a statement that sections that collect water or dont drain properly are not applicable to DCNPS, it could seem like Enhancement 1 doesnt meet the augmented tests and inspections for such sections in GALL-SLR or is incorrect/incomplete.
wetted that cannot be drained or allow water to collect.
7 3.3 3.3-231 SLRA Section 3.3.2-12 includes galvanized traveling water screens. Flow blockage and loss of material is managed by the Open-Cycle Cooling Water System program and long-term loss of material is managed by the One-Time Inspection program.
Based on the program basis document, the suction screen requirements in Section 8.3.3.7 of the 2011 Edition of NFPA 25 apply to the traveling water screens at DCNPS.
It is unclear whether there is a strainer at the fire pump intake and whether it is addressed in the strainer (element) component type in SLRA Table 3.3.2-8.
Please discuss whether there is a strainer at the fire pump intake and whether it is addressed in SLRA Table 3.3.2-8.
8 N/A N/A The program basis document states that Deluge system strainers are inspected after each operation Please identify the specific section(s) in NEIL.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 32 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request or flow test and are inspected at least every 6 years as allowed by NEIL.
The staff notes that Exception 2 does not refer to a NEIL allowance.
9 3.3 3.3-182 SLRA Tables 2.3.3-8 and 3.3.2-8 do not appear to include components associated with the diesel-driven fire pump engine (i.e., heat exchanger channel, shell, tube, and jacket water).
SRP-SLR Table 2.3-2, "Examples of Mechanical Components Screening and Basis for Disposition,"
states the following:
Diesel engine jacket water heat exchanger and portions of the diesel fuel oil system and starting air system supplied by a vendor on a diesel generator skid These are passive, long-lived components having intended functions. They are subject to an AMR for SLR even though the diesel generator is considered active.
In addition, SRP-SLR Table 2.1-6 reflects that heat exchangers are considered passive, long-lived components that are subject to AMR.
The staff notes that SLRA Tables 2.3.3-8 and 3.3.2-8 include flexible connection, piping and piping components, and silencer/muffler exposed to diesel exhaust.
Please address the following:
- 1. Provide a technical basis for why the SLRA does not appear to address components associated with the diesel-driven fire pump engine (i.e., heat exchanger channel, shell, tube, and jacket water).
- 2. Discuss whether there has been any operating experience associated with any components associated with the diesel-driven fire pump engine (e.g., coolant leaks, tube bundle replacement, fouling, etc.), including what caused the degradation.
- 3. Identify and describe periodic maintenance procedures/activities related to the diesel-driven fire pump, including whether periodic chemistry testing is performed.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 33 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request Note: There is a related question under Scoping and Screening.
SLRA AMP B.2.1.21, Selective Leaching Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.21 B-122 SLRA Section B.2.1.21, Selective Leaching, states
[s]ince DNPS is a two-unit site, a reduced periodic visual inspection sample size of eight components per population per unit will be adopted for sample populations that are not percentage-based. This sample size reduction is acceptable because, for the components in the scope of the periodic program, environmental conditions between the units are similar enough such that the aging effects are not occurring differently. Changes to water chemistry practices and to plant equipment and operating conditions (including power rerates) have been performed on both units at approximately the same time. Water chemistry programs monitor various chemistry parameters and require out-of-spec conditions to be corrected under the corrective action program in a timely manner. Raw water systems for both units draw from the same source, the Kankakee River. Therefore, a reduced sample size will provide a representative sample of the condition of the plant and equipment and the existence of the aging effects involved.
The discussion related to the multi-unit site reduction focuses on aqueous environments; however, components susceptible to selective leaching are also exposed to a soil environment.
The staff requests a discussion with respect to soil parameter consistency at DNPS.
2 Various Various The materials within the scope of SLRA Section B.2.1.21 include gray cast iron, malleable iron, and ductile iron. However, AMR Table 2 items which reference the Selective Leaching program only use the material type of cast iron. Therefore, the staff The staff request a discussion with respect to (a) why the type of cast iron is not specified in the individual AMR Table 2 items; and (b) the specific type
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 34 cannot determine which type of cast iron is associated with each AMR Table 2 item.
of cast iron used for buried fire protection system piping.
3 B.2.1.21 B-122 SLRA Section B.2.1.21 states [t]he acceptance criteria are for gray cast iron and ductile iron, the absence of a surface layer that can be easily removed by chipping or scraping.
The staff request a discussion with respect to why malleable iron is also not included.
4 B.2.1.21 B-123 B-124 Exception No. 1 in SLRA Section B.2.1.21 will allow for non-destructive examinations (NDE) in lieu of destructive examinations for gray cast iron, malleable iron, and ductile iron populations. The exception references specific techniques (e.g., ultrasonic, electromagnetic); however, the exception does not specify which technique(s) will be utilized.
On January 12, 2022, the Nuclear Energy Institute (NEI) submitted proposed revisions to AMP XI.M33, Selective Leaching, which introduced allowing for NDE in lieu of destructive examinations (ML22019A291). The staffs audit report (ML22353A608) related to the proposed revisions noted several issues including this effort
[related to NDE research for selective leaching] is an ongoing research project (i.e., not a proven practice) and is applicable to gray cast iron (i.e., only one of the four materials within the scope of GALL-SLR Report AMP XI.M33) and nondestructive examination techniques can be non-conservative (i.e., estimates greater remaining wall thickness) when compared to pit gauges and laser profilometry. In addition, as noted in ER-AA-700-401, Selective Leaching Aging Management, ultrasonic sound attenuation due to grain structure (i.e.,
presence of graphite flakes or nodules) can make cast irons difficult to inspect.
Based on the issues noted above, the staff requests a discussion with respect to the use of NDE in lieu of destructive examinations.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 35 5
B.2.1.21 B-125 The operating experience discussion in SLRA Section B.2.1.21 states [t]he one-time inspections consisted of nine visual inspections (consistent with ASME Section XI VT-1) of susceptible components.
Based on a review of the initial LRA, it appears that these visual inspections were not supplemented with hardness testing. However, it is not clear to the staff if mechanical examination techniques (e.g., chipping, scraping) supplemented these visual inspections.
Based on its review of Dresden Station Aging Management Program Results Book B.1.24 Selective Leaching it does not appear that this is the case, but the staff requests a discussion on this topic to confirm that this is an accurate assessment.
6 Table 3.2.2-3 3.2-81 SLRA Table 3.2.2-3, High Pressure Coolant Injection System, states cast iron pump casings exposed to treated water will be managed for loss of material due to selective leaching using the Selective Leaching program and references NUREG-2192 Table 1 item 3.3.1-72.
The staff requests a discussion with respect to why NUREG-2192 Table 1 item 3.2.1-36 is not cited in lieu of 3.3.1-72. The staff recognizes that NUREG-2192 Table 1 item 3.2.1-36 is only associated with PWRs; however, this MEAP (material/ environment/
aging effect/ program) is applicable in the engineered safety features systems at DNPS.
SLRA Section: B.2.1.24 - Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.24 B-135 The Program Description states, in part, Visual inspections for leakage or surface cracks will be performed to detect cracking of stainless-steel components exposed to a diesel exhaust environment as the detection of staining on the component external surface demonstrates the ability to detect leakage. An evaluation will be performed that demonstrates that As noted in the adjacent column, both the SLRA and program basis document indicate that if visual inspections are used in lieu of surface examinations for detecting leakage or surface cracking in stainless steel components, an evaluation will be
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 36 cracks will be detected prior to challenging the structural integrity or intended function of the component.
The program basis document (DR-PBD-AMP-XI.M38, Rev. 1) states, This program will monitor for cracking in stainless steel components exposed to diesel exhaust.
Visual inspections for leakage or surface cracks can also be performed as an acceptable alternative to conducting surface examinations to detect cracking if an evaluation is performed that demonstrates that cracks will be detected prior to challenging the structural integrity or intended function of the component. If this alternative is used, then the evaluation will be maintained in the program documentation. Stainless steel components exposed to an internal environment of exhaust gas include piping and piping components, expansion joints, and flexible connection[s] associated with the exhaust systems for the emergency diesel generators, station blackout diesel generators, diesel-driven fire protection pumps, and diesel-driven isolations condenser makeup pumps.
The GALL-SLR Volume 2 states the following under the XI.M38 AMP, Visual inspections are conducted where it has been analytically demonstrated that surface cracks can be detected by leakage prior to a crack challenging the structural integrity or intended function of the component. The SLRA includes an overview of the analytical method, input variables, assumptions, basis for use of bounding analyses, and results.
When using this option, cracks can be detected in gas-filled systems by methods such as, but not limited to: (a) for diesel exhaust piping, detecting staining on external surfaces of components; (b) for accumulators and piping connecting the performed and maintained in the program documentation.
However, the GALL-SLR states that any such evaluation(s) are included with the SLRA, including an overview of the analytical method, the input variables, assumptions, and basis for use of the bounding analyses and results.
The SLRA states that the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components AMP is consistent with AMP XI.M38 in the GALL-SLR, with no exceptions or enhancements.
Please provide the evaluations needed to be consistent with the GALL-SLR or revise the SLRA to include an exception and/or enhancement, as appropriate.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 37 accumulators to components, monitoring and trending accumulator pressures or refill frequency; and (c) soap bubble testing when systems are pressurized. The SLRA includes the specific methods used.
2 B.2.1.24
Background:
As shown below in B.2.1.24, the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program only manages hardening and loss of strength of elastomers. Aging effects associated with components within the scope of the Open-Cycle Cooling Water System program, the Closed Treated Water System program and the Fire Water System program are not managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program.
Carbon fiber reinforced polymer (CFRP) is not listed as a material that is managed within the Internal Surfaces in Miscellaneous Piping and Ducting Components program.
B.2.1.24 Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program Description The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program is a new condition monitoring program that will manage loss of material and cracking of metallic components, as well as loss of material, cracking, blistering, hardening and loss of strength of elastomeric materials. Reduction of heat transfer and flow blockage will also be managed. This program will consist of visual inspections of internal surfaces of piping, piping components, ducting, heat exchanger components, polymeric and elastomeric components, and other mechanical components.
Applicable environments include air, condensation, closed cycle cooling water, diesel exhaust, lubricating oil, raw water, treated water, and waste water. Visual The description in B.2.1.24 appears to be in conflict with information provided in Tables 3.4.1 and 3.4.2-1.
A word search of the SLRA did not reveal any Table 2 locations where flow blockage or cracking were being managed for CFRP.
Please clarify how CFRP will be managed in the Condensate System and the specific AERMs that will be managed.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 38 inspections for leakage or surface cracks will be performed to detect cracking of stainless steel components exposed to a diesel exhaust environment as the detection of staining on the component external surface demonstrates the ability to detect leakage. An evaluation will be performed that demonstrates that cracks will be detected prior to challenging the structural integrity or intended function of the component. Except for hardening and loss of strength of elastomers, aging effects associated with components within the scope of the Open-Cycle Cooling Water System (B.2.1.11) program, Closed Treated Water Systems (B.2.1.12) program, and Fire Water System (B.2.1.16) program will not be managed by this program.
Section 3.4.2.1.1 of the SLRA lists CFRP as a material in the Condensate System.
3.4.2.1.1 Condensate System Materials The materials of construction for the Condensate System components are:
- Aluminum Alloy
- Carbon and Low Alloy Steel Bolting
- Carbon Fiber Reinforced Polymer
- Carbon Steel
- Glass
- Stainless Steel
- Stainless Steel Bolting In Table 3.4.1, the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program is called out to manage hardening and loss of strength and loss of material for CFRP components in the Condensate System but flow blockage and cracking are
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 39 not mentioned, despite being listed as aging effects requiring management (AERMs).
In Table 3.4.2-1, the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program is called out to manage hardening or loss of strength, and loss of material for CFRP.
Table 3.4.2-1 does not list flow blockage and cracking as being managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program.
SLRA Section: B.2.1.26 - Monitoring of Neutron-Absorbing Materials Other Than Boraflex Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.26 B-143 and B-144 In SLRA section B.2.1.26 it states that the Monitoring of Neutron-Absorbing Materials Other Than Boraflex program is consistent with GALL-SLR. It also states that the coupons from Unit 3 are being used to cover Unit 2 since it has no coupons. GALL-SLR clearly outlines the surveillance and testing requirements for the absorber materials regardless of whether or not the pool has coupons. It is unclear to the staff how the program is consistent with GALL-SLR when testing of the absorber material is not being done for Unit 2.
Please discuss how the Monitoring of Neutron Absorbing Materials Other Than Boraflex program will be consistent with the GALL-SLR report for Unit 2 when testing is not being conducted on the absorber material or describe your plans to take exception(s) to GALL-SLR.
2 B.2.1.26 B-144 On page B-144 of the SLRA it states that there are 2 coupons remaining in the Unit 3 spent fuel pool. It also states that they are scheduled for removal and testing in 2031 and 2041. Considering the subsequent period of extended operation for Unit 3 will end in 2051 it is unclear to the staff how Constellation will ensure that there are an adequate number of coupons remaining for testing during the subsequent period of extended operation.
Please discuss what provisions are in place to ensure that there are an adequate number of coupons remaining for testing during the subsequent period of extended operation.
3 N/A N/A In the technical justification for why the coupons located in the Unit 3 pool are representative of the in-As shown in the analysis for the Water Chemistry parameters please explain
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 40 service panels in Unit 2 it is unclear to the staff how the coupons from Unit 3 are representative of Unit 2 in terms of cumulative dose and length in service.
how the Unit 3 coupons are representative of Unit 2 in the following areas:
Length of Service Cumulative Dose 4
N/A N/A In the technical justification for why the coupons located in the Unit 3 pool are representative of the in-service panels in Unit 2 there are many references to NEI 16-03-A Rev 1. It is unclear to the staff if Constellation is asking for the implementation of i-LAMP for Unit 2 or whether they are just asking to share coupons between Units 2 and 3.
Please clarify whether the implementation of i-LAMP is being requested for Unit 2 or is the sharing of coupons between Units 2 and 3 being requested.
SLRA Section B.2.1.33: Structures Monitoring Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.33 B-186
- 1. SLRA Section B.2.1.33 refers to ACI 349.3R, ACI 201.1R, and SEI/ASCE 11 in program description and Enhancements 9, 11, and 14 to the Structures Monitoring program. SLRA does not make clear which editions will be used for these codes.
- 2. SLRA Section B.2.1.33 states that the Structures Monitoring program includes elements of the Masonry Walls program and the Water-Control Structures program. SLRA Section B.2.1.32 states that the masonry walls program is implemented by the Structures Monitoring program. SLRA Section B.2.1.34 states that Water-Control Structures program is implemented through the Structures Monitoring program for the associated in-scope structures.
Program
Description:
- 1. Clarify in SLRA Section B.2.1.33 which editions will be used for ACI 349.3R, ACI 201.1R, and SEI/ASCE 11 in the Structures Monitoring program.
- 2. Explain what the word elements entails and discuss how the Masonry Walls and the Water-Control Structures programs are implemented by or through the Structures Monitoring program.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 41 GALL-SLR report considers these three programs as separate AMPs, and GALL-SLR XI.S6 AMP states that the scope may include inspection of masonry walls and water-control structures provided all the attributes of GALL-SLR Report AMP XI.S5 and GALL-SLR Report AMP XI.S7 are incorporated in the attributes of this program.
SLRA does not make clear what the word elements entails in the statement. It is unclear how the applicant implemented these three AMPs.
2 B.2.1.33 Table 3.5.2-9 B-186 to 194 3.5-164
- 1. SLRA is for DNPS, Units 2 and 3. DNPS Unit 1 shares the site and surrounding area with Units 2 and 3 to support operation of Units 2 and 3. SLR-DRE-INDEX Rev. 0, Dresden LR Structural List indicates that only these Unit 1 structures (Intake canal, Crib House, Chimney, Turbine Building) are within the scope of SLR. Enhancement 1 to the Structures Monitoring program add the Unit 1 Turbine building to the scope of the program. SLRA does not make clear what Unit 1 structures are already in the current Structures Monitoring program.
Enhancement 1 to the Structures Monitoring program also add Unit 3 Reactor Building Interlock to scope of the program. SLRA does not make clear whether Unit 2 Reactor Building Interlock is already included in the current Structures Monitoring program.
- 2. SLR-DRE-INDEX Rev. 0 indicates that TR 86 is within the scope of SLRA. However, SLRA lack its information and does not make clear whether it is within the scope.
Scope of Program:
- 1. Verify whether the Unit 1 Intake canal, Unit 1 Crib House, Unit 1 Chimney, and Unit 2 Reactor Building Interlock are already included in the current Structures Monitoring program.
- 2. Clarify what TR 86 and its SCs are, and what AMP(s) will be used to manage their aging effects, and whether the enhancement to AMP(s) is needed. In addition, provide Tables 1 and 2 AMR items for SCs of TR 86.
- 3. Clarify what/where stabilizers and truss arrangement listed in SLRA Table 3.5.2-9 are located and verify whether their aging effects will be managed by the ASME XI, Subsection IWF program instead of the Structures Monitoring program.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 42
- 3. SLRA Table 3.5.2-9 lists Table 2 AMR items associated with Table 1 AMR item 3.5.1-077 for steel components including stabilizers and truss arrangement, which their aging effects are managed by the Structures Monitoring program. UFSAR Figure 3.9-1 shows that stabilizers are used to support Class 1 reactor pressure vessel (RPV), and the ASME XI, Subsection IWF program requires to inspect these stabilizers.
If yes, provide an enhancement to the ASME XI, Subsection IWF program, Table 1 and 2 AMR items. If not, provide a justification. (This breakout session addresses the scope of the Structures Monitoring program only; rest can be addressed in the ASME XI, Subsection IWF program).
3 B.2.1.33 B-188 Enhancement 3 to the Structures Monitoring program describes shortening frequency for inspecting non-segregated bus ducts supports and the Reactor Vessel support skirt ring girder to an interval not to exceed five years. However, SLRA lacks Tables 1 and 2 AMR items for non-segregated bus ducts and their supports.
Reactor vessel support skirt is a Class 1 support, the ASME XI code requires it to be inspected under the ASME XI, Subsection IWF program and other AMPs (IWB, IWC, and IWD), NOT the Structures Monitoring program.
Scope of Program and Detection of Aging effects:
- 1. Clarify what non-segregated bus ducts and their supports are and provide their corresponding Tables 1 and 2 AMR items.
- 2. Clarify whether the reactor vessel support skirt can be removed from Enhancement 3 to the Structures Monitoring program. If yes, delete the wording. If not, provide a justification.
4 B.2.1.33 Table 3.5.2-9 B-188 3.5-164 SLRA Table 3.5.2-2, Component Supports lists Table 2 AMR items for sliding surfaces supporting ASME Class 1 to 3 piping and components, and Class MC components, that their aging effects are managed by the ASME XI, Subsection IWF program.
