IR 05000373/2023012
| ML24017A120 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/18/2024 |
| From: | Robert Ruiz NRC/RGN-III/DORS/RPB1 |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| References | |
| IR 2023012 | |
| Download: ML24017A120 (1) | |
Text
SUBJECT:
LASALLE COUNTY STATION - BIENNIAL PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000373/2023012 AND 05000374/2023012
Dear David P. Rhoades:
On December 15, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution inspection at your LaSalle County Station and discussed the results of this inspection with John Van Fleet, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
The NRC inspection team reviewed the stations problem identification and resolution program to confirm that the station was complying with NRC regulations and licensee standards. The team identified a finding in problem identification, implementation of the process for prioritizing and evaluating these problems, and the effectiveness of corrective actions taken to resolve these problems. Specifically, the team identified a finding with an associated non-cited violation for the failure to follow the requirements of the ASME OM code following a documented test failure of the Unit 1 residual heat removal discharge C heat exchanger relief valve. The details of this issue are discussed in the report.
The team also evaluated the stations effectiveness in identifying, prioritizing, evaluating, and correcting problems, reviewed licensee audits and self-assessments, and its use of industry and NRC operating experience information. The results of these evaluations are in the enclosure.
Finally, the team reviewed the stations programs to establish and maintain a safety-conscious work environment and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews, the team found no evidence of challenges to your organizations safety-conscious work environment. Your employees appeared willing to raise nuclear safety concerns through at least one of the several means available.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
January 18, 2024 If you contest the violation or the significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at LaSalle County Station.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; and the NRC Resident Inspector at LaSalle County Station.
This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Robert Ruiz, Chief Reactor Projects Branch 1 Division of Operating Reactor Safety Docket Nos. 05000373 and 05000374 License Nos. NPF-11 and NPF-18
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000373 and 05000374
License Numbers:
Report Numbers:
05000373/2023012 and 05000374/2023012
Enterprise Identifier:
I-2023-012-0011
Licensee:
Constellation Nuclear
Facility:
LaSalle County Station
Location:
Marseilles, IL
Inspection Dates:
November 27, 2023 to December 15, 2023
Inspectors:
E. Magnuson, Reactor Inspector
J. Meszaros, Resident Inspector
N. Shah, Senior Project Engineer
M. Siddiqui, Reactor Inspector
Approved By:
Robert Ruiz, Chief
Reactor Projects Branch 1
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a biennial problem identification and resolution inspection at LaSalle County Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Comply with ASME Code Requirements Following Test Failure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000373,05000374/2023012-01 Open/Closed
[P.2] -
Evaluation 71152B The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR)
Part 50.55a (f)(4)(ii) for the licensees failure to perform inservice tests to verify the operational readiness of valves whose function is required for safety. Specifically, the licensee failed to perform inservice testing on two additional relief valves from valve group R105, following the set-pressure testing failure of 1E12-F025C, and failed to perform the cause-and-effect evaluation of the testing failure.
Additional Tracking Items
None.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
OTHER ACTIVITIES - BASELINE
71152B - Problem Identification and Resolution Biennial Team Inspection (IP Section 03.04)
- (1) The inspectors performed a biennial assessment of the effectiveness of the licensees Problem Identification and Resolution program, use of operating experience, self-assessments and audits, and safety-conscious work environment.
- Problem Identification and Resolution Effectiveness: The inspectors assessed the effectiveness of the licensees Problem Identification and Resolution program in identifying, prioritizing, evaluating, and correcting problems. The inspectors also conducted a 5-year review of the core standby cooling systems. The inspection also included a selective review of past NRC observations, violations (including non-cited) and findings documented in the licensees Corrective Action Program.
- Operating Experience: The inspectors assessed the effectiveness of the licensees processes for use of operating experience.
- Self-Assessments and Audits: The inspectors assessed the effectiveness of the licensees identification and correction of problems identified through audits and self-assessments.
