IR 05000341/2022003

Integrated Inspection Report 05000341/2022003

ML22311A531
Person / Time
Site:	Fermi
Issue date:	11/08/2022
From:	Billy Dickson NRC/RGN-III/DORS/RPB2
To:	Peter Dietrich DTE Electric Company
References
IR 2022003
Download: ML22311A531 (29)

Text

November 8, 2022

SUBJECT:

FERMI POWER PLANT, UNIT 2 - INTEGRATED INSPECTION REPORT 05000341/2022003

Dear Peter Dietrich:

On September 30, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Fermi Power Plant, Unit 2. On October 19, 2022, 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. One Severity Level IV violation without an associated finding is documented in this report. 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 Fermi Power Plant, Unit 2.

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 Fermi Power Plant, Unit 2. 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, Signed by Dickson, Billy on 11/08/22 Billy C. Dickson, Jr., Chief Reactor Projects Branch 2 Division of Operating Reactor Safety Docket No. 05000341 License No. NPF-43

Enclosure:

As stated

Inspection Report

Docket Number: 05000341 License Number: NPF-43 Report Number: 05000341/2022003 Enterprise Identifier: I-2022-003-0042 Licensee: DTE Electric Company Facility: Fermi Power Plant, Unit 2 Location: Newport, MI Inspection Dates: July 01, 2022 to September 30, 2022 Inspectors: T. Briley, Senior Resident Inspector R. Cassara, Resident Inspector J. Gewargis, Resident Inspector V. Myers, Senior Health Physicist R. Ng, Senior Project Engineer J. Reed, Health Physicist T. Taylor, Senior Resident Inspector Approved By: Billy C. Dickson, Jr., Chief Reactor Projects Branch 2 Division of Operating Reactor Safety Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Fermi Power Plant, Unit 2, 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

Actuator-Yoke Separation on a Main Steam Line Drain Primary Containment Isolation Valve Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.12] - Avoid 71111.12 NCV 05000341/2022003-01 Complacency Open/Closed A self-revealed finding of very low safety significance (Green) with an associated non-cited violations (NCV) of Technical Specification 5.4, "Procedures," occurred when motor-operated valve B2103F019, one of the main steam line drain primary containment isolation valves, became damaged when the actuator separated from the yoke after being operated. The issue was discovered during a startup-related walkdown by licensee personnel at the end of a forced outage. Undersized bolts had been used to attach the actuator to the yoke.

Failure to Submit a Licensee Event Report for the Actuator-Yoke Separation of a Main Steam Line Drain Primary Containment Isolation Valve Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71111.12 NCV 05000341/2022003-02 Open/Closed The inspectors identified a Severity Level IV Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.73 (a)(2)(i)(B), Licensee Event Report System, when the licensee failed to submit a Licensee Event Report (LER) within 60 days of identifying a condition prohibited by Technical Specifications.

Reactor Scram Due to Trip of Reactor Feed Pumps Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green None (NPP) 71152A NCV 05000341/2022003-04 Open/Closed A finding of very low safety significance (Green) with an associated non-cited violation (NCV)of Technical Specification 5.4, "Procedures," was self-revealed on February 4, 2022, when the reactor automatically scrammed from approximately 58 percent power during a plant shutdown to start a refueling outage. Appropriate precautions regarding feed pump suction pressure were not included in the procedure for operating the feedwater system.

Additional Tracking Items

Type Issue Number Title Report Section Status URI 05000341/2022003-03 Seismic Displacement for 71111.18 Open Safety-Related Piping Not Verified

PLANT STATUS

Unit 2 started the reporting period at or near 100 percent reactor power. On September 21, 2022, the unit commenced a planned downpower to approximately 65 percent for maintenance and a rod pattern adjustment. The unit returned to 100 percent reactor power on September 26, 2022, and remained at or near 100 percent reactor power for the remainder of the 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 on-site 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.

71111.01 - Adverse Weather Protection

Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal high temperatures for the following systems:

120kV switchyard, 345 kV switchyard, residual heat removal service water (RHRSW),and control center heating, ventilation, and air conditioning (CCHVAC) during the week ending August 31, 2022

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (2 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) Emergency diesel generator (EDG) 12 during EDG 13 maintenance during the week ending July 23, 2022

(2) Division 2 CCHVAC partial equipment alignment during Division 1 CCHVAC chiller work during the week ending on August 12, 2022

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (4 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) Reactor building fourth floor, recirculation system motor generator area during the week ending July 23, 2022

(2) Control air compressor room during the week ending July 23, 2022

(3) Reactor building second floor north and south quadrants during the week ending September 30, 2022

(4) Auxiliary building fifth floor Division 1 and 2 CCHVAC system equipment rooms during the week ending September 30, 2022

Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the on-site fire brigade response to an announced drill on September 1, 2022, which involved a simulated fire in one of the turbine lube oil rooms.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)

(1) The inspectors observed and evaluated licensed operator requalification training on Engage/Vaporstream, the licensee event notification system, on September 28, 2022.

(2) The inspectors observed simulator training on anticipated transient without scram scenarios, September 15, 2022.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (2 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) Division 1 CCHVAC journal bearing assessment and replacement due to a high bearing oil temperature trip of the chiller on August 3, 2022

(2) Motor operator valve actuator removal and installation practices, actuator to yoke mounting bolts and washers during the week ending September 24, 2022

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (3 Samples)

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

(1) Operability and functionality assessment performed on the Division 1 EDG sequencer trouble alarm received, ending on September 30, 2022

(2) NOVA inverters, Division 1 testability power supply, CARD 22-22886 during the week ending September 17, 2022

(3) Operability of Division 1 channel 'B' turbine building area temperature high primary containment isolation instrumentation erratic indication as documented in CARDs 22-26351 and 22-26412, ending September 30, 2022

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)

(1 Sample)

The inspectors evaluated the following temporary or permanent modifications:

(1) Permanent modification of the new RHRSW and emergency equipment service water (EESW) piping/supports/penetrations

71111.19 - Post-Maintenance Testing

Post-Maintenance Test Sample (IP Section 03.01) (2 Samples)

The inspectors evaluated the following post-maintenance testing activities to verify system operability and/or functionality:

(1) Integrated plant computer system drywell curves not indicating in the complete pressure range during the week ending August 13, 2022

(2) Division 1 B21N117B turbine building area temperature high instrumentation following channel failure during the week ending May 21, 2022

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)

(1) The inspectors completed a review of work hours controls during the RF21 refueling outage, which concludes the outage sample started in the first quarter of 2022.

