IR 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
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.
November 8, 2022 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, 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:
Report Number:
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
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 Aspect Report Section Barrier Integrity Green NCV 05000341/2022003-01 Open/Closed
[H.12] - Avoid Complacency 71111.12 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 Aspect Report Section Not Applicable Severity Level IV NCV 05000341/2022003-02 Open/Closed Not Applicable 71111.12 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 Aspect Report Section Initiating Events Green NCV 05000341/2022003-04 Open/Closed None (NPP)71152A 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 Safety-Related Piping Not Verified 71111.18 Open
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.
REACTOR SAFETY
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,
RADIATION SAFETY
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:
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 Aspect Report Section Barrier Integrity Green NCV 05000341/2022003-01 Open/Closed
[H.12] - Avoid Complacency 71111.12 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 Aspect Report Section Not Applicable Severity Level IV NCV 05000341/2022003-02 Open/Closed Not Applicable 71111.12 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 (Open)
Seismic Displacement for Safety-Related Piping Not Verified URI 05000341/2022003-03 71111.18
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 Aspect Report Section Initiating Events Green NCV 05000341/2022003-04 Open/Closed None (NPP)71152A 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
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
MWC 16-100
Work Control Conduct Manual Implementing Procedure:
Seasonal Readiness
48703549
Evaluate Request for Shutdown 46304 Revisions 1 & 2 for
Impact to Fermi 2
51689384
20KV Relay House Battery Charger Irregularity
63421855
Oil Seepage from Breaker GK
6417741
Enrico Fermi PP Circuit Breaker GH (120KV): Breaker Trip
Coil Test (ITCTRANSMISSION)
64847930
Replace Tagline Switches S4000P003,4,5,6
Work Orders
255931
20KV Mat Fence Line Enhancements
Control Center A/C Water System Functional Operating
Sketch
S
Drawings
Control Center A/C Air System Functional Operating
Sketch
K
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
4B
Division 2 CCHVAC Standby Verification Checklist
103A
Procedures
23.413 Enclosure A
Control Center HVAC Damper Lineup Normal Mode
Division 2 Dampers
103A
Corrective Action
Documents
2-29295
Tracking AIM to Purchase Training Enhancement Items for
Fire Drills
09/02/2022
FB-RB-2-10a
Reactor Building Emergency Equipment Cooling Water,
North, Zone 10, EL. 613'6"
Fire Plans
FP-AB-5-16d
Auxiliary Building Division 1 Control Center Heating,
Ventilating, and Air Conditioning System Equipment
Inspection
Procedure
Type
Designation
Description or Title
Revision or
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"
FP-RB-2-10b
Reactor Building Emergency Equipment Cooling Water,
South, Zone 10, EL. 613'6"
FP-RB-4-17b
Reactor Building Recirculation System Motor Generator
Area Zone 17, Elevation 659'6"
Turbine Building
Fermi 2 Safety Handbook
Miscellaneous
Fire Brigade Drill
Record Form
LP-FP-940-0933/Fire Drill Main Lube Oil Reservoir Room
TB2
09/01/2022
29.100.01 SH 1A
RPV Control-ATWS
29.100.01 SH 5
Secondary Containment and Rad Release
29.100.01 SH1
RPV Control
Classification of Emergencies
Emergency Notifications
63A
EP-290 Enclosure A
Electronic Notification Process
11/01/2021
EP-290 Enclosure B
Nuclear Plant Event Technical Data Form General
Information Requirements
