IR 05000341/2018003
| ML18330A165 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 11/26/2018 |
| From: | Billy Dickson NRC/RGN-III/DRP/B2 |
| To: | Polson K DTE Energy |
| References | |
| IR 2018003 | |
| Download: ML18330A165 (26) | |
Text
UNITED STATES ber 26, 2018
SUBJECT:
ERRATAFERMI POWER PLANT, UNIT 2FINAL SIGNIFICANCE DETERMINATION OF A GREEN FINDING AND NRC INTEGRATED INSPECTION REPORT 05000341/2018003
Dear Mr. Polson:
The U.S. Nuclear Regulatory Commission (NRC) identified an administrative error in the NRC Integrated Inspection Report 05000341/2018003 (ADAMS Accession Number ML18313A189),
dated November 8, 2018. Specifically, a new tracking number was created to track the Green finding associated with the residual heat removal service water valve issue while a tracking number already existed. As a result, the NRC has reissued the report in its entirety with this Section corrected.
The U.S. Nuclear Regulatory Commission (NRC) completed its final significance determination of the apparent violation discussed in NRC Inspection Report 05000341/2018002. This finding involved the licensees failure to identify a condition adverse to quality on the Division 2 residual heat removal service water (RHRSW) outlet flow control valve. Specifically, troubleshooting and the associated post maintenance testing failed to identify and correct a failed anti-rotation key, which resulted in an inoperable Division 2 RHRSW system for longer than its Technical Specification 3.7.1 allowed outage time. The NRC has determined that the final significance of this finding to be Green. On October 23, 2018, the NRC discussed the final significance determination for the apparent violation with Mr. M. Caragher and other members of your staff.
The details of the issue is discussed in the enclosed inspection report.
On September 30, 2018, the NRC completed an integrated inspection at your Fermi Power Plant, Unit 2. On October 4, 2018, the NRC inspectors discussed the results of the integrated inspection with Mr. M. Caragher and other members of your staff. The results of this inspection are also documented in the enclosed report.
Based on the results of this inspection, the NRC has identified two additional issues that were evaluated under the risk significance determination process as having very low safety significance (Green). The NRC has also determined that two violations are associated with these issues.
Because you initiated condition reports to address these issues, these violations are being treated as Non-Cited Violations (NCVs), consistent with Section 2.3.2 of the Enforcement Policy. If you contest the violations or significance of these NCVs, 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 the Fermi Power Plant.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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.
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 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Billy Dickson, Chief Branch 2 Division of Reactor Projects Docket Nos. 50-341 License Nos. NPF-43 Enclosure:
Inspection Report 05000341/2018003 cc: Distribution via LISTSERV
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring licensees performance by conducting an integrated quarterly 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. Findings and violations being considered in the NRCs assessment are summarized in the table below.
List of Findings and Violations Failure to Apply Torque Values Described in Maintenance Procedure for Flexible Couplings on Emergency Diesel Generator 12 Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green P.3 - Resolution 71111.12 Systems NCV 05000341/2018003-01 Closed A finding of very low safety significance with an associated non-cited violation of Technical Specification 5.4.1.a was self-revealed when plant operators discovered a pencil-thick lube oil leak coming from a flexible coupling on emergency diesel generator 12 during planned surveillance testing. Specifically, a lube oil leak developed when the flexible coupling located between the engine driven lube oil pump and the lube oil filter failed due to improper torque applied to the coupling.
Failure to Ensure Electrolytic Capacitors Installed in the Plant Did Not Have Expired Shelf Lives Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green H.7 - 71153 NCV 05000341/2018003-02 Documentation Closed A finding of very low safety significance with an associated non-cited violation of 10 CFR 50,
Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components was self-revealed when the reactor water cleanup system inlet flow square root converter failed, resulting in a failure of the reactor water cleanup differential flow instrument and loss of automatic isolation function of the reactor water cleanup isolation valves. Specifically, electrolytic capacitors were installed in the reactor water cleanup system logic that had expired shelf lives, resulting in failures of the automatic isolation function of the reactor water cleanup system.
Failure to Identify a Condition Adverse to Quality on Division 2 Residual Heat Removal Service Water Outlet Flow Control Valve Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Systems Green H.11 - Challenge 71153 NCV 05000341/2018002-04 the Unknown Closed A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, and TS 3.7.1 Residual Heat Removal Service Water (RHRSW) System, were self-revealed for the licensees failure to identify a condition adverse to quality on the Division 2 RHRSW outlet flow control valve E1150F068B.
Specifically, troubleshooting and the associated post maintenance testing failed to identify and correct a failed anti-rotation key which resulted in an inoperable Division 2 RHRSW system for longer than its TS 3.7.1 allowed outage time.
Additional Tracking Items Type Issue Number Title Report Status Section LER 05000341/2018-003-00 Inoperability of Reactor Water 71153 Closed Cleanup System Isolation Differential Flow High Function LER 05000341/2018-004-00 Inoperability of Reactor Water 71153 Closed Cleanup System Isolation Differential Flow High Function
TABLE OF CONTENTS
PLANT STATUS
INSPECTION SCOPES
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REACTOR SAFETY
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RADIATION SAFETY
...............................................................................................................
OTHER ACTIVITIES - BASELINE
...........................................................................................
INSPECTION RESULTS
............................................................................................................