The enhancement 4 to the Structures Monitoring program includes sliding surfaces within the scope of the program.
SLRA claims AMR item 3.5.1-074 to be not applicable, however, Table 1 AMR item 3.5.1-076 is Scope of Program:
- 1. Clarify where the sliding surfaces supporting ASME related components are located.
- 2. Clarify whether Lubrite sliding surfaces listed in Table 3.5.2-9 and managed under the Structures Monitoring program are only sliding
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 43 applicable and SLRA Table 3.5.2-9 lists Table 2 AMR item associated with Table 1 item AMR 3.5.1-076 for sliding surfaces in the Primary Containment.
It is unclear where sliding surfaces under the ASME XI, Subsection IWF program are located.
surfaces located in the Primary Containment.
5 B.2.1.33 B-188 GALL-SLR AMP XI.S6 includes the preventive actions for proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting.
Section 3.2, Preventive Actions of AMP basis document will continue to address the above preventive actions in the existing procedures for Dresden and provided a list of referenced procedures. The referenced procedures state that these preventive actions are applicable to some other plants, not DNPS site. For example, Section 3.2.18 of MA-AA-736-600, Revision 15 describes these preventive actions for Peach Bottom.
In addition, AMP basis document states that the ASME F1852 and ASTM F2280 bolts are not used at DNPS, which needs to be clarified in SLRA.
Preventative Actions:
- 1. Clarify whether an enhancement to the Structures Monitoring program is needed for proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting. If yes, provide the enhancement. If not, provide the procedure demonstrating that it was implemented in the current Structures Monitoring program.
6 B.2.1.33 AMP basis document 3.5-186 Page 19 of 130
- 1. SRP-SLR report recommends parameters monitored or inspected for visual indications of aggregate reactions, such as map or patterned
- cracking, alkali-silica gel, exudations, surface staining, expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components.
Parameters monitored or inspected and Detection of Aging Effects:
- 1. Clarify whether an enhancement to the Structures Monitoring program is needed for parameters monitored or inspected for visual indications of aggregate reactions. If yes, provide an enhancement. If not, provide the
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 44 The staff reviewed plant procedures ER-AA-450, Revision 11, and ER-DR-450-1001, Revision 0 and other plant procedures, it appears that the existing procedures at DNPS do not have parameters monitored or inspected for visual indications of aggregate reactions recommended by SRP-SLR report.
- 2. GALL-SLR XI.S6 AMP states in the detection of aging effects that the program recommends examining representative samples of the exposed portions of the below-grade concrete, when excavated for any reason.
SLRA Section B.2.1.33 states the Structures Monitoring program includes provisions for opportunistic inspections of accessible below grade concrete structures.
SLRA Section 3.5.2.2.2 states that DNPS examines exposed portions of the below-grade concrete, when excavated for any reason, in accordance with the Structures Monitoring program, i.e. opportunistic inspection.
Section 3.4, Detection of Aging effects of AMP basis document does not address this subject. In addition, the staff could not locate this information in the plant procedures under the Structures Monitoring program.
procedure demonstrating that it was implemented in the current Structures Monitoring program.
- 2. Address the discrepancy of opportunistic inspection between AMP basis document and GALL-SLR XI.S6 AMP.
- 3. Clarify whether there is an existing procedure for opportunistic inspection.
If yes, provide a procedure. If not, provide an enhancement to the Structures Monitoring program.
7 B.2.1.33 B-188 Enhancement 6 to the Structures Monitoring program clarifies procedures to state that evidence of cracking, spalling, scaling, and leaching could indicate the presence of increased porosity and permeability.
Parameters monitored or inspected:
- 1. Clarify aging effects and aging mechanisms for this enhancement to
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 45 The staff searched GALL-SLR reports and identifies the following aging effects and aging mechanisms related to the increased porosity and permeability:
a) Increase in porosity and permeability; cracking; loss of material (spalling, scaling) due to aggressive chemical attack.
b) Increase in porosity and permeability; loss of strength due to leaching of calcium hydroxide and carbonation.
The staff finds that increase in porosity and permeability is already in the plant procedure ER-AA-450, Revision 11. In addition, this enhancement does not make clear of aging effects and aging mechanisms, other factors could also indicate the presence of increased porosity and permeability. The applicant is requested to re-evaluate this enhancement to ensure that enhanced procedure can be adequately implemented during the SPEO.
ensure the consistency with GALL-SLR report.
- 2. Discuss what aging mechanisms could cause increased porosity and permeability using OEs as examples and clarify all potential evidence that could indicate the presence of increased porosity and permeability.
8 B.2.1.33 Table 3.5.2-3 Table 3.5.2-13 B-189 3.5-130 3.5-190 Table 3.5.2-13, Structural commodity group, lists elastomeric penetration Seals, seals, gaskets, and moisture barriers (including caulking, flashing and other sealants), that their aging effect of loss of sealing is managed by the Structures Monitoring program. It is unclear what sealants are included in this line item.
Enhancement 8 to the Structures Monitoring program includes monitoring elastomeric structural sealants, seismic joint fillers, vibration isolators, and bearing pads for cracking, loss of material, and hardening.
Parameters monitored or inspected, Detection of Aging Effects, and Acceptance Criteria:
- 1. Discuss in general where elastomeric structural sealants, seismic joint fillers, and vibration isolators are used, and their aging effects and aging mechanisms.
- 2. Provide Tables 1 and 2 AMR items for elastomeric structural sealants,
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 46 SLRA lacks Table 1 and 2 AMR items for elastomeric structural sealants, seismic joint fillers, and vibration isolators.
seismic joint fillers, and vibration isolators.
9 B.2.1.33 B-189 GALL-SLR XI.S6 AMP states in the Parameters monitored or inspected program element that leakage volumes and chemistry are monitored and trended for signs of concrete or steel reinforcement degradation if through-wall leakage or groundwater infiltration is identified. GALL-SLR XI.S6 AMP states in the Detection of Aging Effects program element that indications of groundwater infiltration or through-concrete leakage are assessed for aging effects.
Section 3.4.c of AMP basis document states, Indication of groundwater infiltration and through-concrete leakage are assessed and evaluated for aging effects as part of the current Structures Monitoring program. The staff could not locate this procedure in the current Structures Monitoring program on the Portal.
Enhancement 10 to the Structures Monitoring program states, Monitor and trend through-wall groundwater leakage infiltration volumes, which is inconsistent with GALL-SLR report.
Parameters monitored or inspected and Detection of Aging Effects:
- 1. Provide the existing procedure for the assessment and evaluation of indication of groundwater infiltration and through-concrete leakage.
- 2. Clarify the wording through-wall groundwater leakage infiltration volumes to ensure its consistency with GALL-SLR report.
10 B.2.1.33 B-190 GALL-SLR XI.S6 AMP states in the monitoring and trending program element that quantitative baseline inspection data should be established per the acceptance criteria described herein prior to the subsequent period of extended operation and previously performed inspections that were conducted using comparable acceptance criteria specified herein are Monitoring and Trending:
- 1. Clarify in the SLRA what structures had inadequate prior inspection results that could not establish a quantitative baseline so that procedures can be revised accordingly and will be implemented adequately during the SPEO.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 47 acceptable in lieu of performing a new baseline inspection.
Enhancement 13 to the Structures Monitoring program states establishing quantitative baseline inspection data prior to the subsequent period of extended operation for structures in which prior inspection results are inadequate to establish a quantitative baseline to allow for effective monitoring and trending.
SLRA does not make clear what structures had inadequate prior inspection results that could not establish a quantitative baseline and how their quantitative baseline inspection data are to be established.
- 2. Explain in the SLRA what acceptance criteria will be used for these structures to establish quantitative baseline inspection data prior to the subsequent period of extended operation.
11 B.2.1.33 B-190 GALL-SLR XI.S6 AMP states in the Acceptance Criteria program element that loose bolts and nuts are not acceptable unless accepted by engineering evaluation.
Enhancement 15 to the Structures Monitoring program clarifies cracked bolts to be accepted by engineering evaluations.
It appears that enhancement 15 to the Structures Monitoring program is inconsistent with the acceptance criteria program elements described in GALL-SLR XI.S6 AMP.
Acceptance Criteria:
Clarify whether cracked bolts can be removed from Enhancement 15 to the Structures Monitoring program. If yes, delete wording cracked bolts. If not, provide a justification.
12 B.2.1.33 B-190 Both chimneys (Unit1, Units 2 and 3) are tapered, reinforced concrete structures.
Operating Experience (Chimneys):
- 1. Discuss the history and timeline of degradations of chimneys and provide
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 48 During onsite audit, the staff observed many discolored spots (surface efflorescence - white and black) and some minor spalling at the exterior surfaces of chimneys from the ground to about 25 feet elevations below the vent stack. The applicant stated that visual inspections of these two chimneys were performed annually using drones and binoculars, etc.
It is unclear what caused these degradations and how aging effects of these chimneys will be adequately managed and their intended functions will be maintained during the SPEO inspection records (IRs and WOs) for Chimney (Unit 1) and Chimney (Units 2 and 3) for the past 10 years.
- 2. Describe corrective actions being taken, if any.
- 3. Provide inspection procedures using both binoculars and drones for visual inspections of chimneys.
- 4. Provide demand/capacity radios of reinforced concrete chimneys at bottom of the stacks and at the bases of chimneys from the structural calculations and clarify their concrete design compressive strength.
- 5. Clarify whether compressive strength of concrete at bases of chimneys using NDT or cores has been tested in the past. If yes, provide testing results.
- 6. Provide root and cause analysis of degradation/aging mechanisms and chemistry composition that is sampled from a location that is representative of the degradations (i.e. discolored spots) and clarify whether evaluation exists and can determine that structural integrity of reinforced concrete chimneys is maintained.
- 7. Explain how their aging effects will be adequately managed and their
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 49 intended functions will be maintained during the SPEO.
- 8. Update SLRA and OE in SLRA Section B.2.1.33 accordingly as needed.
13 B.2.1.33 Table 3.5.2-13 Table 3.5.2-2 3.5.2.2.2.3 B-190 3.5-188 3.5-120 3.5-50 SLRA Table 3.5.2-13 lists Table 2 AMR item associated with AMR item 3.5.1-095 (III.B2.TP-8) for galvanized steel cable trays exposed to air-indoor, uncontrolled environment requiring no aging management. SLRA Table 3.5.2-2 lists Table 2 AMR item associated with AMR item 3.5.1-095 (III.B2.TP-
- 8) for galvanized steel supports of the cable trays exposed to air-indoor, uncontrolled environment requiring no aging management.
SLRA Section 3.5.2.2.2.3, item 3 states that inaccessible areas of concrete elements for Group 6 structures are subject to the aggressive environment.
During onsite audit, the staff observed significant degraded conditions at cable trays and their supports, exposed to air-indoor, uncontrolled environment, at exterior concrete walls below the grade at multiple locations in Units 2 and 3 Crib House, and cables run through multiple south wall penetrations. The staff noted that groundwater can go through cable penetrations at exterior concrete walls and land on cable trays and their supports.
DNPS has plant-specific conditions in Units 2 and 3 Crib House that these galvanized steel cable trays and their supports do require aging management.
SLRA needs to be updated to reflect this condition.
Operating Experience (Cable trays and their supports in Units 2 and 3 Crib House):
- 1. Discuss the history and timeline of degradation (corrosion) and its past corrective actions and/or replacement activities for cable trays and their supports in Units 2 and 3 Crib House and provide their IRs and WOs for the past 10 years.
- 2. Describe acceptance criteria used for corrective actions and how corrective actions were implemented for cable trays and their supports.
- 3. Discuss whether cable trays and their supports in Units 2 and 3 Crib House will be replaced prior to entering the SPEO. If they will be replaced, describe how the design and installation of cable trays and their supports will be different to ensure that corrosion like this would not happen and cable trays and their supports will maintain their intended functions during the SPEO.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 50 Regarding these observed significant degraded conditions in the large portion of cable trays and their supports in Units 2 and 3 Crib House, it is unclear how aging effects were managed and their intended functions will be maintained during the SPEO.
- 4. Evaluate what enhancements to the Structures Monitoring are needed to manage aging effects of the cable trays and their supports in Units 2 and 3 Crib House so that their intended functions will be maintained during the SPEO.
- 5. Provide Table 2 AMR items for SCs related to cable trays and their supports within Units 2 and 3 Crib House, which require aging management.
- 6. Update SLRA and OE in SLRA Section B.2.1.33 accordingly as needed.
14 B.2.1.33 B-190 GALL-SLR XI.S6 AMP states in the Detection of Aging Effects program element that the program includes provisions for more frequent inspections based on an evaluation of the observed degradation.
GALL-SLR XI.S6 AMP also states in the Corrective Actions program element that inspection frequencies are adjusted as determined by the sites corrective action program if any projected inspection results will not meet acceptance criteria prior to the next scheduled inspection.
During onsite audit, the staff observed significant damage and vegetation growth at many areas in Turbine Building roof pavers that could eventually lead to roof leakage.
The staff observed Containment roofs at Units 2 and 3 that were replaced in 2018 and finds them in good condition. Using Units 2 and 3 Containment roof as Operating Experience (Roofs):
- 1. Discuss the history and timeline of what building roofs had been replaced, describe the conditions of building roofs that have not been replaced, and clarify when and what building roofs will be replaced in the future.
- 2. Describe in general what corrective actions of building roofs had been done and provide their IRs and WOs.
- 3. If building roofs with significant degradation are not to be replaced, evaluate whether more frequent inspections are needed by providing a justification and explain how aging effects will be adequately managed
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 51 benchmark, the applicant is requested to evaluate whether rest of building roofs need to be replaced prior to entering SPEO and how their aging effects will be adequately managed by the Structures Monitoring program during the SPEO.
and their intended functions will be maintained during the SPEO.
- 4. Revise SLRA and OE item 4 in SLRA Section B.2.1.33 accordingly as needed.
15 3.5.2.2.2.1 B.2.1.33 3.5-45 B-186 SLRA Section 3.5.2.2.2.1, item 4 defines DNPS site as an aggressive environment and states that high chloride levels are a concern as a potential initiator of reinforcing steel corrosion that could be initially detected as cracking and spalling of concrete. It also states that test results for groundwater and raw water samples taken in 2023 confirm chloride levels above the 500 ppm threshold that defines an aggressive environment, and the groundwater at DNPS is not aggressive with respect to pH or sulfates, in which both statements conflict with each other.
SLRA Section B.2.1.33 states that inspection frequency for the in scope structures will not exceed five years, with provisions for more frequent inspections when conditions are observed that have a potential for impacting an intended function.
During OE audit and onsite audit, the staff noted through-wall leakage or groundwater infiltration including wall penetrations below grade resulted in significant corrosion or degradation of SCs under the Structures Monitoring program, such as: conduit and conduit supports in Unit 2 torus basement area, cable trays and their supports in Units 2 and 3 Crib House, Steel beams in the Unit 3 West Condenser pull pit, concrete basement wall in Units 2 and 3 Crib House, and so on.
Operating Experience (Through-wall leakage or Groundwater infiltration):
- 1. Discuss the applicability of aggressive environment with respect to pH or sulfates.
- 2. Discuss in general the history of through-wall leakage or groundwater infiltration and their degradation identified; corrective actions taken.
Provide IRs and WOs that have not posted on Portal.
- 3. Provide root and cause analysis of through-wall leakage or groundwater infiltration including cable penetrations in the south basement wall in Units 2 and 3 Crib House, describe what actions will be taken to avoid or mitigate future degradation.
- 4. Evaluate whether current condition of interior basement wall in Units 2 and 3 Crib House allows for effective monitoring and trending.
- 5. Evaluate whether one time inspection of interior concrete wall
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 52 During onsite audit, the staff observed significant signs of dark color at multiple locations in Units 2 and 3 Crib House, especially in the fish and leaf removal area, multiple cable penetrations in the south basement wall, and stagnant water at the basement floor trough along the interior south wall. All of these indicate through-wall water leakage.
Enhancement 9 to the Structures Monitoring program states, As part of the engineering evaluations, determine if additional actions are warranted, which might include enhanced inspection techniques and/or increased frequency, destructive testing, and focused inspections of representative accessible (leading indicator) or below grade, inaccessible concrete structural elements exposed to the potentially aggressive environment.
Enhancement 9.c to the Structures Monitoring program states, Develop the initial engineering evaluations prior to the subsequent period of extended operation. SLRA table A.5, item 30 sates program will be enhanced and the initial engineering evaluation will be completed no later than six months prior to the subsequent period of extended operation.
These enhancements to the Structures Monitoring program were approved by the NRC for other plants without significant OEs like DNPS.
GALL-SLR XI.S6 AMP states in the Detection of Aging Effects program element that the program includes provisions for more frequent inspections based on an evaluation of the observed degradation.
GALL-SLR XI.S6 AMP also states in the Corrective Actions program element that inspection frequencies below grade will be performed and any identified degradations will be entered into corrective actions for disposition prior to entering the SPEO.
- 6. Evaluate whether more frequent inspections are needed for interior concrete walls below grade. If yes, clarify the inspection frequency. If not, provide a justification.
- 7. Clarify whether there is an evaluation that determines that the observed degradation such as leaching of calcium hydroxide and carbonation in accessible areas has no impact on the intended function of the concrete structure and clarify the degradation requiring corrective actions.
- 8. Clarify initial engineering evaluations prior to the SPEO and evaluate Enhancement 9 to the Structures Monitoring program to determine what changes in the enhancement can be made to reflect DNPS plant-specific conditions.
- 9. Revise SLRA, Enhancement 9 to the Structures Monitoring program, OE item 2 in SLRA Section B.2.1.33 accordingly as needed.
Note: these breakout questions also apply to the Water-Control Structures program.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 53 are adjusted as determined by the sites corrective action program if any projected inspection results will not meet acceptance criteria prior to the next scheduled inspection.
Based on current conditions of concrete basement walls using Units 2 and 3 Crib House as an example, the applicant is requested to evaluate what enhancements/modifications in SLRA and actions are needed to ensure no loss of structural concrete component intended function during the SPEO.
16 B.2.1.33 B-192 Operating experience, item 2 describes that the Structures Monitoring program identified in June 2013 water in-leakage resulting in corrosion of several conduit and conduit supports, a LPCI support adjacent to the LPCI torus ring header drain valve, and a reactor building equipment drain tank (RBEDT) gallery steel support and states these were acceptable with deficiencies. Reinspection in 2018 showed that three additional supports were degraded and were again acceptable with deficiencies.
