ML25342A161

Consolidated Comments Palisades November 2025 - ACRS Full Committee

ML25342A161
Person / Time
Site:	Palisades
Issue date:	12/08/2025
From:	Advisory Committee on Reactor Safeguards
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ACRS October 10 Meeting Palisades Nuclear Power Plant 2 Minute Verbal Presentation by Arnie Gundersen You are meeting today behind closed doors to review the supposedly non-public proprietary operational assessment of the safety of Palisades. I am not confident that the publics health and safety are well served by your top secret process.

My technical position on the integrity of the Steam Generators is already on the record. After I presented my concerns to the ACRS, new information became available that is deeply concerning. Holtec asked the NRC for eight relief requests because it has now detected stress corrosion cracks in eight dissimilar metal welds in the reactor coolant system! Two in the hot legs, four in the cold legs and two in the pressurizer.

EIGHT reactor coolant welds have experienced SCC in addition to the 3000 flaws that were sleeved in the steam generators. The entire reactor coolant system is degraded because Holtec chose not to maintain EPRI reactor coolant standards for two years.

Those of us who have appeared before the ACRS to criticize NRC Staff decisions have a joke about the ACRS: The ACRS asks the Staff great questions. The problem is that the ACRS accepts crappy answers! Since I only have two minutes, here is just one good question the ACRS might want to ask the Staff:

Good Question: What is the probability that the Steam Generators will spring a leak before the next refueling?

Poor Answer: We dont know but there are systems in place to mitigate the leak when it occurs.

That answer is an admission that the Staff is relying on mitigation when a Reactor Coolant breach is almost inevitable.

Will you accept that crappy answer? After this meeting is finished, will you be able to say that you put safety before expediency on the Palisades resurrection??

1 Additional Written Information to The Advisory Committee on Reactor Safeguards Concerning the Safety of the Palisades Nuclear Plant October 10, 2025 Arnold Gundersen In January 1986, two NASA contract engineers identified that the Challenger Space Shuttle was endangered if it were to be launched in cold weather. Those engineers used all the professional channels available to prevent the launch. But the bureaucratic inertia within NASA to maintain the launch schedule caused those NASA engineers to be overruled. We all know the outcome of that safety lapse. I write to you today in the spirit of those two NASA engineers as I continue to express my safety concerns to the members of the ACRS. You provide the last possible public safety oversight before resurrecting the Palisades nuclear plant.

First, I wanted to thank you for allowing me to share my concerns about the condition of the diminished integrity of the Reactor Coolant System at Palisades for five minutes during the Palisades subcommittee hearing on August 21, 2025. And I also want to thank you for your thoughtful Steam Generator questions to the NRC staff during the full committee meeting of September 3, 2025. I appreciate that the ACRS appears to be taking its oversight of the Palisades resurrection precedent seriously.

That said, new information just placed on the Palisades docket has amplified my previously expressed concerns. I know the NRC staff has not been forthcoming with information for me to analyze as an expert. I fear that the NRC staff has not been forthcoming to the ACRS either.

Never in my 54 year professional career have I been more concerned about the integrity of the reactor coolant pressure boundary than I am about the condition of Palisades. Please let me explain.

All operating nuclear reactors are required to provide detailed Steam Generator (SG) Tube Inspection Reports to the NRC identifying flaws discovered during eddy current inspections. Six months after the inspections are completed, these detailed tube inspection reports become available to experts like me in the Public Document Room (PDR). Based on my prior industry experience, I knew that prolonged corrosive chemical exposure from extended shutdowns is deleterious to the metal components in both the Reactor Coolant and Secondary systems. I suspected that degradation was occurring at Palisades after it was permanently closed by Entergy in May 2022 and acquired by Holtec in June of 2022. But I had no hard data from the PDR to support my concerns. The last detailed Palisades SG tube Inspection Report in the PDR is from the 2020 SG inspections performed by Entergy. Five years of tube inspection data on both the primary and secondary systems is lacking from the PDR.

Since Holtec acquired Palisades, it appears to have used regulatory loopholes to avoid filing years of detailed Steam Generator Tube Inspection Reports indicating the extent of the damage.