2/06/2019
EP-290001
Initial Notification Form Review Checklist
10/22/2021
Procedures
EP-290001
Follow-up Notification Form Review Checklist
09/02/2021
17-25134
Division 1 CCHVAC Chiller High Bearing Oil Temperature
06/08/2017
21-21951
Division 1 CCHVAC Chiller Trip
03/03/2021
21-24247
T4100B009 Division 1 CCHVAC Chiller Trip (MCR Alarm
8D5)
05/15/2021
2-27400
T4100B009 Division 1 CCHVAC Chiller Trip (MCR Alarm
8D5)
06/24/2022
2-27461
FO 22-01 Start Up Walkdown: MOV Actuator
Disconnected from Bonnet
06/27/2022
Corrective Action
Documents
2-28723
Low Oil Flow on from New CCHVAC Sleeve Bearings
08/14/2022
Engineering
2-018
B2103F019 Actuator Separation Impact on Appendix J,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Evaluations
Including bypass, and IST Requirements
20.413.01
Control Center HVAC System Failure
47.306.06
MS Drain Valve PCIV Actuator
ARP 8D5
Division 1 Control Room A/C Trouble
Procedures
VMS25-39
Centrifugal Water Chillers
06/23/2010
Work Orders
58375116
Install TTC and Other Test Equipment in Preparation for
As Found Test
03/11/2022
21-31211
9D21 Division 1 EDG Sequencer Trouble
2/20/2021
2-22199
Frequency Oscillations Occurring for Installed Power
Supplies and Contingency Power Supplies Need
Evaluated
2/21/2022
2-22886
Nova Inverters Found OOT for B21K801B and R31K001
03/01/2022
2-26351
3D56 Testability Logic Channel A/B RPS/Power Failure -
B21N617B Gross Failure Downscale
05/16/2022
Corrective Action
Documents
2-26412
3D56 Testability Logic Channel A/B RPS/Power Failure in
Alarm
05/17/2022
Corrective Action
Documents
Resulting from
Inspection
2-27708
NRC Question - CARD 22-26351 Past Operability
07/05/2022
Visual Annunciator and Sequence Recorder Alarm
Schematic
P
Schematic Diagram Reactor Protection System Testability
Modification
L
EDG Automatic Digital Load Sequencing System Manual
Test Diagram
D
Electrical Schematic EDG Automatic Digital Load
Sequencing System H11P898A
F
EDG Automatic Digital Load Sequencer System
2/18/1985
EDG Automatic Digital Load Sequencing System
Electrical Schematic E.D.G Automatic Digital Load
Sequencing System
Drawings
Automatic Sequencing Cabinet
08/07/1984
Engineering
TE-E21-22-003
2KV Nova Inverters Failed Bench Tests
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Evaluations
TE-R31-22-029
The Class IE Vital Power Distribution (VPD) System Nova
Inverter R31K002 Frequency Oscillation
Miscellaneous
Operator Logs
05/16/2022
Procedures
MMA11
Instrument Testing
DC-0703 Volume
No. III DCD 2
Pipe Supports for Piping Isometric M-3184-1 and M-3184-
DC-0703 Volume
No. IV DCD 1
Pipe Supports for Piping Isometric M-3184-1 and M-3184-
DC-0704 Volume
No. III DCD 1
Pipe Supports for Piping Isometric M-3185-2
DC-0704 Volume
No. IV DCD 1
Pipe Supports for Piping Isometric M-3185-2
DC-0780 Volume
No. I DCD 1
Piping Hanger Calculation per Drawing M-N-2178-2
DC-0781 Volume
No. I DCD 1
Hanger Calculations for M-N-2179-2
DC-0785 Volume
Hanger Calculations for M-N-2183-2
DC-0786 Volume
Piping Hanger Calculations
DC-2586 Volume
No. I DCD 1
Pipe Supports for Piping Isometric M-3359-1
DC-2586 Volume
No. II DCD 1
Pipe Supports for Piping Isometric M-3359-1
DC-2922 Volume
RHR Complex Piping Stress Report SX-08
DC-2923 Volume
RHR Complex Piping Stress Report SX-09
DC-2927 Volume
RHR Complex Piping Stress Report SX-13
DC-2928 Volume
RHR Complex Piping Stress Report
Calculations
DC-2956 Volume
No. lA DCD 1
Pipe Stress Analysis of EESW Supply Line Division 2 to
Heat Exchangers P4400B001D for M-3352-1 and M4630-
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
DC-2957 Volume
No. I DCD 1
Pipe Stress Analysis of EESW Return Line Division 11 (M-
3353-1, M-4631-1 and M-4657-1)
DC-2958 Volume
Pipe Stress Analysis of EESW Supply Line Division 1 to
Plate Heat Exchangers P4400BOO1A and P4400B001C