The SLRA Structures Monitoring basis document (DR-PBD-AMP-XI.S6, Structures Monitoring, Revision 2) references EC 624174, Revision 0, Analysis of LPCI Drain Line 2-1542A-1.5/2-L with and without Supports M-3430-H1 and M-3430-H3.
EC 624174 does not provide enough information for the staffs conclusions on the ability of the applicants proposed AMP to manage the effects of aging in the subsequent period of extended operation.
Operating Experience (EC 624174):
- 1. Provide supporting documents (calculations, inspection reports, WOs, photos, etc.) on the Portal.
- 2. Provide root and cause analysis of through-wall water leakage source and describe what corrective actions were done or will be taken to avoid or mitigate future degradation at this location.
- 3. Discuss this operating experience using illustrated diagrams or constructed drawings/photos and explain the implementing process of the program and its current conditions on the affected concrete walls and various supports.
17 Table 3.5.2-2 Table 3.5.2-5 Table 3.5.2-9 3.5-107 3.5-108 3.5-110
- 1. GALL-SLR report indicates that NUREG-2191 Items III.B1.2.TP-42 and III.B1.3.TP-42 are for concrete components supporting the ASME Class 2 Correct NUREG-2191 Item numbers for Table 2 AMR items associated with AMR items 3.5.1-055, 3.5.1-065, and
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 54 Table 3.5.2-10 Table 3.5.2-12 Table 3.5.2-14 Table 3.5.2-15 3.5-144 3.5-159 3.5-170 3.5-184 3.5-195 3.5-201 and 3 piping and components and the ASME Class MC components, respectively.
SLRA Table 3.5.2-2 on Pages 3.5-107 and 3.5-110 lists Table 2 AMR item associate with AMR item 3.5.1-055 for concrete components supporting ASME Class 2 and 3 piping and components and the ASME Class MC components, using NUREG-2191 Item II.B.1TP-42, which is inconsistent with GALL-SLR report.
- 2. GALL-SLR report utilizes NUREG-2191 Item III.A3.TP-212 for inaccessible concrete components in group 3 structures. However, SLRA Table 3.5.2-5 on Page 3.5-144 lists Table 2 AMR item associated with AMR item 3.5.1-065 for concrete: above grade exterior (inaccessible) by using NUREG-2191 Item III.A6.TP-104.
3.The staff identified similar issues in a) SLRA Table 3.5.2-9 on Page 3.5-159 for Table 2 AMR item associated with AMR item 3.5.1-065 for concrete: interior (inaccessible areas - drywell floor and reactor pedestal) by using NUREG-2191 Item III.A6.TP-104, which primary containment is not Group 6 structure; b) SLRA Table 3.5.2-10, 3.5.2-11, 3.5.2-12, and 3.5.2-14 for Table 2 AMR items associated with AMR item 3.5.1-065 for concrete:
above grade exterior (inaccessible) by using NUREG-2191 Item III.A6.TP-104 (for Group 6 Structures), which Radwaste Structures, Reactor Building, Stacks, Switchyard Structures, and Turbine Building are not Group 6 structures.
- 4. GALL-SLR report indicates that NUREG-2191 Items III.B1.2.TP-8 is for galvanized steel support 3.5.1-095 to be consistent with GALL-SLR report.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 55 members; welds, bolted connections, etc. supporting the ASME Class 2 and 3 components.
SLRA Table 3.5.2-2 on Page 3.5-108 list Table 2 AMR item associate with AMR item 3.5.1-095 for galvanized steel various hangers supporting ASME Class 2 and 3 Piping and Components, using NUREG-2191 Item III.B1.1.TP-8, which is inconsistent with GALL-SLR report.
In conclusion, there are discrepancies between SLRA and GALL-SLR report.
18 2.4.3 2.4.4 Table 3.5.2-3 Table 3.5.2-4 NUREG-2191, Vol. #1 2.4-14 2.4-17 3.5-126 to 130 3.5-132 to 141 pdf pages 134-138 SLRA Section 2.4.3 describes the intake canals and the discharge outfall structures are included within the evaluation boundary of the Cooling Water Structures and determined to be within the scope of license renewal. Cooling Water Structures are Group 6 structures.
SLRA Tables 3.5.2-3 list summary of aging management evaluation. The staff noted that SLRA Table 3.5.2-3 utilizes the following NUREG-2191 Items for Cooling Water Structures:
III.A3.TP-23 III.A3.TP-24 III.A3.TP-26 III.A3.TP-67 III.A3.TP-108 III.A3.TP-261 III.B4.TP-44 In addition, SLRA Section 2.4.4 describes both Crib Houses to be within the scope of license renewal, and Crib Houses are Group 6 structures.
- 1. Evaluate and correct each Table 2 AMR item in SLRA Tables 3.5.2-3 and 3.5.2-4 using NUREG-2191 Items (III.A3.TP-XX, III.B4.TP-44, and III.B5.T-37b).
- 2. Evaluate its impact of potential Table 2 AMR item revisions on SLRA Section 3.5.2.2.2 and revise the SLRA accordingly.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 56 SLRA Tables 3.5.2-4 list summary of aging management evaluation. The staff noted that SLRA Table 3.5.2-4 utilizes the following NUREG-2191 Items for Crib Houses:
III.A3.TP-108 III.A3.TP-302 III.B5.T-37b The staff finds that the NUREG-2191 Items listed above and/or their AMPs for Table 2 AMR items in SLRA Table 3.5.2-3 and 3.5.2-4 are inconsistent with NUREG-2191, Vol.1, Table A6 for Group 6 Structures.
Furthermore, potential revisions of some Table 2 AMR items due to different AMPs may have impact on SLRA Section 3.5.2.2.2.
19 Table 3.5.2-4 Table 3.5.2-2 Table 3.5.2-11 3.5-132 to 141 3.5-110 3.5-179
- 1. SLRA Table 3.5.2-4 lists many Table 2 AMR items for Units 2 and 3 Crib House only. Unit 1 Crib House is also within the scope of SLR. SLRA does not make clear whether there SCs are applicable to Unit 1 Crib House.
- 2. SLRA Table 3.5.2-2 lists only one (1) Table 2 AMR item associated with AMR item 3.5.1-055 (III.B1.1.TP-42) for grout. Other NUREG-2191 Items (III.B1.2.TP-42, III.B1.3.TP-42, III.B2.TP-42, III.B3.TP-42, III.B4.TP-42, and III.B5.TP-42) also list grout materials. SLRA does not make clear whether those NUREG-2191 Items are applicable to SCs in DNPS.
- 3. SLRA Table 3.5.2-11 lists Table 2 AMR item associated with AMR item 3.5.1-066 for hatches/plugs, using both NUREG-2191 Items
- 1. Clarify whether these SCs for Units 2 and 3 Crib House only in SLRA Table 3.5.2-4 are applicable to the Unit 1 Crib House. If yes, provide corresponding Table 2 AMR items accordingly.
- 2. Evaluate the applicability of NUREG-2191 Items (III.B1.2.TP-42, III.B1.3.TP-42, III.B2.TP-42, III.B3.TP-42, III.B4.TP-42, and III.B5.TP-42) and provide Table 2 AMR items as needed.
- 3. Clarify whether NUREG-2191 Item III.A4.TP-26 can be changed into NUREG-2191 Item III.A2.TP-26 for hatches/plugs in SLRA Table 3.5.2-11.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 57 III.A4.TP-26 and III.A2.TP-26, exposed to air-indoor, uncontrolled and air-outdoor environment, respectively. The staff finds that NUREG-2191 Item III.A2.TP-26 covers both air-indoor, uncontrolled and air-outdoor environment. Therefore, NUREG-2191 Item III.A4.TP-26 is used incorrectly.
20 2.4.13 Table 3.5.2-13 2.4-58 3.5-192 SLRA Section 2.4.13 states that structural miscellaneous steel includes platforms, catwalks, grating, stairs, ladders, steel curbs, handrails, permanent scaffolding, kick plates, decking, and sump covers, as well as roof scuttles, blowout panels, fixed louvers, and vents.
SLRA Table 3.5.2-13 lists the component type as Structural Miscellaneous - catwalks, grating, handrails, kick plates, ladders, platforms, stairs, etc.
There is discrepancy between SLRA Section 2.4.13 and Table 3.5.2-13.
Address the discrepancy between SLRA Section 2.4.13 and Table 3.5.2-13 and revise the SLRA accordingly.
LRA Section B.2.1.34, A.2.1.34 - Inspection of Water-Control Structures Associated with Nuclear Power Plants Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
SLRA 2.4.1 2.4-2 SLRA Section 2.4.1 lists Circulating Water Inlet Tunnel within the scope of SLR. However, the staff finds that the SLR-DRE-INDEX Rev. 0 and the Dresden license renewal structural list do not include Circulating Water Inlet Tunnel within the scope of SLR.
In addition, SLRA Section 2.4.1 list UFSAR reference Section 9.2.5, which is related to Ultimate Heat Sink. The provided UFSAR Scope of Program:
Address the discrepancy between the program scope and SLR boundary drawings and clarify whether Circulating Water Inlet Tunnel is indeed in the scope of SLRA.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 58 reference section is not for Circulating Water Inlet Tunnel.
Provide appropriate UFSAR reference section for Circulating Water Inlet Tunnel 2
SLRA 2.4.3 B.2.1.34 2.4-10 B-195 Both SLRA Sections 2.4.3 and B.2.1.34 state that the intake canals (including the associated bridge over the Units 2 and 3 intake canal) and the discharge outfall structure (including ice melt gate and deicing line connecting the discharge headworks to the Units 2 and 3 Crib House forebay) are in scope for the condition monitoring program SLR-DRE-INDEX Rev. 0 lists Units 1, 2 and 3 intake canals, discharge headworks and Crib House forebay within the scope of SLR.
However, Section 2.4.3 and SLRA B.2.1.34 do not make clear of the scope of these structures.
Scope of Program:
- 1. Address the discrepancy of scope between SLR and boundary drawing and clarify if discharge headworks and crib house forebay are within scope of SLR.
- 2. If yes, clarify what SCs are subject to aging management for the unit 2/3 Crib House forebay and discharge headworks.
3 SLRA 2.4.4 B.2.1.34 2.4-20 SLRA Table 2.4-4, Crib Houses lists concrete foundation for Units 2 and 3 only. Unit 1 Crib House is also in-scope, why is the Unit 1 Crib House concrete foundation is not included?
Scope of Program:
Confirm if Unit 1 Crib House foundation should be included on Table 2.4-4 4
B.2.1.34 B-197 GALL-SLR Report AMP XI.S7, Scope of Program program element states in part, the scope of the program also includes miscellaneous steel, such as sluice gates.
There are no Tables 1 and 2 AMR items in SLRA for steel sluice gates.
SLRA does not make clear whether steel sluice gates are present at DNPS site.
Scope of Program:
Clarify whether steel sluice gates are present at the DNPS site.
If yes, clarify whether the enhancement to the Water-Control Structures program is needed to include steel sluice gates and provide Tables 1 and 2 AMR items accordingly.
5 B.2.1.34/
- 1. Enhancement 3 to the Water-Control Structures program describes structural bolting fabricated from ASTM A325 and ASTM A490.
But Section 3.2 in the AMP basis document Preventive Actions:
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 59 states, Structural bolting including ASTM 325, ASTM 1852, and ASTM A490 bolts were not used for components within the scope of Inspection of Water Control Structures. There is a discrepancy between SLRA and PBD.
- 2. GALL-SLR AMP XI.S7 states, The preventive actions emphasize proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting.
Section 3.2 of PBD claims that it will continue to be addressed in the existing procedures for Dresden. The staff reviewed its referenced procedures such as CC-AA-103-1001; PES-S-003; PES-S-010; MA-AA-736-600; MA-AA-410) and finds no implemented procedures at DNPS.
F2280) are used at the water control structures.
- 2. Clarify in SLRA whether the enhancement to the Water-Control Structures program is needed to include proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting.
6 B.2.1.35/
AMP Basis XI.S7 AMP Sec. 3.3 Section 3.3 (d) in the AMP basis document DR-PBD-AMP-XI.S7 states in part, Section C of RG 1.127 that applies to parameters monitored at Dresden is in subsection c.2.a and c.2.e.
GALL-SLR Report AMP XI.S7 Parameters Monitored or Inspected program element states in part, parameters to be monitored for earthen embankment structures include settlement, depressions, sink holes, slope stability, seepage, proper function of drainage systems, and slope protection degradation...
Submerged emergency canals monitored for sedimentation, debris, or instability of slopes.
- 1. Subsection C.2 of RG1.127 does not contain subsections a nor e Parameters Monitored or Inspected:
- 1. Clarify if the referenced subsections in RG 1.127 are supposed to be c.5.a, concrete structures in general, and c.5.e, Cooling-Water Channels and Canals and Intake and Discharge Structures.
- 2. Clarify what parameters are monitored or inspected for earthen embankment structures at DNPS.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 60
- 2. There is no discussion of parameters monitored on earthen embankment structures 7
B.2.1.34/
AMP Basis XI.S7 B-196 GALL-SLR AMP XI.S7 requires periodic inspections to be performed at least once every 5 years.
SLRA Section B.2.1.34 states no exceptions to NUREG-2191.
SLRA Section B.2.1.34 Program Description stated, Inspections will be performed at least once every five years for structural components that are not submerged. Submerged components in the discharge outfall structure are visually inspected at least once every 10 years, which is inconsistent with GALL-SLR report.
Detection of Aging Effects:
- 1. Address the discrepancy of inspection frequency for submerged components in the discharge outfall structure, and provide the enhancement as needed.
- 2. Confirm that all other in scope SCs are inspected at least every 5 years 8
B.2.1.34 B-196 SLRA B.2.1.34 notes, Structures located underwater will not be accessible for evaluation with the same level of visual acuity as structures above water. Inspections will be implemented that establish the condition of these structures by using divers, an equivalent inspection method to divers, or by dewatering, as well as by considering the conditions of the exposed portions of the structures, at the waterline and above, which can serve as an indicator of conditions underwater.
It is unclear what is the condition of the underwater structures for Water-Control Structures, and what procedures were used to inspect the underwater portions.
- 1. Provide procedures for diver inspections on the Portal.
- 2. Provide the divers inspection reports and engineering evaluation(s), if applicable, for the past 10 years at DNPS.
9 B.2.1.34/
AMP Basis XI.S7 AMP Sec. 3.6 Enhancement 8 to the Water-Control Structures program states, Perform inspections to Acceptance Criteria:
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 61 establish quantitative baseline inspection data prior to the subsequent period of extended operation for structures in which prior inspection results are inadequate to establish a quantitative baseline to allow for effective monitoring and trending.
The word structures in the enhancement is vague.
Clarify what structures and components will be inspected to establish quantitative baseline inspection data prior to the subsequent period of extended operation.
10 B.2.1.34/
AMP Basis XI.S7 AMP Sec. 3.6 Enhancements 9 and 10 to the Water-Control Structures program refers to ACI 349.3R in the evaluation criteria, but no version of the ACI 349.3R is listed. SLRA Section B.2.1.34 states that the aging management program is based on the guidance provided in NRC RG 1.127 and American Concrete Institute (ACI) 349.3R-02.
Acceptance Criteria:
Clarify in SLRA whether ACI 349.3R-02 will continue to be used in the enhancements 9 and 10 to the Water-Control Structures program.
11 B.2.1.34/
AMP Basis XI.S7 AMP Sec. 3.6 GALL-SLR AMP XI.S7 Acceptance Criteria program element states in part, Loose bolts and nuts,... are accepted by engineering evaluation or subject to corrective actions.
Enhancement 11 to the Water-Control Structures program states, Clarify that loose bolts and nuts and cracked bolts are not acceptable unless accepted by engineering evaluations, which is inconsistent with GALL-SLR report.
Acceptance Criteria:
Address the discrepancy of cracked bolts between the program and GALL-SLR report.
12 Table 3.5.2-3 Table 3.5.2-4 SLRA Table 3.5.2 items referencing item number 3.5.1-60 states under the AERM column, loss of material and cracking Item 3.5.1-60 states, Loss of material (spalling, scaling) and cracking due to freeze-thaw is same but due to "freeze thaw".
AMR Review:
Clarify if the Table 3.5.2 items referencing Item 3.5.1-060 are for aging related to freeze thaw 13 AMR 3.5.1-062 Table 3.5.2-4 SLRA Table 3.5.2-4 describes the stop logs associated with item number 3.5.1-62 function AMR Review:
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 62 to retain water, and lists the environments air-outdoor and water - flowing.
- 1. Clarify if additional environments such as Water-standing and Raw Water are also applicable to the stop logs 14 Site Audit Observations During the on-site audit, the following structural concerns were noted inside the Unit 2/3 WCS:
- 1. One steel support column has a 3 anchor bolt configuration in the baseplate
- 2. Service water pump support base plate has 4 bolts with less than the required minimum edge distance
- 3. The gutter along the perimeter of the wall base appears to contain standing water with debris accumulated in the gutter. This condition was reported numerous times in ARs on the portal with recommendations to hydroblasting the unclogged drain. It is unclear if gutter and drain maintenance have been performed on schedule.
Operating Experience:
- 1. Provide engineering analysis that demonstrate the column base plate is structurally adequate for the controlling load combinations.
- 2. Verify the water pump support base plate is structurally adequate for the controlling load combinations.
- 3. Provide WO reports that demonstrate the clogged drain has been repaired. Provide WO reports and maintenance procedures that demonstrate the gutters/drain are routinely cleaned.
For SMP AR 04400278 In the AR report, It was noted during the walkdown around 02/04/2021 that the pipe support for the TBCCW is badly degraded and pulling away from the wall.
The closure notes by the engineer stated, the as-found condition of the angles exhibit minor localized separation of the base plate and concrete wall due to rusk pack and concrete spalling and concluded: The field evaluation performed and creation of ECR 450100 meets the intent of this assignment.
Operating Experience:
- 1. Provide evaluation report ECR 450100 on the portal for review.
- 2. Verify if the degradation of the pipe support has not advanced since the field evaluation was performed
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 63 SLRA Sections: A.2.1.29 & B.2.1.29 ASME Section XI, Subsection IWE AMP & AMR Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.29 B-161 to B-163 SLRA B.2.1.29 under Justification for Exception states in part: The program will be enhanced to perform surface examinations or enhanced visual examination (e.g., EVT-1) on accessible portions of one Main Steam (MS)
System and one Feedwater (FW) System drywell penetration subject to cyclic loading but have no CLB fatigue analysis..One penetration for MS system and one for FW system will be selected to represent the remaining population of penetrations for high temperature process lines which include the following There is also a corresponding enhancement 1 to the program in the SLRA.
