IR 05000263/2024002
| ML24226B357 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/14/2024 |
| From: | Richard Skokowski NRC/RGN-III/DORS/RPB3 |
| To: | Hafen S Northern States Power Company, Minnesota |
| References | |
| IR 2024002 | |
| Download: ML24226B357 (1) | |
Text
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT-INTEGRATED INSPECTION REPORT 05000263/2024002
Dear Shawn Hafen:
On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Monticello Nuclear Generating Plant. On July 16, 2024, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations 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 Monticello Nuclear Generating Plant.
August 14, 2024 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 Monticello Nuclear Generating Plant.
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, Richard A. Skokowski, Chief Reactor Projects Branch 3 Division of Operating Reactor Safety Docket No. 05000263 License No. DPR22
Enclosure:
As stated
Inspection Report
Docket Number:
05000263
License Number:
Report Number:
Enterprise Identifier:
I2024002-0062
Licensee:
Northern States Power Company, Minnesota
Facility:
Monticello Nuclear Generating Plant
Location:
Monticello, MN
Inspection Dates:
April 01, 2024 to June 30, 2024
Inspectors:
N. Bolling, Project Engineer
T. McGowan, Resident Inspector
C. Norton, Senior Resident Inspector
J. Reed, Health Physicist
Approved By:
Richard A. Skokowski, Chief
Reactor Projects Branch 3
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Monticello Nuclear Generating Plant, 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 Follow Procedure Resulting in Reactor SCRAM Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000263/202400201 Open/Closed
[H.2] - Field Presence 71153 A self-revealed finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B,
Criterion V, Instructions, Procedures and Drawings, was identified when the licensee failed to follow a surveillance procedure during the functional test of the anticipated transient without SCRAM (ATWS) system, resulting in a reactor SCRAM.
Failure to Promptly Identify a Condition Adverse to Quality resulting in Failure to Recognize Entry into a Condition Governed by a Technical Specification Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000263/202400202 Open/Closed
[H.11] -
Challenge the Unknown 71153 A self-revealed finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B,
Criterion XVI, Corrective Actions, occurred due to the licenses failure to promptly identify a condition adverse to quality (CAQ) resulting in the failure to complete required technical specification (TS) actions. Specifically, on February 28, 2024, following a reactor SCRAM accompanied by a trip of the recirculation pumps, the licensee did not promptly identify portions of the reactor pressure vessel (RPV) exceeded the limiting cooldown rate specified in the Pressure Temperature Limits Report (PTLR). As a result, the licensee did not complete the required TS action to determine the reactor coolant system (RCS) was acceptable for continued operation.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000263/2024001-00 LER 2024001-00 for Monticello Nuclear Generating Plant, Reactor Scram, Containment Isolation, and Cooldown Rate outside of limits following technician adjustment of wrong component 71153 Closed
PLANT STATUS
Monticello began the inspection period at 31 percent of rated thermal power (RTP). The licensee increased power and reached 100 percent of RTP on April 2, 2024. On April 24, 2024, the licensee lowered power to 80 percent of RTP to accommodate wind generation and returned to 100 percent of RTP. On April 25, 2024, the licensee lowered power to 80 percent of RTP to accommodate wind generation and returned to 100 percent of RTP. Monticello remained at or near RTP for the remainder of the inspection period.
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 performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed onsite portions of IPs. 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.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Intake structure fire sprinkler system on April 29, 2024
- (2) Control rod drive SCRAM solenoids energized prior to half SCRAM testing on May 20, 2024
- (3) Standby liquid control on May 21, 2024
- (4) High pressure core injection (HPCI) on May 30-31, 2024
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a complete walkdown of the reactor core isolation cooling (RCIC) system on June 13, 2024.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Fire zone 17, turbine building north cable corridor 941' on April 24, 2024
- (2) Fire zone 16, corridor, turbine building east & west (elevation 911' and 931') on April 24, 2024
- (3) Fire zone 2B, east hydraulic control unit (HCU) area on May 3, 2024
- (4) Fire one 309, control room on May 6, 2024
- (5) Fire zone 19C, feed water pipe chase on May 14, 2024
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during an announced fire drill on April 20, 2024.