EXIT MEETINGS AND DEBRIEFS
............................................................................................ 18
DOCUMENTS REVIEWED
......................................................................................................... 19
PLANT STATUS
Unit 2 began the inspection period at approximately 100 percent rated thermal power. On
July 3, 2018, the unit was reduced to approximately 69 percent rated thermal power to
troubleshoot a failed thyristor associated with the automatic voltage regulator. Rated thermal
power was subsequently reduced to approximately 60 percent to perform associated repairs on
July 4, 2018. The unit was returned to approximately 100 percent rated thermal power on
July 5, 2018. On July 6, 2018, the unit was reduced to approximately 87 percent rated thermal
power for a minor rod pattern adjustment and returned to approximately 100 percent rated
thermal power the same day. On July 14, 2018, the unit was reduced to approximately
percent power to perform a minor rod pattern adjustment and to perform repairs to an
isophase bus duct cooler. On July 15, 2018, the unit was returned to approximately 100 percent
rated thermal power. On July 23, 2018, the unit was reduced to approximately 72 percent rated
thermal power to perform a minor rod pattern adjustment. On July 24, 2018, the unit was
returned to approximately 100 percent rated thermal power. On August 18, 2018, the unit was
reduced to approximately 70 percent rated thermal power to perform a minor rod pattern
adjustment. On August 19, 2018, the unit was returned to approximately 100 percent rated
thermal power. On September 8, 2018, the unit was reduced to approximately 75 percent rated
thermal power to perform a minor rod pattern adjustment. On September 9, 2018, the unit was
returned to approximately 100 percent rated thermal power. On September 20, 2018, the unit
was reduced to approximately 95 percent rated thermal power to troubleshoot elevated moisture
carryover levels. On September 22, 2018, the reactor was shut down for a planned refueling
outage. The unit remained shut down at the end of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in
effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with
their attached revision histories are located on the public website at http://www.nrc.gov/reading-
rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared
complete when the IP requirements most appropriate to the inspection activity were met
consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection
Program - Operations Phase. The inspectors performed plant status activities described in
IMC 2515 Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem
Identification and Resolution. 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
71111.01Adverse Weather Protection
Summer Readiness (1 Sample)
The inspectors evaluated summer readiness of offsite and alternate alternating current (AC)
power systems.
71111.04Equipment Alignment
Partial Walkdown (3 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1) Division 2 non-interruptible air supply system during planned Division 1 non-interruptible
air supply system maintenance during the week ending July 14, 2018;
(2) Division 1 standby gas treatment system during planned Division 2 standby gas
treatment system maintenance during the week ending August 4, 2018; and
(3) Division 2 residual heat removal system after planned maintenance and prior to
establishment for shutdown cooling during the week ending September 29, 2018.
Complete Walkdown (1 Sample)
The inspectors evaluated system configurations during a complete walkdown of emergency
diesel generator systems during the weeks ending September 1, 2018 through
September 15, 2018.
71111.05AQFire Protection Annual/Quarterly
Quarterly Inspection (4 Samples)
The inspectors evaluated fire protection program implementation in the following selected
areas:
(1) Turbine Building First Floor lube oil storage area during the week ending
September 22, 2018;
(2) Turbine Building Basement supplemental cooling chill water chillers during the week
ending September 29, 2018;
(3) Turbine Building Second Floor isophase bus duct area during the week ending
September 29, 2018; and
(4) Radwaste Building Second Floor balance of plant switchgear and uninterruptible
power supply battery rooms during the week ending September 29, 2018.
Annual Inspection (1 Sample)
The inspectors evaluated fire brigade performance on August 22, 2018.
71111.11Licensed Operator Requalification Program and Licensed Operator Performance
Operator Requalification (1 Sample)
The inspectors observed and evaluated a licensed operator graded simulator scenario on
August 18, 2018.
Operator Performance (1 Sample)
The inspectors observed and evaluated operator performance during a reactor recirculation
digital control system module replacement on August 8, 2018; a feedwater digital control
system restart on August 9, 2018; and plant shutdown activities for a planned refueling
outage on September 22, 2018.
71111.12Maintenance Effectiveness
Routine Maintenance Effectiveness (2 Samples)
The inspectors evaluated the effectiveness of routine maintenance activities associated
with the following equipment and/or safety significant functions:
(1) Emergency diesel generator lube oil system following various oil leaks; and
(2) Diesel fire pump.
71111.13Maintenance Risk Assessments and Emergent Work Control (3 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent
work activities:
(1) Planned maintenance during the week ending August 4, 2018, including the Division 2
(2) Planned maintenance during the week of ending August 11, 2018, including emergency
diesel generator 11 safety system outage and feedwater digital control system reset; and
(3) Planned maintenance during the week of ending August 18, 2018, including reactor core
isolation cooling and Division 1 standby gas treatment system.
71111.15Operability Determinations and Functionality Assessments (2 Samples)
The inspectors evaluated the following operability determinations and functionality
assessments:
(1) High diesel fire pump coolant temperature, as documented in CARD 18-23969, during
the week ending July 28, 2018; and
(2) Potential vegetation impact on meteorological tower 10m wind speed and direction as
documented in CARD 18-25518, during the week ending September 22, 2018.
71111.18Plant Modifications (1 Sample)
The inspectors evaluated the following temporary modifications:
(1) Floor plug removal and grating installation on turbine building third floor.
71111.19Post Maintenance Testing (5 Samples)
The inspectors evaluated the following post maintenance tests:
(1) Division 2 standby gas treatment system testing following pre-filter change out, during
the week ending August 4, 2018;
(2) Emergency diesel generator 11 testing following a planned safety system outage
including Flexmaster coupling replacement, during the week ending August 11, 2018;
(3) Reactor core isolation cooling system testing following Division 1 steam header drain pot
to water trap bypass air operated valve (E5150F054) repack, during the week ending
August 25, 2018;
(4) Division 1 standby gas treatment system testing following emergent vortex damper
repair, during the week ending August 18, 2018; and
(5) Division 2 residual heat removal pump B testing following relay replacement, during the
week ending September 29, 2018.
71111.20Refueling and Other Outage Activities (Partial Sample)
The inspectors evaluated refueling outage 19 activities starting September 22, 2018 to
September 30, 2018. The inspectors completed inspection procedure sections 03.01.a
and 03.01.c. This constituted a partial sample.
71111.22Surveillance Testing
The inspectors evaluated the following surveillance tests:
Routine (1 Sample)
(1) Dedicated shutdown panel H21-P623 transfer switch control center isolation test, during
the week ending July 14, 2018.
In-service (3 Samples)
(1) High pressure coolant injection time response and pump operability test at 1025 psi,
during the week ending August 25, 2018;
(2) Division 2 low pressure coolant injection pump and valve operability test, during the
week ending September 1, 2018; and
(3) Division 1 core spray pump and valve test, during the week ending September 15, 2018.
71114.06Drill Evaluation
Emergency Planning Drill (1 Sample)
The inspectors evaluated a graded emergency planning drill on July 24, 2018.
RADIATION SAFETY
71124.03In-Plant Airborne Radioactivity Control and Mitigation
Engineering Controls (1 Sample)
The inspectors evaluated airborne controls and monitoring.
Use of Respiratory Protection Devices (1 Sample)
The inspectors evaluated respiratory protection.
Self-Contained Breathing Apparatus for Emergency Use (1 Sample)
The inspectors evaluated the licensees self-contained breathing apparatus program.