It is not clear from the SLRA the technical justification for adequacy of the proposed sample size of one MS and one FW, and whether it is for each unit or otherwise. It is also not clear how the sample size is consistent is consistent with the representative sample size of 20 percent of the population for each unit recommended in the GALL-SLR Report (e.g.,
AMP XI.M32, XI. M18).
a) Provide the justification for adequacy of the SLRA proposed sample size to detect cracking due to cyclic loading for components where a fatigue analysis or waiver does not exist.
Describe how it would be consistent with the representative sample size recommended in the GALL-SLR Report, and clarify if the sample size applies to each unit.
b) Provide necessary SLRA B.2.1.29 and A.2.1.29 updates, including enhancement 1, consistent with the response above.
2 B.2.1.29 B-163 SLRA includes enhancement to implement one-time supplemental volumetric examination of containment shell surfaces that are inaccessible from one side, if triggered by plant-specific OE of measureable corrosion initiated on the inaccessible side since issuance of the first renewed license.
a) Confirm and explain with the basis whether the triggering OE for this supplemental one-time volumetric examination has occurred to date or not for DNS Unit 2 and/or Unit 3. Update the SLRA accordingly.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 64 It is not clear from the SLRA if the triggering OE has occurred in Dresden Units 2 and/or 3.
3 DR-PBD-AMP-XI.S1, Section 3.5 32 Program Basis Document (PBD) states on page 32: Dresden Unit 3 was chosen for drilling of core bore holes [for augmented UT examinations] because of the existence of conditions that are considered to be the most potentially corrosive of the two units.
It is not clear from the SLRA if the above continues to be the case.
It is also not clear what areas of the DNPS containments are identified for augmented examination in the CISI Program Plan for the current interval and which will continue into the SPEO.
a) Confirm and explain if DNS Unit 3 continues to be representative or bounding of the two DNS units for augmented examination for corrosion in inaccessible side of the DW shell in the sand bed areas.
b) Describe the areas of the DNS Unit 2 and Unit 3 drywell and torus that are identified for augmented examination in the IWE program and whether they will continue into the SPEO.
c) Update the SLRA as necessary based on the above responses.
4 DR-PBD-AMP-XI.S1, Section 3.5; SLRA B.2.1.29 PBD 33; SLRA B-165, B-166 SLRA and PBD Section 3.5 states that: A monitoring program consisting of UT thickness measurements at various locations of the cylindrical and spherical areas of the drywell interior is performed at Dresden Unit 3.
..Forty (40) locations have been inspected since 2008 and only one (1) location required additional attention... This location will continue to be inspected by means of a 1 x 1 grid similar to what was performed in 2022 to continue trending this location.
It is not clear if UT measurements for the cylindrical and spherical shell areas of the a) Clarify if UT examinations of spherical and cylindrical areas of the drywell will continue into the SPEO, and, if so, how they are tracked by the AMP (e.g.,
identified in AMP for augmented examination). Update the SLRA accordingly.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 65 drywell will continue into the SPEO, and whether one or more such locations are identified for augmented examination.
5 3.5.2.2.1.6, B.2.1.29 B-38, B-39 SLRA Section 3.5.2.2.1.6 and Table 3.5.2.2.1.6-1, associated with AMR item 3.5.1-010, identifies the primary containment equipped with stainless steel bellows subject to elevated temperatures potentially above the 140oF threshold for SCC and apparently monitored for SCC.
The FE states: The enhanced ASME Section XI, Subsection IWE (B.2.1.29) program and the 10 CFR Part 50, Appendix J (B.2.1.31) program will be used to manage cracking due to SCC for stainless steel penetration bellows (hot fluid penetrations), vent line bellows, and the associated DMWs. In addition, due to being higher temperature, these penetrations are leading indicators for cyclic load cracking of susceptible drywell shell materials, penetration sleeve, or other locations.
It is not clear from SLRA Section 3.5.2.2.21.6 nor from SLRA B.2.1.29 AMP enhancements, how the stated AMPs (methods used capable of detecting cracking, frequency (periodic, one-time), whether all or sampling-based etc.) will adequately manage cracking due to SCC for the stainless steel penetration bellows, vent line bellows and DMWs.
a) Describe with sufficient technical detail and basis how the SLRA credited AMPs (enhanced IWE and 10 CFR Appendix J) will adequately manage cracking due to SCC for stainless steel hot penetration bellows, vent line bellows and associated DMWs during the SPEO. Include justified examination methods, frequency, and new or revised enhancements as necessary.
b) Clarify the link, between managing cracking due to SCC and cracking due to cyclic loading, as apparently alluded to in the second sentence of the cited FE statement in the Background/Issue column. Clarify how the link is established in the credited AMPs.
c) Update the applicable SLRA sections consistent with the responses to the above.
7 Table 3.5.2-9 and Section 3.5.2.2.1.5 3.5-159 thru 3.5-168; SLRA Table 3.5.2-9 AMR items 3.5.1-027 (CLB fatigue analysis does not exist) lacks clarity regarding components subject to supplemental a) Clarify which components in SLRA Table 3.5.2-9 for which AMR item 3.5.1-027(CLB fatigue
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 66 3.5-28 thru 3.5.-
37 surface or EVT-1 examination, and those not requiring aging management for cracking based on new fatigue waiver analysis performed for SLRA. Also, there are several component types in SLRA Table 3.5.2-9 that credit both AMR items 3.5.1-027 and 3.5.1-009.
It is also not clear if the new fatigue waiver analysis, presented in SLRA Section 3.5.2.2.1.5 (and used as justification for exception to SLRA B.2.1.29 AMP) for drywell structures, penetrations and components that do not have a CLB fatigue analysis, isdocumented in a calculation subjected to the DNPS QA process; such document was not available in the ePortal for verification.
analysis does not exist) is credited are subject to supplemental surface examinations per enhancement 1, and which are the components for which the aging effect does not require management based on the alternate fatigue waiver analysis in 3.5.2.2.1.5 (i.e., the justification for exception to the IWE AMP). For those AMR line items for component types with both 3.5.1-027 and 3.5.1-009 identified in the Table, what are the components for which TLAA is credited.
b) Confirm if the new fatigue waiver analysis presented in SLRA Section 3.5.2.2.1.5 pursuant to related provision in SLR-ISG-2021-03-STRUCTURES is documented in a calculation that has been subject to DNPS QA process. Provide that calculation in the ePortal.
SLRA Section: B.2.1.30 /A.2.1.30 ASME Section XI, Subsection IWF AMP & AMR Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
DR-PBD-AMP-XI.S3 (PBD),
Section 3.1 B-224; PBD p11 The PBD states on page 11: Notable inaccessible drywell component supports at DNPS include drywell steel support skirt and anchor bolts at the bottom of the a) Explain using drawings as illustration the configuration of the RV supports, and what areas are accessible and which areas are inaccessible. Clarify
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 67 drywell, as well as portions of the drywell shear lug stabilizer at the top of the drywell head. Each is either partially or completely embedded in concrete.
From a review of UFSAR Figures 3.9-1 Reactor Vessel Stabilizer and 3.9-2 Reactor Pedestal, the skirt and stabilizer do not appear to be embedded in concrete as claimed, except for the anchor bolts.
It is not clear if the RV supports are inspected in the IWF program and how the effects of aging will be adequately managed during the SPEO.
the cited statement in the PBD regarding the RV support skirt and stabilizer being either partially or completely embedded in concrete.
b) Explain whether the RV supports are inspected using VT-3 visual under the IWF program and provide the results of the most recent two inspections.
c) If not inspected, explain how the effects of aging of the RV supports will be adequately managed during the SPEO consistent with requirements of 10 CFR 54.21(a)(3).
2 A.2.1.30 A-30 The UFSAR supplement states, in part:
The program is implemented through procedures, in accordance with the requirements of ASME Section XI, Subsection IWF, 2017Edition, as approved in 10 CFR 50.55a.
Contrary to a specific code edition (2017) in the above UFSAR supplement statement, 10 CFR 50.55a requires the inspection interval code edition to be consistent with the code edition(s) for subsequent ISI interval during the SPEO in accordance with 10 CFR 50.55a(g)(4),
and therefore this aspect of the UFSAR supplement description appears inconsistent with the regulations.
a) Provide a revised UFSAR supplement statement reflecting the requirement for code edition used during the SPEO consistent with 10 CFR 50.55a(g)(4) for applicable ISI code edition for subsequent inspection intervals.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 68 3
B.2.1.30; PBD 3.1-DNPS B-135; PBD p10 AMP XI.S3 Components Not Used at DNPS:
DNPS Scope of Program element in PBD states:
Supports within the scope of ASME Section XI, Subsection IWF do not have elastomeric vibration elements at DNPS).
Similar statements are also provided in PBD Section 3.3 Parameters monitored or inspected and Section 3.4 Detection of Aging Effects element.
The staff needs the above information on the docket to make its determination of AMP consistency with the GALL-SLR AMP XI.S3, as applicable to DNPS.
a) Update SLRA Section B.2.1.30 to include the information cited in the Background/Issue column of this question regarding use of elastomeric vibration isolation elements.at DNPS.
4 B.2.1.30 / PBD B-171 The SLRA states: All bolting is monitored for cracking regardless of mechanism.
Similar statement is also provided in PBD Section 3.3...
It is not clear how this is performed by the program without surface or volumetric or enhanced examination.
The SLRA also states: The reactor vessel skirt to ring girder bolting consists of high strength bolting (HSB) in sizes greater than 1-inch nominal diameter. SLRA Tables 3.5-1 and 3.5.2-2 AMR item 3.5.1-068 appears to include other ASME Class 1 components.
a) Clarify method used by the program to monitor all bolting for cracking regardless of mechanism prior to loss of function.
b) Clarify if the RV skirt bolting to ring girder is the only (i) HSB used at DNPS and (ii) HSB with nominal diameter greater than 1 used at DNPS.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 69 It is not clear if the RV skirt bolting is the only high strength bolting (HSB) used at DNPS since it is the only bolting sampled for Enhancement 5..
5 Table 3.5-1 and Table 3.5.2-11 for Item 3.5.1-085 3.5-93, 3.5-175 SLRA states that AMR item 3.5.1-085 for managing loss of material of stainless steel structural bolting exposed to treated water is consistent with NUREG-2191 with a generic note B. However, it only uses the Water Chemistry AMP and does not include the IWF AMP (or alternate) also included for the Item in the GALL/SRP-SLR Item. It is not clear how the AMR item is consistent with GALL-SLR.
a) Clarify how the AMR items corresponding to SRP-SLR Table 3.5-1, item 085 is consistent with GALL-SLR and justify the generic note B.
b) Update SLRA as necessary based on the response.
6 (to be discussed with TRP
- 76)
B.2.1.30, 3.5.2.2.2.6, PBD B-171, 3.5-55, 3.5-56 SLRA 3.5.2.2.2.6 under subtitle RPV Support Steel Evaluation, states in part:
Although integrity of the RPV support structures is assured by evaluation of support configuration as presented above, examination of the RPV support structures per the current ASME Section XI, Subsection IWF (B.2.1.30) program will also confirm there is no visible evidence of a loss (or reduction) of fracture toughness due to irradiation embrittlement (e.g.,
cracking) during the SPEO.
The staff notes that, with regard to loss of fracture toughness of RPV steel due to irradiation embrittlement, neither the component nor the material and environment combination is evaluated in NUREG-2191 which would be a case of generic note J.
a) While a plant-specific program may not be necessary, describe how the aging effects due to irradiation embrittlement will be adequately managed by the IWF AMP that is credited for the RPV steel support assembly components for the SPEO?
Provide corresponding Table 2 AMR items, noting that it currently is not included in the GALL-SLR Report, and any related changes that may need to be made to the SLRA.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 70 While a plant-specific AMP may not be necessary, loss of fracture toughness due to irradiation embrittlement remains an applicable new aging effect/mechanism for the RPV steel supports for SLR. Neither the IWF PBD on the ePortal nior the SLRA itself includes loss of fracture toughness due to irradiation embrittlement among the aging effects/mechanisms managed by the program, SLRA Sections 3.5.2.1.2, 3.5.2.1.9, Table 3.5.2-2, and Table 3.5.2-9, do not appear to include AMR items that the aging effect will be managed by the IWF AMP and SLRA B.2.1.30 does not appear to include loss of fracture toughness due to irradiation embrittlement as an aging effect that will be managed by the program.
SLRA Section B.2.1.31 10 CFR Part 50, Appendix J Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.31; Program Basis
- Document, PBD) Section 2.1 PBD Pg. 6 Consistency of Scope of Program Section 2.1 of PBD states, in part: Components and penetrations excluded from testing are identified in various engineering requests (ECR),
Engineering Changes (ECs), and Attachment 6.2.
The of GALL-SLR AMP XI.S4, scope of program element states, The aging effects associated with containment pressure-retaining boundary components within the scope of subsequent license renewal and excluded from Type B or C Appendix J testing must still be managed. Other programs may Scope of Program
- 1. Clarify if Attachment 6.2 is the complete list of components excluded from Type B or C testing at DNPS. If not, provide additional components list and associated AMPs for managing the aging effect of the additional excluded components.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 71 be credited for managing the aging effects associated with these components; however, the component and the proposed AMP should be clearly identified..
It is not clear if all of the components listed in.2 is the complete list of components excluded from Type B or C leak rate testing because various ECR and ECs is mentioned.
SLRA Section B.2.1.4 BWR Vessel ID Attachment Welds Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.4 Appendix C Response to BWRVIP License Renewal Applicant Action Items includes BWRVIP-48-A, BWR Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines and indicates that this report is credited in BWR Vessel ID Attachment Weld program.
LRA Section A.2.1.4 states, in part, the BWR Vessel ID Attachment Welds aging management program is an existing condition monitoring program that manages cracking of the reactor vessel interior attachment welds. This program relies on visual examinations to detect cracking. The examination scope, frequencies, and methods are in accordance with ASME Code,Section XI, Table IWB-2500-The staff noted there is a discrepancy between the information in the documents identified above regarding what will be implemented as part of AMP B.2.1.4 - BWR Vessel ID Attachment Welds Explain what will be implemented as part of AMP B.2.1.4
- BWR Vessel ID Attachment Welds during the SPEO, and discuss whether revisions are necessary either to the SLRA or internal applicant documents.
o If AMP B.2.1.4 - BWR Vessel ID Attachment Welds will be implementing ASME Code,Section XI, Table IWB-2500-1, Examination Category B-N-2 and BWRVIP-48 Revision 2.
The staff noted that use of BWRVIP-48, Revision 2 is inconsistent with the guidance recommended in GALL-SLR.
Additionally, the staff noted that BWRVIP-48, Revision 2 is not an NRC-approved topical report. Provide a technical basis for the use of BWRVIP-48 Revision 2 in lieu of BWRVIP-48-A and explain how adequate aging management will be
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 72 1, Examination Category B-N-2 and BWRVIP-48 Revision 2.
LRA Section B.2.1.4 indicates the following:
the program monitors for cracks induced by SCC, IGSCC, and cyclic loading on the reactor vessel interior attachment welds by detection and sizing of cracks using visual techniques in accordance with the guidelines of BWRVIP-48 Revision 2.
Inspections are performed in accordance with the guidance in BWRVIP-48 Revision 2 and the requirements in ASME Code,Section XI, Table IWB-2500-1, Examination Category B-N-2 to interrogate the components for discontinuities that may indicate the presence of cracking.
The BWR Vessel ID Attachment Welds aging management program is consistent with the ten elements of aging management program XI.M4, BWR Vessel ID Attachment Welds, specified in NUREG-2191.
accomplished in accordance with 10 CFR 54.21(a)(3)
As part of this technical basis, provide an assessment to include but not limited to changes/revisions to inspection scope, coverage, inspection technique, inspection/reinspection frequency, acceptance criteria and reinspection, scope expansion, etc.
Are there any other situations/scenarios in other AMPs in the applicants SLRA that rely/reference topical reports that are inconsistent with those identified in GALL-SLR and were not explicitly identified as an exception.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 73 No exceptions or enhancements to AMP B.2.1.4 were identified in the SLRA The applicants program basis document (DR-PBD-AMP-XI.M4, Rev 1) indicates that the AMP implements ASME Code,Section XI, Table IWB-2500-1, Examination Category B-N-2 and BWRVIP-48 Revision 2.
The program description of GALL-SLR AMP XI.M4 states the program includes inspection and flaw evaluation in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, and the guidance in BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines
[Boiling Water Reactor Vessel and Internals Project (BWRVIP)-48-A]
to provide reasonable assurance of the long-term integrity and safe operation of BWR vessel ID attachment welds.
2 B.2.1.4 Based on initial breakout session and breakout question response, the staff has a few additional discussion topics.
Core spray piping bracket Confirm the inspection technique and frequency o Confirm if for both units.
Jet pump riser brace
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 74 Confirm the inspection technique and frequency Unit 3 - no additional inspections other than Section XI o Expand on the technical basis - large change from -A o What is the difference between unit 3 and unit 2 o Reliance appears solely on OPEX that no additional inspections are warranted.
o Inspection technique that is considered for significant potential exists for increase FIV load Actions dont appear to be definitive and are associated with corrective action, extent of condition or impact of possible design changes. Discuss more with applicant.
Steam dryer support Confirm the inspection technique and frequency o Confirm if for both units.
o Proposed approach - Change in both population and frequency from -A Confirm understanding - unless timing is such there is one before SPEO and one during SPEO?
Expand on the technical basis - large change from -A given that its a change in population and frequency Reliance appears solely on OPEX Plant-specific considerations?
(Design? Material??)
Discussion under New Inspection Requirements, appears to be associated with the extent of condition, corrective actions and follow up inspections - if degradation is found.
Appears many of these actions would have been expected to be considered and implemented as
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 75 appropriate, consistent with QA Program/Corrective Actions Program Inspection technique considered when evidence of significant vibration or movement Actions dont appear to be definitive and are associated with corrective action, extent of condition or impact of possible design changes. Discuss more with applicant.
Non-GALL Mechanical Components Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
Table 3.3.2-20 3.3-291 3.3-295 SLRA Table 3.3.2-20 includes aging management for glass piping elements in four environments. The sodium pentaborate environment is not addressed in the NRC guidance.
Plant-specific Note 1 in Table 3.3.2-20 states, Aging effects on glass exposed to sodium pentaborate solution are established using a treated water environment. The meaning of this statement (i.e., establishing aging effects using treated water) is not clear to the staff.
In addition, the body of Table 3.3.2-20 does not identify the aging effects intended to be managed or the aging management program to be used.
Please address the following regarding the aging management of piping elements in Table 3.3.2-20:
- 1. Identify the aging effect being managed for the sodium pentaborate environment using the Water Chemistry program.
- 2. Clarify or correct the apparent discrepancy between the AMP listed in the table (None) and the AMP identified in plant-specific Note 1 (Water Chemistry program).