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated internal flooding mitigation protections in the lower 4kV room.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during flexible power operations on April 25, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated emergency preparedness exercise on May 7, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (3 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) D53 13 battery charger on April 17, 2024
- (2) Low pressure coolant injection subsystems (LPCI) following division1 inboard injection valve failure on June 27, 2024
- (3) Change out of 12 residual heat removal service water pump on July 30, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (7 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Risk associated with control rod timing and pattern adjustment on April 3, 2024
- (2) Risk associated with group 3 relay replacements on April 22, 2024
- (3) Risk associated with reactor core isolation cooling system valve testing while division 2 auxiliary transformer is unavailable on May 13, 2024
- (4) Licensee assessment and mitigation of risk associated with upcoming ATWS logic maintenance during MT T2/T4 meeting on May 13, 2024
- (5) Risk associated with 2R transformer extended isolation on May 22, 2024
- (6) Risk associated with high pressure coolant injection system testing and challenges to rated core thermal power limit on June 16, 2024
- (7) Risk Associated with division 1 LPCI outboard injection valve inoperable while stuck in mid-stroke on June 28, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (3 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
(1)
===501000083649, emergency diesel generator (EDG) 11 air start 2 air consumption (2)501000085833, FQ2544 drywell floor drain sump monitoring system out of calibration criteria (3)501000087225, MO2012 failed to cycle as expected
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)
=
The inspectors evaluated the following temporary or permanent modifications:
- (1) Plug installed in the standby liquid control system test tank suction
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (4 Samples)
(1)250 VDC battery charger D53 return to service following preventive maintenance, on April 17, 2024
- (2) EDG 11 consolidated testing following preventive maintenance, on April 17-18, 2024
- (3) EDG 12 consolidated testing following preventive maintenance on, May 1-2, 2024
- (4) MO 2012, LPCI outboard injection valve following emergent maintenance on June 29, 2024
Surveillance Testing (IP Section 03.01) (3 Samples)
- (1) Main steam line low pressure group 1 isolation instrument test and calibration on April 10, 2024
- (2) Standby liquid control testing on May 9, 2024
- (3) Low pressure coolant injection loop select testing on May 16, 2024
71114.06 - Drill Evaluation
Required Emergency Preparedness Drill (1 Sample)
- (1) Emergency preparedness drill on May 7,
RADIATION SAFETY
71124.03 - InPlant Airborne Radioactivity Control and Mitigation
Permanent Ventilation Systems (IP Section 03.01) (1 Sample)
The inspectors evaluated the configuration of the following permanently installed ventilation systems:
- (1) Unit 1 control room emergency ventilation filter unit
Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.
71124.04 - Occupational Dose Assessment
Source Term Characterization (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.
External Dosimetry (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.
Internal Dosimetry (IP Section 03.03) (2 Samples)
The inspectors evaluated the following internal dose assessments:
- (1) Internal dose assessment for ID N153067 (2)internal dose assessment for ID 313234
Special Dosimetric Situations (IP Section 03.04) (2 Samples)
The inspectors evaluated the following special dosimetric situations:
- (1) Declared pregnant worker form for ID 237872
- (2) Skin dose assessment for ID
OTHER ACTIVITIES-BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)===
- (1) April 01, 2023 through March 31, 2024
BI02: RCS Leak Rate Sample (IP Section 02.11) (1 Sample)
- (1) April 1, 2023 through March 31, 2024
71152 A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) The human performance deficiencies that contributed to the February 28, 2024, reactor SCRAM
- (2) Failure to lower reactor power prior to performing a preplanned evolution expected to cause an increase in reactor power that could exceed rated thermal power (RTP),minor violation
- (3) Intermediate Range Monitor (IRM) Compliance with IEEE 279, Proposed Criteria for Nuclear Power Plant Protection Systems
- (4) MO2012 Failed to Cycle as Expected and minor 10CFR 50.72 reporting violation
71152 S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the licensees corrective action program to identify potential trends in main steam line high radiation alarms during plant manipulations that might be indicative of a more significant safety issue.
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)
- (1) The inspectors evaluated the licensee response to the unplanned inoperability of the A outboard low pressure coolant injection valve and both divisions of LPCI on June 28, 2024.
Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensees event reporting determinations to ensure it complied with reporting requirements.
- (1) LER 05000263/2024001-00, Reactor Scram, Containment Isolation, and Cooldown Rate outside of limits following technician adjustment of wrong component, (ADAMS Accession No. ML24116A117)
The inspection conclusions associated with this LER are documented in this report under Inspection Results Section
INSPECTION RESULTS
Minor Violation 71152 A Failure to Lower Reactor Power Prior to Performing a Preplanned Evolution Expected to Cause an Increase in Reactor Power That Could Exceed Rated Thermal Power Minor Violation: On June 13, 2024, the inspectors identified during the performance of 025506-IA1, HPCI [high pressure coolant injection system] Pump and Valve Tests, on March 21, 2024, operators did not comply with procedural requirements to reduce reactor prior to running HPCI. Consequently, thermal power spiked several times above RTP.
Procedure C.205, Power Operation, states For preplanned evolutions that could affect reactor power..., it is to be determined if the evolution is expected to cause a transient increase in reactor power. If the evolution is expected to cause a transient increase in reactor power that could exceed 2004 MWt, RTP, action SHALL be taken to reduce power prior to performing the evolution. Referencing IMC 612 Appendix E, example 9.a, Examples of Minor Issues, the inspectors assessed the issue to be minor. Although running HPCI added positive reactivity, the licensee monitored the 2-hour, 1-hour, and 10-minute thermal power averages and would have reduced power had they exceeded any of these average thermal power limits, thus ensuring abundant fuel cladding barrier integrity margin.
