ML24247A112

Summary of Meeting Between Nrc/Region I and C-10 - Questions Regarding Quarterly Inspection Reports for CY2023

ML24247A112
Person / Time
Site:	Seabrook
Issue date:	08/26/2023
From:	Sarah Elkhiamy NRC/RGN-I/DORS
To:	C-10 Research & Education Foundation
References
Download: ML24247A112 (1)

Text

ML24247A112 Documentation of Meeting between C-10 and NRC Region I regarding questions on Quarterly Reports for Calendar Year 2023 Members of the NRC Region I staff met virtually with members of C-10 to respond to questions submitted by the group for the 2023 quarterly inspection reports. The discussions were held on the following dates: October 24, 2023 (1st quarter inspection report), March 19, 2024 (2nd and 3rd quarter inspection reports), and August 7, 2024 (4th quarter inspection report). The members of C-10 were satisfied with the answers and discussion with the resident inspectors and technical staff from the region. This documents all questions raised by C-10 in their letters as satisfactorily answered.

Attached C-10 letters inquiring about 1st thru 4th Quarter 2023 Inspection Reports

July 26, 2023 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 David Lew, Regional Administrator, Region 1 Director, Office of Enforcement 475 Allendale Road Suite 102 King of Prussia, PA 19406-1415 Travis Daun, Senior Resident Inspector Seamus Flanagan, Resident Inspector Seabrook Station 626 Lafayette Road Seabrook, NH 03874 Mel Gray, Branch Chief - Engineering Branch 1 Nicole Warnek, Senior Project Engineer Projects, Branch 2 Joseph DeBoer, Senior Project Engineer Projects, Branch 2 Matt Young, Chief, Reactor Projects, Branch 2 Nik Floyd, Senior Reactor Inspector, Region 1 Sarah Elkhiamy, Senior Project Manager, Region 1

SUBJECT:

Response to NRC 1st Quarter Integrated Inspection Report, NRC Cover letter dated May 11, 2023 (ML23129A193)

Greetings:

This letter is written in response to the above Integrated Inspection Report. We look forward to receiving your responses to the following questions, comments and concerns we have highlighted in this letter.

The NRC introductory cover letter states, No findings or violations of more than minor significance were identified during this inspection. Having reviewed the details reported in this IIR, we cannot agree that all the findings are of only minor significance.

Would you please explain the meaning of the term finding? (See first beginning paragraph, page 8) It seems a finding is different from a violation. We did consult the NRC Inspection Manual Chapter 0609 and Chapter 0612. These address findings.

Is it the case that under the Significant Determination Process, the definition of findings is different from how it is used in this 2023 1st Quarter IIR report?

Following are our specific areas of concern.

Page 5: (71111.15 (4): Electrical vaults W09 and W10 and circulating water/service water pumphouse structures are identified as exceeding their building deformation threshold limits.

Since the inception of the 2016 License Amendment Process, NextEra has repeatedly stated that Alkali Silica Reaction degradation was not a problem because it would be kept within the box (ASLB 2018 hearing transcript p. 521) of stated margins of structural integrity. (see pages 415-416 ASLB hearing transcript) September 2019, we have decades of margin before we would hit that threshold. Only five years of the now-27 more years of contemplated operation of Seabrook Station have elapsed since 2019, and we learn in this report that 10 structures are already exceeding their building deformation threshold limits. How is this not more than minor?

According to this IIR, before this report period there were already seven Seabrook structures identified as needing physical modification or additional analysis to comply with their current license/design basis. As of this report, the NRC placed three additional buildings into the POD/Prompt Operability Determination list (AR 0227 6197). The report states they recently exceeded established thresholds in the two electrical vaults and the circulating water/service water pumphouse. As of now, there are ten Seabrook structures that are acknowledged as exceeding their current license/design basis. How is this not of more than minor significance??