The NRC Staff has even acknowledged that Holtec has failed to provide some Steam Generator inspection details, which is why the NRC staff delayed issuance of the SG sleeving LAR. Here is the NRCs statement about the cause of that schedule delay:

2 NRC staff has estimated that this licensing request will take approximately 940 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.5767e-4 months <br /> to complete. The NRC staff expects to complete this review by September 30, 2025. Due to the eddy current qualification data not being provided by the licensee, the review date is beyond their originally requested date of August 15, 2025. (March 20, 2025, https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML25076A177)

There are only two publicly available documents that discuss the condition of Palisades SG tubes. The first is the September 18, 2024 Preliminary Notification of Occurrence (PNO)

(ML24262A092) issued by the NRC staff based on their concerns after the shocking August 2024 Holtec SG inspection results. The second is a letter containing meeting notes from October 1, 2024 (ML24262A092) between Holtec and the NRC that summarize the August inspection and make vague promises about follow-up analyses. Thats it. If additional information is in the possession of the NRC staff, it should also be in in the PDR, and there is no such information.

That leads me to the conclusion that the NRC staff is not in possession of some critical Steam Generator tube inspection data from 2024 and 2025 or that the staff does not want the public to analyze the condition of the SGs.

In your September 3, 2025 meeting, the NRC staff told the ACRS that approximately 3,000 sleeves were inserted into about 700 tubes since May of 2025. Each sleeve is 18 inches long, which means that 4,500 feet of sleeves (0.85 miles!) were installed. That is an astounding length of sleeving and is not supported by the publicly available flaw data from the September 18 and October 1, 2024 PDR documents. For an expert like me, it would be a simple matter to compare the existing 2020 Entergy Inspection with both the 2024 and 2025 Holtec Inspections to search for trends and their root cause of the increased cracking indications, but none of the 2024 and 2025 inspection data is available. However, it appears likely that the tube damage that was identified and sleeved in 2025 exceeded the tube damage that was identified in 2024.

The general rule for plugging is that tubes are sleeved or plugged when an indication has reached or exceeded 40% through wall. So a 20% indication will not be plugged but will be reexamined during the next refueling outage based on Electric Power Research Institute (EPRI) water chemistry guidelines. But the chemical hideout at Palisades is anything but normal. When Holtec did examine the tubes in 2024, it found some previously unaffected tubes had Stress Corrosion Crack indications exceeding 80% through wall cracks after remaining in cold unpressurized water for two years. Slow, anticipated crack growth that EPRI assumes is not realistic for Palisades. Hence 3,000 sleeves, already a huge number, may be inadequate to prevent additional tube failures because of hideout before the next Palisades Steam Generator inspections.

Traditionally, eddy current testing begins several inches above the tube sheet. The tube sheet is part of the reactor coolant pressure boundary which is where chemical hideout would be expected to be most prevalent. Because of this hideout, it is not clear that either the SG tubes or the SG tube sheet will survive for even half a year after Palisades resurrection is complete.

Now, new information of degradation has become available. In addition to all the steam generator tube and tube sheet indications indicating both SCC and PWSCC in the steam generator, on August 20, 2025 Holtec filed a series of relief requests (ML25232A195 ) indicating

3 that it has discovered Primary Water Stress Corrosion Cracking (PWSCC) in at least eight dissimilar metal welds within Palisades Primary Coolant System. The affected welds include indications in two hot leg welds, four cold leg welds and two pressurizer welds.

The record indicates that Holtec did not take samples of either primary or secondary water chemistry at Palisades for two years and also that it is aware that Palisades was not in compliance with EPRI water quality guidelines. Clearly the absence of adequate water chemistry control at Palisades and its effect on the primary coolant system boundary are issues that deserve the thorough attention of the ACRS before allowing Palisades to set a new licensing precedent. This is a generic issue, as there are other decommissioned reactors now in the queue to be resurrected that have also not maintained adequate water chemistry during closure.

The existing evidence suggests that the reactor coolant pressure boundary degradation detected was caused by inadequate water chemistry control at Palisades, which places the facility in violation of two General Design Criteria:

Criterion 14Reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Criterion 15Reactor coolant system design. The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

The last time a steam generator tube completely ruptured was at Indian Point more than two decades ago. The condition of both the Primary Coolant System and the Steam Generators is even worse at Palisades with extensive SCC and PWSCC already identified. Luckily Indian Points design allowed it to dump the radioactive steam into the condenser where it was contained. Palisades does not have this feature and would use Atmospheric Dumps to discharge radioactivity directly into the atmosphere.

