IR 05000461/2018002
| ML18215A361 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 08/03/2018 |
| From: | Karla Stoedter NRC/RGN-III/DRP/B1 |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| References | |
| IR 2018002 | |
| Download: ML18215A361 (29) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ust 3, 2018
SUBJECT:
CLINTON POWER STATIONNRC INTEGRATED INSPECTION REPORT 05000461/2018002
Dear Mr. Hanson:
On June 30, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Clinton Power Station. On July 10, 2018, the NRC inspectors discussed the results of this inspection with Mr. T. Stoner and other members of your staff. The results of this inspection are documented in the enclosed report.
Based on the results of this inspection, the NRC has identified two issues that were evaluated under the significance determination process as having very-low safety significance (Green).
The NRC also determined that one violation is associated with these issues. Because the licensee initiated condition reports to address this issue, this violation is being treated as a Non-Cited Violation (NCV), consistent with Section 2.3.2 of the Enforcement Policy. The NCV is described in the subject inspection report. Further, the inspectors documented a Licensee Identified Violation which was determined to be of very low safety significance. The NRC is treating this violation as an NCV consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest the violations or significance of the 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 Clinton Power Station. 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 Clinton Power Station.
This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Karla Stoedter, Chief Branch 1 Division of Reactor Projects Docket No. 50-461 License No. NPF-62 Enclosure:
Inspection Report 05000461/2018002 cc: Distribution via LISTSERV
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated quarterly inspection at Clinton Power Station, Unit 1 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. A licensee-identified non-cited violation is documented in Report Section 71124.05.
List of Findings and Violations Failure to Perform an Operability Determination for Suspected Leakage Past Shutdown Service Water Isolation Valves Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.6] - Human 71111.15 -
Systems FIN 05000461/2018002-01 Performance, Operability Closed Design Margins Evaluations The inspectors identified a Green finding for the failure to perform an operability determination in accordance with Procedure OP-AA-108-115, Operability Determinations (CM-1).
Specifically, the licensee failed to determine and document the operability status of the shutdown service water system and the ultimate heat sink after the discovery of leakage past the 1CC075A and 1CC076A isolation valves.
Failure to Establish Adequate Leak Rate Test Procedures for Shutdown Service Water Isolation Valve Testing Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None 71111.22 -
Systems NCV 05000461/2018002-02 Surveillance Closed Testing The inspectors identified a Green finding and a Non-Cited Violation (NCV) of 10 CFR Part 50,
Appendix B, Criterion XI, Test Control, for the failure to ensure testing of the shutdown service water (SX) isolation valves was performed with procedures which: (1) incorporated the requirements and acceptance limits contained in applicable design documents; and (2) included provisions for assuring that all prerequisites for the given test had been met.
Specifically, the licensee failed to establish leak rate test procedures for SX boundary valves 1CC075A and 1CC076A that included provisions for ensuring the required differential test pressure was met during testing.
Additional Tracking Items Type Issue Number Title Report Status Section LER 05000461/2017-007-02 Manual Reactor SCRAM 71153 Closed Due to Loss of Feedwater Heating
TABLE OF CONTENTS
PLANT STATUS
...........................................................................................................................
INSPECTION SCOPES
................................................................................................................
REACTOR SAFETY
..................................................................................................................
RADIATION SAFETY
................................................................................................................
OTHER ACTIVITIES - BASELINE
............................................................................................
INSPECTION RESULTS
..............................................................................................................
EXIT MEETINGS AND DEBRIEFS
............................................................................................ 18 THIRD PARTY REVIEWS .......................................................................................................... 18
DOCUMENTS REVIEWED
......................................................................................................... 18
PLANT STATUS
The unit operated at full achievable power (approximately 99 percent power) for the majority of
the inspection period with the following exceptions:
On April 10, 2018, the unit commenced End-of-Cycle coast down in preparation for Refueling
Outage (RFO) C1R18.