Further, while it is acknowledged in the inspection report that certain of the ten structures are in need of physical modifications to bring them back into conformity with their license/design basis, it is also stated that several may just need to be reanalyzed. This seems to mean that the licensee can change the parameters of what constitutes conformity to the plants license/design basis (see page 7, Evaluation of Three Structures in the Operability Determination, NextEra revised the operability determination Also see page 8, first paragraph, future expansion is expected to remain within the new established operability determination limit.) Our understanding is that the NRC rewrote the Operability Determinations, creating a new document dated February 1, 2023. Is it correct that NRC/NextEra can, at will, change the requirements and by doing so can add margin? Also, the

new version of Operability Determination, dated 2/1/2023, is not appearing on ADAMS. Can you direct us to where we can access this document? A more general question: how does revising the operability determination demonstrate capability to perform intended function?

Page 8: Evaluation of Three Structures in the Operability Determination. The operability determination evaluations also considered impacts to rebar stress and concluded the rebar stress was not approaching the yield limit. Would you advise if this determination was made based on calculations alone, or on field measurements (if so how where those conducted?)

Page 8: Evaluation of Three Structures in the Operability Determination. Would you please confirm how the new limits are calculated by NextEra, reviewed by the NRC, and what the basis is for accepting or rejecting? This is in reference to this excerpt: future expansion is expected to remain within the new established operability determination limits past the next scheduled monitoring inspection with margin. Please explain how you are determining future expansion. Are you merely extrapolating from past data? Is the extrapolation linear? Also, are all new ASR areas subject to be held to these new limits, or just newly found areas such as these three?

Page 8: Status of the Containment Internals Structure. Distress in the containment reactor pit floor, which is now in process of a root cause analysis procedure, to be completed in the summer of 2023: why is there apparent ongoing difficulty in determining and stating the cause of this damage? It seems, over the period of a few inspections, that there is a difference of opinion as to the cause of this damage. It seems clear, even with the contradictory causal factors cited, that at least some of the damage is due to ASR. If that is the case, does this not put this area into Tier 3, which then requires inspection every six months? How can this be accomplished if the area can be inspected only during refueling shutdowns? Through which analytical method or field observation has it been determined that the root cause is (low cycle) fatigue loading caused by thermal stress fluctuation as opposed to ASR?

Page 8: Timing of Corrective Actions This paragraph explains that there are ten structures now in NextEras Operability Determination Process that do not meet the NRC-approved methodology documents requirements. Five of these structures have been in ODP since 2019 and several due dates were extended multiple times. All of these five require physical modification. The paragraph states that the operative procedure (PI-AA-104-1000) calls for cited repairs in the Corrective Action Program to be completed within one to two refueling cycles, which are 18 months in length as you are aware. One intrinsic problem is that there are, apparently, no specific, fixed time deadlines in NRC regulations for NextEra to complete the identified repairs/solutions to listed problems if they are not written as a violation at a higher level than green. The report states that all the ten structures have been demonstrated to be functional with a supporting technical basis, but that such technical approaches were not intended to replace the approved methodology. NextEra has indicated that it plans to complete the work by 2026, years beyond the NRCs stated required timeframe. Does this mean that NextEra can and has changed the analysis numbers to add margin? Is that what supporting

technical basis means? Clearly, this completion timeframe is not within the normal ODP. Again, how is this not more than minor?

Page 9: ASR Monitoring and Expansion Trends. The inspectors noted that several of the extensometer locations were projected to exceed the licensed through-thickness expansion limit before the expiration of renewed Seabrook operating license in 2050, but based on current rates, the projected time to exceedance is greater than 10 years. Is this a result of the NRC inspector doing their own calculation to forecast out 10+ years, or NextEra? Our understanding in general, and mentioned earlier in this paragraph, is that projections need only be made to the next inspection interval, so who is responsible for trending out to the 2050 license end date?

A related question: Is NextEra performing elastic modulus tests on the cores extracted during extensometer installation procedure? As found in the transcript of 2018 Atomic Committee on Reactor Safety (ACRS) meeting (ML18348B117), NextEra reviewed the large-scale testing program data and literature data for multiple properties and determined that reduction in modulus of elasticity was the best material parameter to determine ASR expansion to-date.

We understand that all of you are extremely busy attempting to oversee and regulate Seabrook Station to the best of your ability. That said, we deduce from this inspection report that the situation at the plant is not the best - ASR seemingly moving significantly faster than anticipated, the licensee not keeping up with the Corrective Actions in a timely manner, failures to properly report issues and ongoing problems (containment reactor pit) and another example, at least ten recurrent through-wall water leaks.