Previously, I have seen the ACRS advise the NRC staff and vendor (General Electric) of its concerns that regulatory expediency was placed before public safety. About two decades ago, I was one of a few experts who petitioned the ACRS to evaluate Net Positive Suction Head concerns relating to the request for regulatory relief on Containment Overpressure during Boiling Water Reactor Power Uprates. The ACRS did the right thing then by refusing to allow for the containment overpressure relief which was championed by the NRC staff and GE. I have previously applauded the ACRS personally for making that decision.

My concern initially started with SCC and PWSCC discovered in Palisades SGs but new Holtec relief requests have identified significant PWSCC corrosion at eight other locations within the reactor coolant system. The loss of the reactor coolant pressure boundary can lead to previously unimaginable impacts to the general public. The ACRS must be keenly aware of what could

4 happen in the event of primary coolant system failure or a Steam Generator tube failure due to years of neglect from improper wet layup by Holtec at Palisades.

I pray that you will thoroughly question the integrity of the reactor coolant pressure boundary and steam generator tubes caused by Holtecs failure to meet EPRI primary and secondary water chemistry standards before allowing Palisades to set a new licensing precedent.

Thank you, Arnie Gundersen Expert Witness for Beyond Nuclear, Dont Waste Michigan, et al.

Written Comment for the 729th ACRS Full Committee Meeting October 10, 2025 ACRS Written Comment - Palisades Steam Generator Life Cycle Considerations and Chemical Cleaning ConsiderationsAlan Blind

Dear Members of the Advisory Committee on Reactor Safeguards:

In your September 3, 2025 session, NRC staff, in response to your detailed questions, placed significant weight on the planned steam generator cleaning and emphasized the two-year period during decommissioningwhen chemistry was not programmatically maintainedas the primary path forward to ensure public health and safety. Staff also appeared to rely heavily on Framatomes technical evaluations, and that reliancewithout clear evidence of independent expert reviewshould itself be of concern to the ACRS.

While the decommissioning lay-up period deserves consideration, I respectfully urge the Committee to maintain a broader view that accounts for the full thirty-plus years of operating experience with the current Palisades steam generators. Too much emphasis was placed on assuming that two years without chemistry controls were the sole cause of the unexpected rate and severity observed in the 2024 inspection results. The longer history clearly shows that degradation mechanisms were active well before the recent lay-up period, and it is essential that the operational assessment reflect those cumulative effects. NEI 97-06 requires that all degradation mechanisms be evaluated and that uncertainties be summed in determining the appropriate period of operationmeaning the full prior life cycle of the steam generators must be considered.

Also, There must also be a questioning attitude about the specific chemical cleaning technique selected, its practical contribution to the operational assessment inputs, and whether it can realistically address the full life-cycle challenges of these steam generatorsincluding the embedded copper and lead deposits that have historically, up until 2010, and maybe beyond, proven difficult to remove.

My perspective comes from direct responsibility: in 2010, while serving as Engineering Director at Palisades, I was asked to lead an evaluation of whether secondary-side chemical cleaning could meaningfully extend the life of the replacement steam generators. That studycompleted more than a decade before Entergy elected to shut the plant down, and before many additional years of chemical deposit accumulationreached conclusions that remain directly relevant today.

In 2010, we concluded that without a proven chemical cleaning technique capable of removing the embedded and historically difficult copper and lead deposits, outside diameter stress corrosion cracking from these mechanisms would continue.

Even if cleaning could reduce the rate of degradation, it could not, by itself, assure operation through the 2031 license renewal period. We decided to not chemically clean the Steam Generators.

That finding remains directly applicable today. While chemical cleaning may contribute to life extension, it cannot be viewed as a stand-alone solution. The NRC and the Committee must weigh the broader record of operating experience, the persistence of copper and lead contamination, and the historical difficulty of fully removing hardened deposits when judging the adequacy of the operational assessment.

Operating History The Palisades replacement steam generators, Combustion Engineering Model 2530, were installed in late 1990 and entered service in 1991. The tubing is mill-annealed Alloy 600, a material widely recognized in industry for its susceptibility to corrosion. From the outset, over 300 tubes in each generator were preventively plugged, with additional plugging accumulating in every refueling cycle. By 2009, effective plugging rates reached ~5% in both units. Predictive models in the 2010 Entergy study showed that without chemical cleaning, Palisades would likely exceed its 15% plugging limit by the mid-2020s, short of the 2031 license renewal period.

Unique Issues Identified in the 2010 Study The 2010 chemical cleaning evaluation identified several unique challenges at Palisades:

Persistent sludge pile on the tubesheets containing hardened deposits of copper and lead, historically at Palisades, resistant to mechanical lancing.