- Safety-Conscious Work Environment: The inspectors assessed the effectiveness of the stations programs to establish and maintain a safety-conscious work environment.
INSPECTION RESULTS
Assessment 71152B Assessment of the Corrective Action Program Effectiveness of Problem Identification Overall, the station was effective at identifying issues at a low threshold and was properly entering them into the corrective action program (CAP) as required by station procedures.
During interviews, workers were familiar with how to enter issues into the CAP and stated that they were encouraged to use it to document issues. During plant walkdowns, the team observed that issues were being identified in the field and that they were being properly addressed in the CAP. The team determined that the station was generally effective at identifying negative trends that could potentially impact nuclear safety. For the areas reviewed, the team did not identify any issues in problem identification.
Effectiveness of Prioritization and Evaluation of Issues In-depth reviews of a risk-informed sampling of action requests (ARs), work orders (WOs),and root and apparent cause and condition evaluations were completed. The team determined that the licensee had established a low threshold for entering deficiencies into the CAP, that the issues were generally being appropriately prioritized and evaluated for resolution, and that corrective actions (CAs) were implemented to mitigate the future risk of issues occurring that could affect overall system operability and/or reliability.
The inspectors noted that issues were properly screened, with most either classified as Conditions Adverse to Quality (CAQ) or Non-Corrective Action Program (NCAP) items.
Through a selective review of CAP and NCAP items, the inspectors found no issues either with the assigned level of evaluation or the proposed corrective actions. Issues having potential operability concerns were properly addressed through the screening process and during control room observations, and accompaniment of non-licensed operators during daily rounds; the inspectors did not identify any significant operator workarounds or similar deficiencies.
While most CAP products were acceptable, the inspectors identified some minor examples where the process requirements were not met. These included, but were not limited to:
- A Corrective Action Program Evaluation (CAPE) for AR 4011747 did not have a required effectiveness review as an assigned action.
- AR 4468596 had an incorrect significance level assigned.
- The effectiveness review for root cause evaluation (RCE) associated with AR 4406976 was assigned within 3 months after implementation of the corrective actions to prevent recurrence, which was too early to assess their effectiveness.
The inspectors noted that similar examples were noted in recent licensee self-assessments and audits of the CAP and that actions were being taken by the licensee to reinforce the CAP standards and expectations. The licensee also captured the above inspector-identified examples in the CAP.
The inspectors did a selective review of issues identified by the NRC either documented as observations, or for which findings or other enforcement was issued. These issues were properly documented and screened in the CAP and corrective actions were appropriate and timely scheduled.
Issue evaluations were generally sound and of good quality. Most issues were screened as low significance and were assigned a work group evaluation (the lowest level of review);more significant issues were assigned a CAPE, or if highly significant, a root cause evaluation. Through a selective review, the inspectors verified that the assigned evaluations were consistent with the significance of the issue as defined in the licensees process.
Most evaluations were generally thorough and consistent with the expectations in licensee procedures; however, the inspectors did identify some CAPEs and RCE evaluations where the quality and depth of the review was less than thorough. Examples included:
- The CAPE for AR 4479498, 1A RR Pump Failed to Downshift and Tripped to 0 Speed, concluded that the event was caused by loose terminal studs on the high speed to low speed auto transfer relay (i.e., GE HFA relays). Specifically, during maintenance on these relays, there was no action to verify the stud tightness prior to returning to service, resulting in some of the connections either being loose or becoming so over time. As a corrective action, the associated maintenance procedures/work packages for the GE HFA relays were revised to include steps to verify the tightness of the terminal stud connections. The inspectors noted that the extent of condition review for this CAPE did not consider other relays that were susceptible to the same condition. As a result, the licensee subsequently identified other relays that should also have been addressed by similar corrective actions.