DC-2959 Volume
No. I DCD 1
Pipe Stress Analysis of EESW Return Line Division 1 from
Plate Heat Exchangers P4400B00IA and P4400B001C
DC-2965 Volume
Piping Stress Report for RHR-01 & 06
DC-2966 Volume
Piping Stress Report RHR 03/19
DC-6766 Volume
No. I DCD 1
Pipe Stress for Division 1 EESW and RHRSW Supply and
Return Lines
DC-6771 Volume
No. I
Division 1 RHR Complex and Reactor Building RHRSW
and EESW Penetrations Evaluations
SS-0026 Volume
No. II DCD 6
Reactor/Auxiliary Building-Final Load Verification for
Concrete Walls
2-24369
NRC Identified: EESW/RHRSW Design Specification
3071-517 does Not Acknowledge the use of Later ASME
Section III Codes used in Piping and Support Calculations
for the Systems
03/29/2022
2-26894
NRC Identified: Typo Error on Reference Document
Number in Calculations
06/07/2022
2-27033
NRC Identified: Evaluation of Potential Rattle Space
Violation
06/10/2022
2-27114
NRC Identified: Wrong Calculation Version Submitted to
ARMS
06/14/2022
2-27182
NQA - NRC Identified: Documentation Error on
Nondestructive Examination (NDE) Reports
06/16/2022
Corrective Action
Documents
Resulting from
Inspection
2-28927
NRC Identified, Legacy Pipe Support Calculation Used an
Improper Basis for Allowable Stress
08/22/2022
Engineering
Changes
80028
Division 1 Tie-in Buried Pipe Replacement for RHRSW
and EESW Systems
C
Miscellaneous
RHR Service Water Buried Piping Replacement EDP
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Repair/Replacement
Program 19-002
80026, 80027, 80028. 80029
Repair/Replacement
Program 19-003
Emergency Equipment Service Water Buried Piping
Replacement EDP 80026, 80027, 80028, 80029
Repair/Replacement
Program 21-041
RHR Service Water Buried Piping Replacement EDP
80028
Repair/Replacement
Program 21-044
EESW Division 1 Buried Piping Replacement EDP 80028
Design Specification
3071-517
The Detroit Edison Company Design Specification for
RHR Complex Fermi 2
E
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
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
(Hx). Division1 Yard
07/15/2021
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
NDE Reports
21-QCR-0166
MT on 24" RHRSW Replacement Piping
10/14/2021
Inspection
Procedure
Type
Designation
Description or Title
Revision or
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
2/26/2022
2-QCR-0384
MT on 28" Excavation on Repair Weld
03/11/2022
2-QCR-0389
03/13/2022
2-QCR-0410
MT Excavation of Weld for RHRSW Replacement Pipe
03/08/2022
2496756
ECN-80026 / 28-Division 1 - Phase 4-RF21-RHRSW -
Buried Pipe-PMT Hydro (Final Whole System)
03/21/2022
Work Orders
2497382
ECN - 80026 / 28 - Division 1 - Phase 4 - RF 21 - EESW -
Buried Pipe - PMT Hydro (Final Whole System)
03/21/2022
2-25696
DWSIL Curve in IPCS Uses Wrong Drywell Pressure Input
04/26/2022
2-26351
3D56 Testability Logic Channel A/B RPS/Power Failure -
B21N617B Gross Failure Downscale
05/16/2022
2-26412
3D56 Testability Logic Channel A/B RPS/Power Failure in
Alarm
05/17/2022
2-26690
SCR Requested for IPCS DWSIL Drywell Pressure
Calculation
05/31/2022
2-28608
IPCS Incorrect Input to DWSIL Ineffectively
Communicated to Operations and Training Personnel
08/10/2022
Corrective Action
Documents
2-28837
DWSIL in Simulator Functions as Designed, Not as
Indicated by 22-25696
08/18/2022
Corrective Action
Documents
Resulting from
Inspection
2-27708
NRC Question - CARD 22-26351 Past Operability
07/05/2022
Drawings
Schematic Diagram Reactor Protection System Testability
Modification
L
Engineering
Evaluations
Operator Logs
05/16/2022
400-23842-F03-28
Safety Parameter Display System [Confidential]
Procedures
MMA11
Instrument Testing
2-21098
Potential FMS Gap in Hours Tracked in Fatigue
Management
2/04/2022
Corrective Action
Documents
2-27315
Fatigue Management - Fatigue Assessment Not