Copper transport legacy from admiralty brass and copper-nickel condenser and heater tubing, which was replaced only after the steam generators were already in service.

Lead contamination originated from low-pressure turbine rupture discs, which were not replaced until 2012more than 20 years after the steam generators entered operation.

Top of tubesheet collars and tube scale deposits, containing copper and lead, nearly impossible to remove mechanically.

The report concluded that unless copper and lead deposits were removed, long-term service to 2031 was unlikely.

Chemical Cleaning Techniques Four principal methods were reviewed in 2010:

EPRI-SGOG process: most thorough and documented, capable of targeting copper/lead/iron; but costly and operationally intrusive.

ASCA (Advanced Scale Conditioning Agent): less expensive, widely used, but originally intended for maintenance rather than full sludge removal.

High-temperature chemical cleaning (HTCC): aggressive but risky; prior use at Waterford 3 created iron redeposition issues.

Deposit Minimization Treatment (DMT, AREVA, now Framatome):

newer, lower cost, but limited operating experience and uncertain effectiveness against copper/lead.

Unique Considerations at Palisades Because Palisades replacement generators entered service with pre-existing copper/lead contamination and hardened deposits, not all cleaning methods may be fully effective. The ACRS needs to question NRC staff on the particular cleaning method Holtec plans to use. The ASCA and DMT methods in particular may not penetrate hardened sludge piles or collars, raising questions about their adequacy

for the Palisades application. Even the most effective process (EPRI-SGOG) was predicted to provide only partial mitigation, not a guarantee of service through 2031.

Conclusion and Recommendation As you meet with NRC staff on October 10, I urge the Committee to:

1.

Acknowledge, as you have already done, the role of poor lay-up chemistry during decommissioning, but also weigh the prior three decades of degradation mechanisms that are equally important to todays operational assessment.

2.

Question the NRC staff closely on the basis for confidence in the specific cleaning method selected, particularly whether it can address Palisades hardened sludge piles and copper/lead contamination.

3.

Probe whether staff has critically evaluated vendor claims and benchmarked outcomes from other Alloy 600 plants.

4.

Ask the NRC staff how it expects the wide range of degradation mechanisms and they uncertainties to be quantified in the operational assessment.

Closing Remarks In closing, I urge the Committee to keep in mind the unique circumstances at Palisades. Steam Generator Life-cycle management must be considered in full, not just the two-year period of unmaintained secondary chemistry during decommissioning, but the entire thirty-plus years of operation that preceded it. The upcoming operational assessment will necessarily be far more complex than those at other plants, with a wide range of uncertainties for each degradation mechanism.

As required by NEI 97-06, these uncertainties must be treated quantitatively summed as the square root of the sum of the squaresnot minimized or considered in isolation. Finally, Palisades licensing basis for a steam generator tube rupture accident relies solely on the atmospheric dump valves for rapid depressurization when offsite power is available, and on primary system feed-and-bleedan operator-induced LOCAwhen it is not. While this approach is licensed and permissible, it must weigh heavily on any judgment about the acceptability of additional operating periods given the condition and age of these steam generators.

Respectfully submitted, Alan Blind, Retired Baroda, Michigan

Written Comment for the 729th ACRS Full Committee Meeting October 10, 2025 Re-Examining NRC Staffs Tube Integrity Assessment Response to ACRS: A Thought Experiment for ACRS Consideration Alan Blind At the September 3 meeting, in response to the ACRS request to NRC Staff to address the most severe cracks found in the October 2024 inspection, NRC staff stated that Palisades met tube integrity requirements during its last cycle because all 22 suspect tubes passed in-situ pressure testing, showing no rupture or leakage at three times normal operating differential pressure.

While NRC staffs statement is an assessment of tube integrity on the inspection date, I present this thought experimenta back-of-the-envelope review of the most degraded tube, R73C94not as a definitive analysis, but to highlight that the staffs response to the ACRS appeared to lack in-depth intellectual curiosity and

focus on conservative decision making. I respectfully suggest that the Committee probe this issue further in your October 10 meeting.

During Holtecs 2024 inspections, Tube R73C94 was identified as the most degraded and stood out with its long axial and deep penetration crack. It exhibited a 73% through-wall axial crack, 1.7 inches long, located at a tube support platea known site for corrosion and flow stagnation.