The licensee entered this issue into their corrective action program, QIM 501000086704, NRC ID: Licensed Power Limit.
On June 17, 2021, the inspectors observed the performance of 025506-IA1, HPCI Pump and Valve Tests. The procedure had been changed to include steps to lower reactor power if the transient increase in reactor power could exceed rated thermal power (RTP). After determining that thermal power could increase above RTP, the operators lowered reactor power before starting the surveillance. During the evolution, no spikes above RTP occurred.
The inspectors identified no issues of concern and determined the corrective actions were effective. Future corrective actions include performing a procedure review for other evolutions that could cause transient increases in RTP and adding steps to lower reactor power if the transient increase in reactor power could exceed RTP.
The inspectors assessed that corrective actions adequately maintain margin to RTP during HPCI testing. No issues of concern were identified.
Screening: The inspectors determined the performance deficiency was minor. This is consistent with IMC 612 Appendix E, example 9.a, Examples of Minor Issues.
Enforcement:
This failure to comply with Licensee Procedure C.205, Power Operation, which describes for preplanned evolutions that could affect reactor power, action shall be taken to reduce power prior to performing the evolution, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
Observation: Short Term Actions to Correct the Human Performance Deficiencies that Contributed to the Reactor SCRAM on February 28, 2024 71152 A On February 28, 2024, while performing 0278B, [Anticipated Transient without SCRAM]
ATWS-Recirculation Trip for Reactor Pressure and Level Trip Unit Test and Calibration, technicians failed to use proper human performance (HU) tools and manipulated the wrong potentiometer causing an ATWS SCRAM and recirculation pump trip. The corrective action program (CAP) document 501000082361, Reactor SCRAM No. 142: ATWS Trip, identified that supervision failed to ensure technicians understood expected HU tool usage.
Specifically, flagging, peer checking, and hard match were used inappropriately. There was no supervisor in the field. In addition, the procedure did not specify the expected level of human performance tool usage.
On May 29, 2024, the inspectors observed the performance of 0278A, ATWS-Recirculation Trips for Reactor Pressure and Level Trip Unit Test and Calibration, on the opposite ATWS division. The inspectors observed that the procedure contained steps directing the use of the expected HU tools; the technicians employed the expected HU tools; and the supervisor was at the job site.
The inspectors assessed the short-term actions to address the HU issues that contributed to the February 28, 2024, reactor SCRAM were adequate. No issues of concern were identified.
Observation: MO2012, Division 1 Low Pressure Coolant Outboard Injection Valve Failed to Cycle as Expected 71152 A On June 27, 2024, while performing OSP-RHR0556, [residual heat removal] RHR Water Fill Verification, MO2012, the division 1 outboard low pressure coolant injection (LPCI)injection valve failed in mid-stroke. The motor turned in the open and closed directions, but the clutch mechanism would not engage to stroke the valve. Manual valve operation was likewise impeded. The failure of MO2012 in mid-stroke prevented both LPCI divisions from fulfilling their safety functions. At 0110 on June 28, 2024, in accordance with TS 3.5.1 D, the licensee declared both LPCI divisions inoperable and entered a 72-hour shutdown limiting condition of operation (LCO). The licensee documented the condition in QIM 501000087255, MO2012 Failed to Cycle as Expected, and commenced troubleshooting and repair activities. Troubleshooting revealed that a broken spring in the MOV declutching mechanism had lodged in a position that precluded valve movement. The valve was repaired, tested and returned to service. The licensee exited the 72-hour shutdown LCO at 1533 on June 29, 2024. The licensee is performing a root cause analysis to determine why the valve failed.
The inspectors determined the corrective actions to repair MO2012 were adequate to return the valve to service and exit the LCO. Further evaluation of the valve failure mechanism will be completed during the assessment of the associated LER.
Observation: Intermediate Range Monitor Compliance with IEEE 279, Proposed Criteria for Nuclear Power Plant Protection Systems 71152 A On March 3, 2024, during the reactor startup and following notching rods to criticality, the inspectors observed the IRM 12 HI HI light illuminate and seal in on the C36 intermediate range neutron monitor cabinet. The licensee documented this in corrective action program document (CAP) 501000082463. The associated annunciator, 5A-21, IRM A HI HI/INOP did not alarm, nor was a half SCRAM received concurrent with annunciator 5B-4, REACTOR AUTO SCRAM CHANNEL A for an IRM HI HI condition. In addition, the IRM 12 HI HI warning light on C05 was not received (expected at 119.375 of 125 units). IRMs were monitored and IRM 12 continued to track with the others; no IRMs were observed to be erratic. Reactor startup continued from criticality to the point of adding heat. When the IRM 12 indications on C36 were observed, operators verified expected current IRM conditions. The IRM recorder on C05 was observed to have a peak indication of 118.8 units. Operators verified that OPS-NIS0042, IRM Functional Test completed on February 29, 2024, indicated no discrepancies.