71114.06 - Drill Evaluation

Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)

The inspectors evaluated:

(1) Emergency preparedness drill on August 16,

71124.06 - Radioactive Gaseous and Liquid Effluent Treatment

Walkdowns and Observations (IP Section 03.01) (1 Sample)

The inspectors evaluated the following radioactive effluent systems during walkdowns:

(1) Reactor building ventilation

71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &

Transportation

Radioactive Material Storage (IP Section 03.01)

The inspectors evaluated the licensees performance in controlling, labeling and securing the following radioactive materials:

(1) Radioactive materials in the radioactive waste on-site storage facility

(2) Radioactive materials in warehouse G

Radioactive Waste System Walkdown (IP Section 03.02) (1 Sample)

The inspectors walked down the following accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality:

(1) Resin processing equipment

Waste Characterization and Classification (IP Section 03.03) (2 Samples)

The inspectors evaluated the following characterization and classification of radioactive waste:

(1) Bead resin

(2) Oil waste

Shipping Records (IP Section 03.05) (4 Samples)

The inspectors evaluated the following non-excepted radioactive material shipments through a record review:

(1) Radioactive waste shipment, EF2-22-034, of low specific activity bead resin in a general design package

(2) Radioactive waste shipment, EF2-22-047, of waste class 'A' bead resin in a type 'B' package

(3) Radioactive waste shipment, EF2-22-018, of dry active waste in a general design package

(4) Radioactive waste shipment, EF2-21-006, of waste class 'A' bead resin in a type 'A' package

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:

MS08: Heat Removal Systems (IP Section 02.07) ===

(1) July 1, 2021 through June 30, 2022

MS09: Residual Heat Removal Systems (IP Section 02.08) (1 Sample)

(1) July 1, 2021 through June 30, 2022

MS10: Cooling Water Support Systems (IP Section 02.09) (1 Sample)

(1) July 1, 2021 through June 30, 2022 BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (1 Sample)

(1) October 1, 2021 through June 30, 2022

OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)

(1) October 1, 2021 through June 30, 2022 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)

(1) October 1, 2021 through June 30, 2022

71152A - 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) Follow-up to selected CARDs implying production-over-safety, during the week ending August 20, 2022

(2) Follow-up to a reactor scram caused by a perturbation in the feedwater system, during the week ending September 24, 2022

(3) Reactor scram caused by mayflies, during the week ending September 30,

INSPECTION RESULTS

Actuator-Yoke Separation on a Main Steam Line Drain Primary Containment Isolation Valve Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.12] - Avoid 71111.12 NCV 05000341/2022003-01 Complacency Open/Closed A self-revealed finding of very low safety significance (Green) with an associated non-cited violations (NCV) of Technical Specification 5.4, "Procedures," occurred when motor-operated valve B2103F019, one of the main steam line drain primary containment isolation valves, became damaged when the actuator separated from the yoke after being operated. The issue was discovered during a startup-related walkdown by licensee personnel at the end of a forced outage. Undersized bolts had been used to attach the actuator to the yoke.

Description:

On June 27, 2022, the licensee was performing a walkdown of the reactor building steam tunnel near the end of a forced outage. The actuator for motor-operated valve B2103F019 was observed near the top of the valve stem and in continuous operation. The actuator had detached from the valve yoke and had walked up the valve stem during the operation. The motor remained energized, causing the valve stem to shake back and forth. The licensee secured power to the valve actuator, took control of the valve, and, using the manual handwheel, walked the actuator back down the valve stem to the valve yoke. Afterward, the licensee installed bolting to reconnect the actuator and valve yoke. The licensee considered the valve inoperable. The licensee performed an engineering evaluation to support leaving the valve in the condition for the remainder of the cycle. The valve is one of two in a series that perform a containment isolation function for main steam line drains. The valves are normally closed during operation, and the licensee manipulates them during plant startups and shutdowns. Per technical specifications, the licensee verified that the upstream valve was closed, allowing for continued operation.

Initial investigation revealed a lack of proper thread engagement on the bolts that held the actuator to the yoke. With insufficient thread engagement, valve operating forces allowed the actuator to become detached from the valve yoke and move up the threads of the stem when operators attempted to close the valve from the control room. The valve initially indicated correctly, but the actuator had separated and moved up the valve stem in the field. The abnormal configuration resulted in the control system continuing to attempt to close the valve.

With the actuator continuously running, valve position indication fluctuated between open and closed for nearly 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The licensee identified this condition after-the-fact when the licensee reviewed plant computer data. The licensee performed an extent of the condition review and determined that other valves that had similar work performed during the outage remained operable. The inspectors reviewed the licensee's efforts and did a field walkdown of one of the valves.

Later investigation by the licensee revealed that improper thread engagement existed due to maintenance staff using the wrong length bolts to connect the actuator to the yoke. During the outage, the licensee performed thrust testing on the valve. This test included separating the actuator and the valve yoke, inserting a test device, and reconnecting all three together using the normally installed bolts along with four additional bolts (the additional bolts for the test device are shorter than the four normally installed bolts). Upon reassembly following testing, maintenance personnel incorrectly used the shorter bolts to reattach the actuator to the yoke, and the licensee did not find the other four bolts.

Corrective Actions: The licensee secured the actuator back in place on the yoke, performed an engineering evaluation, and performed an investigation.