Crack Origin and Timing of Degradation Outer Diameter Stress Corrosion Cracking (ODSCC) requires three elements: a susceptible material, an aggressive environment, and sustained stress. The stress component is particularly important, arising from operating pressure differentials across the tube wall as well as thermal stresses from repeated heat-up and cool-down cycles during normal plant operation. During the shutdown and decommissioning phase, when Palisades steam generators were depressurized and at ambient conditions, these stress drivers were absenteven though water chemistry controls had also lapsed. This means that the severe crack found in Tube R73C94 could not have grown significantly during the decommissioning interval; instead, it must be conservatively assumed that the defect was already present at

the time of plant shutdown and entry into decommissioning. This point is important within my thought experiment because it indicates Palisades ended power operations already carrying a tube in near-failure condition, making the margin to rupture even smaller than NRC staff acknowledged.

Basic Concepts: Failure Mechanics and Stress Environment Stress Orientation and Crack Growth: Hoop (circumferential) stress is the dominant stress in thin-wall, pressurized tubes and acts perpendicular to axial cracks. Axial ODSCC forms along the tube axis due to stress concentration, stagnant chemistry, and thermal conditions near the TSP. In R73C94, the crack approached through-wall depth. As the wall thins, remaining ligament stresses increase non-linearly, leading to imminent structural failure.

Likely Failure Characteristics:

• Burst-Type Rupture - Failure would occur rapidly, resulting in a high-pressure rupture. Though the crack is axial, the final rupture would be circumferential due to internal pressure acting radially.

• Leak-Before-Break Unlikely - The high crack length and depth offered

This 4.5-month estimate represents a bounding upper limit. It does not account for the non-linear intensification of hoop stresses as the ligament thins, which could plausibly accelerate failure and lead to rupture sooner under actual operating conditions.

Key Implication Tube R73C94 was approaching structural failure with insufficient margin for detection or operator intervention. The most probable failure mode was a sudden, circumferential rupture at high pressurea scenario that would have posed a serious challenge for operator response and triggered a General Emergency declaration with potential offsite evacuation under Palisades licensing basis.

Thus, while tube integrity could be demonstrated at the time of the October 2024 inspection, the underlying data show Palisades was operating on the edge of a steam generator tube rupture. Continued operation for only a few additional months could have resulted in a markedly different outcome.

I respectfully submit this perspective for the Committees consideration: tube integrity assessments should not be viewed only as point-in-time

demonstrations, but also in the broader context of time-to-failure and the limited margins shown in the October 2024 results.

Alan Blind

Written Comment for the 729th ACRS Full Committee Meeting October 10, 2025 Reassessing NRC Benchmarks for Palisades Steam Generator Tube Integrity: CE Fleet Experience, 2024 Inspection Results, and the Role of Copper/Lead Oxides and Metal Deposits For Future At Power Operations Alan Blind Introduction In this comment, I will first explain why Beaver Valley is not a valid benchmark for Palisades steam generator performance, given the major design differences between Westinghouse and Combustion Engineering (CE) units. I will then compare Palisades operating history to the broader CE fleet, showing that Palisades has already exceeded the end-of-life service years of all other CE steam generators. Next, I will discuss why the extensive tube degradation found during Palisades 2024 inspection should not have been unexpected, based on known Alloy 600 behavior and the well-documented exponential growth of cracks.

Finally, I will highlight the unresolved issue of copper and lead deposits in Palisades steam generators, including the risk that these deposits may have oxidized during the uncontrolled chemistry period in decommissioning, and propose specific questions for the ACRS to raise with NRC staff at the November 10 meeting.

Note: This paper refers to copper and lead deposits in Palisades steam generators. For a more complete explanation of this issueincluding chemical cleaning considerationsI respectfully direct the Committee to my separate comment titled Palisades Steam Generator Life Cycle Considerations and Chemical Cleaning Considerations Benchmarking At the September 3 meeting, an ACRS member asked NRC staff what other plants inform the Palisades design and SG tube evaluations. Staff replied that only one other plant with Alloy 600 tubing remains for benchmarking: Beaver Valley. The staff appeared to place some weight on this benchmark as a validation of their evaluation of Palisades vendors analysis of steam generator tube condition.

This comparison is deeply problematic. Beaver Valley is a Westinghouse three-loop PWR, not a CE two-loop plant like Palisades. That difference matters technically when thinking about ODSCC mechanisms:

Loop & system layout: Westinghouse SGs at Beaver Vally are designed for three-loop primary systems; CE plants like Palisades operate with two large steam generators and four reactor coolant pumps, a different hydraulic and thermal environment.