- The CAPE for AR 4407880, Spurious Group 4 Signal Causes VG/VR Actuations, identified that the event was primarily due to a technician not using the appropriate controls (i.e., self-check, questioning attitude), when performing maintenance on the system, resulting in the spurious signal; however, the CAPE also discussed issues with the work planning and coordination as having played a significant role. These work planning issues were not identified as a contributing cause and therefore, no action was assigned to identify and address any organizational learnings.
The licensee captured each of these examples in the CAP and had assigned appropriate corrective actions to address the concern.
The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV), for the failure to follow ASME code requirements following the set-pressure testing failure of residual heat removal pump discharge relief valve 1E12-F025C. This issue is discussed in more detail later in the report.
Assessment 71152B Five Year Review of the Core Standby Cooling System (CSCS)
The inspectors performed an expanded 5-year review of the Units 1 and 2 CSCS; specifically, by performing system walkdowns; evaluating condition reports and work orders; and interviewing personnel responsible for working on the system. The inspectors also reviewed a sample of aging management actions associated with the Unit 1 high pressure core spray (HPCS) room cooler piping, including condition reports, piping replacement plans, and work orders. Overall, the inspectors determined that the licensee was effectively managing issues associated with this system.
Assessment 71152B Assessment of Operating Experience and Self-Assessments and Audits Assessment of Operating Experience and Self-Assessment and Audits Based on the samples reviewed, the team determined that licensee performance in the use of Operating Experience (OE) and Self-Assessments and Audits adequately supported nuclear safety.
Use of Operating Experience The licensee routinely screened industry and NRC OE information for station applicability.
Based on these initial screenings, the licensee-initiated actions in the CAP to fully evaluate the impact, if any, to the station. When applicable, actions were developed and implemented in a timely manner to prevent similar issues from occurring. During interviews, licensee staff stated that operating experience lessons-learned were communicated during work briefings and department meetings and incorporated into plant operations.
The inspectors did identify one example of OE that was reviewed by licensee staff and deemed as not applicable to the site but was lacking sufficient documentation to support that conclusion. Specifically, AR 467770467770 OPEX Review 52007 Vulnerability. The licensee documented this issue in the CAP and reopened the evaluation to add the supporting basis. The inspectors reviewed the additional information and had no further issues.
Although the use of operating experience was seen as valuable, the inspectors identified some instances where it was not properly evaluated as a contributing cause in CAPE or root cause evaluations. Specifically, both the CAPE and root cause guidance procedures stated that OE should be reviewed to determine if there were similar industry events that could provide insights on the issue, including corrective actions, and whether this OE was a missed opportunity for the licensee to have taken prior action to prevent the event. The inspectors identified several examples where the licensee had identified potentially applicable OE but had not fully documented associated learnings from the event, or whether it was a missed opportunity. The inspectors also noted that the licensee had made a similar observation in a prior self-assessment and had identified an adverse trend. This trend was documented in AR 4723092 and the inspectors examples were added to the AR. The inspectors reviewed the AR and noted that the corrective actions, while in-progress, appeared appropriate.
No findings or violations were identified.
Self-Assessments and Audits
The inspectors reviewed several audits and self-assessments and deemed those sampled as thorough and intrusive with regards to following up with the issues that were identified.
No findings or violations were identified.
Assessment 71152B Assessment of the Safety-Conscious Work Environment Assessment of Safety-Conscious Work Environment The team reviewed the stations programs to establish and maintain a safety-conscious work environment and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews, workers at the station expressed freedom to raise and enter safety concerns through any one of the various avenues available to them, and the team encountered no indications of a chilled work environment.
Workers expressed favorable opinions of the Employee Concerns Program (ECP) during interviews. While most workers felt no need to engage the ECP, the inspectors noted that there were still several issues documented in the program. Through a selective review, the inspectors concluded that these issues were appropriately handled and identified no adverse trends. The inspectors did note however, that some staff were unaware of where the ECP offices were onsite. This was despite several signs posted throughout the station describing the ECP office location. This was discussed with the ECP coordinator who planned to address this during routine plant department outreach meetings.