The inspectors questioned whether the condition described above complied with IEEE 2791968, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, Criterion 4.16, Completion of Protective Action Once It Is Initiated, which states, The protection system shall be so designed that, once initiated, a protection system action shall go to completion. Return to operation shall require subsequent deliberate operator action.
The licensee initiated CAP 501000085948 to address the failure to address the 10CFR50.55 a requirement for compliance with IEEE 279 standards in the resolution of CAP 501000082463.
The licensee documented in condition evaluation 500000329671, IEEE279 Compliance/CAP 82463, that during the startup on March 3, 2024, elevated indication nearing the HI HI setpoint was observed on IRM12. Review of the recorder data showed a maximum indication of 118.8, which is below the trip setpoint of 120. The IRM chassis handles both the indicating light and trip output via separate internal circuitry. While it is ideal for the light indication and trip output circuits to actuate simultaneously, in this event, the light indication circuit actuated and sealed-in while the trip output did not actuate. At no time was a protective action initiated by the sensor, nor was one required based on the indications. The anomaly was that the light indication actuated prematurely. Following the event, testing was performed to verify proper time response of the IRM signal from the preamplifier, through the IRM drawer, and into the RPS logic. The equipment functioned properly. ISP-NIS0043, IRM Time Response Procedure, was performed and satisfied the acceptance criteria which is based upon the General Electric Reactor Protection System Design Specification 257HA344 which identifies IEEE 279 within the codes and standards.
The inspectors did not identify any issues of concern with Monticello Nuclear Generating Plants IRMs compliance with IEEE 279, Criterion 4.16.
Observation: Non-Conservative Main Steam Line High Radiation Setpoint 71152 S The inspectors performed a detailed review of corrective action programs (CAPs) related to main steam line high radiation alarms during plant manipulations. During the review, inspectors identified that since at least December 2019, the licensee had not adequately implemented the main steam line high radiation (MSLHR) setpoint into their procedures and/or setpoint specifications. Specifically, the licensee had not verified that the MSLHR setpoint was in accordance with design basis requirements which resulted in a non-conservative alarm setpoint.
The licensees design basis states in part, the licensee utilizes hydrogen water chemistry (HWC) which has a significant effect on background radiation levels in the MSLs.
Occasionally, the need to modify the hydrogen injection program may arise. The licensee therefore intends to adjust the radiation monitor setpoints as necessary to account for the background effects of HWC changes. Set points will be selected to maintain a margin that is at least equivalent to a 1.5 X full power background level setpoint. The inspectors reviewed current and historical main steam line radiation levels through December 2019 and determined that the current setpoint of 1400 mrem/hr does not represent a margin equivalent to a 1.5 X full power background level setpoint. The licensees procedure 1414 Rev 18, Main Steam Line Radiation Monitor Test and Calibration, did not include steps to ensure compliance with the 1.5 X full power background requirement and instead used a fixed setpoint of 1400mrem/hr.
The licensee has entered this issue into their corrective action program under QIM 501000084421. Additionally, the licensee has other leading indication of fuel failure; therefore, no findings or violations of NRC requirements of more-than-minor safety significance were identified by the inspectors in the course of this review.
Failure to Follow Procedure Resulting in Reactor SCRAM Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000263/202400201 Open/Closed
[H.2] - Field Presence 71153 A self-revealed finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified when the licensee failed to follow a surveillance procedure during the functional test of the anticipated transient without SCRAM (ATWS) system resulting in a reactor SCRAM.
Description:
On February 28, 2024, at 0839 (CST) while in Mode 1 at 100 percent power, an inadvertent ATWS system initiation resulted in a trip of both reactor recirculation pumps, depressurization of SCRAM air header, and control rod insertion. Control room operators took appropriate action to insert a manual SCRAM. The SCRAM was caused by maintenance personnel not following a surveillance procedure. During the performance of step 115 of the 0278B surveillance procedure to adjust the A channel setpoint adjust potentiometer, the instrumentation and controls (I&C) technician incorrectly adjusted the potentiometer on the adjacent module which caused a trip of the C channel. The trip of both the A and C channels caused an ATWS initiation of alternate rod insertion and a trip of both reactor recirculation pumps.
Corrective Actions: Surveillance procedure 0278B was revised to include mandatory human performance error reduction techniques.
Corrective Action References: CAP 501000082261, "Reactor SCRAM No. 142: ATWS Trip"
Performance Assessment:
Performance Deficiency: The licensee failed to follow a station procedure affecting quality, resulting in a reactor SCRAM. The I&C technician manipulated the incorrect component during the prescribed calibration activity. This was within the licensees ability to foresee and prevent; therefore, this was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the reactor SCRAM was caused by maintenance personnel not following the procedure during the performance of the logic system functional test for ATWS recirculation pump trip instrumentation. The licensee manipulated the incorrect component, causing the reactor SCRAM.
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 answered No to transient questions, Did the finding cause a reactor trip AND the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss of condenser, loss of feedwater)? Other events include high-energy line breaks, internal flooding, and fire? Therefore, the finding was screened as having very low safety significance (Green).