On April 30, 2018, the licensee shut down the reactor for RFO C1R18. Following completion of
outage activities, operations personnel placed the reactor in Mode 2 and commenced a reactor
startup on May 21, 2018. On May 22, 2018, the reactor entered Mode 1 and operations
personnel synchronized the main generator to the grid. The reactor returned to full power
operation on May 25, 2018.
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)
(1) The inspectors evaluated summer readiness of offsite and onsite alternating
current (AC) power systems.
Impending Severe Weather (1 Sample)
(1) The inspectors evaluated readiness for impending adverse weather conditions for
impending severe thunderstorms on May 2, 2018.
71111.04Equipment Alignment
Partial Walkdown (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1) Residual Heat Removal (RHR) A while in shutdown cooling mode with RHR B out of
service on May 7, 2018;
(2) Division 2 Emergency Diesel Generator (EDG) (after configuration control event) on
May 17, 2018;
(3) RHR B (after Division 2 EDG configuration control event) on May 18, 2018; and
(4) High Pressure Core Spray (HPCS) System (after Division 2 EDG configuration control
event) on June 5, 2018.
71111.05AQFire Protection Annual/Quarterly
Quarterly Inspection (4 Samples)
The inspectors evaluated fire protection program implementation in the following selected
areas:
(1) Fire Zone CB-6, Main Control Room ComplexElevation 800-0 on April 23, 2018;
(2) Fire Zone A-3b, RHR C Pump RoomElevation 707--6 on June 4, 2018;
(3) Fire Zone C-1, Containment DrywellElevations 723-1, 737-0, 755-0, 778-0,
707-0, and 768-0 on May 18, 2018; and
(4) Fire Zone R-1c, Radwaste Building Basement-SouthElevation 702 on June 20, 2018.
71111.07Heat Sink Performance
Heat Sink (1 Sample)
(1) The inspectors evaluated RHR A heat exchanger performance testing Clinton Power
Station (CPS) 2700.20 on May 28, 2018.
71111.08In-service Inspection Activities (1 Sample)
The inspectors assessed the effectiveness of the licensees programs for monitoring
degradation of the reactor coolant system boundary, risk-significant piping system boundaries,
and the containment boundary by reviewing the following activities from April 30, 2018, to
May 4, 2018:
(1) Surface magnetic particle examination of support skirt weld, RPV-SK, ASME Class 1,
Category B-K/B10.10;
(2) Surface magnetic particle examination of welded attachments 1SD270-SP16-WA,
ASME Class 2, Category C-C/C3.20;
(3) Volumetric ultrasonic examination of risk-based weld 1-RH-26-2-3 in the RHR system,
ASME Class 2, Category R-A/R1.13-3;
(4) Volumetric phased array ultrasonic examination of dissimilar metal weld 1-RT-36-6A
per BWRVIP-75-A, SIL-455, and ASME Section XI, Appendix VIII;
(5) Relevant indication reported during the inspection of dissimilar metal weld N1B-W-1
in the recirculation system accepted for continued service per Engineering
Change 619802; and
(6) Pressure boundary welds for ASME Class 1 valve replacements following local leak rate
testing failures per Work Orders 01142657 and 01504438.
71111.11Licensed Operator Requalification Program and Licensed Operator Performance
Operator Requalification (1 Sample)
(1) The inspectors observed and evaluated just-in-time training for reactor startup during
RFO C1R18 on May 14, 2018.
Operator Performance (1 Sample)
(1) The inspectors observed and evaluated the plant shutdown for planned RFO C1R18 on
April 29, 2018.
71111.13Maintenance Risk Assessments and Emergent Work Control (4 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent
work activities:
(1) Planned Green risk due to credited operator action to maintain reactor core isolation
cooling (RCIC) availability during RCIC tank work on April 10, 2018;
(2) Planned Green risk due to emergency reserve auxiliary transformer (ERAT) inoperability
for static volts amperes resistance (VAR) compensator work on June 5, 2018;
(3) Unplanned Yellow risk due to inoperability of HPCS for spurious operation of injection
valve on June 20, 2018; and
(4) Planned Yellow risk due to RCIC inoperability for maintenance on June 25, 2018.