Based on the interview results and a review of the ECP issues, the most common issues at the station involved resources and a perception that lower-level CAP issues were not addressed timely. Licensee management was fully aware of these issues, as they had been identified prior in various station self-assessments and audits and in the previous NRC PI&R biennial inspection. The inspectors noted that the licensee had taken action to address these issues. The inspectors also noted that these issues had not had a significant impact on station performance nor had they discouraged licensee staff from raising concerns.
Overall, the inspectors found no evidence of challenges to the licensees safety-conscious work environment, as licensee employees were willing to raise nuclear safety concerns through at least one of several means available.
No violations or findings were identified.
Failure to Comply with ASME Code Requirements Following Test Failure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000373,05000374/2023012-01 Open/Closed
[P.2] -
Evaluation 71152B The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR)
Part 50.55a (f)(4)(ii) for the licensees failure to perform inservice tests to verify the operational readiness of valves whose function is required for safety. Specifically, the licensee failed to perform inservice testing on two additional relief valves from valve group R105, following the set-pressure testing failure of 1E12-F025C, and failed to perform the cause-and-effect evaluation of the testing failure.
Description:
The residual heat removal (RHR) pump discharge relief valves (E12-F025A/B/C) at LaSalle station were normally closed valves and had a safety function to open to protect the RHR pump discharge piping in the event of an overpressure condition. The relief valves were set to relieve pressure at 500 psig, as listed in the updated final safety analysis report (UFSAR)
Section 6.3.2.2.12. The valves were also required to close to provide containment isolation for the RHR system.
The licensees American Society of Mechanical Engineers (ASME) Code of Record for Operation and Maintenance of Nuclear Power Plants, (OM) was the 2004 Edition with Addenda though 2006. The licensee categorized the relief valves as Category C valves and tested the relief valves in accordance with Mandatory Appendix 1 of the OM Code.
Mandatory Appendix I Paragraph I-1350(c)(1), required, for each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either +/- tolerance limit of the Owner-established set-pressure acceptance criteria of I-1310(e) or +/- 3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group.
Additionally, Mandatory Appendix I Paragraph I-1350(c)(3) states, in part, the Owner shall evaluate the cause and effect of valves that fail to comply with the set-pressure acceptance criteria established in I-1350(c)(1).
On May 20, 2022, the 1E12-F025C failed to meet the as-found set-pressure testing acceptance criteria under Work Order (WO) 05240622-04, EWP MM Perform Testing of Relief Valve 1E12-F025C. The licensee had replaced the valve with a pre-tested relief valve and entered the issue into the corrective action program (CAP) as AR 4501066, 1E12-F025C WO#5240622-04.
The inspectors reviewed licensee Procedure LMP-GM-06, Bench Testing/Setting of ASME OM Class 2 and 3 Safety Relief Valves, Revision 36, and noted step E.2.3.3 stated, Initiate an Issue Report. Issue Report must specify that an Action Tracking Item be created for a Causal Evaluation of the failed as-found test. The Inspectors requested the casual evaluation of the failed as-found set-pressure test, and requested the WOs which performed the inservice testing of two additional relief valves as a result of the 1E12-F025C testing failure.
In response to the inspectors questions, the licensee determined they had failed to perform inservice testing on two additional relief valves, and had failed to perform the causal evaluation of the testing failure. The licensee determined they had inadequate WO closure documentation for WO 5240622-04 and had failed to determine the additional scope expansion testing requirements under AR 4501066.
Corrective Actions: The licensee entered this issue into their CAP and planned to perform the additional testing during outage L1R20.