We at C-10 empathize with your work burden, while at the same time feeling strongly that the above questions and the further information that your answers will provide are vital in keeping the public informed on the situation at Seabrook Station. Thank you for your anticipated responses!

Sincerely, Patricia Lang Skibbee C-10 Board of Directors President/Research Team 978 502 4782

October 30, 2023 via electronic mail To:

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 David Lew, Regional Administrator, Region 1 Director, Office of Enforcement 475 Allendale Road Suite 102 King of Prussia, PA 19406-1415 Travis Daun, Senior Resident Inspector Seamus Flanagan, Resident Inspector Seabrook Station 626 Lafayette Road Seabrook, NH 03874 cc:

Mel Gray, Branch Chief - Engineering, Branch 1 Nicole Warnek, Senior Project Engineer Project, Branch 2 Joseph DeBoer, Senior Project Engineer Projects, Branch 2 Matt Young, Chief, Reactor Projects, Branch 2, Division of Operating Reactor Safety Niklas Floyd, Senior Reactor Inspector, Region 1 Sarah Elkhiamy, Senior Project Manager, Region 1

SUBJECT:

C-10 Research & Education Foundation communication regarding ML23220A209 Seabrook Station - Integrated Inspection Report 05000443/2023002

Dear Seabrook Station NRC Inspectors:

C-10 has reviewed the 2nd quarter Integrated Inspection Report and there are three areas of the document that prompted us to reach out to you for further information.

Topic 1: The finding Failure to Establish Continuous Fire Watch on Fire Door documented on page 10.

Context 1.1: This was classified as an NCV and described as a cross-cutting issue in the Teamwork category, since it demonstrated a lack of communication/coordination among NextEra staff work teams.

In this case, fire door C107 was left open for extension cords to go through the door opening to supply electricity for work being done on the other side of the door. Per the IIR No fire watch was seen at the door or in the essential switchgear room or the stairwell the door goes to.

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The NRC inspectors notified the Control Room. The workgroup was briefed by the control room on the requirement to station a continuous fire watch while the door was open for the cords but when the work group left, they removed all the cords but missed one that was still in the door. This was left unnoticed by the work group until the inspectors discovered it. The failure to establish a continuous fire watch violated section 11 of the Technical Requirements Manual, which requires such fire watch be established within one hour of the breached fire door.

C-10 believes this demonstrates a lack not only of communication, but of disregard of basic safety procedure.

Question 1.1:

Can the NRC share with C-10 the explanation that NextEra provided as to how this mistake occurred, and how they plan to improve on it moving forward considering the briefing from the control room immediately prior to this mistake occurring, did not appear to sufficiently prevent it?

Context 1.2:

C-10 keeps a record of all fire safety violations at Seabrook Station. Alarmingly during the twelve-month period from July 1, 2022 through June 30, 2023, the NRC cited NextEra for three separate fire safety violations (ML22304A180, ML23030A482, ML23220A209). The fact that these occurred in a relatively short period of time is indicative of an ongoing pattern. As we understand it, these previous violations would have entered the Corrective Action Program, and required a Root Cause Analysis and Corrective Action Plan be completed by NextEra.

Question 1.2 What is the long-term solution to what appears to be a disturbingly regular occurrence of repeated fire safety violations at Seabrook Station? Will the NRC be applying a higher level of scrutiny to NextEras Root Cause Analyses and Corrective Action Plans on this newest NCV, given that the previous Corrective Action Plans have clearly not produced an improved result or prevention of future violation?

Topic 2: The failure of NextEra to follow regulations (in three separate instances) regarding their failure to report issues to the NRC within eight hours (or whichever amount of time is applicable to the incident) of occurrence.

Context 2.1:

On three occasions within the inspection period - April 12, 2023 and two separate times on May 6, 2023 - valid actuation signals were triggered within various plant systems.

NextEra states that it didnt think these were reportable incidents because the plant was at zero power at those times. However 10 CFR 50.72 requires that such incidents be reported within eight hours of occurrence even when the plant is at zero power. Similar to the fire-safety issues outlined above, this is also indicative of a pattern.