Support structure architecture: Westinghouse SGs use broached or quatrefoil tube support plates; CE SGs (including Palisades Model 2530s) use egg-crate lattice supports with vertical/diagonal straps and a stay-cylinder region. This geometry dictates how and where ODSCC initiates, as well as how inspections and repairs must be performed.

Tube bundle geometry: Westinghouse SGs use all U-bend tube bundles, while CE SGs combine both: at Palisades, rows 1-18 are U-bends and rows19-165 are square bends, supported by egg-crate lattices, vertical straps, and diagonal straps. This difference in tube geometry creates very different stress and wear environments, as well as distinct inspection challenges.

Tube scale and collars unique to CE: At Palisades, tube scale and top-of-tubesheet (TTS) scale collars remain in place on the outside of tubes. These collars, along with copper/lead scale, are extremely hard deposits nearly impossible to remove mechanically because of the CE designed tri-pitch tube geometry. They directly promote ODSCC in dents, freespan, and egg-crate regions. Westinghouse SGs do not face this same combination of tri-pitch geometry and persistent areas of scaling, including for Palisades, copper/lead collars.

Materials: Both plants may have Alloy 600 tubing, but Palisades specific mill-annealed Alloy 600 (0.75 OD x 0.042 wall), coupled with CE support geometry, has shown a much higher susceptibility to ODSCC than most Westinghouse designs.

Because material, geometry, and scale/collar buildup drive ODSCC progression, Beaver Valley cannot be considered a technically valid benchmark for Palisades.

A More Meaningful Benchmark: The CE Fleet Experience

Rather than relying on Beaver Valley, the more meaningful benchmark is the actual fleet history of CE plants with steam generators of similar mechanical design.

Following are the nominal steam generator effective full-power years (EFPY), as of the year 2010, at which other CE units reached end-of-life and replaced their steam generators:

Palo Verde 1 - 15 EFPY Palo Verde 2 - 14.8 EFPY Palo Verde 3 - 20 EFPY Arkansas Nuclear One 2 - 15.5 EFPY Calvert Cliffs 1 - 19 EFPY Calvert Cliffs 2 - 18 EFPY St. Lucie 1 - 16 EFPY St. Lucie 2 - 20 EFPY Fort Calhoun - 25 EFPY San Onofre 2 - 21 EFPY

San Onofre 3 - 21 EFPY Waterford 3 - 23 EFPY In every case, CE units retired their steam generators well before reaching Palisades present replacement steam generator operating age. Palisades today is estimated at ~28 EFPY of operation on its replacement steam generators, already exceeding the highest steam generator service life of any other CE unit.

Note: I do not indicate the tube alloy used at each of these CE plants, as that information is not always reliably available. It is well understood that allow 600 tubes, like Palisades, are the most susceptible alloy. However, regardless of alloy differences, the fact remains that the entire CE fleet reached end of SG life well before Palisades current estimated EFPY age. This makes the CE fleet record the most meaningful benchmark for ACRS consideration of Palisades tube condition.

Why the 2024 Palisades Inspection Results Were Predictable Viewed against this fleet record and the well-established behavior of Alloy 600 tubing, Palisades 2024 inspection resultsshowing widespread and exponential

increase in cracking and tube degradation at roughly 28 EFPYshould not have come as a surprise. Let me explain:

Established damage modes: Alloy 600 steam generator tubing is prone to both mechanical wear (caused by interaction with support structures such as egg-crates, straps, and tube-to-tube contact) and stress corrosion cracking (both primary water SCC and outside-diameter SCC). These modes are well documented across the CE fleet.

Exponential growth of cracks: Once initiated, cracks do not grow in a linear fashion; instead, they accelerate, often increasing dramatically after a relatively quiet incubation period. This behavior aligns with industry findingsfor example, A Review on the ODSCC of Steam Generator Tubes by H. Chung et al. (2013) describes ODSCC as a thermally activated process that accelerates once initiated. See Summary of Literature Search.

Role of deposits: At Palisades, persistent copper and lead deposits in sludge piles, tube scale, and TTS collars have created a corrosive environment that promotes ODSCC. These deposits act as stress concentrators and chemical initiators, ensuring that once cracking began, its progression would be rapid.