71124.04Occupational Dose Assessment
Source Term Characterization (1 Sample)
The inspectors evaluated the licensees source term characterization.
External Dosimetry (1 Sample)
The inspectors evaluated the licensees external dosimetry program.
Internal Dosimetry (1 Sample)
The inspectors evaluated the licensees internal dosimetry program.
Special Dosimetric Situations (1 Sample)
The inspectors evaluated the licensees performance for special dosimetric situations.
OTHER ACTIVITIES - BASELINE
71151Performance Indicator Verification (3 Samples)
The inspectors verified licensee performance indicators submittals listed below:
(1) MS08: Heat Removal Systems - 1 Sample, July 1, 2017 - June 30, 2018;
(2) MS09: Residual Heat Removal Systems - 1 Sample, July 1, 2017 - June 30, 2018; and
(3) MS10: Cooling Water Support Systems - 1 Sample, July 1, 2017 - June 30, 2018.
71152Problem Identification and Resolution
Annual Follow-Up of Selected Issues (1 Sample)
The inspectors completed a review of the licensees corrective action program focused on
the licensees closure of the NRC findings and violations. The inspectors also performed a
limited assessment of the Safety Conscious Work Environment amongst the licensees Site
Leadership Team (department directors and their direct reports). The inspectors used NRC
Appendix C Inspection Procedure (IP) 93100 Safety-Conscious Work Environment Issue of
Concern Follow-up for guidance in completing this independent assessment.
71153Follow-Up of Events and Notices of Enforcement Discretion
Licensee Event Reports (2 Samples)
The inspectors evaluated the following licensee event reports which can be accessed at
https://lersearch.inl.gov/LERSearchCriteria.aspx:
(1) Licensee Event Report (LER) 05000341/2018-003, Inoperability of Reactor Water
Cleanup System Isolation Differential Flow High Function.
(2) Licensee Event Report (LER) 05000341/2018-004, Inoperability of Reactor Water
cleanup System Isolation Differential Flow High Function.
INSPECTION RESULTS
71111.12Maintenance Effectiveness
Failure to Apply Torque Values Described in Maintenance Procedure for Flexible Couplings
on Emergency Diesel Generator 12
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green P.3 - Resolution 71111.12 -
Systems NCV 05000341/2018003-01 Maintenance
Closed Effectiveness
A finding of very low safety significance with an associated non-cited violation of Technical Specification 5.4.1.a was self-revealed when plant operators discovered a pencil-thick lube oil
leak coming from a flexible coupling on emergency diesel generator 12 during planned
surveillance testing. Specifically, a lube oil leak developed when the flexible coupling located
between the engine driven lube oil pump and the lube oil filter failed due to improper torque
applied to the coupling.
Description:
On April 20, 2018, the licensee was performing a routine slow start surveillance of emergency
diesel generator 12 (EDG-12), when plant operators noted a pencil-thick lube oil leak from
the flexible coupling fastener located between the engine driven lube oil pump and the lube oil
filter with the engine running in idle. Plant operators subsequently shut down the engine,
discontinued the surveillance, and EDG-12 was declared inoperable.
The licensee performed an investigation and found the flexible coupling fastener was torqued
to 120 in/lbs. Maintenance procedure 35.307.008, Emergency Diesel Generator - Engine
General Maintenance, Enclosure X, Revision 44 required a torque value of 240-260 in/lbs for
the size of piping the fastener was on. The coupling was last disturbed in 2011, and the
maintenance procedure at that time did not contain information regarding torque values for
flexible couplings. A similar flexible coupling fastener failed in 2016 due to inadequate work
instructions for torqueing flexible couplings (NCV 05000341/2016004-01, ADAMS Accession
Number ML17030A328), and corrective actions were developed to use the vendor
recommended values that had already been added to the maintenance procedure as
X in 2014. However, the corrective actions did not require all flexible couplings to
be checked to ensure they were appropriately torqued. Opportunities existed for the licensee
to ensure these flexible couplings were properly torqued according to vendor
recommendations, either through scheduled maintenance online or during refueling and
forced outages. Therefore, on April 20, 2018, another flexible coupling that was not checked
as an extent of condition failed due to an under torqued condition.
Corrective Action: The licensee entered this issue into their corrective action program. The
flexible coupling was repaired and torqued to the requirements listed in the maintenance
procedure. The EDG was successfully tested and returned to an operable condition.
Corrective Action Reference: CARD 18-23217
Performance Assessment:
Performance Deficiency: The licensee failed to ensure that vendor recommended
torque specifications prescribed in maintenance procedure 35.307.008 were applied to
flexible coupling fasteners. Specifically, a flexible coupling fastener on EDG-12 failed
during a planned surveillance test due to being torqued to 120 in/lbs, vice the
recommended 240-260 in/lb. This resulted in the EDG to be shut down and declared
Screening: The inspectors determined the performance deficiency was more than minor
because it adversely affected the Equipment Performance attribute of the Mitigating Systems
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences (i.e., core damage).
Specifically, the failure to torque the flexible coupling fastener to prescribed vendor
recommended torque values led to the shutdown and inoperability of EDG-12 during a
planned surveillance test.
Significance: The inspectors assessed the significance of the finding using the IMC 0609,
Appendix A, and answered No to all the Section A questions of Exhibit 2. Therefore, this
performance deficiency screened as Green.
Cross-Cutting Aspect: The finding had a cross-cutting aspect in the Resolution component of
the Problem Identification and Resolution cross-cutting area, which states that the
organization takes effective corrective actions to address issues in a timely manner
commensurate with their safety significance. Specifically, a similar failure on another flexible
coupling occurred in 2016, but the licensee failed to ensure all flexible couplings were
appropriately torqued to the specifications included in the revised maintenance procedure.
[P.3]
Enforcement:
Violation: Technical Specification 5.4.1.a requires, in part, that written procedures be
established, implemented, and maintained covering the applicable procedures recommended
in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements,
Revision 2, February 1978. Regulatory Guide 1.33, Appendix A, Section 9.a, states
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, written procedures were not established and implemented for
maintenance that can affect the performance of safety-related equipment appropriate to the
circumstances since 2011. Specifically, there were no torque requirements specified for
flexible coupling maintenance in 2011. As a result, on April 20, 2018, EDG-12 was shut
down and declared inoperable when a flexible coupling failed during a planned surveillance
test.