Corrective Action References: AR 04720092, PI&R Issue - Missed Relief Valve Tests
Performance Assessment:
Performance Deficiency: The licensees failure to perform inservice tests to verify the operational readiness of relief valves, whose functions are required for safety, was a performance deficiency and contrary to 10 CFR 50.55a (f)(4)(ii). Specifically, the licensee failed to perform inservice testing on two additional relief valves from the R105 valve group, following the set-pressure testing failure of the 1E12-F025C valve, and failed to perform the cause-and-effect evaluation of the testing failure.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when a relief valve fails as-found set-pressure testing, the licensee is required to test additional relief valves as a means to address any generic issues which could apply to similar valves in the valve group. The licensee failed to perform set-pressure testing on two additional valves within the valve group, and therefore failed to ensure the availability, reliability, and capability of those additional valves to perform their intended safety function.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined the finding was of very low safety significance (Green) because although the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), the SSC maintained its operability and PRA functionality.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, when the 1E12-F025 set-pressure failure occurred, the organization did not thoroughly evaluate the issue to ensure compliance with all applicable ASME OM Code requirements.
Enforcement:
Violation: Title 10, CFR Part 50, 50.55a (f)(4)(ii) states, in part, Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section 18 months before the start of the 120-month interval.
The licensees ASME Code of Record for Operation and Maintenance of Nuclear Power Plants, is the 2004 Edition with Addenda though 2006.
Paragraph ISTC-5240, Safety and Relief Valves, states, safety and relief valves shall meet the inservice testing requirements of Mandatory Appendix I. Mandatory Appendix I, Paragraph I-1350(c)(1), states, for each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either +/- tolerance limit of the Owner-established set-pressure acceptance criteria of I-1310(e) or +/- 3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group. Additionally, Mandatory Appendix I Paragraph I-1350(c)(3) states, in part, the Owner shall evaluate the cause and effect of valves that fail to comply with the set-pressure acceptance criteria established in I-1350(c)(1).
Contrary to the above, as of December 15, 2023, the licensee failed to perform inservice tests to verify the operational readiness of valves whose function is required for safety, in accordance with the requirements of the 2004 ASME OM Code with Addenda through 2006. Specifically, following the set-pressure testing failure of safety-related relief valve 1E12-F025C, the licensee failed to perform inservice testing on two additional valves from the same (R105) valve group, and failed to evaluate the cause and effect of this valve that failed to comply with the set-pressure acceptance criteria established in I-350(c)(1).
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On December 15, 2023, the inspectors presented the biennial problem identification and resolution inspection results to John Van Fleet, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
297044
TSP#3 RPS Relay 1C71A-K10C Did Not Deenergize as
Required
24099401
Results of Failure Analysis on RPV Limit Switch
20487
2C71A-K106G Relay Did Not Open During LOS-RP-Q2
4171578
Roll Up IR for RPS TSV Limit Switch Issues
265960
Large Steam Leak Found Downstream of 1G33-F040
4404903
2DG01A Heat Exchanger Tubes Material Degradation
4406519
NRC ID: Failure to Write IR 4406975
Damage in 2B33-F060B Rx Recirculation Flow Control Valve
4407109