Summary of Literature Search: End-of-Life and Rapid Exponential Crack Growth Increasing Stress Intensity Factor The growth of a crack is primarily driven by the stress intensity factor (K) at its tip.

Because K depends directly on crack size, as a crack becomes longer or deeper the stress at its tip increases. This accelerates the growth rate and creates a positive feedback loop: a larger crack produces a higher stress intensity, which in turn drives the crack to grow even faster.

Crack Coalescence During the incubation period, numerous small, isolated microcracks can form in high-stress regions, such as tube support plate crevices. As these microcracks extend, they begin to link and merge into larger, more significant cracks. This coalescence dramatically increases the effective crack size, producing a sudden surge in growth rate and a rapid increase in the number of detectable defects during inspections.

Worsening Chemical Environment The aggressive chemical environment within crevicesnecessary for ODSCC initiationintensifies over time. Localized boiling, driven by heat transfer from the primary coolant, concentrates impurities inside the narrow crevice region, sometimes by factors as high as 10. Continued tube operation allows sludge and deposits to accumulate, restricting water flow and further concentrating chlorides, sulfates, and caustics. The result is a progressively more aggressive microenvironment that accelerates crack initiation and growth.

Sources Chung, H., Hong-Deok Kim, Seungjin Oh, et al. A Review on the ODSCC of Steam Generator Tubes in Korean NPPs. Nuclear Engineering and Technology, Vol. 45, No. 4, August 2013.

Cizelj, L. Trends of Degradation in Steam Generator Tubes, 1998.

U.S. NRC, NUREG/CR-5752, ANL-99/4, Assessment of Current Inspection and Monitoring of Steam Generator Tubes.

Palisades Copper and Lead Deposits: The Unresolved Question of Oxidation Beyond these fleet-wide degradation patterns, Palisades faces an additional, unresolved risk: the extent and persistence of copper and lead accumulations within its steam generators. These deposits are concentrated in precisely the locations most vulnerable to ODSCCdents, freespan regions, egg-crate supports, and the tubesheet transition. Historically these deposits were present in metallic formhence hard to remove.

Basic chemistry shows why this distinction matters. Metallic copper and lead are relatively stable, but when exposed to oxygenated water or uncontrolled secondary-side chemistry, they can oxidize to form compounds such as copper oxides or lead oxides. In oxidized form, these compounds are well known to promote electrochemical reactions that degrade Alloy 600 tubing and accelerate ODSCC initiation and growth.

During Palisades decommissioning period, for nearly two years, secondary-side chemistry was not controlled or monitored. This raises a critical unresolved issue:

whether some portion of the long-standing copper and lead metallic deposits in Palisades steam generators may have shifted from metallic to oxidized form under

these conditions. If so, the potential for localized aggressive environments around the tube surfaces will be a factor for future power operations, when stresses are again introduced.

This is an important consideration that requires further explanation. Specifically, NRC staff should clarify to the ACRS:

Whether copper and lead deposits at Palisades have oxidized following the extended period of uncontrolled secondary chemistry; What impact any such oxidation would have on ODSCC progression during the approved operating interval; and How these risks will be addressed in the upcoming operational assessments before the next full inspection of the steam generator tubes.

In short, the chemistry of these deposits is no longer an academic question. It is central to whether Palisades can reasonably be assumed to maintain tube integrity until the next mandated inspection.

Conclusion

Taken together, these factors meant that Palisadesoperating beyond the demonstrated steam generator service lives of all other CE plantswas virtually certain to experience extensive tube degradation by the time of its 2024 steam generator tube inspection. Perhaps, in 2020 Entergy understood this very well, and it was a factor in their decision to discontinue future operations?

The October 2024 inspection results validated long-standing industry knowledge about Alloy 600 tubing rather than representing an unforeseen development. They confirm that the aging process for these steam generators follows the well-documented trajectory of slow initiation followed by rapid, exponential crack growth.

When viewed through the broader lens of CE plant steam generator experience clearly showing the industry end of CE steam generator life, and the extensive research on rapid, exponential crack growth after prolonged years of operation, the ACRS must press the NRC to explain how continued operation in this high-risk region of accelerated crack growth can be considered acceptable in the operational assessment when defining the permissible operating period before the next shutdown and inspection.

Moreover, the current 2024 Palisades data available today indicates that these accelerated growth rates are continuing unabated, as the research indicated. The literature shows that little can be done to halt degradation once it has been set in motion years earlier, given the inherent metallurgy, tube geometry, and crevice chemistry conditions driving the process. This reality underscores the urgency of adopting truly conservative decision-making goin forward.