Disposition: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
71152Problem Identification and Resolution
Observation 71152Annual Sample Review
On August 24, 2018, the inspectors completed a partial review of the licensees corrective
action program focused on the licensees closure of the NRC findings and violations.
Additionally, the inspectors conducted a limited assessment of the Safety Conscious Work
Environment (SCWE) amongst Fermis Site Leadership Team (SLT). The inspectors used
NRC Appendix C IP 93100 Safety-Conscious Work Environment Issue of Concern Follow-up
for guidance in completing this independent assessment. Specifically, the inspectors
focused on completing the inspection requirements contained in Sections 02.04 and 02.05 of
IP 93100. These sections direct inspectors to first review the licensees SCWE-related
policies, communications, and training materials and then assess SCWE by interviewing
and/or conducting focus groups with selected site personnel.
The inspectors reviewed Fermis Nuclear Safety Management Policy Statement, Revision 6,
Fermi Employee Concerns Program, Revision 4, and General Management Policy
Statement on maintaining a SCWE, Revision 10. The inspectors also reviewed the results of
pulse surveys by the Employee Concern Program coordinator conducted in June and
December 2017 and Gallup surveys (surveys conducted DTE wide) conducted in Spring and
Winter of 2017 and in Spring of 2018. Based on a review of these documents, the inspectors
did not identify any trends that would be indicative of an issue with SCWE or issues of
concern.
The inspectors also reviewed the SLTs documented reviews of products from the Nuclear
Safety Culture Monitoring Panel. Those reports appeared balanced with no indication of an
adverse work environment. The SLTs 2018 review indicated a flat or declining trend in plant
overall performance; that trend was recognized and discussed with the inspectors by a
member of the SL
a review by the Independent Review Group:
Site management acknowledges that the quality and rigor of Fermi causal evaluations needs
to significantly improve with regard to the identification and understanding of the underlying
organizational and behavioral shortfalls that are contributing to station events and
performance gaps.
The inspectors conducted interviews of eleven members of the licensees site leadership
team. Additionally, the inspectors completed a focus group discussion. The focus group was
made up of four supervisors from the Maintenance Department. The inspectors also
observed various work group and plant management meetings.
The inspectors did not identify a pervasive chilled work environment at Fermi. Additionally, no
violations of regulatory requirements were identified during the review of closure and follow-up
actions associated with NRC findings and violations.
71153Follow-Up of Events and Notices of Enforcement Discretion
Failure to Ensure Electrolytic Capacitors Installed in the Plant Did Not Have Expired Shelf
Lives
Cornerstone Significance Cross-Cutting Aspect Report
Section
Barrier Integrity Green H.7 - Documentation 71153 -
NCV 05000341/2018003-02 Follow-up of
Closed Events and
Notices of
Enforcemen
t Discretion
A finding of very low safety significance with an associated non-cited violation of 10 CFR 50,
Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components
was self-revealed when the reactor water cleanup system inlet flow square root converter
failed, resulting in a failure of the reactor water cleanup (RWCU) differential flow instrument
and loss of automatic isolation function of the RWCU isolation valves. Specifically, electrolytic
capacitors were installed in the RWCU system logic that had expired shelf lives, resulting in
failures of the automatic isolation function of the RWCU system.
Description:
On May 27, 2018, and again on May 31, 2018, the capacitors on the square root converter of
the RWCU logic system failed, resulting in a failure of the RWCU differential flow instrument
and loss of automatic isolation function of the RWCU isolation valves. Both failures involved
capacitors that had surpassed their shelf life of 10 years. Licensee procedures contained
shelf life values for electrolytic capacitors. However, the procedures did not direct personnel
to record the date of manufacture codes of capacitors and to track those codes to ensure they
would not be installed in the plant prior to their shelf life expiration.
For the May 27, 2018 failure, the square root converter was installed on December 12, 2017,
and was a commercially dedicated component as it was originally a non-qualified component.
No capacitors were changed out during the upgrade process. Per CARD 18-24204, the
capacitor that failed had a date code from 2006. During the upgrade process under
procedure MMM04, Engineering Evaluation Disposition, the electrolytic capacitors were not
checked to ensure their date codes were still valid for further usage nor was there procedural
direction to do so. Corrective actions for this failure included revising the upgrade process to
consider replacing electrolytic capacitors. The licensee submitted Event Notification 53429 and Licensee Event Report 05000341/2018-003 to report the event as a
condition that could have prevented the fulfillment of a safety function needed to control the
release of radioactive material and as a condition that could have prevented the fulfillment of
a safety function needed to mitigate the consequences of an accident.
For the May 31, 2018 failure, the square root converter electrolytic capacitors were procured
on June 9, 2015. However, the electrolytic capacitors received were stamped with a date
code indicating they were manufactured in 1985. At the time of procurement in 2015,
licensee receipt inspection procedure W-ELCAP, Revision B, did not have a requirement to
verify date of manufacture codes. Revision C of the procedure did have a requirement to
check date code markings. However, that procedure was issued on May 24, 2016, and no
action was created to ensure existing stock met the requirements of the receipt inspection
procedure. Date codes were also not checked prior to installation in the plant to ensure the
electrolytic capacitors were still acceptable for use. The licensee submitted Event Notification 53435 and Licensee Event Report 05000341/2018-004 to report the event as a
condition that could have prevented the fulfillment of a safety function needed to control the
release of radioactive material and as a condition that could have prevented the fulfillment of
a safety function needed to mitigate the consequences of an accident.
Corrective Action: The licensee entered this issue into their corrective action program.
Corrective actions included revising the receipt inspection procedure W-ELCAP to Revision D
to set a clear and distinct acceptance date of manufacture at the time of receipt, inspect and
verify all inventory for capacitors with expired shelf life, and update the upgrade process to
consider shelf life date codes.
Corrective Action Reference: CARD 18-24368
Performance Assessment:
Performance Deficiency: The licensee failed to ensure that capacitors installed in the plant
for safety related applications did not have expired shelf lives. Specifically, capacitors
installed in the square root converter of the RWCU logic failed resulting in a failure of the
RWCU differential flow instrument and loss of automatic isolation function of the RWCU
isolation valves. Date codes were not inspected as part of the process for commercially
dedicating a capacitor by use of procedure MMM04 for the May 27th failure and the
procurement of capacitors by use of W-ELCAP, Revision B for the May 31st failure.