Corrective Action References: CARD 22-27461

Performance Assessment:

Performance Deficiency: Procedure 46.306.06, "MOV Diagnostic Testing with the QUIKLOOK 3 System," was not performed correctly. The procedure referred to "provided" and "original" bolts when directing installation and removal of the test device. In the restoration section of the procedure, step 5.8.3 directs the restoration of subcomponents to the original configuration. The licensee did not install the original bolts to connect the valve yoke to the actuator.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control 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. Specifically, use of incorrect bolts resulted in a safety-related containment isolation valve becoming inoperable.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.

Specifically, the finding screened to Green based on answering 'no' to both questions in Section C of Exhibit 3 of IMC 0609. The finding did not create an actual open pathway in containment, nor did it involve hydrogen igniters.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, individuals performing the valve reassembly failed to recognize and/or check for the minimum amount of thread engagement when bolting the actuator to the yoke. The licensee determined with the incorrect bolts installed, minimum actual thread engagement would have existed and could have been noticed. Further, the licensee did not take appropriate actions to validate the correct size bolts when following the step to restore the valve to its "original configuration."

Enforcement:

Violation: Technical Specification (TS) 5.4, Procedures, requires, in part, that the applicable procedures recommended in Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements (Operation), are established, implemented, and maintained. Section 9 of RG 1.33 states, in part, that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, on April 11, 2022, Procedure 46.306.06, "MOV Diagnostic Testing with the QUIKLOOK 3 System," a procedure that affects the performance of safety-related equipment, was not properly performed. Specifically, the procedure directed the original bolts to be installed between the actuator and yoke after testing, and they were not. Shorter bolts were installed, and as a result the yoke and actuator separated during valve operation, rendering the valve inoperable. Compliance with technical specifications was restored on June 29, 2022, when the licensee performed the required actions to close and deactivate the other containment isolation valve in series.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Submit a Licensee Event Report for the Actuator-Yoke Separation of a Main Steam Line Drain Primary Containment Isolation Valve Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71111.12 Applicable NCV 05000341/2022003-02 Applicable Open/Closed The inspectors identified a Severity Level IV Non-Cited Violation (NCV) Title 10 of the Code of Federal Regulations (10 CFR) Part 50.73 (a)(2)(i)(B), Licensee Event Report System, when the licensee failed to submit a Licensee Event Report (LER) within 60 days of identifying a condition prohibited by Technical Specifications.

Description:

On June 26, 2022, the licensee was performing a post forced outage walkdown of the steam tunnel on the first floor of the reactor building and discovered the actuator for B2103F019, the outboard main steam line drain primary containment isolation valve (PCIV), was disconnected from the yoke of the valve and shaking. On June 27, 2022, the licensee documented the discovery and inoperability in Condition Assessment Resolution Document (CARD) 22-27461. On July 21, 2022, the licensee completed a past operability evaluation and determined that B2103F019 was inoperable from March 27, 2022, when maintenance personnel installed the incorrect actuator bolts during the most recent refueling outage. The incorrect bolting installation is described in this report as NCV 05000341/2022003-01.

Additionally on July 21, 2022, the licensee completed a reportability evaluation that concluded the actuator-yoke separation was not reportable to the NRC based on the safety function of isolating the primary containment flow path being maintained with the operable inboard PCIV.

The inspectors agreed with the conclusion regarding the safety function. However, per technical specifications, the inoperable outboard PCIV would have required the flow path to be isolated with a closed and deactivated valve in the flow path within four hours. If the licensee did not isolate the flow path within that time frame, the technical specifications would have required the plant to be in Mode 3 (Shutdown) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Since the licensee became aware of the past inoperability on July 21, 2022, the licensee had 60 days from that date to submit the LER per 10 CFR Part 50.73 (a)(2)(i)(B). The licensee failed to submit the LER in time by not recognizing the reporting criterion.

Corrective Actions: The licensee documented the failure to report in the corrective action program and reassessed the issue of reportability.

Corrective Action References: CARD 22-30092, CARD 22-27461

Performance Assessment:

None. The inspectors determined this violation was associated with a minor performance deficiency since it only dealt with reporting requirements. A finding of very low safety significance (Green) associated with the incorrect bolts is discussed in this report as NCV 05000341/2022003-01.

Enforcement:

The Reactor Oversight Processs (ROPs) significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.

Severity: Based on the NRC Enforcement Policy dated January 14, 2022, Section 6.9, Subsection d, Number 9, lists a failure to make a required report per 10 CFR Part 50.73, LER System, as an example of a Severity Level IV violation.

Violation: 10 CFR Part 50.73, "Licensee Event Report System," states, in part, holders of an operating license for a nuclear power plant shall submit a LER for any event of the type described in this paragraph within 60 days after the discovery of the event.

Section (a)(2)(i)(B) of 10 CFR Part 50.73, "Licensee Event Report System," states, in part, that the licensee shall report any operation or condition which was prohibited by the plant's Technical Specifications.

Contrary to the above, since September 20th, 2022 (60 days from the completion of a past operability assessment regarding PCIV B2103F019), through the date of the exit meeting for this report (October 19, 2022), the licensee failed to submit a LER. At the time of the exit meeting, the licensee had a corrective action to draft and submit the LER.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Unresolved Item Seismic Displacement for Safety-Related Piping Not Verified 71111.18 (Open) URI 05000341/2022003-03

Description:

Updated safety analysis report Table 3.2-1 delineates the residual heat removal service water (RHRSW) piping is designed to American Society of Mechanical Engineers (ASME)

Section III, Subsection ND, 1971 edition. ASME Section III Subsection ND-3611 states, in part, The requirements for acceptability of class 3 piping systems are that they shall be designed in accordance with the rules of NC-3600 except as otherwise permitted in this Sub article. ASME Section III Subsection NC-3622, states, in part, The provisions of NB-3622 shall apply except that, in addition ASME Section III Subsection NB-3622.1 requires impact forces caused by either external or internal conditions shall be considered in the piping design.