- 3. Clarify the meaning of the statement in plant-specific Note 1 that, Aging effects on glass
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 76 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request exposed to sodium pentaborate solution are established using a treated water environment.
Further Evaluations in SLRA Sections 3.2, 3.3, 3.4 Stress Corrosion Cracking and Loss of Material (pitting, crevice) for Stainless Steel, Nickel Alloys, and Aluminum Alloys Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
3.2.2.2.2 3.2-13 SLRA Section 3.2.2.2.2 on Page 3.2-13 addresses AMR Item 3.2.1-107 in two places, one at the top of the page with items 3.2.1-004 and -048, and one in the middle of the page. Both refer to piping, piping components, or tanks exposed to air or condensation. The latter occurrence notes that this item is for insulated components, which is consistent with the description in Table 3.2.1. It is not clear why item 3.2.1-107 is mentioned in the first instance with other AMR items for uninsulated components.
Please explain the reason for addressing item 3.2.1-107 in two places in SLRA Section 3.2.2.2.2 and revise the SLRA if appropriate.
2 3.4.2.2.7 Table 3.4.1 3.4-19 3-4-51 Item 3.4.1-109 addresses cracking due to stress corrosion cracking (SCC) for aluminum piping, piping components, and tanks exposed to air, condensation, raw water, and waste water. The discussion of 3.4.1-109 in Section 3.4.2.2.7 states that this item is not applicable. However, Table 3.4.1 states that this item is not used and that aluminum tanks exposed to air or condensation are addressed by another item (3.4.1-102).
Due to the inconsistency in the wording (not applicable vs. not used), the staff is not completely clear on how item 3.4.1-109 is being addressed.
For AMR item 3.4.1-109:
- a. Please clarify the discrepancy between SLRA Section 3.4.2.2.7 (not applicable) and Table 3.4.1 (not used).
- b. For any applicable components, please clarify how they are addressed in the aging management plans, including components not susceptible to SCC based on the material, and
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 77 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request components addressed by other AMR items.
- c. Please revise the SLRA for consistency between Section 3.4.2.2.7, Table 3.4.1, and the relevant Tables 3.4.2-X.
SLRA Section 3.5.2.2.2.6: Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request C1 3.5.2.2.2.6 UFSAR 3.5-53 through 3.5-58 Figures 3.9-1 to 3
This breakout question is to facilitate clear staff understanding of the configuration, elevations, and dimensions of structures and components (SCs) that fall under the scope of SLRA Section 3.5.2.2.6.
DWG No. 718E699 reactor vessel support, sht1, Rev 002.pdf on the Portal shows that top of grout above reactor concrete pedestal is higher than anchor bolts for ring girder, i.e. these anchor bolts are covered by the grout.
a) Provide Figure in SLRA to show as-constructed as-operated relative configuration of RPV, active core, BSW (steel columns, reinforced concrete, and its liner), penetrations through BSW, RV stabilizer, reactor vessel support skirt, ring girder, reactor concrete pedestal, and drywell concrete floor with descriptions, including the following:
- 1. Top of BSW elevation.
- 2. Center elevation of RV stabilizer.
- 3. Top and bottom of active core region elevation.
- 4. Top of ring girder elevation.
- 5. Top of reactor concrete pedestal elevation.
- 6. Top of drywell concrete floor elevation.
- 7. Thickness of liner, BSW and reactor pedestal.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 78 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request
- 8. Inner diameter dimension of BSW and reactor concrete pedestal.
- 9. Compressive strength of reinforced concrete of pedestal and within the BSW.
b) Clarify the top of grout elevation condition above reactor concrete pedestal at Units 2 and 3.
C2 Table 3.5.1 Table 3.5.2-9 UFSAR Section 12.3.2.2.1 3.5-100 3.5-159 12.3-4 SLRA Table 3.5.1, AMR item 3.5.1-097 is related to managing reduction of strength; loss of mechanical properties due to irradiation for Group 4: Concrete (reactor cavity area proximate to the reactor vessel):
reactor (primary/biological) shield wall; sacrificial shield wall; reactor vessel support/pedestal structure.
SLRA Table 3.5.2-9 shows Table 2 AMR item associated with AMR item 3.5.1-097 for concrete:
near reactor vessel (reactor shield wall) for managing reduction of strength; loss of mechanical properties. SLRA Table 3.5.2-9 lacks Table 2 AMR item for managing aging effects of the reactor concrete pedestal.
UFSAR Section 12.3.2.2.1 describes reactor shield wall, it states, The reactor shield wall consists of a hollow cylinder of ordinary concrete having a 2-foot thick wall and circumscribing the reactor vessel.Reinforcing steel is used in the concrete to give structural strength. It does not make clear whether concrete fill within the BSW is structural or non-structural.
a) Provide Table 2 AMR item associated with AMR item 3.5.1-097 for managing reduction of strength; loss of mechanical properties due to irradiation of the reactor concrete pedestal.
b) Clarify in SLRA whether concrete fill within the BSW is structural or non-structural and provide the justification.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 79 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request C3 3.5.2.2.2.6 3.5-55 RPV Support Steel Evaluation in SLRA Section 3.5.2.2.2.6 states, The reactor shield wall in turn is anchored at the base to the top of the reactor pedestal and restrained at the top by a horizontal tubular truss system designed to permit axial expansion. The truss system transmits lateral loads through drywell shear lug mechanisms to the concrete structure (part of the reactor building) outside the drywell to limit horizontal vibration and to resist seismic and jet reaction forces.
The staff reviewed UFSAR Figure 3.9-1 and notes that it does not make clear what the axial expansion, the truss system, and the drywell shear lug mechanism are.
a) Discuss/describe the lateral loads transfer arrangement between RPV and BSW/Drywell in more details using detailed construction drawings.
b) Clarify the axial expansion, the truss system, and the drywell shear lug mechanism and revise SLRA as needed.
C4 Table 3.5.1 2.3.1.2 Table 3.1.2-2 UFSAR 3.5-77 2.3-6 3.1-88 Figure 3.9-2 SLRA utilizes AMR 3.5.1-055 for managing aging effect of reduction in concrete anchor capacity for building concrete or grout in supports at locations of expansion and grouted anchors. SLRA does not make clear of which NUREG-2191 Item Number is used for managing aging effect of reduction in concrete anchor capacity for concrete and grout at location of anchor bolts for ring girder shown on UFSAR Figure 3.9-3.
SLRA Section 2.3.1.2, Reactor Vessel discusses the reactor vessel support skirt. SLRA Table 3.1.2-2 shows Table 2 AMR item associated with AMR 3.1.1-113 for reactor vessel support skirt that its aging effect is managed by the ASME Section XI, Subsections IWB, IWC and IWD programs. The ASME XI code requires the ASME XI, Subsection IWF program to inspect supports for Class 1 BPV.
a) Clarify Table 2 AMR item in SLRA for managing aging effect of reduction in concrete anchor capacity for concrete and grout at location of anchor bolts for ring girder shown on UFSAR Figure 3.9-3.
b) Clarify whether reactor vessel support skirt including weldments is managed under the ASME Section XI, Subsection IWF program. If yes, provide Tables 1 and 2 AMR items and verify whether enhancements to AMP are needed. If not, provide the justification why it is not managed under the ASME XI, Subsection IWF program.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 80 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request SLRA does not make clear whether reactor vessel support skirt is managed under the ASME Section XI, Subsection IWF program.
The staff noted narrow space between reactor vessel skirt and BSW shown on UFSAR Figure 3.9-2, SLRA does not make clear how aging effects of the SCs within this space are managed during the SPEO.
c) Describe the access (opening size and location) to space between reactor vessel skirt and BSW and explain in SLRA how aging effects of SCs such as reactor vessel support skirt, liner, grout, and bolts, anchor bolts etc.
within this space are adequately managed during the SPEO.
S1 3.5.2.2.2.6 3.5-55 to 58 Reactor Shield Wall Structural Steel Evaluation in SLRA Section 3.5.2.2.2.6 states, Inspection of accessible portions of the steel liners per the current Structures Monitoring (B.2.1.33) program will also confirm there is no visible evidence of a loss (or reduction) of fracture toughness due to irradiation embrittlement (e.g., cracking) during the SPEO.
Similarly, SLRA Section 3.5.2.2.2.6 states the current ASME Section XI, Subsection IWF (B.2.1.30) program will also confirm there is no visible evidence of a loss (or reduction) of fracture toughness due to irradiation embrittlement (e.g., cracking) during the SPEO.
The staff reviewed the Structures Monitoring program basis document and its procedures (ER-AA-450, Rev. 11 and ER-DR-450-1001, Rev. 0) and finds that current Structures Monitoring program procedure does not include the inspection of a loss (or reduction) of fracture toughness/cracking due to irradiation embrittlement. Additionally, the staff could not AMR line items for the purpose of ensuring/confirming that there will be no visible evidence of a loss (or reduction) of fracture toughness due to irradiation embrittlement (e.g.,
cracking) during the SPEO.
a) Discuss the evaluation of the aging effects and aging mechanisms for the steel components of biological shield wall, RV steel structural supports, and their assembled components that fall under the scope of SLRA Section 3.5.2.2.2.6.
b) Provide Table 2 AMR line items in the SLRA to ensure that there will be no visible evidence of a loss (or reduction) of fracture toughness due to irradiation embrittlement (e.g.,
cracking) during the SPEO for both the Section XI IWF AMP and the Structures Monitoring AMP.
c) Provide necessary plant-specific enhancements to selected GALL-SLR AMPs (e.g., AMP XI.S3, ASME Section XI, Subsection IWF, and/or AMP XI.S6, Structures Monitoring) to ensure these aging effects are adequately managed during the SPEO.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 81 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request SLRA appears to lack aging management of reduction of fracture toughness/cracking due to irradiation-induced combined mechanisms for the steel components of biological shield wall (e.g.
columns, liners, openings/penetrations, and doors),
reactor vessel (RV) steel structural supports, and their assembled components (e.g., RPV skirt; ring girder; sole plates; RV seismic restraints; welds; bolted connections; support anchorage to reinforced concrete pedestal) that fall under the scope of SLRA Section 3.5.2.2.2.6.
S2 3.5.2.2.2.6 3.5-56 to 56 On these pages of the SLRA, the applicant discussed the radiation exposure levels above and below reactor vessel beltline region. The SLRA states that the lateral stabilizers more than 14 feet above the upper edge of the beltline region and the top of the RV support skirt is more than 13 feet below the lower edge of the beltline region.
Show drawings that confirm the vertical distances of the lateral stabilizers and top of the RV support skirt from the beltline region. Clarify what is considered the top of the RV support skirt. This is related to Question C1.
S3 3.5.2.2.2.6 3.5-57 On this page of the SLRA, the applicant stated that the 1/4-inch thick steel plates that form the inner and outer liners for the DNPS reactor shield wall are fabricated from steel conforming to ASTM A36 low carbon steel, and that based on information provided in NUREG-1509, Tables 4-1 and 4-2, the initial nil-ductility transition temperature (NDTT) plus 1.3 standard deviation for this material is 39°F.
The staff noted that this initial NDTT value of 39°F is the NDT + 1.3 value for carbon-manganese as-hot rolled steel in Table 4-1 of NUREG-1509, which Confirm that there were no plant-specific certified material testing reports that contain the plant-specific initial NDTT of the ASTM 36 steel, and that therefore, the recommended value in Table 4-1 of NUREG-1509 was used.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 82 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request maybe used if plant-specific initial NDT values are not available.
S4 3.5.2.2.2.6 3.5-57 The SLRA states that the reactor vessel base metal to which the reactor skirt and stabilizer bars are connected is SA-302 Grade B (modified with Code Case 1339).
UFSAR Table 5.1-1 and 5.2-2 seem to be inconsistent with regard to the code case associated with SA-302 Grade B.
UFSAR Table 5.1-1 states Code Case 1339 (consistent with SLRA), while UFSAR Table 5.2-2 states Code Case 1335. Explain the inconsistency between UFSAR Table 5.1-1 and 5.2-2 with regard to the code case associated with the reactor vessel base metal.
S5 3.5.2.2.2.6 3.5-53 to 58 The SLRA does not have summary of operating experience (OE) of reactor vessel (RV) SCs (concrete structures, bolts, ring girder, skirt, shield wall steel, etc.). WO 00497712 reports two vertical cracks in the Unit 3 bioshield wall weld.
The staff reviewed the inspection results for the following RV support structures/components and associated work orders (WOs) and need discussion of them for general clarity:
Unit 2 RV skirt, WO 0189966 (pg. 149) and WO 04997402 (pgs. 50 and 51)
Unit 2 stabilizers, WO 04997403 (pgs. 36 and 37)
Unit 3 RV skirt, WO 05236469 (D3R27-VT-056 and D3R27-MT-009)
Unit 3 stabilizers, WO 01608051 (pgs. 10 and 19) and WO 05121032 (D3R27-VT-052 and D3R27-VT-057) a) Provide summary of OE of RV SCs (concrete structures, bolts, ring girder, skirt, shield wall steel, etc.), including discussion of two vertical cracks in Unit 3 bioshield wall weld. This information would need to be docketed because per the guidance in NUREG-1509, prior to any analysis the condition of the RV support system must be assessed.
b) During the breakout session, the staff is requesting that the applicant show the pages of the WOs shown and discuss such topics as (but not limited to) the ASME Code,Section XI Item No. being inspected, whether the specific item is inspected in Structures Monitoring or IWF AMP.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 83 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request The staff could not find WOs for inspection of the Unit 2 bioshield wall.
c) Show and provide in portal WOs of inspection results of the Unit 2 bioshield wall.
S6 3.5.2.2.2.6 3.5-57 The SLRA states a permissible lowest service temperature of 91.5°F for the reactor shield wall steel.
a) Confirm that the normal operating temperature in the reactor cavity in the region of the reactor shield wall steel exposed to high radiation (i.e., in the core region elevation) is much greater than 91.5°F.
b) In this confirmation, identify in UFSAR Figures 3.11-1 to 3.11-5 that show the environmental data zones the annulus between the RPV outer wall surface and the shield wall inner wall surface at the elevation of the reactor core region.
c) Also, in this confirmation, discuss the average drywell temperature of 135°F in the portal in file Item 10 part 4.docx.
S7 Table 3.5.2-9 Table 3.5.2-2 3.5-164 3.5-109 UFSAR Figure 3.9-2 shows reactor vessel to biological shield stabilizer and biological shield to containment stabilizer, which are used to provide lateral support for the ASME Class 1 RPV. Table IWF-2500-1 (F-A) of the ASME XI, Subsection IWF requires to inspect these supports.
SLRA Table 3.5.2-9 lists Table 2 AMR item associated with AMR item 3.5.1-077 for carbon steel stabilizers and truss arrangements, which their aging a) Clarify whether carbon steel stabilizers and truss arrangements discussed in Table 2 AMR item associated with AMR item 3.5.1-077 are that reactor vessel to biological shield stabilizer and biological shield to containment stabilizer shown on UFSAR Figure 3.9-2.
b) Clarify whether the ASME XI, Subsection IWF program will be used
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 84 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request effects are managed by the Structures Monitoring program, which conflict with the ASME XI code.
In addition, SLRA Table 3.5.2-2 lists supports for ASME Class 2 and 3 Piping and Components (Support members; welds; bolted connections; support anchorage to building structure). However, SLRA Table 3.5.2-2 does not present Table 2 AMR items for supports for ASME Class 1, Piping and Components (e.g., RPV support)
It appears that SLRA and the ASME XI code has a discrepancy on managing aging effects of lateral support for the ASME Class 1 RPV.
to manage aging effects due to irradiation embrittlement of carbon steel stabilizers and truss arrangements. If yes, evaluate what enhancements are needed to the ASME XI, Subsection IWF program and provide Table 2 AMR items. If not, provide the justification why their aging effects are not managed by the ASME XI, Subsection IWF program.
c) Clarify whether Table 2 AMR items are needed for supports for ASME Class 1, Piping and Components (Support members; welds; bolted connections; support anchorage to building structure). If not, provide the justification why they are not needed.
SLRA Section 3.5.2.2, AMR Results for Which Further Evaluation is Recommended by the GALL-SLR Report Question Number LRA Section LRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
3.5.2.2.1.2 3.5-22 SLRA FE Section 3.5.2.2.1.2 states that station areas that bound high temperature considerations are the drywell general area and reactor shield wall piping penetration local area. However, it lacks the information of elevated temperatures at reactor shield wall piping penetration local area and how and what AMPs are
- 1. Describe the piping and the elevated temperatures at the shield wall penetrations.
- 2. Describe what types of insulation and AMPs are used to manage the effects of elevated temperatures at local concrete
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 85 used to manage the effects of elevated temperatures at local areas during the subsequent period of extended operation.
areas during the subsequent period of extended operation.
2 3.5.2.2.2.1.3 Table 3.5.1 3.5-43 3.5-73 SLRA only lists Table 2 AMR items associated with AMR 3.5.1-044 for switchyard structures and yard structures and claims Table 1 AMR item 3.5.1-046 to be not used.
FE 3.5.2.2.2.1, item 3 addresses aging effects of cracking and distortion due to increased stress levels from settlement and reduction in foundation strength, and cracking due to differential settlement and erosion of porous concrete subfoundations, both of them are applicable aging effects to the DNPS structures.
If Table 1 AMR item meets the following condition, Table 1 AMR item is not used.
The component, material, and aging effect combination are addressed by a different Table 1 AMR item.
SLRA lacks Table 2 AMR items associated with AMR 3.5.1-044 for all the DNPS structures except switchyard structures and yard structures.
- 1. Provide Table 2 AMR items associated with AMR 3.5.1-044 for all DNPS structures in both accessible and inaccessible concrete areas.
- 2. Clarify whether the component, material, and aging effect combination of Table 1 AMR item 3.5.1-046 are addressed by a different Table 1 AMR item. If yes, state the Table 1 AMR item.
- 3. Explain why the Group 4 structures are not subject to these aging effects evaluated in FE 3.5.2.2.2.1, item 3.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 86 SLRA does not make clear whether the component, material, and aging effect combination of Table 1 AMR item 3.5.1-046 are addressed by a different Table 1 AMR item.
GALL-SLR item III.A3.TP-30 covers all concrete structures and components that their aging effects of cracking and distortion due to increased stress levels from settlement are managed by the Structures Monitoring program.
However, Table 2 AMR items associated with AMR 3.5.1-044 have only addressed effects of aging for the inaccessible concrete areas.
FE 3.5.2.2.2.1, item 3 states, This Table 3.5-1 Item Number does not apply to the Group 4 structures due to the Mark I containment design, but SLRA does not make clear why the Group 4 structures are not subject to these aging effects.