Screening: The inspectors determined the performance deficiency was more than minor
because it adversely affected the SSC and Barrier Performance attribute of the Barrier
Integrity cornerstone objective to provide reasonable assurance that physical design barriers
(fuel cladding, reactor coolant system, and containment) protect the public from radionuclide
releases caused by accidents or events. Specifically, the licensee did not ensure that
capacitors installed in the plant did not have expired shelf lives, which led to a loss of the
automatic isolation function of the RWCU isolation valves.
Significance: The inspectors assessed the significance of the finding using the IMC 0609,
Appendix A, and answered No to all the Section B questions of Exhibit 3. Therefore, this
performance deficiency screened as Green.
Cross-Cutting Aspect: The finding had a cross-cutting aspect in the Documentation
component of the Human Performance cross-cutting area, which states that the organization
creates and maintains complete, accurate and up to-date documentation. Specifically, when
the receipt inspection procedure was updated to include date codes, the licensee did not
evaluate the current stock to ensure no items were procured with already expired shelf lives
and the engineering evaluation procedure did not consider evaluating existing components
when upgrading component classifications. [H.7]
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion VIII, Identification and Control of
Materials, Parts, and Components requires that measures shall be established for the
identification and control of materials, parts, and components, including partially fabricated
assemblies. These measures shall assure that identification of the item is maintained by heat
number, part number, serial number, or other appropriate means, either on the item or on
records traceable to the erection, installation, and use of the item. These identification and
control measures shall be designed to prevent the use of incorrect or defective material,
parts, and components.
Contrary to the above, between May 24, 2016 and May 27, 2018, the licensee did not have
measures to prevent the use of incorrect or defective material, parts, and components.
Specifically, between May 24, 2016, when the receipt inspection procedure was updated to
include date codes, and the failure on May 27, 2018, the licensee did not record or track date
codes to ensure that electrolytic capacitors were not stored and installed in the plant that
exceeded their shelf lives. As a result, capacitors with expired shelf lives were installed in the
RWCU system and failed on May 27, 2018 and again May 31, 2018 resulting in a loss of the
automatic isolation capabilities of primary containment isolation valves.
Disposition: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Identify a Condition Adverse to Quality on Division 2 Residual Heat Removal
Service Water Outlet Flow Control Valve
Cornerstone Significance Cross-Cutting Report Section
Aspect
Mitigating Green H.11 - Challenge 71153 - Follow-Up
Systems NCV 05000341/2018002-04 the Unknown of Events and
Closed Notices of
Enforcement
Discretion
A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, and TS 3.7.1 Residual Heat Removal Service
Water (RHRSW) System, were self-revealed for the licensees failure to identify a condition
adverse to quality on the Division 2 RHRSW outlet flow control valve E1150F068
- B.
Specifically, troubleshooting and the associated post maintenance testing failed to identify
and correct a failed anti-rotation key which resulted in an inoperable Division 2 RHRSW
system for longer than its TS 3.7.1 allowed outage time.
Description:
On May 5, 2017, while securing Division 2 RHRSW from routine biocide treatment of the
Division 2 residual heat removal reservoir, control room operators noted abnormal indications
on the Division 2 RHRSW outlet flow control valve (E1150F068B) while the system was still in
operation. The valve indicated fully closed with approximately 8,500 gallons per minute of
flow. The Division 2 RHRSW system was shut down and declared inoperable. Initial
troubleshooting identified that the limit switches were not actuating when expected during
hand wheel operation of the valve. The limit switch assembly was replaced, limits were reset,
and the post maintenance test was performed without the system operating (static no-flow
condition). The test passed, and the valve was returned to service on May 7, 2017.
On May 22, 2017, while placing Division 2 RHRSW in service for routine biocide treatment of
the Division 2 residual heat removal reservoir, the Division 2 RHRSW outlet flow control valve
failed to fully open as evidenced by the reduced flow and full open valve indication.
Troubleshooting discovered the direct cause was the failure of the anti-rotation bushing stem
key. High vibration caused the tack welds of the key to fail. The licensee also concluded that
previous troubleshooting on the limit switch issue, discovered while the system was in
operation for routine biocide treatment on May 5, 2017, for the Division 2 RHRSW outlet flow
control valve was inadequate and did not identify the failure of the anti-rotation key. As a
result, the Division 2 RHRSW outlet flow control valve was returned to service on
May 7, 2017, and subsequently failed on the next on-demand stroke on May 22, 2017.
Division 1 RHRSW was available throughout the event except on two occasions. On
May 9, 2017, and May 11, 2017, the Division 1 RHRSW was unavailable for mechanical draft
cooling tower nozzle cleaning activities.
The licensee submitted LER 05000341/2017-003-00 to report this event in accordance with
CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS 3.7.1 and 10 CFR 50.73(a)(2)(v)(B)
as a condition that could have prevented the fulfillment of the safety function of structures or
systems that are needed to remove residual heat.
The inspectors concluded that all potential failure mechanisms were not evaluated during
troubleshooting on May 5, 2017 and the anti-rotation key failure was within the licensees
ability to foresee and correct based on the anti-rotation key failure most likely being the cause
of the abnormal indications identified on May 5, 2017. Initial licensee troubleshooting
activities did not identify a true fault mechanism and post maintenance testing was not
performed under dynamic conditions which were present at the time of failure.
Corrective Actions: The failed anti-rotation key and associated tack welds were replaced and
preventative maintenance procedures were updated to perform periodic anti-rotation key and
associated tack weld inspections. Additionally, an evaluation was performed to assess
additional training on the expectations for troubleshooting performance.
Corrective Action References: CARD 17-24236 and 17-24655
Performance Assessment:
Performance Deficiency: The licensees failure to identify a condition adverse to quality on
Division 2 RHRSW outlet flow control valve E1150F068B was a performance deficiency.
Specifically, troubleshooting and the associated post maintenance testing failed to identify
and correct a failed anti-rotation key, which resulted in an inoperable Division 2 RHRSW
system for longer than its TS 3.7.1 allowed outage time.
Screening: The inspectors determined the performance deficiency was more than minor
because it adversely affected the Equipment Performance attribute of the Mitigating Systems
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences (i.e. core damage).
Specifically, Division 2 RHRSW was not fully capable of performing its intended safety
function from May 3, 2017 (last successful run) to May 24, 2017.
Significance: A Senior Reactor Analyst (SRA) completed a detailed risk evaluation using the
Fermi Standardized Plant Analysis Risk (SPAR) Model, version 8.52. The following
assumptions were made to characterize the degraded condition that resulted from the
performance deficiency.