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Question 2.1:

Will NextEra be investigating whether there have been additional mistakes made by Seabrook Station Employees related to mandated reporting to the NRC that are not yet known to the NRC and therefore also not by the public? This seems like it could be a logical part of the Root Cause Analysis, to determine if it is a systemic issue or more of an issue with certain individual personnel repeating the mistake.

Topic 3: Item 2 in the section 71111.15 - Operability Determinations and Functionality Assessments.

Context 3.1:

B emergency diesel following incorrect fluid addition to crankcase oil sump (AR 02458425) on April 23, 2023. This error is not explained with any detail or cited as a violation in the Inspection Results section, but it seems to us to be of great significance given that it could impact the operability of the emergency diesel generator.

Question 3.1.1:

How was this error discovered?

Question 3.1.2:

Was there an operational issue with the generator resulting from the adding of the incorrect fluid?

Question 3.1.3:

What action can/will the NRC accept from NextEra to prevent future errors of this type?

Question 3.1.4:

We understand that there are some areas of work within a nuclear plant which require a partner system or require that a certain-level supervisor be present to observe, so that one person is not left alone in certain work situations. It makes sense that requiring the presence of two or more workers allows the opportunity for them to aid/monitor each other. Is this fluid addition scenario one of those work situations in which two or more workers must be present? If yes, were the required personnel present during this work?

Thank you for your consideration on these items and we look forward to your responses.

Kindest Regards, Sarah Abramson Executive Director C-10 Research & Education Foundation 11 Chestnut St., Amesbury, MA 01913 Ph: (978) 465-6646 Email: sarah@c-10.org C-10 Research & Education Foundation / 11 Chestnut St., Amesbury MA 01913 / www.c-10.org / (978) 465-6646 Page 3 of 3

March 12, 2024 via electronic mail To:

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 David Lew, Regional Administrator, Region 1 Director, Office of Enforcement 475 Allendale Road Suite 102 King of Prussia, PA 19406-1415 Travis Daun, Senior Resident Inspector Eben Allen, Resident Inspector Seabrook Station 626 Lafayette Road Seabrook, NH 03874 cc:

Mel Gray, Branch Chief - Engineering, Branch 1 Nicole Warnek, Senior Project Engineer Project, Branch 2 Joseph DeBoer, Senior Emergency Preparedness Inspector Matt Young, Chief, Reactor Projects, Branch 2, Division of Operating Reactor Safety Niklas Floyd, Senior Reactor Inspector, Region 1 Sarah Elkhiamy, Senior Project Manager, Region 1

SUBJECT:

C-10 Research & Education Foundation communication regarding ML23307A018 Seabrook Station - Integrated Inspection Report 05000443/2023003

Dear Seabrook Station NRC Inspectors:

C-10 has reviewed the 3rd quarter Integrated Inspection Report and there are three areas of the document that prompted us to reach out to you for further information.

Item 1 - (Page 6): Inspection Review (IP Section 02.01-02.03) (1 Sample) 71114.04 (Page 6): NextEra submitted seven proposed changes to the Emergency Action Level and Emergency Plan. It is stated in the report that this evaluation does not constitute NRC approval.

Questions:

1.1: Do any of these proposed changes overlap with changes proposed in the License Amendment Request for Common Emergency Plan from October 2022 (ML22278A031)

(with subsequent documents submitted by FLP providing additional information:

C-10 Research & Education Foundation / 11 Chestnut St., Amesbury MA 01913 / www.c-10.org / (978) 465-6646 1

ML22343A254 in December 2022, ML23219A102 in August 2023, and ML23332A005 in November 2023)?

1.2: Our understanding is that these changes would have been put into effect immediately by NextEra, and that the NRC will review them during the Annual EP Safety Inspection, the results of which will be published in the inspection report that follows. Is that correct, or is the NRC review of these items now complete?

Item 2: 71152A (Page 9) Observation: Review of NextEras Evaluation and Corrective Actions of Safety-Related Structures Affected by Alkali-Silica Reaction (ASR): Status of the Consolidated Prompt Operability Determination (POD)

Questions 2.1: Which structures do not meet demand-to-capacity ratio?