3 3.5.2.2.2.1.4 3.5.2.2.2.3.3 3.5-44 3.5-49 SLRA FE Sections 3.5.2.2.2.1, item 4 and 3.5.2.2.2.3, item 3 state that some minor leaching of calcium hydroxide has been observed on DNPS reinforced concrete structures as well as DNPS Group 6 concrete structures.
FE 3.5.2.2.2.1, item 4 states, This Table 3.5-1 Item Number does not apply to the Group 4 structures due to
- 1. Provide ARs and WOs associated with these aging effects in the Portal and discuss this OE to determine whether observed leaching of calcium hydroxide and carbonation in accessible areas has any impact on the intended function of the concrete structure.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 87 the Mark I containment design, but SLRA does not make clear why the Group 4 structures are not subject to these aging effects.
- 2. Explain why the Group 4 structures are not subject to these aging effects evaluated in FE 3.5.2.2.2.1, item 3.
4 3.5.2.2.2.2 3.5-46 SLRA FE Section 3.5.2.2.2.2 addresses reduction of strength and modulus of concrete due to elevated temperatures could occur in Group 1-5 concrete structures.
SLRA lacks detailed information to justify why concrete structural components in Groups 1-5 are not subject to general area temperature greater than 150°F or local area concrete temperature greater 200°F.
- 1. Provide detailed information for why ventilation systems and AMP are used to maintain general area normal temperatures less than 150°F.
- 2. Describe the piping and the elevated temperatures at the penetrations through concrete walls and describe what types of insulation and AMPs are used to manage the effects of elevated temperatures at local concrete areas during the subsequent period of extended operation.
5 3.5.2.2.2.3.1 Table 3.5.2-1 3.5-46 3.5-103 SLRA FE Section 3.5.2.2.2.3, item 1 states that structural reinforced concrete has not exhibited significant loss of material (spalling, scaling) and cracking due to freeze-thaw in accessible areas of in scope reinforced concrete structures.
However, SLRA does not make clear whether observed loss of material (spalling, scaling) and cracking due to freeze-thaw in accessible areas have any impact on the intended function of the concrete structure.
Table 3.5.2-1 provides summary of aging management evaluation for the
- 1. Provide evaluation to determine whether observed loss of material (spalling, scaling) and cracking due to freeze-thaw in accessible areas have any impact on the intended function of the concrete structure.
- 2. Provide Table 2 AMR items for addressing aging effects of loss of material (spalling, scaling) and cracking due to freeze-thaw for the circulating water inlet tunnel.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 88 circulating water inlet tunnel. SLRA lacks Table 2 AMR item for addressing aging effects of loss of material (spalling, scaling) and cracking due to freeze-thaw.
NURE-2192 Item III.A6.TP-110 has air-outdoor, groundwater/soil environments, however, SLRA lacks Table 2 AMR items associated with Table 1 AMR item 3.5.1-049 for concrete components exposed to air-outdoor environment.
item 3.5.1-049 for concrete components exposed to air-outdoor environment.
6 3.5.2.2.2.1 3.5-41 SLRA Section 3.5.2.2.2.1 states that DNPS does not have any concrete structures categorized as Groups 7 and 8 per NUREG-2191. However, there are many fuel oil, water, and contaminated condense storage tanks in the yard structures shown on the license renewal drawing composite site plan. SLRA does not make clear whether these storage tanks are categorized as Groups 7 and 8 with concrete components.
- 1. Clarify whether fuel oil, water, and contaminated condense storage tanks in the yard structures are categorized as Groups 7 and 8.
- 2. Clarify whether these storage tanks have concrete components such as concrete foundations.
SLRA TLAA 4.2.6 Reactor Pressure Vessel AXIAL Weld Failure Probability Analyses Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.2.5 TLAA 4.2 Basis Document - Section 4.2.5 for RPV Circumferential Weld Failure Probability Analyses contains Table B, which provides verification of acceptable RTmax in BWRVIP-329-A for Dresden Unit 2. Table B indicates that for Unit 2 - the neutron Please clarify which RPV material is the limiting axial weld for Unit 2 - discuss whether revisions to either the
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 89 fluence data and RPV material property data is based on the information in SLRA Tables 4.2.1.1-1 and 4.2.3-1, respectively.
Based on the information in SLRA Table 4.2.3-1 it appears the axial weld in Unit 2 Shell Ring 2 identified as Electroslag (ES) with Heat No./Lot 34A167/3496 would be the limiting axial weld based on the ART value for 1/4 T.
Additionally, footnote 5 for SLRA Table 4.2.3-1 states The bounding ART in the table is the ESW material with Heat No.
34A167/3496 and ART of 103 °F.
With respect to SLRA Table 4.2.1.1-1:
It identifies the neutron fluence for the limiting axial weld in Unit 2 at the 1/4 T and 0T locations as 3.88E+17 n/cm2 and 5.61E+17 n/cm2, respectively.
The 1/4 T neutron value identified in this table is consistent with value used for Unit 2 Shell Ring 2 axial weld identified as Electroslag (ES) with Heat No./Lot 34A167/3496 and supports this material being the limiting axial weld for Unit 2.
With respect to SLRA Tables 4.2.1.1-1 and 4.2.1.1-2:
C2873-1 is identified in both tables for the limiting Shell 2 axial weld for Units 2 and 3 respectively.
Heat No. C2873-1 appears to be associated with Peach Bottom 2 based on available information in BWRVIP-86, Rev 1-A and BWRVIP-321, Rev 1-A.
Based on the information in Table B in Section 4.2.5 of the TLAA 4.2 Basis Document and SLRA Section 4.2.3 - the staff noted a discrepancy in identification of the limiting axial weld in Table B of the TLAA 4.2 Basis document and SLRA Table 4.2.3-1. Further, the staff understands that the additional information relevant in comparing plant-specific information for Unit 2 and BWRVIP-329-A may not impacted.
SLRA or TLAA 4.2 Basis document is necessary Discuss the relevance of the Peach Bottom 2 RPV material with Heat No. C2873-1 for the Dresden SLRA.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 90 SLRA Section 4.1 - TLAA Identification Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
Question 1 TLAA Basis Document - Part 1 - Identification - 4 -
Page 8 of 37 -
Basis that ECCS suction strainer TLAA in initial LR is no longer a TLAA for SLR due to ongoing period UT inspections from ongoing program. It states that acceptability of the ECCS suction strainer to perform their intended function is an ongoing program of performing and evaluating UT measurements, rather than an analysis to project acceptability through the period of extended operation.
Operating experience for One-Time inspection program states Follow-up inspections were completed but ultrasonic testing thickness measurements were performed on different locations of the ECCS suction strainer flange such that the results of the follow-up inspections could not be used to establish an accurate corrosion rate. Therefore, to ensure aging of the ECCS suction strainer flanges is adequately managed, periodic inspections on a four (4) year frequency were established.
Operating experience for Bolting Integrity Program discusses VT-3 for all four Unit 3 Emergency Core Cooling System (ECCS) suction strainers.
What ongoing program manages the ECCS Suction Strainer?
Is this ongoing program an AMP for SLR - if so, which program manages for which aging effects - describe the activities to manage galvanic corrosion.
Describe the details of the inspections for the ECCS Suction Strainer during the SPEO
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 91 Based on the available information its not clear what inspections are being performed and how this supports that an analysis is not relied upon for acceptability of the component through the PEO and SPEO 2
TLAA Basis Document - Part 1 - Identification - 4 - Page 11 of 37 -Drywell Shell Thickness reduction in sand pocket Based on the available information it is not clear what inspections are being and will be performed and support that an analysis is not relied upon for acceptability of the component through the PEO and SPEO Provide EC 638774 -
Is there no drywell shell degradation - or is it so minimal that any corrosion and corrosion rate exceeds the time period beyond the SPEO?
Is the corrosion analysis no longer part of the CLB and no longer being credited for a safety determination or safety conclusion?
PMID 00004804 - is this still and will continue to be an ongoing activity for UT of the Drywell Liner - Is this activity a part of an AMP for SLR? If so, which one?
Which, if any, AMPs will manage the Drywell Shell in sand pocket region during the SPEO.
3 Question 4 TLAA Basis Document - Part 1 - Identification - 4 - Page 20 of 37 -Drywell Shell Thickness reduction - Upper spherical and cylindrical areas. The applicant indicates that future inspections of Drywell shell will continue.
Based on the available information its not clear what inspections are being performed and how this supports that an analysis is not relied upon Provide EC 638775 (referenced on page 20 of attachment 14) -
Discuss whether there will be ongoing inspections during the SPEO If so, describe the details of the inspections and which AMP will these inspections be associated with
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 92 for acceptability of the component through the PEO and SPEO 4
TLAA Basis Document - Part 1 - Identification - 4 - Page 34 of 37 - Jet Pump Sensing Line Clamps. The applicant indicated that that these components are not related to safety of plant per 10 CFR 50.2 and the modification components are not safety related.
How are the jet pump sensing lines and the associated modifications scoped in for LR?
Are they both non-safety affecting safety?
Are the components subject to AMR - if so -
identify the SLRA Table/Page these components are addressed in GE Recommendation GE-NE-0000-0065-3643-R0 was cited - Is this 10-year re-inspection an ongoing activity and will it continue to be in the SPEO. If so, which AMP is the inspection associated with?
Have the inspections been going on since installation in 2001/2002?
5 TLAA 4.1 Basis document - Attachment 8 - Item 374 on spreadsheet - Indicates flaw for ISO condenser (found in 2018) was assessed for 40 years. An assessment for 40 years is beyond the SPEO for Dresden. This analysis was identified as a TLAA in SLRA Section 4.7.8.
The staff does not have a concern that it was identified as a TLAA.
Applying a similar logic used by the applicant for other analyses during the TLAA identification process- (e.g., Attachment 14 - Unit 2 N20A nozzle overlay, Unit 3 Jet Pump Slip Joint Repair Clamps) - It appears to be an inconsistent approach Clarify and/or provide a discussion on the inconsistent approach for identifying TLAAs for the Dresden SLRA.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 93 SLRA Section: 4.2.1 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.2.1.2 4.2-8 10 CFR 54.21(c)(1)(ii) - The Unit 2 and Unit 3 reactor vessel internal component fluence analyses have been satisfactorily projected through the SPEO.
The technical reviewer noted the projected fluence values in SLRA Table 4.2.1.2-1 appear to be different by an order of magnitude to results from other BWR sites in their SLRA.
Explain the results of the projected reactor vessel internal component fluence analyses and any benchmarking performed by the applicant to ensure that the General Electric Hitachi (GEH) Discrete Ordinates Transfer (DORT) methodology is adequate to project the values through the SPEO.
SLRA Section B.3.1.1, Fatigue Monitoring Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.3.1.1 B-242 The operating experience (OE) section of SLRA Section B.3.1.1 (Fatigue Monitoring AMP) addresses the operating experience evaluation regarding the feedwater nozzle in relation to NRC Regulatory Issue Summary (RIS) 2008-30.
The OE evaluation indicates that FatiguePro fatigue monitoring software performs stress-based fatigue monitoring and related calculation on four feedwater nozzle locations and the reactor pressure vessel (RPV) support skirt. The OE evaluation also explained that the fatigue monitoring and related calculations use a simplified, single stress term.
The OE evaluation further explained that, in response to RIS-2008-30, confirmatory EAF
- 1. Clarify how the applicant determined that the single stress term methodology was conservative than the methodology that considered the six stress components (e.g., by comparing the environmentally assisted cumulative usage factor (CUFen) values between the methodologies). As part of the discussion, describe the CUFen values calculated by the two methodologies.
- 2. Provide justification for why the use of FatiguePro Version 4, which considers the six stress components, is not an enhancement to the Fatigue
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 94 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request analyses of the feedwater nozzles were performed, which determined that the single stress term methodology was conservative and, therefore, exercised appropriate engineering judgement.
However, the applicant did not clearly discuss how it determined that the single stress term methodology was conservative than the methodology that considered the six stress components (e.g., by comparing the environmentally assisted cumulative usage factor (CUFen) values between the methodologies).
In addition, the OE evaluation indicates that, in place of the currently used FatiguePro Version 3.0 that uses the single stress term methodology for stress-based fatigue monitoring, FatiguePro Version 4.0, which considers the six stress components, will be used for subsequent license renewal.
However, the staff noted that the use of the FatiguePro Version 4.0 is not identified as an enhancement to the Fatigue Monitoring AMP.
Monitoring AMP. If such justification cannot be provided, identify a relevant program enhancement. As part of this discussion, describe the schedule for implementing the use of FatiguePro Version 4.0.
- 3. Describe the periodic inspections that are performed on the feedwater nozzle and RPV support skirt and clarify whether the inspection locations include the limiting fatigue locations of these components. In addition, clarify whether the inspection results indicate the absence of fatigue cracking in these components.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 95 SLRA Section 4.3.1, Transient and Cumulative Usage Projections For 80 Years Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.3.1 4.3-2 SLRA Section 4.3.1 addresses the transient cycle and cumulative usage projections for 80 years of operation.
SLRA Section 4.3.1 indicates that, since most nuclear power plants, including DNPS Units 2 and 3, have experienced a significant declining trend in accumulation of transients over time, transient trends based on recent operating experience provide an accurate basis for future projections.
SLRA Section 4.3.1 also explains that the extrapolated rates starting after December 31, 2022 for Unit 2 and June 30, 2022 for Unit 3 were weighted (by 75 percent) with the recent 10-year occurrence rate rather than the overall occurrence rate.
However, the following items are not clear to the staff: (1) whether the weighting factor for the overall occurrence rate (i.e., cycle accumulation rate since the start of plant operation) is 25 percent; and (2) whether the cycle estimations based on the applicants approach are in agreement with the actual transient cycles.
- 4. Clarify whether the weighting factor for the overall occurrence rate (i.e.,
cycle accumulation rate since the start of plant operation) is 25 percent. If not, describe the weighting factor for the overall occurrence rate and the technical basis of the weighting factor.
- 5. Clarify whether the cycle estimations based on the applicants approach are in agreement with the actual transient cycles. If not, provide justification for why the applicants approach for cycle projections using the specified weighting factors is adequate.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 96 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 2
4.3.1 4.3-2 SLRA Tables 4.3.1-1 and 4.3.1-2 indicate the Unit 2 turbine roll and increase to rated power and the Units 2 and 3 main steam fill during flood-up transient are projected to exceed the design transient cycles before 80 years (i.e., 160 and 20 cycles, respectively). The transient cycle is also called the transient occurrence.
In comparison, the applicant dispositioned the fatigue waiver TLAAs for reactor pressure vessel components in accordance with 10 CFR54.21(c)(1)(ii), as discussed in SLRA Section 4.3.4.
The applicant also dispositioned the fatigue TLAA for the Unit 2 jet pump riser repair/mitigation clamps in accordance with 10 CFR54.21(c)(1)(ii), as discussed in SLRA Section 4.3.6.2.
In addition, the applicant dispositioned the fatigue TLAAs for the isolation condensers and the Unit 2 replacement core spray piping in accordance with 10 CFR54.21(c)(1)(i), as discussed in SLRA Sections 4.3.7 and 4.7.5, respectively.
Given the TLAA dispositions for the components discussed above in accordance with (i) or (ii) (i.e., not using the Fatigue Monitoring AMP), the staff needs clarification on whether the 80-year projected cycles of the Unit 2 turbine roll and increase to rated power transient and the Units 2 and 3 main steam fill during
- 1. Given the TLAA dispositions for the components discussed in the issue section in accordance with (i) or (ii)
(i.e., not using the Fatigue Monitoring AMP), clarify whether the 80-year projected cycles of the Unit 2 turbine roll and increase to rated power transient and the Units 2 and 3 main steam fill during flood-up transient, which exceed the design cycles, may affect the validity of these TLAA dispositions (e.g., affecting the validity of the fatigue waiver or resulting in the 80-year CUF values greater than 1.0).
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 97 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request flood-up transient, which exceed the design cycles, may affect the validity of these TLAA dispositions (e.g., affecting the validity of the fatigue waiver or resulting in the 80-year CUF values greater than 1.0).
SLRA Section 4.3.2, ASME Section III, Class 1 Fatigue Analyses Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.3.2 4.3.5 4.3-13 SLRA Section 4.3.2 addresses the fatigue TLAAs for the ASME Code Section III, Class 1 piping systems.
SLRA Section 4.3.2 explains that the Fatigue Monitoring AMP (SLRA Section B.3.1.1) includes requirements that initiate corrective actions if any cumulative usage factor (CUF) or environmentally adjusted CUF (CUFen) values exceed 80 percent of the ASME Section III acceptance criterion (i.e., 1.0).
In comparison, UFSAR Section 5.4.1.2.2 indicates that Dresden Unit 2 pipe whip restraints 0201A-G-105 and 0202B-G-106 have been removed from the
- 6. Given that UFSAR Section 5.4.1.2.2 addresses the HELB location postulation per BTP MEB 3-1 for the recirculation piping (i.e., part of the reactor coolant system), clarify whether the HELB postulation for the recirculation piping uses the CUF criterion.
- 7. In light of the HELB analysis of the recirculation piping, clarify whether the applicants fatigue monitoring needs to include Class 1 high-energy line locations in order to
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 98 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request reactor recirculation discharge and suction piping utilizing the relaxation in arbitrary intermediate pipe rupture requirement from NRC Generic Letter 87-11 and Branch Technical Position (BTP) MEB-3-1 (ADAMS Accession No. ML19137A335).
The staff noted that BTP MEB 3-1 includes the high-energy line break (HELB) postulation criterion for Class 1 piping involving a CUF threshold (i.e., 0.1).
However, SLRA Section 4.3.2 does not clearly discuss whether the fatigue monitoring needs to include Class 1 high-energy line locations to confirm the continued validity of the HELB location postulation against the CUF criterion (0.1).
confirm the continued validity of the HELB location postulation by using the CUF criterion (CUF of 0.1). If so, clarify whether the Fatigue Monitoring AMP needs to be enhanced to perform fatigue monition against the CUF criterion for Class 1 HELB location postulation. If not, provide the basis for why such fatigue monitoring in relation to HELB analysis is not needed.
- 8. Based on the discussion above, clarify whether the SLRA (e.g.,
SLRA Sections 4.3.2 and 4.3.5) and related discussion regarding the HELB TLAA need to be revised to include the Class 1 HELB location postulation involving the CUF criterion. If so, provide the additional discussion on the evaluation and disposition for the Class 1 HELB TLAA.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 99 SLRA Section 4.3.3, Environmental Fatigue Analyses for RPV and Class 1 Piping Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.3.3, Table 4.3.1-3 4.3-15 4.3-11 SLRA Section 4.3.3 addresses the environmental fatigue (also called environmentally assisted fatigue or EAF) for the reactor pressure vessel and Class 1 piping systems.