71111.15Operability Determinations and Functionality Assessments (7 Samples)
The inspectors evaluated the following operability determinations and functionality
assessments:
(1) Action Request (AR) 4121864: Unexpected Main Control Room Annunciator 5065-4B
for E12-F042B in Manual Override;
(2) AR 4053236: Unexpected CCW Tank Level Increase During 1RIX-PR004 Calibration;
(3) AR 4140689: EDG 1A Tripped During CPS 9080.01;
(4) AR 4133436: Low Thickness Readings Discovered on A RHR Minimum Flow Piping;
(5) AR 4141038: NRC Has A Question on Completion for Step 8.2.1.5 of 3309.01;
(6) AR 4138010: NRC Identified Review of a Job in Progress; and
(7) AR 4137330: 9861.02D017 Leakage Needs Evaluated, RCIC Steam Supply.
71111.18Plant Modifications (1 Sample)
The inspectors evaluated the following temporary modification:
(1) Engineering Change 624166: Evaluation of the Impact to 1E12F064B During Field Work
to Replace Portions of 1RH19BA and 1RH19 BB (in support of B RHR operability
during RFO C1R18).
71111.19Post Maintenance Testing (9 Samples)
The inspectors evaluated the following post maintenance tests:
(1) Testing of 828 Containment Airlock Interlock after repair on March 30, 2018;
(2) Testing of the Division 2 4160 to 480V transformer replacements on May 9, 2018;
(3) Testing of RHR B after pipe replacement on May 13, 2018;
(4) Testing of HPCS Minimum Flow Valve 1E22-F012 after repair May 16, 2018;
(5) Testing of the reactor coolant system pressure boundary after maintenance on
May 13, 2018;
(6) Testing of the Division 1 SX Room Cooler Fan after repair on May 4, 2018;
(7) Testing of the standby liquid control system after maintenance on May 19, 2018;
(8) Testing of reactor recirculation flow control valve B after maintenance on May 15, 2018;
and
(9) Testing of the Division 1 EDG after the replacement of the reverse power relay on
May 25, 2018.
71111.20Refueling and Other Outage Activities (1 Sample)
(1) The inspectors evaluated RFO C1R18 activities from April 30, 2018 to May 22, 2018.
71111.22Surveillance Testing
The inspectors evaluated the following surveillance tests:
Routine (4 Samples)
(1) CPS 9813.01: Control Rod SCRAM Time Testing on June 5, 2018;
(2) CPS 9080.21: EDG 1A Integrated Test on June 7, 2018;
(3) CPS 9051.01: HPCS Pump & HPCS Water Leg Pump Operability on April 13, 2018;
and
(4) CPS 9053.05: RHR/LPCS Valve Operability on May 4, 2018.
In-service (2 Samples)
(1) CPS 9843.01V015: Leak Rate Testing of RCIC Head Spray on May 7, 2018; and
(2) CPS 9861.09D006: Leakage Test on Valve 1CC075A and 1CC076A on April 23, 2018.
Containment Isolation Valve (1 Sample)
(1) CPS 9843.01D002: Category A Valve Leak Rate Test Via Flowmeter on June 5, 2018.
71114.06Drill Evaluation
Emergency Planning Drill (1 Sample)
(1) The inspectors evaluated a full scale drill of Crew B on April 3, 2018.
RADIATION SAFETY
71124.01Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (1 Sample)
The inspectors evaluated radiological hazards assessments and controls.
Instructions to Workers (1 Sample)
The inspectors evaluated worker instructions.
Contamination and Radioactive Material Control (1 Sample)
The inspectors evaluated contamination and radioactive material controls.