The inspectors reviewed calculation no. DC-2966 Volume Number IA DCD 2, Piping Stress Report RHR 03/19, Revision 0. The licensee performed this calculation to analyze the Division 1 RHRSW supply and return piping inside the reactor building. The inspectors noted that the maximum displacement (based on the seismic loading condition) for the piping was 1.233 inches. The licensee did not perform a physical inspection to determine whether the maximum displacement was acceptable and verify that no external impact forces exist between the piping and a system, structure, or component (SSC).

This issue is unresolved because the inspectors cannot determine whether there is a violation and will need information based on a physical inspection performed by the licensee to validate if the maximum piping displacement impacts any SSCs.

Planned Closure Actions: The inspectors will review the physical inspection information when it becomes available from the licensee to determine whether a violation exists.

Licensee Actions: The licensee plans to perform a physical inspection to validate if the maximum piping displacement impacts any SSCs.

Corrective Action References: CARD 22-27033, NRC Identified: Evaluation of Potential Rattle Space Violation, dated 06/10/2022.

Reactor Scram Due to Trip of Reactor Feed Pumps Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green None (NPP) 71152A NCV 05000341/2022003-04 Open/Closed A finding of very low safety significance (Green) with an associated non-cited violation (NCV)of Technical Specification 5.4, "Procedures," was self-revealed on February 4, 2022, when the reactor automatically scrammed from approximately 58 percent power during a plant shutdown to start a refueling outage. Appropriate precautions regarding feed pump suction pressure were not included in the procedure for operating the feedwater system.

Description:

On February 4, 2022, the plant started a downpower to begin a refueling outage. The plant had been at approximately 58 percent power for several weeks due to the planned power profile before the outage. Both turbine-driven reactor feedwater pumps (RFPs) provided flow to the reactor and maintained the reactor level. One of the first evolutions scheduled was to secure one of the RFPs. Despite not performing the just-in-time training (JITT) practice session for this evolution (since the licensee scheduled this evolution for the night shift), the day shift crew decided to start the downpower and secure an RFP.

Additionally, while the crew that practiced the evolution lowered power to approximately 50 percent before securing an RFP, the day shift crew started securing an RFP at around 58 percent power. While procedurally allowed (the maximum power to secure an RFP was 60 percent), starting at a higher power resulted in a lower RFP suction pressure. Reactor feedwater pump suction pressure is an important parameter because if the pressure gets too low, an RFP can trip, and if the pressure gets too high, perturbations can start in the heater drains system.

While lowering the speed of the south RFP, the minimum flow valve started to open gradually before suddenly going full open. A sudden pressure drop at the suction of the RFPs occurred, causing the north RFP to trip. Operators attempted to raise feed flow with the south RFP, but had to trip it manually because vibration levels had increased. With the loss of both RFPs, the reactor scrammed on a low reactor water level condition (Level 3). The crew stabilized the plant without other injection systems using the condensate and feedwater systems. The operators maintained pressure control with the turbine bypass valves. No significant complications occurred during the scram, and the operators stabilized the plant in hot shutdown. On the following shift, the plant continued into the refueling outage.

The licensee conducted a root cause analysis to evaluate the event. The inspectors reviewed the licensee's root cause analysis. The licensee determined that the minimum flow control valves operated as expected, and a gap in the procedure for removing an RFP from service existed. Specifically, procedure 23.107, "Reactor Feedwater and Condensate Systems," did not alert operators to the behavior of the RFP minimum flow control valves and the impact they could have on suction pressure. In 2001, the licensee changed the method of removing an RFP from service.

The new method (which existed in the procedure during the February 4, 2022, event)involved taking manual control of the RFP speed. While lowering the speed, as flow decreased, the minimum flow control valve would start to open. However, given Fermi's plant-specific design, the minimum flow control valve would suddenly go fully open at some point, causing an approximate 250-pound pressure drop at the RFP suction. Before the procedure change in 2001, operators adjusted a bias setting to get the minimum flow control valve to open before taking manual control of the RFP. This method avoided the sudden opening of the minimum flow control valve and the resultant suction pressure drop. With the procedure change, the licensee failed to recognize all impacts on the system, including the sudden pressure drop that would occur at the RFP suctions due to the new behavior of minimum flow control valves. As a result, the licensee did not include appropriate parameters in the procedure for operators to observe while securing an RFP, especially from higher initial power levels.

Corrective Actions: The licensee performed a root cause analysis. The licensee created actions to revise procedures to account for feed system behavior while removing RFPs from service.

Corrective Action References: CARD 22-21157

Performance Assessment:

Performance Deficiency: The inspectors identified that the licensee failed to maintain a procedure for operation of the reactor feedwater system that included operating parameters appropriate for a quality procedure as described in Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements (Operation). Specifically, critical parameters associated with operating the feedwater system were not included. The licensee is committed to RG 1.33 via TS 5.4, Procedures, and operation of the feedwater system is described as a safety-related process per RG 1.33.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality 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, a latent procedural deficiency resulted in a 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 finding screened as Green, or very low safety significance, based on answering 'no' to question B of Exhibit 1 of IMC 0609. Specifically, the normal feed and condensate systems remained able to control reactor water level while the turbine bypass valves controlled pressure following the scram (i.e., no loss of main condenser vacuum).

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance. Specifically, the procedure change occurred in 2001.

Enforcement:

Violation: Technical Specification (TS) 5.4, Procedures, requires, in part, that the applicable procedures recommended in RG 1.33, Quality Assurance Program Requirements (Operation), are established, implemented, and maintained. Section 4 of RG 1.33 lists startup, shutdown, and operation of a boiling water reactor feedwater system as an applicable procedure. RG 1.33 further states ANSI N18.7-1976/ANS-3.2, "Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants" requires preparation of the procedures. ANSI N18.7-1976/ANS-3.2 contains requirements for the content of procedures listed in RG 1.33. Section 5.3.2, Procedure Content, requires, in part, that procedures shall identify plant conditions that must exist prior to use. Further, that precautions should be established to alert the individual performing a task to those important measures which should be used to protect equipment and avoid abnormal situations.