Unit 2 Refuel Bridge Mast Issue During Fuel Moves
4407880
Spurious Group 4 Signal Causes VG/VR Actuations
03/10/2021
4407906
MSL Tunnel Temp Exceeds 135 F on U1
03/10/2021
4435863
NOSA-LAS-21-05 NOS Audit of LaSalle CAP
4455617
2A RPS Bearing Replacement 4.0 Critique
4460065
NRC MD 8.3 Revisions - NRC Incident Investigation Program
4468596
RM - U2 Radiochemistry Samples Ind. Presence of Fuel
Defect
2/28/2021
4468963
Level 3 OPEX Review for IRIS# 515757
447352
Level 3 OPEX Review for IRIS 515751
4474379
Level 3 OPEX Review for IRIS 513205
4478972
NOS ID: Parts Control Requires Management Attention
4479498
RM-1A RR Failed to Downshift and Tripped to 0 Speed
4480457
Level 3 OPEX Review for IRIS 513205
4480462
Level 3 OPEX Review for IRIS 518226
4481815
Trend IR 1B HX Partition Plate
4483962
Eval Required to Leave Scaffold in Place (RWCU Scaffolding)
4485486
Level 3 OPEX Review for IRIS# 515457
4487513
1WTO1PB 1B TBCCW Pump Degrading Seal
4487864
NOS ID: Gaps within CAPE 4407880-10 and Associated
Actions
03/28/2022
Corrective Action
Documents
4496476
Trend - Incorrect IR Significance Level
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
4498422
LPCS Water Leg Pump Leak
4500996-42
Perform Risk Assessment for the Eaton Breakers
4501066
4507165
NOS ID: Continuing Issues with Parts Control in EMD Shop
22651
B Diesel Fire Pump Coolant Leak
23547
NRC ID: DBAI Design Analysis (EMD) Discrepancy
27147
NRC ID: DBAI Diesel Fuel
27147
NRC ID DBAI: Diesel Fuel Oil Level Calculation Errors
4530043
Shepard Calibrator Unsecured
4541790
Tagouts Self-Assessment Gap #1
2/09/2022
4541799
Tagouts Self-Assessment Gap #2
2/09/2022
4548783
FP SA Review of Completed Fire Protection Permits
01/17/2023
4548974
01/18/2023
4549245
01/19/2023
4549300
01/19/2023
4549302
01/19/2023
4549364
01/19/2023
4549541
01/20/2023
4549546
01/20/2023
4549548
01/20/2023
4549551
01/20/2023
4549568
01/20/2023
4549615
FP SA Fire Brigade Participation 2021
01/20/2023
4551562
1A DG Speed Control Governor Not Responding
4556006
NRC ID: Fire Door Issue
4558566
Level 3 OPEX Review for IRIS# 543717
4559142
2E12-F064B Wont Stay Closed
4559305
U2 Auto Scram
03/04/2023
4564302
Trend in Non-Discretionary Clock Resets 2022 to 2023
4667295
Level 3 OPEX Review for IRIS #546271
04/03/2023
4671310
Site Trend in CAP Products Going Overdue
4674503
LaSalle County SO and EOC EMNET/NARS Phone Failure
4677770
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
4678390
2A Diesel B/D Starting Air Motors Inline Oiler Not Working
4679014
ILT 21-1 NRC Preliminary Results
4681955-08
Review of Fleet Events Requiring IRIS Reporting
4684120
Maintenance Trend: MMD Attention to Detail
4689109
23 LaSalle Safety Culture Assessment
4689109-08
Biennial Fleet Safety Culture Self-Assessment
4689631
Level 3 OPEX Review for IRIS #561354
07/10/2023
4697410
Ops Obj 1 & 2 SA - Simulator UPS
08/19/2023
4697411
Ops Obj 1& 2 SA - Simulator Post-Event Testing
08/19/2023
4697412
Ops Obj 1& 2 SA - EO-I Qualification Book Cues
08/19/2023
4697413
Ops Obj 1 & 2 SA - Objective Alignment
08/19/2023
4697414
Ops Obj 1 & 2 SA - Operator Fundamentals Reference
08/19/2023
4697415
Ops Obj 1 & 2 SA - Objective Implementation
08/19/2023
4697416
Ops Obj 1 & 2 SA - PBIG Development
08/19/2023
4702449
Level 3 OPEX Review for IRIS# 561471
09/14/2023
11/07/2018
Potential Floatable Material (Snake) Found in U1 RB
08/28/2021
Eng. Trend - Leadership Behaviors in Department
Improvement
05/09/2022
0VE02CA Tripped
05/20/2022
Trend in DC Battery Corrosion Issues
06/15/2022
L1M24 Critical Path Delay - 35 Hours
07/27/2022
NRC RIS 2022-22 Op Leakage Inconsistent w/Op Eval Proc
11/14/2022
Level 3 OPEX Review for IRIS# 543773
01/19/2023
Potential Trend: L2R19 Failed Welds
03/02/2023
RM - LOS-RP-SA4 1C71A-K10G Didnt De-Energize as
Expected
09/07/2023
1HP55A-4" Pipe Section Replacement
11/30/2023
Scope Add Process Not Implemented Properly
01/15/2016
LOS-LP-Q1 and A RHR Inoperable is This Necessary?