SLRA Table 4.3.1-3 describes the limiting EAF locations (also called sentinel locations) including the NUREG/CR-6260 locations and other plant-specific locations that may be more limiting than the NUREG/CR-6260 locations.
However, the staff noted that SLRA Tabel 4.3.1-3 does not clearly identify the following NUREG/CR-6260 locations as limiting locations: (1) reactor vessel bottom head; and (2) residual heat removal (RHR)
Class 1 piping.
- 9. Clarify which specific limiting locations in SLRA Table 4.3.1-3 correspond to the following NUREG/CR-6260 locations: (1) reactor vessel bottom head; and (2) residual heat removal (RHR) Class 1 piping. If these locations are bounded by the other limiting locations, provide the following information regarding these NUREG/CR-6260 locations and bounding limiting locations to confirm the bounding nature of the other limiting locations: (1) material; (2) environmental fatigue correction factor (Fen); and (3) environmentally adjusted cumulative usage factor (CUFen).
2 4.3.3, Table 4.3.1-3 4.3-15 4.3-11 SLRA Table 4.3.1-3 identifies the Pipe SRVDL A, B, C, D and E locations as limiting EAF locations.
However, SLRA Table 4.3.1-3 does not clearly describe which specific piping system and component are associated with these limiting locations.
In addition, the staff noted that SLRA Table 4.3.1-3 does not address the limiting locations related to high energy line break (HELB) location postulation that
- 1. Clarify which specific piping system and component are associated with the limiting locations discussed in the issue section (SRVDL A, B, C, D and E locations).
- 2. Clarify whether the applicants existing fatigue analyses in the current licensing basis include an analysis that evaluates the HELB location postulation based on a CUF criterion of 0.1.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 100 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request involves a cumulative usage factor (CUF) criterion of 0.1. The staff needs clarification on whether the applicants existing fatigue analyses in the current licensing basis include an analysis that evaluates the HELB location postulation based on a CUF criterion of 0.1.
3 4.3.3, Table 4.3.1-3 4.3-15 4.3-11 SLRA Table 4.3.1-3 describes the limiting EAF locations and their CUFen values.
However, SLRA Table 4.3.1-3 does not clearly describe the thermal zones evaluated in the EAF analysis and specific materials of fabrication associated with the limiting EAF locations.
In addition, the SLRA does not clearly discuss whether a thermal zone may be bounded by another thermal zone in the EAF analysis so that a limiting location may not be identified for a thermal zone.
- 1. Clarify the thermal zones and materials of fabrication associated with the limiting EAF locations described in SLRA Table 4.3.1-3.
- 2. Clarify whether a thermal zone may be bounded by another thermal zone in the EAF analysis so that a limiting location may not be identified for a thermal zone. In addition, discuss, if any, specific thermal zones bounded by another thermal zone in terms of the identification of limiting EAF locations and the applicants approach used to determine that a thermal zone is bounded by another thermal zone.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 101 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 4
4.3.3 4.3-15 LRA Section 4.3.3 addresses the screening EAF evaluation to determine the plant-specific EAF locations that may be more limiting the NUREG/CR-6260 location.
The LRA section explains that the screening evaluation uses bounding environmental fatigue correction factor (Fen) and CUFen values. However, the LRA does not clearly discuss how the bounding Fen and CUFen values are calculated.
Specially, the following items are not clear to the staff: (1) how the bounding temperature is calculated for each location (e.g., use of the maximum temperature of the thermal zone for the component location or the component-specific maximum temperature); (2) how the bounding strain rate is calculated; and (3) how bounding sulfur content is calculated for the components fabricated with carbon or low alloy steel.
In addition, the LRA indicated that the detailed EAF evaluation was performed to refine the CUFen values to reduce the excessive conservatism associated with the screening CUFen values. However, the LRA does not clearly discuss how the applicant refined the CUFen values for the limiting locations.
- 1. Discuss the following items related to the applicants approach to calculate the bounding Fen and CUFen values in the screening evaluation: (1) how the bounding temperature is calculated for each location (e.g., use of the maximum temperature of the thermal zone for the component location or the component-specific maximum temperature); (2) how the bounding strain rate is calculated; and (3) how bounding sulfur content is calculated for the components fabricated with carbon or low alloy steel.
- 2. Discuss how the detailed EAF evaluation refined the CUFen values by reducing the excessive conservatism associated with the screening CUFen values. As part of the discussion, clarify the applicants approach used in the detailed EAF evaluation in terms of determining the transient temperature, strain rate and evaluated cycles, as compared to the approach used in the screening evaluation.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 102 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 5
4.3.3 4.3-16 The following report indicates that in the detailed EAF analysis for the limiting locations, the Fen calculations use the average temperature approach considering the threshold temperature for a material type (Structural Integrity Associates (SIA) 2200483.305P, Environmentally Assisted Fatigue Calculations for Sentinel Locations at Dresden, Revision 1).
However, the SLRA does not clearly discuss the following: (1) whether the average temperature approach is used only for simple, linear transients; and (2) if not, why the conservatism of the applicants approach is comparable to that of the modified rate approach described in NUREG/CR-6909, Rev. 1, Section 4.4 (i.e., plant-specific demonstration of the adequacy of the applicants approach).
- 1. Clarify the following: (1) whether the average temperature approach is used only for simple, linear transients; and (2) if not, why the conservatism of the applicants approach is comparable to that of the modified rate approach described in NUREG/CR-6909, Rev.
1, Section 4.4 (i.e., plant-specific demonstration of the adequacy of the applicants approach).
SLRA Section 4.3.4, ASME Section III, Class 1 Fatigue Waivers Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.3.4 4.3-17 SLRA Section 4.3.4 addresses the fatigue waiver TLAAs for ASME Code Section III, Class 1 components (e.g., reactor pressure vessel steam outlet nozzle and vent nozzle).
Specifically, SLRA Table 4.3.4-2 describes the number of transient cycles assumed in the fatigue waiver reevaluation for 80 years of operation for each
- 10.
Describe the 80-year projected cycles in comparison with the cycles assumed in the 80-year fatigue waiver reanalysis to confirm that the transient cycles evaluated in the fatigue waiver reanalysis are bounding for the 80-year projected cycles.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 103 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request of the ASME Code Section III, N-415.1 criteria for fatigue waiver (i.e., N-415.1(a) through (f)).
However, SLRA Table 4.3.4-2 does not clearly describe the 80-year projected cycles in comparison with the cycles assumed in the 80-year fatigue waiver reanalysis.
In addition, SLRA Table 4.3.4-2 does not describe the following information on the N-415.1(f) criterion regarding significant mechanical load fluctuations: (1) transients assumed in the fatigue waiver reanalysis, (2) transient cycles assumed in the existing waiver analysis, (3) transient cycles assumed in the fatigue waiver reanalysis, (4) 80-year projected cycles of the evaluated transients and (5) whether the fatigue waiver criterion is met in the fatigue waiver reanalysis.
- 11.
Describe the following information on the N-415.1(f) criterion regarding significant mechanical load fluctuations: (1) transients assumed in the fatigue waiver reanalysis, (2) transient cycles assumed in the original waiver analysis, (3) transient cycles assumed in the fatigue waiver reanalysis, (4) the 80-year projected cycles of the evaluated transients to confirm that the transient cycles evaluated in the fatigue waiver reanalysis are bounding for the 80-year projected cycles and (5) whether the fatigue waiver criterion is met in the fatigue waiver reanalysis.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 104 SLRA Section 4.3.5, ASME Section III, Class 2 & 3, And ANSI B31.1 Allowable Stress Analyses And Associated Helb Analyses Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.3.5 4.3-20 SLRA Section 4.3.4 addresses the fatigue TLAAs for the non-Class 1 piping systems (i.e., ASME Section III Class 2 and 3 and ANSI B31.1 piping systems). The TLAAs are related to the allowable stress analyses and high energy line break (HELB) analyses for the piping systems.
SLRA Section 4.3.4 indicates that some non-Class 1 high energy piping locations were selected for HELB analyses at intermediate piping locations where either circumferential or longitudinal stresses assessed for normal and upset plant conditions exceeded 0.8 (Sh + SA) or expansion stresses exceeded 0.8 SA. In these analyses, Sh is the stress calculated by the rules of NC-3600 and ND-3600 of the ASME Section III code, and SA is the allowable stress range for expansion calculated by the rules of NC-3600 of the ASME Section III or ANSI B31.1 code.
The SLRA also explains that the allowable stress (Sh) and stress range (SA) are a function of the stress range reduction factor for cyclic conditions, which depends on transient cycles.
In contrast, the staff noted that, as discussed in ASME Code Section III, NC-3611.1, Sh is the basic material allowable stress at maximum (hot)
- 1. Given that the Sh values are material properties as a function of temperature but are not time-dependent, explain why SLRA Section 4.3.5 indicates that Sh is a function of the stress range reduction factor that depends on transient cycles.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 105 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request temperature from the allowable stress tables (i.e.,
ASME Code Section III, Appendix I, Tables I-7.1 through I-7.3). therefore, the staff noted that the Sh values are material properties as a function of temperature but are not time-dependent.
2 4.3.5 4.3-20 SLRA Section 4.3.4 addresses the fatigue TLAAs for the non-Class 1 piping systems (i.e., ASME Section III Class 2 and 3 and ANSI B31.1 piping systems). The TLAAs are also related to the allowable stress and HELB analyses for the piping systems.
The TLAAs regarding allowable stress analyses rely on the implicit fatigue analysis provisions in the ASME Code Section III and ANSI B31.1 code.
These provisions allow no reduction in the allowable stress range for thermal expansion stresses below 1.0 if the number of equivalent full temperature cycles does not exceed 7000 cycles.
In addition, SLRA Table 4.3.5-2 describes the conservative 80-year projected cycles for the non-Class 1 piping systems that are affected by the transients other than the reactor coolant system (RCS) transients in SLRA Tables 4.3.1-1 and 4.3.1-
- 2.
- 1. Discuss how the 80-year projected cycles listed in SLRA Table 4.3.5-2 were determined (e.g., based on piping system design information, plant operation procedures, test requirements, UFSAR information and specific system-level knowledge). As part of the discussion, clarify how the pre-operational testing cycles were determined for the piping systems.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 106 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request However, SLRA Section 4.3.5 does not clearly describe how the 80-year projected cycles were determined (e.g., based on piping system design information, plant operation procedures, test requirements, UFSAR information and specific system-level knowledge).
SLRA Section 4.3.6, Reactor Pressure Vessel Internals Fatigue Analyses Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.3.6.2 4.3-27 SLRA Section 4.3.6.2 addresses the thermal fatigue TLAA for the DNPS Unit 2 jet pump riser repair/mitigation clamps (also called jet pump riser brace (JPRB) clamps).
SLRA Section 4.3.6.2 provides the following explanation regarding the JPRB clams. During the fall 2001 refueling outage for DNPS Unit 2, a crack was detected in the JPRB for jet pumps 9/10. The SLRA also explains that a mechanical clamping system designed to structurally replace these welds was installed on both jet pump risers in 2003.
However, the SLRA does not clearly discuss which specific welds of the jet pump assembly are structurally replaced by the JPRB clamps.
In addition, the staff noted that typically two jet pumps share one riser pipe. However, the SLRA
- 2. Clarify which specific welds of the jet pump assembly (jet pumps 9/10) are structurally replaced by the JPRB clamps. As part of the discussion, clarify the specific location of the JPRB where the mitigation clamp is installed.
- 3. Given that typically two jet pumps share one riser pipe, clarify whether the mechanical clamps were installed on the two jet pump riser pipes of the jet pumps 9/10.
- 4. Discuss the following items: (1) how many jet pumps are installed inside the Unit 2 reactor pressure vessel; (2) how many jet pump riser pipes and how many JPRBs are installed in the Unit 2 reactor pressure
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 107 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request explains that the mechanical clamp system was installed on both jet pump risers of the jet pumps 9/10, indicating that the jet pumps 9/10 have two riser pipes rather one riser pipe.
SLRA Section 4.3.6.2 further explains that the JPRB clamps were also installed on the other 19 JPRBs of Unit 2 to preclude high cycle fatigue cracking concerns. However, the SLRA section does not clearly discuss the following items related to the JPRB clamps: (1) how many jet pumps are installed inside the Unit 2 reactor pressure vessel; (2) how many jet pump riser pipes and how many JPRBs are installed inside the Unit 2 reactor pressure vessel; (3) frequency and method of the periodic inspections for the JPRB clamps; and (4) whether the results of these inspections confirm the absence of cracking in the clamps due to fatigue or other degradation mechanisms (e.g., stress corrosion cracking).
vessel; (3) frequency and method of the periodic inspections for the JPRB clamps; and (4) whether the results of these inspections confirm the absence of cracking in the clamps due to fatigue or other degradation mechanisms (e.g.,
2 4.3.6.2 4.3-27 SLRA Section 4.3.6.2 addresses the thermal fatigue TLAA for the DNPS Unit 2 jet pump riser repair/mitigation clamps (also called jet pump riser brace (JPRB) clamps).
SLRA Section 4.3.6.2 indicates that the number of startup transient cycles assumed in the 40-year design analysis was scaled up to a 50-year design life in the fatigue TLAA (i.e., fatigue reevaluation for the clamps).
- 2. Clarify whether the startup transient evaluated in the fatigue TLAA for the JPRB clamps is equivalent to the heatup transient listed in SLRA Table 4.3.1-1. In addition, clarify whether the design cycles and 80-year projected cycles for the startup transient (SLRA Section 4.3.6.2) are equivalent to those for the heatup transient, respectively (SLRA Table 4.3.1-1).
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 108 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request However, the staff needs clarification as to whether the startup transient is equivalent to the heatup transient listed in SLRA Table 4.3.1-1 that describes the reactor coolant system design transients and their design cycles and 80-year projected cycles.
- 3. If the startup transient and the associated design and 80-year projected cycles are not equivalent to the heatup transient and the associated design and 80-year projected cycles (SLRA Table 4.3.1-1), clarify the definition of the startup transient and the associated design cycles and 80-year projected cycles as evaluated in SLRA Section 4.3.6.2.
SLRA Section 4.3.7, Fatigue Analysis of the Isolation Condensers Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.3.7 4.3-27 SLRA Section 4.3.7 addresses the fatigue TLAA for the isolation condenser.
SLRA Section 4.3.7 explains the specification for the isolation condensers and the supporting system piping and components requires analysis for 250 shutdown depressurization occurrences in the original design life of 40 years.
SLRA Section 4.3.7 also indicates that 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 SLRA further explains that the limiting thermal shock event is caused by an isolation event that raises the fluid
- 5. Clarify whether the shutdown depressurization transient, which involves 250 cycles required to be evaluated by the specification, is equivalent to the combined pressure, thermal, and thermal shock transient, which involves 280 cycles as the maximum allowable cycles for the limiting location (tube-to-tubesheet junction). If not, explain the difference between these transients in terms of the transient definition and clarify why these two different transients are used in the fatigue analysis of the isolation condenser as design transients.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 109 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request on the secondary side of the isolation condenser to boiling at atmospheric pressure, or 212 °F, followed by injection of makeup water as cold as 70 °F into the secondary side.
However, it is not clear to the staff whether the shutdown depressurization transient, which involves 250 cycles required to be evaluated by the specification, is equivalent to the combined pressure, thermal, and thermal shock transient, which involves 280 cycles as the maximum allowable cycles for the limiting location (tube-to-tubesheet junction).
In addition, SLRA Tables 4.3.1-1 and 4.3.1-2, which address the design transients for Units 1 and 2 respectively, indicate that the number of design cycles for the operation of isolation condenser transient is 240 cycles. This number is slightly less than the 250 cycles of the shutdown depressurization transient discussed above. The staff needs to resolve this apparent inconsistency between the design cycles (i.e., 240 cycles versus 250 cycles).
- 6. Resolve the apparent inconsistency between the 240 design cycles in SLRA Tables 4.3.1-1 and 4.3.1-2 (i.e., cycles of the operation of isolation condenser transient) and the 250 design cycles in SLRA Section 4.3.7 (i.e., cycles of the shutdown depressurization transient).
- 7. If there is any difference among the shutdown depressurization transient, operation of isolation condenser transient and combined pressure, thermal and thermal shock transient discussed above, provide justification for why these design transients for the fatigue analysis of the isolation condenser are not equivalent to one another.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 110 SLRA Section TLAA: 4.2.3 Reactor Pressure Vessel Adjusted Reference Temperature (ART) Analyses Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.2.3 4.2-23 Table 4.2.3-1 lists ART values for Unit 2 Beltline materials, including plates and welds. The table presents, in order, the Shell Ring 1 plates, Shell Ring 2 plates, Ring 2 axial welds, and moves to Unit 1 Shell Ring 2 Axial Welds, before finishing with Unit 2 Shell Ring 1 to Shell Ring 2 Girth/Circumferential Welds.
Please clarify if the heading Unit 1 Shell Ring 2 Axial Welds is correct. If it is not, please provide the correct heading for weld materials listed under this.
2 4.2.3 4.2-25 Table 4.2.3-1 presents three electroslag (ES) axial welds in the Unit 1 Shell Ring 2 Axial Welds section, with a note (1) on each ES weld stating that while the azimuthial angle 317° ES weld is the controlling weld, welds at 77° and 197° were included for completeness.
Table 4.2.3-3 provides similar information for Unit 3, with ES welds called out for both Rings 1 and 2. Only the controlling ES weld is listed for each shell ring, 317° for Ring 1 and 142° for Ring 2.
Notes 1 and 3 on Table 4.2.3-3 list two other ES welds for each of Ring 1 and 2 for Unit 3, with reference to a document containing their fluence values and material properties. Is there a reason these four welds were not included in Table 4.2.3-3 for completeness similar to Table 4.2.3-1?
Fire Protection Scoping and Screening Section:
2.3.3.8, Fire Protection System Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
Section 2.3.3.8 2-3-75 Table 2.3.3-8 of the SLRA excludes the following fire protection components from the scope of subsequent license renewal and from being subject to an aging management review (AMR):
Halon 1301 storage bottles Pump casing (pressure maintenance jockey pump)
Verify whether the listed system/components are within the scope of subsequent license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). If they are not within the scope of license renewal and
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 111 Filter housing Orifice Standpipe risers Intake traveling screen/trash rack Floor drains for removal of fire-fighting water Station transformer fire suppression system and components Seismic support for standpipes system piping Passive components in diesel driven fire pump engine, Heat exchanger (diesel fire water pump cooler) shell side components.
Heat exchanger (diesel fire pump water pump cooler) tubes.
not subject to an AMR, the staff requests that the applicant justify their exclusion.