Contrary to the above, from October 25, 2001, until April 19, 2022, Procedure 23.107, "Reactor Feedwater and Condensate Systems," was not maintained. Specifically, appropriate parameters (namely RFP suction pressure) were not included as prerequisites in the section for securing an RFP.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Observation: Response to Select CARDs in the Corrective Action Program 71152A The inspectors noted several Condition Assessment Resolution Documents (CARDs) written by licensee personnel near the end of the 2022 refueling outage that appeared to express concerns with production-over-safety. Some examples included questioning a policy of having operators man a confined space rescue team (CSRT) while they were also on the fire brigade, and questioning why a reactivity management senior reactor operator (RMSRO) was not stationed in the control room during a portion of the reactor startup. Further, the inspectors noted that the licensee had not potentially addressed a CARD from the 2020 refueling outage dealing with safety concerns associated with hand-barring the main turbine.

Additionally, the inspectors wanted to follow-up on an issue they identified during the refueling outage regarding how the licensee changed a test procedure. In consultation with regional management, the inspectors decided to review the licensee's follow-up to the various issues in the corrective action program.

The inspectors discovered no findings nor violations. However, the inspectors identified several observations that may contribute to the number of CARDs written questioning whether production or safety is the overarching priority when the licensee make decisions. In some cases, the inspectors noted a difference in the resolution documented in the CARDs and understood by the initiators, versus what had actually been done. Regarding the concern regarding the CSRT and fire brigade manning, operations management indicated they found additional personnel to man both positions, and this was not documented in the CARD nor communicated to the concerned operator. Additionally, while the practice may have been allowed, the inspectors noted an unbiased third party, such as a member of the safety department, could have weighed in and documented the acceptability.

For the hand-barring of the main turbine, the licensee took actions to address the concerns involving multiple departments, but the licensee did not document those actions nor was a formal work order created despite that being one of the corrective actions. For the RMSRO issue, while the licensee eventually staffed an additional senior reactor operator (SRO) from outside the shift, the initial attempt to resolve the question utilized a watchstander on shift to pick up the additional duty. If the assumption was that the plant was actually in the condition described by the concerned operator (an ongoing reactor startup), this would not have been allowed by the reactivity management procedure, MOP 19. As in the fire brigade example above, an unbiased third party may have been a more appropriate choice to evaluate the CARD. The inspectors noted a corrective action to benchmark other facilities and assess whether the licensee could add clarity to the procedure.

Regarding a change to how control rod scram time-testing was going to be performed (due to the normal measuring system being degraded), the inspectors questioned the use of a work order to allow use of other equipment versus following the more formal procedure change process. Ultimately, the licensee changed the procedure and wrote a CARD to explore the use of work orders. As a result, no findings nor violations were identified. The licensee also created an action to add clarity to their procedure change process.

Observation: Review of Reactor Scram due to Mayflies 71152A On June 24, 2022, the reactor automatically scrammed due to an electrical disturbance associated with the 345kV electrical distribution system on site. A large swarm of mayflies caused an electrical fault in the switchyard containing the main generator disconnects. As a result, the main turbine tripped, which caused the reactor to shut down.

Mayflies caused a similar electrical fault in 2020. The inspectors reviewed the circumstances, corrective actions from the 2020 event, and the evaluation performed for the 2022 event. The inspector identified no findings or violations. The 2020 event did not result in a reactor scram, and the licensee attributed the event to installing new LED lights near the 345kV switchyard.

Following that event, the licensee took actions to secure more lighting around the site, including the switchyard where the 2022 fault occurred. Unlike the 2020 event, where mayflies blanketed one of the electrical insulators and caused a short to ground due to excessive lighting in the area, the licensee determined that in 2022, a large swarm of mayflies traversed the site and, due to the density, caused a fault from ground to a suspended conductor despite the switchyard being completely dark.

The inspectors reviewed the immediate response to the trip, repairs, and proposed corrective actions. The inspectors identified no issues.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

On October 19, 2022, the inspectors presented the integrated inspection results to Mr. P. Dietrich, Chief Nuclear Officer, and other members of the licensee staff.

On August 19, 2022, the inspectors presented the Radiation Protection Baseline inspection results to Mr. P. Dietrich, Chief Nuclear Officer, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.01 Procedures MWC 16-100 Work Control Conduct Manual Implementing Procedure: 2

Seasonal Readiness

Work Orders 48703549 Evaluate Request for Shutdown 46304 Revisions 1 & 2 for 0

Impact to Fermi 2

51689384 120KV Relay House Battery Charger Irregularity 0

63421855 Oil Seepage from Breaker GK 0

6417741 Enrico Fermi PP Circuit Breaker GH (120KV): Breaker Trip 0

Coil Test (ITCTRANSMISSION)

64847930 Replace Tagline Switches S4000P003,4,5,6 0

255931 120KV Mat Fence Line Enhancements 0

71111.04 Drawings 6M721-5736-2 Control Center A/C Water System Functional Operating S

Sketch

6M721-5736-3 Control Center A/C Air System Functional Operating K

Sketch

Procedures 23.307 Emergency Diesel Generator System 135

23.413 Control Center HVAC 103A

23.413 Attachment Control Center HVAC System Valve Lineup 103A

23.413 Attachment Control Center HVAC System Electrical Lineup 103A

23.413 Attachment Control Center HVAC System Instrument Lineup 103A

23.413 Attachment Division 2 CCHVAC Standby Verification Checklist 103A

4B

23.413 Enclosure A Control Center HVAC Damper Lineup Normal Mode 103A

Division 2 Dampers

71111.05 Corrective Action 22-29295 Tracking AIM to Purchase Training Enhancement Items for 09/02/2022

Documents Fire Drills

Fire Plans FB-RB-2-10a Reactor Building Emergency Equipment Cooling Water, 4

North, Zone 10, EL. 613'6"

FP-AB-5-16d Auxiliary Building Division 1 Control Center Heating,

Ventilating, and Air Conditioning System Equipment

Inspection Type Designation Description or Title Revision or

Procedure Date

Room, Zone 16, EL. 677'6"