06/21/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Performed as Required for Post Even - Injured Knee
Miscellaneous
Drill Scenario for August 16, 2022 Emergency
Preparedness Drill
n/a
Classification of Emergencies
Procedures
08-27501
Closure of Barnwell Waste Management Facility - Class
B/C Waste Minimization Strategy
11/11/2008
Corrective Action
Documents
20-32274
Evaluate Excess Radioactive Material Stored Outside the
Plant Radiologically Controlled Area (RCA)
11/17/2020
Engineering
Evaluations
Scaling Factor
Report and Waste
Stream Sample
Results from GEL
Laboratories, LLC,
January 12, 2022
10CFR61 Analysis for Multiple Waste Streams
01/12/2022
Miscellaneous
Performance of Source Leak Testing
01/04/2021
MMM06
Material Receipt, Identification, and Status
MMM10
Radioactive Material Procurement and Accountability
MRP21
Radwaste Shipping Operations
MRP24
Fermi-2 10CFR61 Compliance Manual
7A
Procedures
MRP26
4B
NPRP-22-0044
Quick Hit Self-Assessment: Radioactive Solid Waste
Processing and Radioactive Material, Handling, Storage,
and Transportation
05/24/2022
Self-
Assessments
Quick Hit Self-
Assessment Report
- Part 37 Security
Plan
10CFR37 Self-Assessment
09/23/2021
EF2-22-018
Radioactive Waste Shipment of Dry Active Waste in a
General Design Package
03/11/2022
EF2-21-006
Radioactive Waste Shipment, EF2-21-006, of Waste Class
'A' Bead Resin in a Type 'A' Package
06/16/2021
Shipping Records
EF2-22-034
Radioactive Waste Shipment Documents for Bead Resin
04/29/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
in a General Design Package
EF2-22-047
Bead Resin Waste Shipment: UN2916 Shipment in a Type
'B' Package
05/26/2022
Corrective Action
Documents
21-27287
23.206 Procedure Enhancement and Clarification
08/18/2021
Miscellaneous
Fermi 2 MSPI Basis Document
71151
Procedures
44.020.231
NSSSS - RCIC Steam Line Flow, Trip System 'A'
Functional Test
19-20003
SOP 23.107 Correction Required
01/01/2019
20-29211
Safety of Personnel Barring the Main Turbine
08/11/2020
21-21815
RIN Superintendent Continues to Disregard Traits of a
Healthy Nuclear Safety Culture
2/25/2021
2-23605
Union Safety Concern - MTG Barring Device Needs
Improvement
03/14/2022
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
Testing is Required on Various Control Rods
05/01/2022
2-26246
Reactivity Management SRO Not Stationed
05/12/2022
2-27456
Ground Fault on Y-Phase on Output from Main Unit
Transformer to CM and CF Output Breakers
06/27/2022
2-27473
Procedure Revision for Mayfly Infestation Preparation Plan
27.322
06/27/2022
2-27499
Add Circuits to 27.322 Mayfly Infestation
06/28/2022
Corrective Action
Documents
2-27633
NSRG 22-01-15; Safety Oversight Subcommittee Action -
Develop Comprehensive Strategy to Address Potential
SCWE Issue
07/01/2022
2-26236
NRC Identified - Question on Use of WO to Document
Scram Time Testing
05/11/2022
Corrective Action
Documents
Resulting from
Inspection
2-28815
NRC Identified - Observations Related to Incomplete
Written CARD Responses to Employee Concerns
08/17/2022
Administrative Controls and Quality Assurance for the
Operational Phase of Nuclear Power Plants
1976
Miscellaneous
Event Notification
Main Turbine Trip
06/24/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2-003
Memorandum to
Pete Dietrich from
Eric Olson
First Trimester 2022 Nuclear Safety Culture Report
06/28/2022
Organizational
Effectiveness Cause
Evaluation CARD
20-27403
Turbine Trip Resulting in Reactor Scram
06/24/2022
Organizational
Effectiveness Cause
Evaluation CARD
20-27545
Loss of 345kv Due to Mayfly Infestation
07/02/2020
Root Cause
Evaluation Report
2-21157, Reactor SCRAM on Loss of Feed
SS-OP-202-22013
RF-21 Shutdown JITT
2.000.03
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
Procedures
MOP19
Reactivity Management
27