Cross-Cutting Aspect: H.2 - Field Presence: Leaders are commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations. Deviations from standards and expectations are corrected promptly. Senior managers ensure supervisory and management oversight of work activities, including contractors and supplemental personnel.
Leaders were not present in the field ensuring maintenance personnel were reinforcing standards and expectations.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established 0278B, revision 31, ATWS-Recirc Trip for Reactor Pressure and Level Trip Unit Test and Calibration, as the implementing procedure for performing a functional test of the ATWS logic circuitry, an activity affecting quality.
Procedure 0278B, revision 31, at Step 115 states, IF performing procedure during the 1st quarter of the year, THEN EXERCISE TRIP ADJ potentiometer [A channel potentiometer]
R11 10 complete cycles.
Contrary to the above, on February 28, 2024, while performing Procedure 0278B, revision 31, Step 115, the I&C specialist adjusted C channel potentiometer instead of A channel potentiometer. The failure to manipulate the correct component resulted in a reactor SCRAM.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Promptly Identify a Condition Adverse to Quality Resulting in Failure to Recognize Entry into a Condition Governed by a Technical Specification Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000263/202400202 Open/Closed
[H.11] -
Challenge the Unknown 71153 A self-revealed finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Actions, occurred due to the licenses failure to promptly identify a condition adverse to quality (CAQ) resulting in the failure to complete required technical specification (TS) actions. Specifically, on February 28, 2024, following a reactor SCRAM accompanied by a trip of the recirculation pumps, the licensee did not promptly identify portions of the reactor pressure vessel (RPV) exceeded the limiting cooldown rate specified in the Pressure Temperature Limits Report (PTLR). As a result, the licensee did not complete the required TS action to determine the reactor coolant system (RCS) was acceptable for continued operation.
Description At 0839 on February 28, 2024, the licensee experienced a plant SCRAM with a trip of the reactor recirculation pumps and entered Mode 3. Unknown to the operators, RPV cooldown rates in the bottom head violated the PTLR cooldown limits. Operators were unable to identify the violations of the PTLR limits because their procedure, 0118, Reactor Vessel Temperature Monitoring, did not account for the recirculation pump trips and RPV temperature stratification, which are activities that adversely affected quality. Subsequently, they transitioned from mode 3 to mode 4 at 2332 on February 28, 2024, from mode 4 to mode 2 at 1610 on March 2, 2024, entered mode 1 at 0234 on March 3, 2024, and reached 100 percent of RTP at 2030 the same day. On March 8, 2024, the licensee determined that within an hour of the reactor SCRAM, the RPV bottom head had exceeded the PTLR cooldown limit of 100F/hour with at a rate of 129F/hour. At 1400 on March 8, 2024, the licensee entered TS LCO 3.4.9 and initiated the A.2 required action to determine that the RCS was acceptable for continued operation. After obtaining an evaluation from Structural Integrity Associates that there was sufficient margin in the PTLR bottom head limit to accommodate the cooldown rates experienced, the licensee exited TS LCO 3.4.9 condition A at 2000 on March 08, 2024.
In addition to procedure 0118, Reactor Vessel Temperature Monitoring, not accounting for recirculation pump trips and RPV temperature stratification, which are activities that adversely affected quality, subsequent licensee processes collectively failed to promptly identify that the cooldown rate had exceeded the PTLR cooldown rate limit in three key areas, a condition adverse to quality. First, operators were not taking into account that the bottom head was cooling down at a faster rate than the beltline when cooling the reactor to enter mode 4.
Second, engineering did not identify the faster cooldown rate of the reactor bottom head while supporting operations as they cooled the reactor to enter mode 4. Third, contrary to TS 3.4.9.C, the licensee failed to promptly identify that the PTLR cooldown rate limit had been exceeded, and a determination that the RCS was acceptable for continued operation prior to going to mode 2 from mode 4.
Corrective Actions: In addition to determining there was sufficient margin in the PTLR curve to accommodate the cooldown rates experienced, the licensee changed C.4B.01.04.B, Abnormal Procedures-Trip of Two Recirculation Pumps, to note temperature stratification could result in excessive cooldown of the RPV bottom head. On June 25, 2024, inspectors interviewed reactor operators, revealing the licensee had not changed procedure 0118, Reactor Vessel Temperature Monitoring. Following the procedure, operators would not identify excessive bottom head cooldown rates following a reactor SCRAM with a trip of both recirculation pumps because an imbedded computer program did not account for the RPV temperature stratification that occurs without recirculation. Therefore, the inspectors questioned the effectiveness of the licensees corrective actions. As a result, the licensee documented this in QIM 501000087051, Gap in Station Response to CAP. The licensee implemented procedure changes to make procedure 0118 more explicit and correct the computer program that monitors RPV temperature rate of change to account for RPV temperature stratification following the trip of both recirculation pumps.