- The Division 2 RHRSW outlet flow control valve (E1150F068B) would fail to fully open
on demand. The valve was not recoverable either from the control room or through
local manual operation due the failure mechanism.
- Approximately 2800 gallons per minute RHRSW flow was observed when the valve
failed to fully open. The licensee performed thermal-hydraulic calculations using the
Modular Accident Analysis Program Version 5.03 to assess containment parameters
and the probabilistic risk assessment (PRA) success criteria for continued core cooling
with degraded containment cooling. The licensee evaluation was documented as
Technical Evaluation TE-E11-17-052, Revision B. The licensee determined that
RHRSW flow in the degraded condition was sufficient to prevent core damage due to
containment failure in sequences where the standby feedwater (SBFW) system or a
combination of other systems such as the high pressure coolant injection (HPCI) and
low pressure core spray (LPCS), reactor core isolation cooling (RCIC) and LPCS, or
RCIC and control rod drive (CRD) systems were successful in providing core cooling.
The NRC reviewed the licensees evaluation and agreed with the revised containment
cooling success criteria when the SBFW system was successful in providing core
cooling but did not fully review or agree with the revised success criteria for the other
combinations of injection systems. The SBFW can provide core cooling at both high
and low reactor coolant system pressures and has a suction source outside of
containment. The SBFW system would not be expected to be impacted by higher
containment temperatures resulting from degraded containment cooling as compared
to systems that have a suction source inside containment, or would otherwise be
affected by harsh environments in the primary or secondary containment.
- The exposure time used in the evaluation was 18 days. On May 7, 2018, the licensee
returned the system to service without identifying and correcting the valve failure. The
valve failure was discovered on May 22, 2018, repaired and returned to service on
May 24, 2018. The exposure period is based on the period from May 7, 2018 through
May 24, 2018.
The SRA modified the SPAR model with the assistance of Idaho National Laboratory (INL) to
model the potential for common cause failure of the RHRSW flow control valves in both
divisions (E1150F068A and E1150F068B). The degraded condition was modeled as a failure
of the E1150F068B to open, which results in the failure of one division of suppression pool
cooling. Using the results of the licensee technical evaluation, the SRA reviewed the
quantitative SPAR model results and removed the core damage sequences involving
successful injection with the SBFW system.
For the exposure period, after removing the sequences with successful SBFW core cooling,
the change in core damage frequency from internal events was less than 1E-7/yr. The
dominant core damage sequence involved a loss of offsite power, successful injection with
the reactor core isolation cooling system and successful depressurization, followed by failure
of containment heat removal, leading to failure of late injection with low pressure systems.
Cross-Cutting Aspect: The finding had a cross-cutting aspect in the Challenge the Unknown
component of the Human Performance cross-cutting area, which states that individuals stop
when faced with uncertain conditions. Risks are evaluated and managed before proceeding.
Specifically, troubleshooting quickly associated an abnormality as an indication issue without
stopping and performing more thorough investigation to determine why the limit switch
settings were not correct upon discovery. Static post maintenance testing following limit
switch replacement was accepted as completion of troubleshooting without challenging and
understanding the true cause. (H.11)
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, states in part, that measures shall
be established to assure that conditions adverse to quality, such as failures, malfunctions,
deficiencies, deviations, defective material and equipment, and non-conformances are
promptly identified and corrected.
Technical Specification 3.7.1 Residual Heat Removal Service Water (RHRSW) System
requires, in part, that Division 1 and Division 2 RHRSW subsystems shall be operable in
Modes 1, 2, and 3. Technical Specification 3.7.1 requires that if one or more required
RHRSW subsystems are inoperable for reasons not associated with the RHRSW pumps for
more than 7 days, action must be taken to place the unit in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and
Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Contrary to the above, from May 5, 2017 to May 24, 2017, a condition adverse to quality was
not promptly identified and corrected. Specifically, an anti-rotation key failure on the
Division 2 RHRSW outlet flow control valve, a component important to safety, was not
identified and corrected.
From May 5, 2017 to May 24, 2017, Division 2 RHRSW subsystem was inoperable and action
was not taken to restore the subsystem to service in seven days or place the unit in Mode 3
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Disposition: Because this violation was of very low safety significance and was entered into
the licensees corrective action program, this violation is being treated as a Non-Cited
Violation, consistent with Section 2.3.2 of the Enforcement Policy.
The disposition of this finding and associated non-cited violation, closes
AV 05000341/2018002-04 as discussed in NRC Inspection Report 05000341/2018002,
(ADAMS Accession Number ML18227A091).
EXIT MEETINGS AND DEBRIEFS
The inspectors confirmed that proprietary information was controlled to protect from public
disclosure. No proprietary information was documented in this report.
- On July 13, 2018, the inspectors presented the radiation protection program inspection
results to Mr.
- K. Polson, Senior Vice President, and other members of the licensee staff.
- On October 04, 2018, the inspectors presented the quarterly integrated inspection
results to Mr.
- M. Caragher, and other members of the licensee staff.
- On October 29, 2018, the inspectors presented the final significance determination
for the apparent violation with Mr. M. Caragher and other members of the licensee
staff.