FP-AB-5-16e Auxiliary Building Division 2 Control Center Heating,

Ventilating, and Air Conditioning System Equipment

Room, Zone 16, EL. 677'6"

FP-AB-BMT-4 Control Air Compressor Room, Zone 4, Elevation 551'0" 5

FP-RB-2-10b Reactor Building Emergency Equipment Cooling Water, 5

South, Zone 10, EL. 613'6"

FP-RB-4-17b Reactor Building Recirculation System Motor Generator 5

Area Zone 17, Elevation 659'6"

FP-TB Turbine Building 11

Miscellaneous Fermi 2 Safety Handbook 13

Fire Brigade Drill LP-FP-940-0933/Fire Drill Main Lube Oil Reservoir Room 09/01/2022

Record Form TB2

71111.11Q Procedures 29.100.01 SH 1A RPV Control-ATWS 18

29.100.01 SH 5 Secondary Containment and Rad Release 14

29.100.01 SH1 RPV Control 19

EP-101 Classification of Emergencies 44

EP-290 Emergency Notifications 63A

EP-290 Enclosure A Electronic Notification Process 11/01/2021

EP-290 Enclosure B Nuclear Plant Event Technical Data Form General 12/06/2019

Information Requirements

EP-290001 Initial Notification Form Review Checklist 10/22/2021

EP-290001 Follow-up Notification Form Review Checklist 09/02/2021

71111.12 Corrective Action 17-25134 Division 1 CCHVAC Chiller High Bearing Oil Temperature 06/08/2017

Documents 21-21951 Division 1 CCHVAC Chiller Trip 03/03/2021

21-24247 T4100B009 Division 1 CCHVAC Chiller Trip (MCR Alarm 05/15/2021

8D5)

2-27400 T4100B009 Division 1 CCHVAC Chiller Trip (MCR Alarm 06/24/2022

8D5)

2-27461 FO 22-01 Start Up Walkdown: MOV Actuator 06/27/2022

Disconnected from Bonnet

2-28723 Low Oil Flow on from New CCHVAC Sleeve Bearings 08/14/2022

Engineering 22-018 B2103F019 Actuator Separation Impact on Appendix J, 0

Inspection Type Designation Description or Title Revision or

Procedure Date

Evaluations Including bypass, and IST Requirements

Procedures 20.413.01 Control Center HVAC System Failure 29

47.306.06 MS Drain Valve PCIV Actuator 2

ARP 8D5 Division 1 Control Room A/C Trouble 14

VMS25-39 Centrifugal Water Chillers 06/23/2010

Work Orders 58375116 Install TTC and Other Test Equipment in Preparation for 03/11/2022

As Found Test

71111.15 Corrective Action 21-31211 9D21 Division 1 EDG Sequencer Trouble 12/20/2021

Documents 22-22199 Frequency Oscillations Occurring for Installed Power 02/21/2022

Supplies and Contingency Power Supplies Need

Evaluated

2-22886 Nova Inverters Found OOT for B21K801B and R31K001 03/01/2022

2-26351 3D56 Testability Logic Channel A/B RPS/Power Failure - 05/16/2022

B21N617B Gross Failure Downscale

2-26412 3D56 Testability Logic Channel A/B RPS/Power Failure in 05/17/2022

Alarm

Corrective Action 22-27708 NRC Question - CARD 22-26351 Past Operability 07/05/2022

Documents

Resulting from

Inspection

Drawings 6I721-2080-27 Visual Annunciator and Sequence Recorder Alarm P

Schematic

6I721-2155-21 Schematic Diagram Reactor Protection System Testability L

Modification

6I721-2714-20 EDG Automatic Digital Load Sequencing System Manual D

Test Diagram

6I721-2714-22 Electrical Schematic EDG Automatic Digital Load F

Sequencing System H11P898A

6I721-2714-23 EDG Automatic Digital Load Sequencer System 12/18/1985

6I721-2714-24 EDG Automatic Digital Load Sequencing System 1

6I721-2714-25 Electrical Schematic E.D.G Automatic Digital Load 1

Sequencing System

6I721-2714-6 Automatic Sequencing Cabinet 08/07/1984

Engineering TE-E21-22-003 2KV Nova Inverters Failed Bench Tests 0

Inspection Type Designation Description or Title Revision or

Procedure Date

Evaluations TE-R31-22-029 The Class IE Vital Power Distribution (VPD) System Nova 0

Inverter R31K002 Frequency Oscillation

Miscellaneous Operator Logs 05/16/2022

Procedures MMA11 Instrument Testing 27

71111.18 Calculations DC-0703 Volume Pipe Supports for Piping Isometric M-3184-1 and M-3184- 0

No. III DCD 2 2

DC-0703 Volume Pipe Supports for Piping Isometric M-3184-1 and M-3184- 0

No. IV DCD 1 2

DC-0704 Volume Pipe Supports for Piping Isometric M-3185-2 0

No. III DCD 1

DC-0704 Volume Pipe Supports for Piping Isometric M-3185-2 0

No. IV DCD 1

DC-0780 Volume Piping Hanger Calculation per Drawing M-N-2178-2 0

No. I DCD 1

DC-0781 Volume Hanger Calculations for M-N-2179-2 0

No. I DCD 1

DC-0785 Volume Hanger Calculations for M-N-2183-2 0

No. IA DCD 1

DC-0786 Volume Piping Hanger Calculations 0

No. IA DCD 1

DC-2586 Volume Pipe Supports for Piping Isometric M-3359-1 0

No. I DCD 1

DC-2586 Volume Pipe Supports for Piping Isometric M-3359-1 0

No. II DCD 1

DC-2922 Volume RHR Complex Piping Stress Report SX-08 0

No. IA DCD 1

DC-2923 Volume RHR Complex Piping Stress Report SX-09 0

No. IA DCD 1

DC-2927 Volume RHR Complex Piping Stress Report SX-13 0

No. IA DCD I

DC-2928 Volume RHR Complex Piping Stress Report 0

No. IA DCD 1

DC-2956 Volume Pipe Stress Analysis of EESW Supply Line Division 2 to 0

No. lA DCD 1 Heat Exchangers P4400B001D for M-3352-1 and M4630-

Inspection Type Designation Description or Title Revision or

Procedure Date

DC-2957 Volume Pipe Stress Analysis of EESW Return Line Division 11 (M- 0

No. I DCD 1 3353-1, M-4631-1 and M-4657-1)