SLRA Section 4.7.5, Unit 2 Core Spray Replacement Piping Fatigue and Leakage Assessment Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.7.5 4.7-8 SLRA Section 4.7.5 addresses the fatigue and leakage TLAAs for Dresden Unit 2 core spray replacement piping.
The applicant explained that, in November 2009, all four lower sections of the core spray system were replaced and thus removed all known piping flaws of the core spray piping inside the Unit 2 reactor pressure vessel.
The applicant also indicated that, since the fatigue and leakage analyses for the replacement piping assumed the 40-year service life since the replacement (November 2009) until November 2049, the analyses were reevaluated to extend the evaluation period by additional 5 years (total 45
- 8. Describe the transients and transient cycles used in the existing 40-year fatigue analysis for the core spray pipe flange to confirm that the 40-year cycle accumulation rate reasonably represents the cycle accumulation rate anticipated for the reevaluation period (i.e., 45 years).
- 9. As baseline information, discuss the degradation mechanism that caused cracks in the previous replaced piping (e.g., intergranular stress corrosion cracking).
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 112 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request years) beyond the end of subsequent period of extended operation for Unit 2 (i.e., December 2049).
In relation to the existing 40-year fatigue evaluation, the applicant indicated that the pipe flange location is the limiting location that has the maximum cumulative usage factor (CUF) of the replacement core spray piping.
However, the applicant did not clearly describe the transients and transient cycles used in the existing 40-year fatigue analysis for the core spray pipe flange to confirm that the 40-year cycle accumulation rate reasonably represents the cycle accumulation rate anticipated for the reevaluation period (i.e., 45 years).
2 4.7.5 4.7-8 SLRA Section 4.7.5 addresses the fatigue and leakage TLAAs for Dresden Unit 2 core spray replacement piping.
The applicant explained that the. In November 2009, all four lower sections of the core spray system were replaced and thus removed all known piping flaws of the core spray piping inside the Unit 2 reactor pressure vessel.
The applicant also indicated that, since the fatigue and leakage analyses for the replacement piping assumed the 40-year service life since the replacement (November 2009) until November
- 1. In relation to the leakage reassessment, clarify the following item: (1) 45-year allowable corrosion acceptance criteria (e.g.,
allowable thickness reduction) for the leakage paths fabricated with stainless steel, alloy 718 and alloy X-750; (2) 40-year and 45-year projected corrosion extents (e.g.,
thickness reductions) for the leakage paths fabricated with the materials discussed above; and (3) how the 45-year corrosion acceptance criteria were determined to meet the leakage acceptance
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 113 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 2049, the analyses were reevaluated to extend the evaluation period by additional 5 years (total 45 years) beyond the end of subsequent period of extended operation for Unit 2 (i.e., December 2049).
The applicant indicated that the leakage reassessment concluded that additional 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 existing calculated leakage.
However, the applicant did not clearly describe the following items related to the conclusion of the leakage reassessment: (1) 45-year allowable corrosion acceptance criteria (e.g., allowable thickness reduction) for the leakage paths fabricated with stainless steel, alloy 718 and alloy X-750; (2) 40-year and 45-year projected corrosion extents (e.g., thickness reductions) for the leakage paths fabricated with these materials; and (3) how the 45-year corrosion acceptance criteria were determined to meet the leakage acceptance criteria for the leakage paths fabricated with these materials.
criteria for the leakage paths fabricated with the materials discussed above.
- 2. Describe, if any, additional information supporting the applicants conclusion that additional 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 existing calculated leakage.
SLRA Section AMP: B.2.1.7 - BWR Vessel Internals Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.7 B
B-48 Throughout the SLRA and the associated Program Basis Document (PBD), the applicant references Clarify why the use of alternate revisions of BWRVIPs than those cited
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 114 DR-PBD-AMP-XI.M9 Rev. 1 4 - 34 BWRVIPs as the basis for aging management of the noted components. Specifically, the applicant proposes to utilize BWRVIP-76 Revision 1-A, BWRVIP-25 Revision 1-A, BWRVIP 18 Revision 2-A, BWRVIP-41 Revision 4-A, and BWRVIP-183 Revision A in lieu of the revisions of those documents cited in GALL-SLR.
Although the revisions being proposed for use during the subsequent period of extended operation are approved for generic use by the NRC, they are still variances from the guidance in GALL-SLR.
in GALL-SLR is not noted as an exception to the BWR Vessel Internals AMP.
2 B.2.1.7 DR-PBD-AMP-XI.M9 Rev. 1 B-46 12, 15 The SLRA and the associated PBD, state the applicant will use BWRVIP-180 Revision 1 in lieu of Revision 0 which is referenced in GALL-SLR.
Provide technical justification for the use of BWRVIP-180 Revision 1.
Discuss whether the SLRA would need to be revised to provide this technical basis 3
B.2.1.7 DR-PBD-AMP-XI.M9 Rev. 1 B-45 11, 46 GALL-SLR recommends the use of BWRVIP-02-A, Revision 2 for repair design criteria of the core shroud.
The BWR Vessel Internals AMP in the SLRA and Page 11 of the PBD state that the repair design criteria of BWRVIP-02-A were used. However, reference 4.2.5 on page 46 of the PBD cites BWRVIP-02 Revision 2-A.
Specify which revision of BWRVIP-02 is used since the SLRA appears to reference an earlier revision.
If it is an earlier revision than BWRVIP-02 Revision 2-A, provide technical rationale for the use of an earlier revision than specified by GALL-SLR.
4 B.2.1.7 DR-PBD-AMP-XI.M9 Rev. 1 B-46 14 GALL-SLR notes that inspection and evaluation guidelines for the jet pump assembly utilize BWRVIP-41 and BWRVIP-138 Revision 1-A.
The SLRA doesnt specify whether BWRVIP-138 Revision 1-A is used. However, page 14 of the PBD states that BWRVIP-138 is used as part of the BWR Vessel Internals AMP.
Specify if BWRVIP-138 Revision 1-A is used as part of the BWR Vessel Internals AMP. If not, provide technical justification for the difference from the GALL-SLR guidance.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 115 5
B.2.1.7 B-46 The SLRA states:
The DNPS, Unit 2 and Unit 3 original equipment General Electric steam dryers were replaced with General Electric steam dryers in 2007 and 2006, respectively. DNPS inspections and evaluations are performed in accordance with the BWRVIP-139 Revision 1-A and GEH 0000-0085-8689 R0. The repair design criteria in BWRVIP-181-A would be utilized in preparing a repair plan for the steam dryer.
The applicability of BWRVIP-139 Revision 1-A is limited to the configurations described in Section 2.3 of the report which has resulted in some applicants needing to take exception to the use of BWRVIP-139 Revision 1-A from GALL-SLR guidance. However, section 2.3.15 of BWRVIP-139 Revision 1-A includes the configuration of the replacement steam dryer design at Dresden Units 2 and 3.
Confirm that the configuration of the replacement steam dryers is in accordance with section 2.3.15 of BWRVIP-139 Revision 1-A.
6 BWRVIP-76 Revision 1-A (AAI 8)
C-15 Throughout the SLRA, the applicant cites BWRVIP-100-A as part of their BWR Vessel Internals AMP in accordance with GALL-SLR guidance. In response to BWRVIP-76 Revision 1-A (AAI 8), the applicant cites BWRVIP-100 Revision 1-A.
BWRVIP-100 Revision 1-A was subject to a Part 21 notice (ML21084A164) due to non-conservatism.
Confirm whether BWRVIP-100 Revision 1-A is/was used as part of the BWR Vessel Internals AMP at Dresden Units 2 and 3.
If BWRVIP-100 Revision 1-A is/was used, explain how Dresden responded to the Part 21 notice.
7 BWRVIP-315 (Condition 5)
Limitation 2 C-23, C-24 BWRVIP-315-A Section 4.5.1 Limitation 2 states:
owners submitting an application for operation beyond 60 years (e.g., an SLRA in the U.S.) should either commit to implementing a future version of BWRVIP-47 that addresses extended operations or Given that NRC does not have a regulatory footprint related to the practices and guidelines in BWRVIP-94 and NEI 03 discuss how limitation 2 in BWRVIP-315-A can be addressed for the Dresden SLRA such that there is some NRC regulatory durability (e.g.,
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 116 propose a set of plant-specific activities to manage age-related degradation of CRGTs.
In response, the SLRA states:
new or revised inspection recommendations in BWRVIP-47 are required to be implemented in accordance with BWRVIP-94-R4: BWR Vessel and Internals Project, Program Implementation Guide and NEI 03-08, Guideline for the Management of Material Issues. Future revisions of BWRVIP-47 that address DNPS extended operations will be implemented as applicable.
BWRVIP-94 and NEI 03-08 are not NRC approved documents.
an enhancement/statement regarding the future implementation of BWRVIP-47 in Section A.2.1.7 of the FSAR or a Subsequent License Renewal Commitment in Table A.5?
SLRA Section TLAA: 4.7.4 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
4.7.4 A.4.7.4 4.7-7 A-68 The applicant states in the TLAA evaluation and the UFSAR that, DNPS has made a commitment to maintain the inspection of these nozzles within the ISI Program.
Commitment 5 of NUREG-1796 contained a commitment to incorporate the BWR Feedwater Nozzle inspections into the ISI Program but Table A.5 of the SLRA doesnt contain a specific commitment regarding BWR Feedwater Nozzles.
Confirm that the statement regarding the implementation of a commitment to inspect the BWR Feedwater nozzles is referring to Commitment 5 of NUREG-1796 or to Commitment 1 of Table A.5 of the current application. If not, please specify which subsequent license renewal commitment is being referenced.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 117 SLRA Section TLAA: 4.7.6 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
Appendix C C-11 BWRVIP-74-A (AAI 14) states:
Subsection IWB-3600 to Section XI of the ASME Code states that flaw indications that exceed the size of allowable indications defined in IWB-3500 may be evaluated by analytical procedures, such as described in Appendix A of Section Xl of the ASME Code in order to calculate its growth until the next inspection or the end of service lifetime of the component. Therefore, components that have indications that have been previously analytically evaluated in accordance with IWB-3600 until the end of the 40-year service period shall be re-evaluated for the 60-year service period corresponding to the LR term. This Is Renewal Applicant Action Item 14.
In response to BWRVIP-74-A (AAI 14) the applicant states There are no components within the ASME Code Class 1 reactor coolant pressure boundary with indications that have been previously analytically evaluated until the end of the 60-year service period.
SLRA Section 4.7.6 contains a TLAA associated with the DNPS Unit 2 closure flange to upper shell circumferential weld (2-SC4-FLG). It appears that this flaw evaluation in SLRA Section 4.7.6 should have been identified per BWRVIP-74-A (AAI 14).
Explain why TLAA 4.7.6 isnt reported in Appendix C in response to BWRVIP-74-A (AAI 14)?
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 118 SLRA Section B.2.1.23: External Surfaces Monitoring of Mechanical Components Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.23 SLRA Appendix B.2.1.23; B.2.1.10 Dresden Program Basis Document DR-PBD-AMP-XI.M36, Rev. 1 -
Page 17 GALL-SLR The External Surfaces Monitoring of Mechanical Components AMP (XI.M36) states in program element 4, Detection of Aging Effects, that, This program manages the aging effects of loss of material, cracking, hardening or loss of strength, reduced thermal insulation resistance, loss of preload for HVAC closure bolting, and reduction of heat transfer due to fouling using visual inspections.
Dresden SLRA Appendix B.2.1.23, External Surfaces Monitoring of Mechanical Components is silent on loss of preload for HVAC closure bolting but claims consistency without enhancement or exception to the GALL-SLR.
Appendix B.2.1.10, Bolting Integrity states that, Heating and ventilation system bolted joints are managed by the External Surfaces Monitoring of Mechanical Components (B.2.1.23) program.
Dresden Program Basis Document DR-PBD-AMP-XI.M36, Rev. 1 On page 17 of 65, it states that, Loss of preload for HVAC closure bolting in scope of this program will not be monitored as it is not an applicable aging effect.
Please address:
- 1. B.2.1.23 claim consistency with the GALL-SLR yet does not address loss of preload for HVAC closure bolting.
- 2. The program basis document states that loss of preload is not an applicable aging effect but does not provide a justification for that position.
- 3. B.2.1.10 claims that loss of preload is managed by B.2.1.23.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 119 SLRA Sections 3.6.2.3.1, B.2.1.36 - Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
3.6.2.3.1 Table 3.6.1 3.6-14 3.6-28 DNPS SLRA Table 3.6.1 states that AMR 3.6.1-022 is not applicable because there are no aging effects to be managed for bakelite; phenolic melamine or ceramic; or molded polycarbonate insulation material in fuse holders (not part of active equipment) exposed to an air - indoor controlled or uncontrolled environment in Electrical Commodities.
See section 3.6.2.3.1.
The staff notes that SLRA section 3.6.2.3.1 appears not to address aging effects on the insulation materials of the Dresden in-scope fuse holders (not part of active equipment).
Discuss the AMR results for the insulation material of the fuse holders (not part of active equipment) to demonstrate that there are no aging effects to be managed for the insulation material.
2 B.2.1.36 B-208 NUREG 2191 XI.E1 Program Description states:
Adverse localized environments (ALE) are identified through the use of an integrated approach. This approach includes, but is not limited to: (a) the review of EQ program radiation levels, temperatures, and moisture levels; (b) recorded information from equipment or plant instrumentation; (c) as-built and field walk down data (e.g., cable routing data base); (d) a plant spaces scoping and screening methodology; and (e) the review of relevant plant-specific and industry operating experience (OE).
The staff notes that DNPS SLRA section B.2.1.36 program description appears not to include how the ALEs will be identified at DNPS.
Provide a description of how the ALE will be identified at DNPS and include it in SLRA section B.2.1.36.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 120 3
B.2.1.36 B-208 NUREG 2191 XI.E1 Program Description states:
The cable condition monitoring portion of the AMP utilizes component sampling for cable and connection electrical insulation testing, if deemed necessary. The following factors are considered in the development of the electrical insulation sample: the environment including identified adverse localized environments (high temperature, high humidity, vibration, etc.), voltage level, circuit loading, connection type, location (high temperature, high humidity, vibration, etc.) and the electrical insulation composition. The component sampling methodology utilizes a population that includes a representative sample of in-scope electrical cable and connection types regardless of whether or not the component was included in a previous aging management or maintenance program. The technical basis for the sample selection is documented.
The staff notes that DNPS SLRA section B.2.1.36 program description appears not to include a description of the component sampling for cable and connection electrical insulation.
Provide a description (factors considered and methodology for sampling) of the component sampling for cable and connection electrical insulation testing and include the description in the SLRA section B.2.1.36.
4 B.2.1.36 B-209 DNPS SLRA section B.2.1.36 provides the following operating experience:
In May 2021, the 2B reactor recirculation (RR) pump tripped resulting in a power reduction at DNPS Unit 2.
A corrective action program evaluation (CAPE) was performed to address cause and extent of condition.
Provide the CAPE and work order or issue report for this operating experience related to the 2B RR pump trip.
5 B.2.1.36 NUREG 2191 X1.E1 Parameters Monitored or Inspected element states:
Provide the most limiting moisture environment and its basis.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 121 An ALE is a plant-specific condition; therefore, the applicant should clearly define the most limiting temperature, radiation, and moisture environments and their basis. The applicant should also inspect for adverse localized environments for each of the most limiting cable and connection electrical insulation plant environments (e.g., caused by temperature, radiation, moisture, or contamination).
The staff notes that DNPS Program basis document DR-PBD-AMP-XI.E1, Parameters Monitored or Inspected element appears not to include the limit for a moisture environment.
Clarify if DNPS will inspect for ALEs for each of the most limiting cable and connection electrical insulation plant environments (e.g., caused by temperature, radiation, moisture, or contamination).
Include the above actions (most limiting moisture environment and inspection for ALEs for each of the most limiting environments) in the Parameters Monitored or Inspected element of DR-PBD-AMP-XI.E1 for consistency with NUREG 2191 AMP XI.E1.
6 B.2.1.36 NUREG 2191 AMP XI.E1 Detection of Aging Effects element states:
Cable and connection insulation are evaluated to confirm that the dispositioned corrective actions continue to support in-scope cable and connection intended functions during the subsequent period of extended operation.
The staff notes that DNPS Program basis document DR-PBD-AMP-XI.E1, Detection of Aging Effects element appears not to include the evaluation of cable and connection insulation to confirm that the dispositioned corrective actions continue to support in-scope cable and connection intended functions during the subsequent period of extended operation.
Clarify if DNPS will evaluate the Cable and connection insulation to confirm that the dispositioned corrective actions continue to support in-scope cable and connection intended functions during the subsequent period of extended operation.
Include the above action in the Detection of Aging Effects element of DR-PBD-AMP-XI.E1 for consistency with NUREG 2191 AMP XI.E1.
7 The aging management program (AMP) Effectiveness Review (AR 04080012-12) includes an update that states:
-Clarify if these cable damage issues were related to cable in-scope of license renewal.
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 122 Potential Aging Evaluation has been performed for cable damage issues that were found via several IRs.
Refer to evaluation located in ATI 04203856-02.
-Provide the aging evaluation in ATI 04203856-02.
SLRA Section / SLRA Section B.2.1.38/39/40: XI.E3A, B, C Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)
Discussion Question / Request 1
B.2.1.38/39/40 B-220 OE Section 2:
During this assessment, it was identified that manhole inspection and dewatering was not being performed in accordance with procedure guidance.
It was not clear that inspection in SBO MH-1 that trays that appeared to be totally enclosed were adequately assessed to verify there was no residual water remaining inside them, immersing cables inside. This question could be generically applicable depending on other trays such as this in other underground cable banks serviced by manholes.
In addition, corrosion of SBO MH-1 cover and frame was significant and probably contributed to excessive stormwater in leakage to the manhole. This was communicated with the NRC Civil team.
2 B.2.1.38 B-219 OE Section 1:
During annual inspections of the service water pump duct bank, several instances were identified where there was a small build-up of calcium deposits or debris.
Water leakage from service water pump duct bank has been collecting in cable trays in the service water pump room causing pooling and significant corrosion of cable trays. It was not apparent that issue reports had been written to address potential cable
Dresden Nuclear Power Station, Units 2 and 3 Subsequent License Renewal Application Breakout Audit Questions 123 degradation due to submersion.
This question submitted to follow-up on prepared issue reports.
3 B.2.1.38/39/40 B-217,
Remote level monitoring instrumentation is installed in five manholes at DNPS and provides a means to monitor water level so that dewatering activities can be performed on an as-needed basis to prevent cables from becoming submerged.
The licensee was asked about ongoing maintenance and testing of these manhole level monitoring systems. This question was submitted to follow-up on answers to this request.