Radiological Hazards Control and Work Coverage (1 Sample)
The inspectors evaluated radiological hazards control and work coverage.
High Radiation Area and Very High Radiation Area Controls (1 Sample)
The inspectors evaluated risk-significant high radiation area and very high radiation area
controls.
Radiation Worker Performance and Radiation Protection Technician Proficiency (1 Sample)
The inspectors evaluated radiation worker performance and radiation protection technician
proficiency.
71124.05Radiation Monitoring Instrumentation
Walk Downs and Observations (1 Sample)
The inspectors evaluated radiation monitoring instrumentation during plant walk downs.
Calibration and Testing Program (1 Sample)
The inspectors evaluated the licensees calibration and testing program.
OTHER ACTIVITIES - BASELINE
71151Performance Indicator Verification (4 Samples)
The inspectors verified licensee performance indicator submittals listed below:
(1) MS06: Emergency AC Power Systems - 1 Sample (04/01/2017-03/31/2018)
(2) MS07: High Pressure Injection Systems - 1 Sample (04/01/2017-03/31/2018)
(3) MS08: Heat Removal Systems - 1 Sample (04/01/2017-03/31/2018)
(4) OR01: Occupational Exposure Control Effectiveness - 1 Sample (04/01/2017-
03/31/2018)
71152Problem Identification and Resolution
Semiannual Trend Review (1 Sample)
The inspectors reviewed the licensees corrective action program (CAP) for trends that might
be indicative of more significant safety issues, CAP implementation issues and issues
regarding the timeliness of operability determinations.
Annual Follow-Up of Selected Issues (2 Samples)
The inspectors reviewed the licensees implementation of its CAP related to the following
issues:
(1) AR 4082490: Reactor SCRAM from Trip of 1AP07EJ; and
(2) AR 4116223: Blown Fuses during 9080.23 8.4 for Fast Transfers.
71153Follow-Up of Events and Notices of Enforcement Discretion
Licensee Event Reports (1 Sample)
The inspectors evaluated the following licensee event reports which can be accessed at
https://lersearch.inl.gov/LERSearchCriteria.aspx:
(1) Licensee Event Report (LER) 05000461/2017-007-02, Manual Reactor SCRAM Due to
Loss of Feedwater Heating. The original LER was closed in NRC Integrated Inspection
Report 05000461/2018001 (ML18108A219).
INSPECTION RESULTS
71111.15Operability Determinations and Functionality Assessments
Failure to Perform Operability Determination for Suspected Leakage Past SX Isolation Valves
Cornerstone Significance Cross-cutting Report
Aspect Section
Mitigating Green [H.6] - Human 71111.15
Systems FIN 05000461/2018002-01 Performance,
Closed Design Margins
The inspectors identified a Green finding for the failure to perform an operability determination
in accordance with Procedure OP-AA-108-115, Operability Determinations (CM-1).
Specifically, the licensee failed to determine and document the operability status of the SX
system and the ultimate heat sink (UHS) after the discovery of suspected leakage past
the 1CC075A and 1CC076A isolation valves.
Description: The safety-related SX system and UHS at Clinton Power Station were designed
to remove heat from equipment following a design basis event (DBE). The SX system
contained isolation valves that separated it from non-safety related piping to ensure the
system, and the UHS, would be able to perform their safety function following a DB
- E.
Calculation IP-M-0563, Determination of Allowable Leak Rates and Loss of UHS Volume
from Shutdown Service Water Boundary Valves, identified valves 1CC075A and 1CC076A
as part of the SX isolation valves whose leakage could contribute to the loss of SX system
cooling capability or loss of volume from the UHS.