01/29/2016
1A RHR Inoperable Due to Low Pressure
2/13/2016
Indication Identified on 1DG02A-10"
2/18/2016
Line 1DG05A-4" Was Not Cut in The Correct Location Per EC
2/19/2016
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Rejectable Indication in 1DG02A-10" Weld 3 WO 01326732-02
2/20/2016
Division 2 CSCS Issues
2/27/2016
LOS-RH-SR1 2B RHR PMP Seal Cooler Flowrate Pre-Test
Data
07/15/2016
Low Flow Going Through 2B RHR Pump Seal Cooler
07/18/2016
License Renewal (LR) Inspections for Selective Leaching
07/21/2016
Pipe Downstream of 2E12-F407A is Degraded
09/14/2016
Maintenance Rule Unavailability Exceeded for 2-CSCS-DG-06
09/15/2016
(A)(1) Action Plan Not Completed Per Schedule
09/14/2016
Troubleshoot Cause of Flow Increase
2/16/2016
2E12-F386B, Hand Wheel Found Detached Upon Arrival
2/08/2017
Maintenance Rule (A)(1) Determination for LAS-0-DG-01
03/15/2017
Engineer Evaluate Trend in DG Cooling Water Flow
03/22/2017
IR Not Written in a Timely Manner
04/07/2017
NDE 1VY03A 2.5" Supply Header UT Readings Below Min.
04/14/2017
NOS ID: Non-Consequential Errors in CSCS Hydraulic Model
06/22/2017
Level 3 OPEX Review Requested LER 3412017003 Fermi 2
09/27/2017
Data Points Do Not Support Welding for 2HP55BB Line
01/29/2018
LAS-2-VY-02 Mrule Hours Criteria Exceeded
2/15/2018
NOS ID: Active Leak from Drain Line
04/24/2018
Preconditioning Eval Request for LOS-RH-Q2, ATT. 1A
10/12/2018
VT-3 Exams Specified on Supports Exempted from
Examination
2/19/2019
2B DG Coolant Leak
03/27/2019
CMO Failed to Perform GL 89-13 RHR HX Inspection
04/05/2019
Existing Cooler Housing Has Sag
06/14/2019
Isolated pit identified on CSCS discharge pipe 1RH83BA-24"
10/09/2019
Replace Piping 1DG23B-6"
10/16/2019
Flooring and Piping Support Degradation in U2 CSCS Pump
Room
10/16/2019
Visual Exam Results for ISI Component HPCS DG Cooler 1
2/10/2020
Pipe Wall Leak Immediately Downstream of 1FC045B
03/02/2020
Extent of condition UT inspection on 2DG23B-6"
05/26/2020
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Localized Corrosion Spots Identified on 1HP54A Piping
10/27/2020
Extent of Condition UT Inspection on 1HP54A-4"
11/06/2020
2A RHR Pump Seal Cooler Flow Low
2/12/2021
2DG01A Heat Exchanger Tubes Material Degradation
2/26/2021
NOS ID Final Weld not Inspected in WO 4677538-29
03/23/2021
UT inspection on 1RH83BB-24"
09/01/2021
NOS ID: Insulation on Valve 2FC045A Crushed
2/10/2022
NOS ID: Pipe Support Severely Corroded, Actions Untimely
2/10/2022
Level 3 OPEX Review for IRIS# 459111
06/01/2022
Level 3 OPEX Review for IRIS# 493320 and 500237
06/15/2022
Level 3 OPEX Review for IRIS# 529028
07/18/2022
Snake in U1 RB Lower Raceway
09/06/2022
LOS-DG-SR7 Acceptance Criteria Not Met
11/21/2022
Level 3 OPEX Review for IRIS# 541158
01/19/2023
Level 3 OPEX Review for IRIS# 542416
01/19/2023
results on
UT Inspection Results on 1HP52A-10"
10/28/2020
233263