DOCUMENTS REVIEWED
71111.01Adverse Weather Protection
- CARD 18-25406; Potential Inconsistency in UFSAR Chapter 8 Descriptions of 120 kV Power
Lines; 07/16/18
- CARD 18-24696; Spurious 11D31 (120kV Relaying System Failure) and 11D28 (120kV
Breaker GM Trouble Alarms); 06/14/2018
- Procedure 20.300.Offsite; Loss of Offsite Power; Revision 12A
- Procedure 20.300.SBO; Loss of Offsite and Onsite Power; Revision 25A
- Procedure 20.300.120kV; Loss of 120kV; Revision 18
- 20.300.SBO Bases; Loss of Offsite and Onsite Power Bases; Revision 8
- 20.300.120kV Bases; Loss of 120kV Bases; Revision 3
- Procedure 20.300.345kV; Loss of 345kV; Revision 16
- Procedure 20.300.GRID; Grid Disturbance; Revision 7
- 20.300.345kV Bases; Loss of 345kV Bases; Revision 0
- CARD 18-24164; SOP and DCS Logic is not Compatible; 05/24/2018
71111.04Equipment Alignment
- Drawing 6I721-25491; Standby Gas Treatment System Diagram Instrumentation and Controls;
Revision P
- Drawing 6M721-2015; Station and Control Air; Revision CN
- Drawing 6M721-4615; Interruptible and Non-Interruptible Control Air; Revision AG
- Drawing 6M721-5734; Emergency Diesel Generator System Functional Operating Sketch;
Revision BF
- Drawing 6M721-5737; Standby Gas Treatment System; Revision AA
- Drawing 6M721N-2046; Diesel Generator System Division 1 RHR Complex; Revision AF
- Drawing 6SD721-2500-01; One Line Diagram Plant 416V System Service; Revision BO
- Procedure 23.129; Station and Control Air System; Revision 114
- Procedure 23.205; Residual Heat Removal System; Revision 137
- Procedure 23.307; Emergency Diesel Generator System; Revision 123
- Procedure 23.404; Standby Gas Treatment System; Revision 57
71111.05AQFire Protection Annual/Quarterly
- Fermi Fire Drill Evaluation LP-FP-940-12YX; Motor Generator Sets, Fourth Floor Reactor
Building, El 6596; Revision 0
- MOP10; Fire Brigade; Revision 9
- Procedure 20.000.22; Plant Fires; Revision 46
- Procedure FP-RB-4-17b; Reactor Building Recirculation System Motor Generator Area,
Zone 17, El 6596; Revision 4
71111.11Licensed Operator Requalification Program and Licensed Operator Performance
- Procedure 23.138.01; Reactor Recirculation System; Revision 114
- WO 51147599; Multiple RR DCS Alarms in MCR; 06/28/2018
- WO 51192733; Main Steam Line Flow Failed for Steam Lines A and D on DCS Flat Panel
Display; 07/03/2018
71111.12Maintenance Effectiveness
- CARD 17-00626; Oil Leak from Lube Oil Heater; 12/24/2017
- CARD 17-26959; Lube Oil Leaking from Coupling; 08/19/2017
- CARD 17-29635; Oil Leak; 12/03/2017
- CARD 18-00399; EDG 11 Lube Oil Heater Has an Oil Leak; 04/02/2018
- CARD 18-20928; Fuel Oil Leaks Observed on EDG 14; 02/01/2018
- CARD 18-23085; Oil Leak Near Oil Filter; 04/16/2018
- CARD 18-23217; Solid Stream Pencil Thick Oil Leak on Clamped Flange Outlet of EDG 12
Filter; 04/20/2018
- CARD 18-24354; Oil Leak from Flange Supplying EDG 11 LO Filter; 06/01/2018
- CARD 18-26426; NRC Identified Discrepancy; 08/24/2018
- WO 46468420; Verify Torque on Flexible Coupling on EDG 12 to 90-100 inch - Pounds Per
Action 16-25666-12; 11/04/2016
- WO 50476401; Oil Leak Near Oil Filter; 04/20/2018
71111.13Maintenance Risk Assessments and Emergent Work Control
- CARD 18-25977; Dip Switch out of Position; 08/08/2018
- CARD 18-26024; Indicated Instantaneous CTP Exceeded 3486MWth While Transferring into
and Out of Emergency Bypass on the Feedwater DCS Controllers; 08/09/2018
- CARD 18-26093; RHR Handrail and Panel Damage Due to Fork Truck Operation; 08/13/2018
- Risk Management Plan; Feedwater MFP Module Replacement; 08/06/2018
- Risk Management Plan; MDCT Division 1 Nitrogen Tie In; 08/18/2018
- Risk Management Plan; Perform 44.030.253 Reactor Water Level (Level., 2 and 8) Division 1
Channel C Functional Test; 02/07/2017
- Risk Management Plan; Risk Management Plan for the Performance of 24.321.07 (72CF
Throwover Test); 05/11/2015
- Risk Management Plan; RR Pump B MG Set B Speed Controller Replacement; 08/15/2018
71111.15Operability Determinations and Functionality Assessments
- CARD 04-23268; Potential Tree Removal from Areas Adjacent to 60 Meter Meteorological
Tower; 07/21/2004
- CARD 10-29358; Request for Tree Clearance Near 60 Meter Meteorological Tower;
10/20/2010
- CARD 18-00383; DFP FO Gauge Sticks; 05/26/2018
- CARD 18-23969; High Coolant Temperature During Diesel Fire Pump Run; 05/18/2018
- CARD 18-23990; Failed PMT WO 50530433 DFP; 05/19/2018
- CARD 18-24008; Gauge Reading Pressure with DFP Shut Down; 05/20/2018
- CARD 18-24124; High Iron in DFP Angle Drive Oil Sample; 05/23/2018
- CARD 18-25518; Trees on the South Side of 60m Met Tower Need to be Cleared to 10 times
Distance to Height Ratio, According to RG-1.23; 07/19/2018
- CARD 18-23902; Diesel Fire Pump Failed to Start; 05/16/2018
- TE-D40-18-052; Past Operability Evaluation, Impact of Vegetation on Meteorological Tower;
Revision 0
71111.18Plant Modifications
- CARD 00-14755; Elevated TB Temps, Contingency Action Recommendation; 8/7/00
- CARD 00-17713; Remove Steam Tunnel Plugs to Support TBHVAC Fan Maintenance; 8/7/00
- CARD 18-22345; Guidance to Remove TB3 to Mezzanine Floor Plugs Removed in Rev 65;
3/20/18
- Document Change Request 00-1656; Turbine Building Hearing, Ventilation, and Air
Conditioning; 10/10/00
- Document Change Request 18-0068; Turbine Building Hearing, Ventilation, and Air
Conditioning; 3/14/18
- Drawing 6A721-2004-02; Auxiliary Bldg. & Turbine Hse. Steam Tunnel & Cable Tray Area -
Plans & Sections; Revision G
- Drawing 6A721-2017; Turbine House 3rd Floor Plan, El. 643-6; Revision AF