2.2: When is combining structures that do not meet the demand-to-capacity ratio acceptable?

2.3: What was the nature of the physical modifications that were made that restored the margin to the Mechanical Penetration Area?

Item 3 (Page 10): 711152A Observation: Review of NextEras Evaluation and Corrective Actions of Safety-Related Structures Affected by Alkali-Silica Reaction (ASR); Evaluation of Two Structures in the POD Context 3.1: Corrective Actions indicates that ASR margins have been exceeded in the two structures that were inspected. In this case the inspectors determined the planned corrective actions were to physically modify (aka retrofit) the structures to bring structural elements into conformance with the licensing and design basis. The inspectors noted the planned retrofit approaches and concepts were primarily targeted to increase the structural capacity for the exceeded limit state and were not based on increasing or altering the structural stiffness. The retrofit designs also considered the need for continued access to monitor for the effects of ASR after retrofit.

Questions:

3.1.1: Why were only 2 structures inspected out of all the structures identified as having ASR?

3.1.2: Inspectors were only able to walk down the accessible portions of the POD structures; what about the inaccessible parts? How and when are they monitored?

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Context 3.2: The inspectors further determined the POD evaluations considered impacts to rebar stress and concluded the rebar stress was either not approaching the yield limit or established enhanced monitoring consistent with ASR-related license conditions to show that rebar yielding would be detected.

Questions:

3.2.1: What type of enhanced monitoring will show rebar yield?

3.2.2: When/if the enhanced monitoring shows rebar yield to the point that its out of margin, what is the course of action?

Context 3.3 (Page 11): Structural damping for the safe shutdown earthquake unusual load combination increased from 7 percent in the licensing basis to 10 percent in the POD. The inspectors noted a damping level of 10 percent is allowed by the standard in American Society of Civil Engineers 43, Seismic Design Criteria for Structures Systems and Components in Nuclear Facilities, based on the corresponding level of cracking and high stress levels in the structure under the safe shutdown earthquake load.

The engineering standards referenced in the IIR from The American Society of Civil Engineers 43, Seismic Design Criteria for Structures, Systems and Components in Nuclear Facilities state The goal of this standard is to ensure that nuclear facilities can withstand the effects of earthquake ground-shaking with desired performance.

Our review of the standard cited in the report concluded that there is no mention of alkali silica reaction or its impact on structures in nuclear facilities. The 10 percent damping level that is allowed by the standard is specific to concrete performing within design standards and not impacted by ASR. If standards are to be used, more current standards which are peer-reviewed and offered by expert professional societies, American Concrete Institute (ACI) Committee Report 349.3R, Evaluation of Existing Nuclear Safety-Related Concrete Structures, and American Society for Testing and Materials (ASTM) Standard C856-11, Standard Practice for Petrographic Examination of Hardened Concrete are highly recommended.

C-10 is concerned that the true nature of ASR on structural integrity is not accounted for. It appears that NextEra is only comparing the current result to the latest monitoring record inspection. So, the 1.1. ratio or crack growth are not absolute values that go back to the mandated start of monitoring (not the date monitoring was implemented). Questions:

Questions:

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3.3.1: It appears that the criterion is continually revised during every inspection based only on the previous inspection. Is this correct?

3.3.2: How are NextEra and the NRC documenting and tracking the absolute growth of the cracks to monitor total cracking and expansion progression?

What we learned in this report with respect to ASR is that buildings are now being retrofitted to meet design basis, inspection intervals are being increased, calculation bases are being shifted and are not executed in a timely manner according to the agreed upon schedule. What does the trending analysis data indicate for the current timeline of when operability limits will be reached?

At the in-person June 2023 Annual Public Safety Meeting, NRC inspectors stated that some structures in the POD are trending to exceed the allowable limits in more than 10 years but before 2050. What is the current status of NextEras plan to address this looming timeframe?

In a meeting with C-10 on October 24, 2023 the NRC inspectors explained that re-analyzing is a stop gap before structure modifications, but that it cant go on forever. If retrofitting will not be complete until 2026 (approximately 2 years from now), what is guaranteeing the safety of these structures and the citizens residing in the EPZ and beyond for the next 2 years?

Thank you for your consideration on these items and we look forward to your responses.