PI&R Identified Issue - Gaps Identified with RCR 4406976
20087
PI&R Issue - IR 4468596 Incorrect Significance Level
20092
PI&R Issue - Missed Relief Valve Tests
20279
PI&R Issue - Legacy Effectiveness Reviews Not Assigned
21459
PI&R Issue - NRC PI&R Identified LVL 3 OPX Issue
22116
PI&R Issue - NRC PI&R Question Regarding
CAPE 4479498-07
22154
PI&R Issue - Revise Policy Guidance 139 for Quorum
23257
PI&R Identified Issue - Clarify CA from CAPE 4011747
23263
PI&R Identified Issue - Gaps Identified with RCR 4406976
2/13/2023
23313
PI&R Identified Issues - IR How Discovered Incorrectly Coded
23516
PI&R Inspection Issue - Cause from CAPE not Addressed
by CA
2/14/2023
23611
PI&R Inspection Identified - HPCS Degraded Piping
2/14/2023
Corrective Action
Documents
Resulting from
Inspection
24263
PI&R - NRC Ops Review Lists No Operability Concerns
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
633646
Lost Parts Evaluation L2R18
2
637157
Line Kill for RT Heat Exchanger Vent Leaks
2VY03A Cooler Tubing Inlet Riser Repair Evaluation
05/23/2018
Engineering
Changes
TCC 634296
TCCP Temp Setpoint Change for 2B RHR WS Strainer
DPS 2E12-N501 From 4 psid to 6 psid
LaSalle Nuclear Safety Culture Review Meeting Minutes
10/9/2023
IST-LAS-BDOC-
V-26
LaSalle County Generating Station - Inservice Testing Basis
Document - 1E12-F025C
03/01/2019
Miscellaneous
BWROG Reactivity Controls Review Committee (RCRC)
Guidelines for Excellence
Technical Report
AH1543-
41060080
Long Range Guided Wave Ultrasonic Pipe Screening Results
07/05/2022
Technical Report
AH1548-
41076266
Long Range Guided Wave Ultrasonic Pipe Screening Results
07/05/2022
Technical Report
AM1261-343593,
AM1295-344407
Long Range Guided Wave Ultrasonic Pipe Screening Results
09/29/2011
Technical Report
AM3116-430677
Long Range Guided Wave Ultrasonic Pipe Screening Results
11/21/2012
Technical Report
AM4103-535476
Long Range Guided Wave Ultrasonic Pipe Screening Results
11/20/2013
NDE Reports
Technical Report
AM4103-535476R
Long Range Guided Wave Ultrasonic Pipe Screening Results
01/13/2014
Safety Conscious Work Environment
Revision 4
LEP-HC-103
Refuel Bridge Preventative Maintenance
Revision 18
LGA-002
Secondary Containment Control
LOS-NB-R2
Reactor Vessel Leakage Test
LOS-RD-SR3
Control Rod Operations
Revision 27
Reactivity Management Administration
Operating Experience Program
Revision 6
Procedures
Root Cause Analysis Manual
Revision 7
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action Program Evaluation Manual
Effectiveness Review Manual
Revision 2
Self-Assessments
NOSA-LAS-22-01
Maintenance Functional Area Audit Report
2/23/2033
1741625-03
EM EWP Special Test 2C71-K4B1 Agastat Relay MR-90
03/09/2021
5166843-01
RXS - Perform Foreign Material Inspection of Jet Pumps
2/17/2023
5168984-03
RXS - Inspect & Clean Unit 2 Bottom Head Drain
2/27/2023
Work Orders
256192-01
RXS Perform Ultrasonic Fuel Cleaning in L2R19
2/23/2023