- Drawing 6A721-2017-01; Turbine Building 3rd Floor Plan, El. 643-6; Revision K
71111.19Post Maintenance Testing
- CARD 18-25955; NQA FME Barrier Located Above EDG-11 Was not Installed as an
Effective Barrier; 08/08/2018
- CARD 18-26135; Division 1 SGTS Exhaust Fan Supply vortex Damper Failed to Operate;
08/14/2018
- CARD 18-26209; E5150F054 Friction Value Lower Than Expected After Valve Repack;
08/17/2018
- CARD 18-27450; When Attempted to Start B RHR Pump for Surveillance, the Pump Tripped
on a Z51 Device (Over Current); 09/28/2018
- Procedure 24.204.06; Division 2 LPCI and Suppression Pool Cooling/Spray Pump and Valve
Operability Test; Revision 78
- Procedure 24.307.45; Emergency Diesel Generator 11 Fast Start Followed by Load Reject;
Revision 17
- Procedure 34.307.001; Emergency Diesel Generators Inspection and Preventative
Maintenance; Revision 79
- WO 46533988; Perform 43.404.001 Division 1 Standby Gas Treatment Filter Performance
Test; 08/14/2018
- WO 46534020; Perform 43.404.001 Division 1 SGTS Charcoal Sample Withdrawal;
11/16/2016
- WO 46585850; Replace SGTS Pre-Filters in T4600D002; 11/29/2016
- WO 46697020; Perform 24-Month PM Tasks Per 34.307.001 on Emergency Diesel
Generator- 11; 12/09/2016
- WO 47213441; Perform 24.206.01 RCIC System Pump Operability and Valve Test at 1000
PSIG; 08/16/2018
- WO 47277010; Perform General PM T4600C003 SGTS Division 1 Exhaust Fan;
03/13/2017
- WO 49050601; Division 1 Standby Gas Treatment System Flow Out of Spec High; 10/30/2017
- WO 49202148; Packing Leak; Repack E5150F054; 11/16/2017
- WO 51443116; 43.404.002 Division 2 SGTS Filter Performance; 07/31/2018
- WO 51525895; Repair / Replace Coupling EDG 11 Standby LO Pump; 08/08/2018
71111.20Refueling and Other Outage Activities
- CARD 18-27161; Received 3D156 Rx Water Level Low During Shutdown; 09/22/2018
- CARD 18-27181; 30 Inch Drywell Hatch Cover O-Ring Dropped Through Hatch into Drywell
Upper Elevation; 09/22/2018
- MOP05-200; RPV Water Inventory Control; Revision 0
- MOP19; Reactivity Management; Revision 26A
- MWC13; Outage Nuclear Safety; Revision 18
- Procedure 22.000.03; Power Operation 25% to 100% to 25%; Revision 103A
- Procedure 22.000.04; Plant Shutdown From 25% Power; Revision 84
- Procedure 22.000.05; Pressure/Temperature monitoring During Heatup and Cooldown;
Revision 50
- Procedure 23.623; Reactor Manual Control System; Revision 72
71111.22Surveillance Testing
- CARD 18-26489; E1100F074 RHRSW to RPV Emergency Line Drain Partially Plugged;
08/28/2018
- CARD 18-26495; E1100F078 Did Not Indicate Open Property During 24.204.06; 08/28/2018
- Drawing 6M721-5708; High Pressure Coolant Injection System; Revision AQ
- Procedure 24.202.01; HPCI Pump and Valve Operability Test at 1025 PSI; Revision 115
- Procedure 24.202.08; HPCI Time Response and Pump Operability Test at 1025 PSI;
Revision 18
- Procedure 24.203.02; Division 1 CSS Pump and Valve Operability, and Automatic Actuation;
Revision 59
- Procedure 24.204.06; Division 2 LPCI and Suppression Pool Cooling/Spray Pump and Valve
Operability Test; Revision 77
- Procedure 24.321.05; Dedicated Shutdown Panel H21-P623 Operability Test EF2 System
Transfer; Revision 30
- Procedure 42.321.14; Dedicated Shutdown Panel H21-P623 Transfer Switch Control Center
Isolation Test; Revision 25
- Procedure 43.000.005 ;Visual Examination Piping and Components (VT-2); Revision 36
- Procedure 43.202.001; HPCI Leakage Monitoring Test; Revision 28
- WO 47379727; Perform 47.208.01 Sec-6.3 RHR Division 2 RHRSW Crosstie Valve
E1150F073; 08/28/2018
- WO 47379768; Perform 24.204.05 Sec-5.4 Division 2 RHR Local Valve Position Indication
Verification; 08/28/2018
71114.06Drill Evaluation
- Blue Team 2018 Drill Package; 07/24/2018
71124.03In-Plant Airborne Radioactivity Control and Mitigation
- Laboratory Report Compressed Air/Gas Quality Testing; 4/16/2018
- 50.59 Screen 16-0220; EDP 37719; 12/13/2016
- CARD 18-21052; NRC Violation Failure to Perform Fit Testing on Self-Contained Breathing
Apparatus Respirators; 02/07/2018
- NIOSH Reverence TN-21331; Letter from NIOSH to Mine Safety Appliances Company;
7/25/2018
- Posi3 USB Test Results for HAWK008, HAWK014, and HAWK030; 7/21/2017
- Radiation Protection Respirator Qualification Report; 7/10/2018
71124.04Occupational Dose Assessment
- NPRP-17-0173; Annual Prospective Internal Dose Evaluation; 12/22/2017
- Pregnancy Declaration Form; 6/4/2018
- DLR/Secondary Dosimetry Comparison Resolution; 07/01/2017 to 12/31/2017
71151Performance Indicator Verification
- Fermi 2 RHR Performance Indicators; July 2017
71153Follow-Up of Events and Notices of Enforcement Discretion
- MMM04; Engineering Evaluation Disposition; Revision 20
- MMM13; Storage Maintenance Program; Revision 10
- MMM04; Engineering Evaluation Disposition; Revision 19
- CARD 18-24380; Steam Leak; 06/01/2018
- Fermi Operator Log
- LERs 2018-003 and 2018-004; Inoperability of Reactor Water Cleanup System Isolation
Differential Flow-High Function; Revision 0
- CARD 18-24368; Degraded Capacitor Caused Malfunction of Square Root Converter in
RWCU System; 06/01/2018
- CARD 18-24204; G33R609 RWCU Pump discharge Flow Gage Appears to Have Failed;
05/27/2018
- EN 53435; RWCU System Isolation Differential Flow - High Function was Declared Inoperable
- CARD 18-24380; Steam Leak; 06/01/2018
- CARD 18-24368; Degraded Capacitor Caused Malfunction of Square Root Converter in
RWCU System; 06/01/2018
- CARD 18-24204; G33R609 RWCU Pump Discharge Flow Gage Appears to Have Failed;
05/27/2018
23