DC-2958 Volume Pipe Stress Analysis of EESW Supply Line Division 1 to 0

No. IA DCD 1 Plate Heat Exchangers P4400BOO1A and P4400B001C

DC-2959 Volume Pipe Stress Analysis of EESW Return Line Division 1 from 0

No. I DCD 1 Plate Heat Exchangers P4400B00IA and P4400B001C

DC-2965 Volume Piping Stress Report for RHR-01 & 06 0

No. IA DCD 2

DC-2966 Volume Piping Stress Report RHR 03/19 0

No. IA DCD 2

DC-6766 Volume Pipe Stress for Division 1 EESW and RHRSW Supply and 0

No. I DCD 1 Return Lines

DC-6771 Volume Division 1 RHR Complex and Reactor Building RHRSW 0

No. I and EESW Penetrations Evaluations

SS-0026 Volume Reactor/Auxiliary Building-Final Load Verification for 0

No. II DCD 6 Concrete Walls

Corrective Action 22-24369 NRC Identified: EESW/RHRSW Design Specification 03/29/2022

Documents 3071-517 does Not Acknowledge the use of Later ASME

Resulting from Section III Codes used in Piping and Support Calculations

Inspection for the Systems

2-26894 NRC Identified: Typo Error on Reference Document 06/07/2022

Number in Calculations

2-27033 NRC Identified: Evaluation of Potential Rattle Space 06/10/2022

Violation

2-27114 NRC Identified: Wrong Calculation Version Submitted to 06/14/2022

ARMS

2-27182 NQA - NRC Identified: Documentation Error on 06/16/2022

Nondestructive Examination (NDE) Reports

2-28927 NRC Identified, Legacy Pipe Support Calculation Used an 08/22/2022

Improper Basis for Allowable Stress

Engineering 80028 Division 1 Tie-in Buried Pipe Replacement for RHRSW C

Changes and EESW Systems

Miscellaneous ASME RHR Service Water Buried Piping Replacement EDP 1

Inspection Type Designation Description or Title Revision or

Procedure Date

Repair/Replacement 80026, 80027, 80028. 80029

Program 19-002

ASME Emergency Equipment Service Water Buried Piping 2

Repair/Replacement Replacement EDP 80026, 80027, 80028, 80029

Program 19-003

ASME RHR Service Water Buried Piping Replacement EDP 2

Repair/Replacement 80028

Program 21-041

ASME EESW Division 1 Buried Piping Replacement EDP 80028 2

Repair/Replacement

Program 21-044

Design Specification The Detroit Edison Company Design Specification for E

3071-517 RHR Complex Fermi 2

NDE Reports 21-QCR-0052 MT Welds on Replacement Pipe for RHRSW 05/26/2021

21-QCR-0053 MT Welds on Replacement Pipe for RHRSW 06/01/2021

21-QCR-0054 MT Weld on Replacement Pipe for RHRSW 06/10/2021

21-QCR-0071 NDE of RHRSW Replacement Pipe 06/16/2021

21-QCR-0072 MT of RHRSW Replacement Buried Pipe 06/22/2021

21-QCR-0076 MT of RHRSW Replacement Buried Pipe 06/23/2021

21-QCR-0078 MT of RHRSW Replacement Buried Pipe 06/29/2021

21-QCR-0083 MT of RHRSW Replacement Buried Pipe 07/07/2021

21-QCR-0084 MT of RHRSW Replacement Buried Pipe 07/07/2021

21-QCR-0089 MT Final Weld for 14" Supply Piping 06/30/2021

21-QCR-0090 28" RHRSW & EESW Return Piping Division 1 07/14/2021

21-QCR-0100 MT of Piping Iso EESW Supply to EECW Heat Exchanger 07/15/2021

(Hx). Division1 Yard

21-QCR-0102 28" RHRSW & EESW Return Piping Division 1 07/15/2021

21-QCR-0125 MT of RHRSW Replacement Buried Pipe 10/04/2021

21-QCR-0143 MT of 14" EESW Supply to EECW Division 1 09/29/2021

21-QCR-0144 24" RHRSW Supply to RHR Division 1 09/29/2021

21-QCR-0147 28" RHRSW & EESW Return Piping Division 1 10/05/2021

21-QCR-0159 MT of RHRSW Replacement Buried Pipe 10/12/2021

21-QCR-0166 MT on 24" RHRSW Replacement Piping 10/14/2021

Inspection Type Designation Description or Title Revision or

Procedure Date

21-QCR-0169 MT of RHRSW Replacement Buried Pipe 10/21/2021

21-QCR-0174 MT on 28" RHRSW Replacement Piping 10/19/2021

2-QCR-0317 MT of FW-E11-4647-09 02/26/2022

2-QCR-0384 MT on 28" Excavation on Repair Weld 03/11/2022

2-QCR-0389 Final MT on 28" Repair Weld 03/13/2022

2-QCR-0410 MT Excavation of Weld for RHRSW Replacement Pipe 03/08/2022

Work Orders 52496756 ECN- 80026 / 28- Division 1 - Phase 4-RF21- RHRSW - 03/21/2022

Buried Pipe- PMT Hydro (Final Whole System)

2497382 ECN - 80026 / 28 - Division 1 - Phase 4 - RF 21 - EESW - 03/21/2022

Buried Pipe - PMT Hydro (Final Whole System)

71111.19 Corrective Action 22-25696 DWSIL Curve in IPCS Uses Wrong Drywell Pressure Input 04/26/2022

Documents 22-26351 3D56 Testability Logic Channel A/B RPS/Power Failure - 05/16/2022