During a review of operator logs on April 23, 2018, the inspectors noted the licensee had
manually manipulated valves 1CC075A and 1CC076A in an attempt to obtain better isolation
between the SX system and the component cooling water (CCW) system during routine
maintenance activities. Through discussions with licensee personnel, the inspectors were
informed that valves 1CC075A and 1CC076A were known to leak, and that this condition had
been previously captured in the CAP as AR 4053236, Unexpected CCW Tank Level
Increase During 1RIX-PR004 CAL, dated September 18, 2017. The AR described leakage
past the 1CC075A and/or 1CC076A valves as the suspected cause for an unexpected
increase in CCW tank level during routine maintenance activities. Through a review of the
AR, the inspectors noted that the Operable Basis section of the document had not been
completed. Therefore, within the AR, the operability status of the SX system and the UHS
had not been determined since September 18, 2017, when valves 1CC075A and 1CC076A
were suspected to be leaking by. Nonetheless, the Operators maintained the operational
status of the SX system and the UHS as operable.
Through a review of Procedure OP-AA-108-115, Operability Determinations (CM-1), the
inspectors noted the licensee was required to not only enter degraded conditions into the
CAP, but to also determine and document the operability status of the affected structure,
system, or component (SSC) in accordance with the CA
- P. Specifically, the procedure stated
the following:
- Section 4.1.1 stated, Enter degraded conditions, non-conforming conditions, and the
discovery of an unanalyzed condition into the Corrective Action Program.
- Section 4.1.4, stated, in part, Determine and document the operability status of the
affected SSC in accordance with the CA
- P.
- Section 4.1.5, stated, Immediately determine operability from a detailed examination
of the deficiency. Operability should be determined immediately upon discovery that
an SSC subject to Technical Specifications is in a degraded or nonconforming
condition. The determination should be made without delay and in a controlled
manner using the best available information. The SRO should not postpone the
determination until receiving the results of detailed evaluations. In most cases the
decision can be made immediately and appropriately documented on the IR. In other
cases, the decision shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> even though complete information
may not be available.
The inspectors discussed the failure to determine and document the operability status of the
SX system and the UHS after the discovery of leakage past the 1CC075A and 1CC076A
valves with the licensee. The licensee generated AR 4130127, 1CC075A and 1CC076A
Leaking By, on April 24, 2018, to again capture the valve leakage in the CAP. The initial
operability basis provided in the AR addressed the impact of the valves leakage on the SX
systems cooling capability since September 18, 2017, but failed to address the impact on the
UH
- S. On April 26, 2018, the operability basis in the AR was updated to address the impact of
the leakage past the valves on the UHS. The inspectors questioned the basis for operability
provided in the AR because it did not consider the potential increase in leakage past the
valves due to an increase in differential pressure across the valves following a DBE if the
non-safety related CCW system was lost. The licensee again updated the operability basis
on May 16, 2018, to address the potential increase in leakage past the valves following a
DBE. The inspectors reviewed the operability determination and didnt identify any additional
concerns.
Corrective Actions: The licensees immediate corrective actions included evaluating the
suspected leakage past the 1CC075A and 1CC076A valves to ensure operability of the SX
system and UHS was being maintained.
Corrective Action Reference: Action Request 4130127, 1CC075A and 1CC076A Leaking
By
Performance Assessment:
Performance Deficiency: The inspectors determined the licensees failure to determine
and document the operability status of the SX system and the UHS after the discovery
of suspected leakage past the 1CC075A and 1CC076A valves was contrary to
Procedure OP-AA-108-115 and was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Mitigating Systems Cornerstone attribute of equipment
performance, and it adversely affected the cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent undesirable
consequences. Specifically, the failure to determine the operability impact on the SX system
and the UHS after the discovery of suspected leakage past the 1CC075A and 1CC076A
valves did not ensure the availability, reliability, and capability of the SX system and the UHS
to perform their safety functions following a DBE.