This violation closes LER 2024001-00.
Corrective Action References:
QIM 501000082683, RPV Bottom Head Cooldown Rate Greater than 100 F/hour QIM 501000087051, Gap in Station Response to CAP [NRC identified gap in station response to RPV bottom head cooldown rate greater than 100 F/hour]
Performance Assessment:
Performance Deficiency: Failure to promptly identify a CAQ resulting in failure to recognize entry into a condition governed by a TS LCO is a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the RCS Equipment and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The licensee started the reactor and increased power to RTP without knowledge the cooldown rate for the bottom head and vessel drain line had exceed PTLR limits. The plant operated at RTP for 5 days without a determination that the RCS was acceptable for continued operation.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria. After determining the PTLR cooldown rate for the bottom head had been exceeded, the licensee entered TS LCO 3.4.9.A. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the licensee received an initial operability summary for the reactor pressure vessel bottom head from Structural Integrity Associates, Inc. that concluded the bottom head had sufficient margin to support the cooldown rates experienced without effect on the bounding operational PT Limit Curves. After reviewing the full operability summary from Structural Integrity Associates, the NRC Vessels and Internals Branch assessed the transient did not significantly challenge RCS integrity and the finding screened to Green.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. During uncertain plant conditions, with the recirculation pumps tripped and RPV stratification, the licensee failed to challenge whether the cooldown rate was within the PTLR limit.
Enforcement:
Violation: 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires, in part, measures be established to ensure that conditions adverse to quality, such as deficiencies and nonconformances, are promptly identified.
10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, requires, in part, activities affecting quality shall be prescribed by documented procedures, of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Procedures shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. The licensee established procedure 0118, Reactor Vessel Temperature Monitoring, Revision 23, as the implementing procedure for reactor vessel temperature monitoring, an activity affecting quality.
Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits, states RCS pressure, RCS temperature, RCS heat up and cooldown rates, and the recirculation pumps starting temperature requirements shall be maintained within limits specified in the PTLR.
Condition A states, Requirements of LCO not met in Mode 1, 2 or 3, Required Action A.2 states, Determine RCS is acceptable for continued operation, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Condition C states, Requirements of LCO not met in Mode 1, 2 or 3, Required Action C.1 initiate action to restore parameter(s) within limits immediately and Required Action C.2 states, Determine RCS is acceptable for continued operation, prior to entering Mode 2 or 3.
Contrary to the above, on February 28, 2024, the licensee failed to promptly identify a CAQ.
Specifically, the licensee failed to identify prior to going from mode 4 to mode 2, that the RPV bottom head had exceeded the cooldown rate limit specified in the PTLR without determining that the RCS was acceptable for continued operation.
Additionally, the licensee failed to meet the requirements of 10 CFR 50, Appendix B, Criterion V, when procedure 0118, Reactor Vessel Temperature Monitoring, Revision 23, did not include appropriate quantitative and qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the procedure did not account for pump trips and RPV temperature which adversely affected quality.
Subsequently, the licensee started the reactor in violation of Technical Specification requirements 3.4.9.A.2 and TS 3.4.9.C.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Minor Violation 71153 Late required 8-hour 10CFR50.72 Notification Minor Violation: On June 27, 2024, while performing OSP-RHR0556, RHR [Residual Heat Removal] Water Fill Verification, MO2012, the division 1 outboard low pressure coolant injection (LPCI) injection valve failed in mid-stroke. At 0110 on June 28, 2024, in accordance with TS 3.5.1 D, the licensee declared both LPCI divisions inoperable. The licensee notified the inspectors of the event at 0400 on June 28, 2024, but at 1529 on June 28, 2024, was 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 19 minutes late in making a 10 CFR 50.72(b)(3)(v), required 8-hour notification for Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:...(B) Remove residual heat...
This minor violation was licensee identified.
Screening: The inspectors determined the performance deficiency was minor. Although the licensee official notification was late, the NRC ability to respond was not impeded.
Enforcement:
This failure to comply with 10CFR 50.72 8-hour reporting requirements constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On July 16, 2024, the inspectors presented the integrated inspection results to Shawn Hafen, and other members of the licensee staff.