B21N617B Gross Failure Downscale

2-26412 3D56 Testability Logic Channel A/B RPS/Power Failure in 05/17/2022

Alarm

2-26690 SCR Requested for IPCS DWSIL Drywell Pressure 05/31/2022

Calculation

2-28608 IPCS Incorrect Input to DWSIL Ineffectively 08/10/2022

Communicated to Operations and Training Personnel

2-28837 DWSIL in Simulator Functions as Designed, Not as 08/18/2022

Indicated by 22-25696

Corrective Action 22-27708 NRC Question - CARD 22-26351 Past Operability 07/05/2022

Documents

Resulting from

Inspection

Drawings 6I721-2155-21 Schematic Diagram Reactor Protection System Testability L

Modification

Engineering Operator Logs 05/16/2022

Evaluations

Procedures 400-23842-F03-28 Safety Parameter Display System [Confidential] 2

MMA11 Instrument Testing 27

71111.20 Corrective Action 22-21098 Potential FMS Gap in Hours Tracked in Fatigue 02/04/2022

Documents Management

2-27315 Fatigue Management - Fatigue Assessment Not 06/21/2022

Inspection Type Designation Description or Title Revision or

Procedure Date

Performed as Required for Post Even - Injured Knee

71114.06 Miscellaneous Drill Scenario for August 16, 2022 Emergency n/a

Preparedness Drill

Procedures EP-101 Classification of Emergencies 44

EP-301-01 Technical Support Center 24

71124.08 Corrective Action 08-27501 Closure of Barnwell Waste Management Facility - Class 11/11/2008

Documents B/C Waste Minimization Strategy

20-32274 Evaluate Excess Radioactive Material Stored Outside the 11/17/2020

Plant Radiologically Controlled Area (RCA)

Engineering Scaling Factor 10CFR61 Analysis for Multiple Waste Streams 01/12/2022

Evaluations Report and Waste

Stream Sample

Results from GEL

Laboratories, LLC,

January 12, 2022

Miscellaneous Work Order Performance of Source Leak Testing 01/04/2021

54316238

Procedures MMM06 Material Receipt, Identification, and Status 17

MMM10 Radioactive Material Procurement and Accountability 14

MRP21 Radwaste Shipping Operations 24

MRP24 Fermi-2 10CFR61 Compliance Manual 7A

MRP26 Process Control Program 4B

Self- NPRP-22-0044 Quick Hit Self-Assessment: Radioactive Solid Waste 05/24/2022

Assessments Processing and Radioactive Material, Handling, Storage,

and Transportation

Quick Hit Self- 10CFR37 Self-Assessment 09/23/2021

Assessment Report

- Part 37 Security

Plan

Shipping Records EF2-22-018 Radioactive Waste Shipment of Dry Active Waste in a 03/11/2022

General Design Package

EF2-21-006 Radioactive Waste Shipment, EF2-21-006, of Waste Class 06/16/2021

'A' Bead Resin in a Type 'A' Package

EF2-22-034 Radioactive Waste Shipment Documents for Bead Resin 04/29/2022

Inspection Type Designation Description or Title Revision or

Procedure Date

in a General Design Package

EF2-22-047 Bead Resin Waste Shipment: UN2916 Shipment in a Type 05/26/2022

'B' Package

71151 Corrective Action 21-27287 23.206 Procedure Enhancement and Clarification 08/18/2021

Documents

Miscellaneous MSPI Fermi 2 MSPI Basis Document 9

Procedures 44.020.231 NSSSS - RCIC Steam Line Flow, Trip System 'A' 41

Functional Test

71152A Corrective Action 19-20003 SOP 23.107 Correction Required 01/01/2019

Documents 20-29211 Safety of Personnel Barring the Main Turbine 08/11/2020

21-21815 RIN Superintendent Continues to Disregard Traits of a 02/25/2021

Healthy Nuclear Safety Culture

2-23605 Union Safety Concern - MTG Barring Device Needs 03/14/2022

Improvement

2-25476 Operations Management Committed to Production 04/21/2022

2-25798 Illusion of Safety 04/28/2022

2-25884 Reactor Engineering Needs to Know if Scram Time 05/01/2022

Testing is Required on Various Control Rods

2-26246 Reactivity Management SRO Not Stationed 05/12/2022

2-27456 Ground Fault on Y-Phase on Output from Main Unit 06/27/2022

Transformer to CM and CF Output Breakers

2-27473 Procedure Revision for Mayfly Infestation Preparation Plan 06/27/2022

27.322

2-27499 Add Circuits to 27.322 Mayfly Infestation 06/28/2022

2-27633 NSRG 22-01-15; Safety Oversight Subcommittee Action - 07/01/2022

Develop Comprehensive Strategy to Address Potential

SCWE Issue

Corrective Action 22-26236 NRC Identified - Question on Use of WO to Document 05/11/2022

Documents Scram Time Testing

Resulting from 22-28815 NRC Identified - Observations Related to Incomplete 08/17/2022

Inspection Written CARD Responses to Employee Concerns

Miscellaneous ANSI N18.7-1976 Administrative Controls and Quality Assurance for the 1976

Operational Phase of Nuclear Power Plants

Event Notification Main Turbine Trip 06/24/2022

Inspection Type Designation Description or Title Revision or

Procedure Date

2-003

Memorandum to First Trimester 2022 Nuclear Safety Culture Report 06/28/2022

Pete Dietrich from

Eric Olson

Organizational Turbine Trip Resulting in Reactor Scram 06/24/2022

Effectiveness Cause

Evaluation CARD

20-27403

Organizational Loss of 345kv Due to Mayfly Infestation 07/02/2020

Effectiveness Cause

Evaluation CARD

20-27545

Root Cause 22-21157, Reactor SCRAM on Loss of Feed 0

Evaluation Report

SS-OP-202-22013 RF-21 Shutdown JITT 0

Procedures 22.000.03 Power Operation 25% to 100% to 25% 107

23.107 Reactor Feedwater and Condensate Systems 71, 89, 90,

155

27.322 Mayfly Infestation Preparation Plan 17, 23

MOP19 Reactivity Management 27

26