Suggestions for ACRS Questions at the November 10 Meeting Given the CE fleet experience, the unique design differences of Palisades, and the predictable progression of ODSCC, I respectfully suggest that the Committee ask NRC staff at the November 10 meeting to address the following questions:

Benchmarking Basis: Why is Beaver Valleya Westinghouse three-loop plant with fundamentally different steam generator design features considered a meaningful benchmark for Palisades, when the CE fleet record provides a more relevant basis for comparison?

Fleet Experience Envelope: Does NRC staff agree that Palisades, at ~28 EFPY, is already operating beyond the demonstrated service lives of all

other CE steam generators, and if so, what technical justification supports allowing further operation?

Deposit-related Degradation: How is NRC staff accounting for the persistent copper/lead scale, tube scale, and top-of-tubesheet collars unique to CE SGsfeatures that are absent in Westinghouse designs but known to accelerate ODSCC at Palisades?

Predictability of ODSCC: Given the well-established, non-linear progression of ODSCC once initiated, does NRC staff agree that Palisades 2024 inspection results were predictable, and if so, how is this being factored into the operational assessment and inspection interval planning?

Alan Blind

Five-Minute Verbal Comment - ACRS Full Committee, October 10, 2025 Alan Blind Good afternoon, Members of the Committee.

I appreciate the opportunity to speak. I will summarize my written comments today. Let me emphasize up front: the October 2024 Palisades inspection results confirm that accelerated crack growth is continuing unabated. There is nothing Holtec can do now to reverse this. The literature shows this is exactly what happens once the degradation process has been set in motion years earlier. Given the inherent metallurgy, tube geometry, and crevice chemistry of Alloy 600 steam generator tubing, very little can be done to halt it once underway.

By way of background, I served as the Engineering Director at Palisades.

In that role, I managed the benchmarking of Palisades steam generator remaining life against the full population of Combustion Engineering plants. That benchmarking effort, completed in 2010, showed clearly that Palisades was at the front edge of the CE fleet experience curve, where the rate of outside-diameter stress corrosion cracking accelerates dramatically what we described as a third-derivative growth curve.

We also accounted for plant-specific history. Decisions made in the 1990s introduced persistent lead and copper deposits into Palisades steam generators. These scale collars and sludge piles are extremely difficult to remove and created localized conditions that exacerbated ODSCC.

Combined with the known morphology of Alloy 600 cracking and the CE fleet benchmarking, we projected that Palisades would reach the 15%

tube-plugging limit around 2025. The Holtec October 2024 inspection results confirmed that projection almost exactly.

Those results show rapid expansion of both the rate and severity of ODSCC cracks. Several tubes approached or exceeded 90 percent through-wall depth, with one tube R73C94 within months of rupture under operating conditions. This is not random chance, and not simply the result of the recent period without secondary water chemistry controls. It is the textbook exponential growth phase of Alloy 600 cracking a progression that began decades ago, and one we already understood and anticipated in 2009.

I applaud the Committee for shining a light on the Operational Assessment, and I recognize that in your September 22 letter to the Commission you highlighted the uncertainties introduced by the extended layup without chemistry control. That is indeed a concern, and perhaps

one that chemical cleaning may help address. But the broader and more fundamental issue is this: the accelerated crack growth rates documented in 2024 show that Palisades has entered the high-risk, end-of-life phase where failures can occur suddenly and no credible mitigation strategy exists. The lack of proper wet layup during decommissioning did not cause this the process was already well underway.

Fleet experience we reviewed in 2009 confirms this. Every other CE plant replaced its steam generators well before reaching this point. None attempted to continue operation in the exponential growth region of Alloy 600 degradation. Yet here we are in 2025, and Palisades is the outlier.

So the central question before you, as advisors to the Commission, is this:

how can continued operation in this region of accelerated crack growth be considered acceptable in an Operational Assessment that defines the permissible interval before the next inspection? The slope of the exponential curve is approaching infinity.

In my view, conservative decision-making is the only defensible path forward. The data, the published research, the CE fleet benchmarking, and now the Palisades 2024 inspection all point to the same conclusion: once initiated, this degradation cannot be arrested. Palisades has crossed into a

region of unacceptable risk, and that fact must shape both NRC staffs review and the Committees recommendations.

Thank you.