- On May 15, 2024, the inspectors presented the radiation protection baseline inspection results to S. Scott, Director of Nuclear Licensing and Regulation, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
NH 36665
P&ID Service Water System & Make Up Intake Structure
NH36245
Control Rod Hydraulic System
NH36251 (M125)
NH36252 (M126)
NH36253
Standby Liquid Control System
Drawings
NH36666
P&ID Screen Wash, Fire, & Chlorination System, Intake
Structure
25502-III
SBLC Quarterly Pump and Valve Tests
25508-IA1
RCIC Quarterly Pump and Valve Tests
Procedures
215413
RCIC System Prestart Valve Checklist
Work Orders
0007001342440020
SCRAM Solenoid Energization Test
STRATEGY
A.302-B
Fire Zone 2B East HCU Area
STRATEGY A.309
Fire Zone Control Room Strategy A.309
STRATEGY A.316
Corridor, Turbine Building East & West (Elevation 911'
and 931')
STRATEGY A.317
Fire Zone 17, Turbine Building North Cable Corridor 941'
Fire Plans
STRATEGY
A.319-C
Fire Zone 19C Feed Water Pipe Chase
Procedures
2176
Fire Drill Procedure
Procedures
B.09.0605
Operation of HELB Flood Barrier Lower 4kV
Miscellaneous
Full Scale Drill Timeline
05/07/2024
2300
Reactivity Adjustment
OPS MAN B.0509-
MPR/EPR Adjustment
Procedures
OPS MAN C.205
B.2 Power Adjustments
501000084213
Ground Alarm From D53
04/17/2024
501000084235
DC Ground Identified Upon D53 Restore
04/17/2024
Corrective Action
Documents
501000087225
MO2012 Failed to Cycle as Expected
06/28/2024
214
RHR Service Water Pump Replacement
Procedures
8151
Heavy Load Movement
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Ops Man
B.06.0502
Condensate and Reactor Feedwater, Description of
Equipment
Ops Man
B.06.0505
Condensate and Reactor Feedwater System Operation
2000025803
Trouble Shooting Plan: MO2012
06/28/2024
7000899300080
700130629
D53, Circuit Board & CAP RPLC 4525PM
04/17/2024
Work Orders
7001409820010
HIP HTR 14B Level Dump Control
501000085669
2R Transformer Window Extended
05/22/2024
Corrective Action
Documents
501000087225
MO2012 Failed to Cycle as Expected
06/28/2024
Corrective Action
Documents
Resulting from
Inspection
501000086704
NRC ID: Licensed Power Limit
06/13/2024
Miscellaneous
2407
Operations Memo: Pre-Planned Evolutions Affecting Core
Thermal Power
06/14/2024
25508-1A1
RCIC Quarterly Pump and Valve Test
2300
Reactivity Adjustment
2300
Reactivity Maneuvering Steps
04/03/2024
C2.05
Power Operation
FGOP-RSK01
Configuration Risk Monitor User Guide
Procedures
FPOP-RSK01
Risk Monitoring and Risk Management
7000815560020
Relays K10B/C304B K65B
7000815570020
Relays K10D/C304D K65D
Work Orders
7001374740010
Group 3 Primary Containment Isolation 16A-K6 AD, Install
Banana Jacks
04/22/2024
501000085833
FQ2544 Out of Calibration Criteria
05/22/2024
501000087255
MO2012 Failed to Cycle as Expected
06/28/2024
Corrective Action
Documents
50100083649
EDG 11 Air Start 2 Air Consumption
04/03/2024
0533
Containment Sump Flow Measurement Instrumentation
Procedures
410002-OCD
EDG 11 2 Starting System
Drawings
NH36253
Standby Liquid Control System
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Engineering
Evaluations
CR 602000032882
Procedure Change Request, SBLC Comprehensive Pump
and Valve Tests
11/10/2023
Procedures
25502-III1A
SBLC Comprehensive Pump and Valve Tests
501000084213
Ground Alarm From D53
04/17/2024
501000084235
DC Ground Identified Upon D53 Restore
04/17/2024
501000084251
Swing Check Valve Non Functioning 11 EDG
04/17/2024
Corrective Action
Documents
501000084253
Discrepancy Between EDG Test Steps
04/17/2024
0054B
Main Steam Line Low Pressure Group 1 Isolation
Instrument Test and Calibration (RX In Run)
2502-111
Standby Liquid Control Quarterly Pump and Valve Test
25PM
No. 13 & 16 Battery Charger Preventive Maintenance
B.09.0905
H.1 250 VDC Battery Charger Transfer
ISP-RHR055201
Reactor Recirculation Loops DP LPCI Select Interlock
Channel Functional Test
OPS MAN
B.09.0802
OPS MAN
B.09.0805
D.2 12 Emergency Diesel Generator Startup
OPS MAN
B.09.0806
OSP-EDG005012
Emergency Diesel Generator Consolidated Testing
Procedures
OSP-EDG055011
Emergency Diesel Generator Consolidated Testing
7000899300080
MO2012 PMT
06/29/2024
7001308010030
Emergency Diesel Generator Speed Sensing Panel 1
MT:1:DGN:DG2/SSP1
7001309450020
Emergency Diesel Generator MT:1:DGN:G3B
7001356670020
Work Orders
70013962920
Place D 53 Charger in Service
04/17/2024
Procedures
5790101-02
Emergency Action Level Matrix
501000071354
HEPA Problems During DOP Testing
03/14/2023
501000072278