Kindest Regards, Sarah Abramson Executive Director C-10 Research & Education Foundation 11 Chestnut St., Amesbury, MA 01913 Ph: (978) 465-6646 Email: sarah@c-10.org C-10 Research & Education Foundation / 11 Chestnut St., Amesbury MA 01913 / www.c-10.org / (978) 465-6646 Page 4 of 4

May 21, 2024 via electronic mail To:

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Attn: Travis Daun, Senior Resident Inspector Eben Allen, Resident Inspector Seabrook Station 626 Lafayette Road Seabrook, NH 03874 Cc:

Matt R. Young, Chief Projects Branch 2 Division of Operating Reactor Safety David Lew, Regional Administrator, Region I Mel Gray, Chief, Engineering Branch 1 Nicole Warnek, Senior Project Engineer Project, Branch 2 Sarah Elkhiamy, Senior Project Manager, Region 1

SUBJECT:

C-10 Research & Education Foundation communication regarding Seabrook Station - Integrated Inspection Report 05000443/2023004 (ML24043A002)

Dear Seabrook Station NRC Inspectors:

C-10 has reviewed the 2023 4th quarter Integrated Inspection Report and there are four areas of the document that prompted us to reach out to you for further information.

Item 1: Motor oil issue timeline We would like to request some clarity regarding the motor oil issue and associated timeline for resolution.

A. Can you please advise if these dates are correct, and if not what the correct dates are?

On page 6 the report states that an Operability Determination was performed following replacement of motor oil on RHR Pump A on November 14, 2023.

However, further down on the same page the report states that a post-maintenance test was conducted November 1, 2023, following replacement of motor oil on RHR Pump A.

On page 9 the report states that the motor oil on RHR Pump A was replaced on October 10, 2023 and that a sample of the oil was sent off for analysis. However the report then states that the oil analysis was received in September 2023.

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While it was discouraging to learn that the apparent cause for the use of the incorrect type of oil (decanting oil into mis-labeled container(s)) was identified as a trend, it was encouraging to know that a better system is now in place.

B. How is the NRC monitoring that this better system is being followed by plant staff?

Item 2: Review of Seabrooks Evaluation and Corrective Actions for Radiation Monitoring System Failures (71152A)

On page 10 the report notes that there have been 9 failures between 2021 - 2023 of one or more of the containment gaseous radioactivity monitors. It further states that the station has been investigating a long term action to replace their radiation monitoring system.

A. Given the number of failures, has the NRC provided NextEra with guidance on a timeline to implement a new radiation monitoring system?

B. Does the NRC require a certain level of reliability for such radiation monitoring systems, such as a limit to how many failures this system can endure in a certain time period before action is required by the licensee?

C. For clarity, is the intent of NextEra to replace all radiation monitors on site at the plant, or just those containment gaseous radioactivity monitors? We are very interested in understanding the reliability and current status of the real-time radiation monitoring of the air that is done on-site outdoors at Seabrook Station.

Item 3: Remedial Training and Re-examinations On page 4 it was noted that the inspectors evaluated the effectiveness of remedial training conducted by the licensee, and reviewed the adequacy of re-examinations for licensed operators who did not pass a required requalification examination.

A. How common is it for licensed operators to fail the examination and require requalification at Seabrook Station, and how does Seabrook Station compare to other licensees in this area?

Item 4: Observation: Semiannual Trend Review On page 9 it was noted that the inspectors evaluated the operator challenges program including an audit of control room deficiencies, control board notifications, and operator workarounds.

A. What type of deficiencies and workarounds are acceptable in the control room?

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B. Further down the report notes that the station has made significant progress toward correcting long standing video alarms in the control room. What was the issue with the video alarms and how much more work is necessary to resolve the issue and consider it closed?

Thank you for your consideration on these items and we look forward to your responses.

Kindest Regards, Sarah Abramson Executive Director C-10 Research & Education Foundation 11 Chestnut St., Amesbury, MA 01913 Ph: (978) 465-6646 Email: sarah@c-10.org C-10 Research & Education Foundation / 11 Chestnut St., Amesbury MA 01913 / www.c-10.org / (978) 465-6646 Page 3 of 3