Unexpected Airborne Area SJAE Room
04/14/2023
501000073091
Delta Suit Implementation Question
05/03/2023
Corrective Action
Documents
501000080705
SSA: SCBAs Will Miss Annual Flow Testing
01/10/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
501000081147
Airborne Area Entry Controls
01/24/2024
Corrective Action
Documents
Resulting from
Inspection
501000085183
FPRP-RPP01 Clarification
05/09/2024
FireHawk M7
Responder Air
Mask
Vendor Manual
Form 5646
Service Air Composition Test - Grade D Test
07/20/2023
Form 5646
Service Air Composition Test - Grade D Test
03/19/2024
Miscellaneous
Form 740 L
Monthly SCBA Inspection
07/06/2023
FPRP-RPP01
Respiratory Protection Program
FPRP-RPP02
Respirator Fit Testing and Portacount Fit Tester Operation
R.02.04
Analysis of Airborne Radioactivity Samples
R.05.02
Respirator Maintenance
R.05.03
Respirator Issuance
Procedures
R.05.07
SCBA Inspection and Functional Check
Work Orders
700077705
HEPA Filter Efficiency Test
09/22/2023
Dose Assessment
for ID 313217
Internal Dose Assessment
04/21/2023
Calculations
Dose Assessment
for ID N153067
Internal Dose Assessment
05/01/2023
501000070703
DLR Spike Data Results Unacceptable
2/15/2023
501000072780
Contaminated Worker Released from PA
04/25/2023
Corrective Action
Documents
501000079747
Individual Entered RCA Without SRD
2/07/2023
501000085046
Incorrect RWP Use During 1R31
05/07/2024
501000085074
Prudency of CAP Requirements for WBC
05/08/2024
501000085135
Changing Radiological Conditions
05/09/2024
Corrective Action
Documents
Resulting from
Inspection
501000085367
71124.03/04 Minor Violation
05/15/2024
Dose assessment
for ID 248904
Skin Dose Assessment
10/03/2023
Miscellaneous
Form 5833 for ID
237872
Unborn Child Protection Program
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
NVLAP Certification
NVLAP Certificate of Accreditation for Mirion Technologies
10/03/2023
Procedures
R.14.13
Electronic Dosimeter Operations
Technical Basis
Document 19003
Neutron DLR, SRD, and Rem Ball Characterization
Technical Basis
Document 22001
Neutron Factor for Drywell Startup / Power Entries
Radiation
Surveys
Technical Basis
Document 23001
Prospective Dose Assessment and Dose Reporting
Threshold Summary
QF0445
NRC Data Collection and Submittal - Reactor Coolant
System Total Leakage (April 2023 Through March 2024)
71151
Miscellaneous
QF0445
NRC and Data Collection and Submittal - Safety System
Functional Failures (April 2023 Through March 2024)
501000082361
Reactor SCRAM No. 142 ATWS Trip Cause Evaluation
03/08/2024
Corrective Action
Documents
501000082361
Reactor SCRAM No. 142: ATWS SCRAM
2/28/2024
501000082463
IRM 12 HI HI Light on C36 During Startup
03/03/2024
Corrective Action
Documents
Resulting from
Inspection
501000085948
IEEE 279 Compliance/CAP 82463
05/29/2024
Engineering
Evaluations
500000329671
IEEE 279 Compliance/CAP 82463
06/25/2024
Miscellaneous
257HA344 AB
General Electric Design Specification, Reactor Protection
System
25506-1 IA1
HPCI Quarterly Pump and Valve Tests.
114
278A
ATWS - Recirc Trips for Reactor Pressure and Level Trip
Unit Test and Calibration
71152 A
Procedures
FPGDOC03
Procedure And Work Instruction Use and Adherence
Engineering
Changes
Revise MSLRM Alarm Setpoint
Condition Eval, QIM 501000084421, Main Steam Line
Radiation Monitor Set Points for 1.5*Nom Background
Radiation
DPR22
Amendment No.83
Monticello Nuclear Generating Plant-Amendment No. 83 to
Facility Operating License No. DPR22
08/18/1992
71152 S
Miscellaneous
Safety Evaluation for Elimination the Boiling Water Reactor
October
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Main Steam Line Isolation Valve Closure Function and
Scram Function of the Main Steam Line Radiation Monitor
1992
1414
Main Steam Line Radiation Monitor Test and Calibration
Procedures
7702
Instrument Information Sheet-PCR 01349750
501000082327
278B Trip/Cal: NUS Trip Unit Pots
2/28/2024
501000082338
ATWS SCRAM While Performing 0278B
2/28/2024
501000082361
Reactor Scram No. 142 ATWS Trip
2/28/2024
501000082683
Reactor Bottom Head Cooldown Rate Greater Than 100/hr
03/08/2024
Corrective Action
Documents
501000087225
MO2012 Failed to Cycle as Expected
06/28/2024
Corrective Action
Documents
Resulting from
Inspection
501000087051
Gap in Station Response to CAP
06/25/2024
0118
Reactor Vessel Temperature Monitoring
278B
ATWS-RECIRC Trip for Reactor Pressure and Level Trip
Unit Test and Calibration
Procedures
QF0431
(FPPACSE01)
Cause Evaluation Template
21