Significance: The inspectors determined the finding affected the Mitigating Systems
Cornerstone and assessed its significance using SDP Appendix A, The Significance
Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening
Questions. The finding screened as having very low safety significance (Green) because the
inspectors were able to answer No to all of the associated screening questions. Specifically,
the licensee evaluated the suspected leakage past the 1CC075A and 1CC076A valves and
determined that operability of the SX system and UHS were maintained because overall SX
boundary valve leakage was still within allowable limits.
Cross-cutting Aspect: The finding had a cross-cutting aspect in the Design Margins
component of the Human Performance cross-cutting area, which states that the licensee:
(1) operates and maintains equipment within design margins; (2) carefully guards margins
and changes them only through a systematic and rigorous process; and (3) places special
attention on maintaining fission product barriers, defense-in-depth, and safety-related
equipment. Specifically, the licensee failed to place special attention on maintaining
safety-related valves 1CC075A and 1CC076A to ensure overall SX system leakage margin
was being carefully guarded and changed only through a systematic and rigorous process.
(H.6)
Enforcement: The inspectors did not identify a violation of regulatory requirements
associated with this finding because the procedure the licensee failed to follow was a
self-imposed standard.
71111.22Surveillance Testing
Failure to Establish Adequate Leak Rate Test Procedures for SX Isolation Valve Testing
Cornerstone Significance Cross-cutting Report Section
Aspect
Mitigating Systems Green None 71111.22
Closed
The inspectors identified a Green finding and a NCV of 10 CFR Part 50, Appendix B,
Criterion XI, Test Control, for the failure to ensure the testing of SX isolation valves was
performed with procedures which: (1) incorporated the requirements and acceptance limits
contained in applicable design documents; and (2) included provisions for assuring that all
prerequisites for the given test had been met. Specifically, the licensee failed to establish
leak rate test procedures for SX boundary valves 1CC075A and 1CC076A that included
provisions for ensuring the required differential test pressure was met during testing.
Description: The safety-related SX system and UHS at Clinton Power Station were designed
to remove heat from equipment following a DBE. The SX system contained isolation valves
that separated it from non-safety related piping to ensure the system, and the UHS, would be
able to perform their safety function following a DBE. Valves 1CC075A and 1CC076A were
part of the SX isolation valves required to close to separate the SX system from the
component cooling water system and to provide SX cooling flow to a fuel pool cooling and
cleanup heat exchanger.
The 1CC075A and 1CC076A valves were considered American Society of Mechanical
Engineers (ASME) Operation and Maintenance (OM) Code Category A valves as described in
IST-CPS-BDOC-V-01, Clinton Inservice Testing Program Bases DocumentComponent
Cooling Water. These Category A valves were required to be leak rate tested in accordance
with the ASME OM Code-2004, Paragraph ISTC-3630, Leakage Rate for Other Than
Containment Isolation Valves. Paragraph ISTC-3630, Subsection (b)(5) required valves
1CC075A and 1CC076A, which were butterfly valves, to be leak rate tested at full maximum
function pressure differential because they did not qualify for reduced pressure testing as
allowed by Subsection (b)(4) of the paragraph.
Through a review of leak rate test procedures CPS 9861.09, Shutdown Service Water
Boundary Valve Leak Rate Testing, and CPS 9861.0D006, Leakage Test on Valve
1CC075A and 1CC076A, the inspectors noted the differential pressure expected following a
DBE was not required to be established during leak rate testing of the valves. The test
procedures, as written, would allow testing of the valves at lower pressures than those
encountered following a DBE. Testing these butterfly valves at a lower differential pressure
than the expected differential pressure following a DBE would provide non-conservative leak
rate results. Therefore, acceptable results obtained during leak rate testing of the valves
would not necessarily support operability of the SX system and the UHS. The inspectors
reviewed the recorded data from the previous performances of this surveillance and did not
identify any invalid tests.
Corrective Actions: The licensees immediate corrective actions included a review of the most
recent leak rate test for valves 1CC075A and 1CC076A, performed on January 19, 2017, to
ensure the required differential test pressure was met during performance of the test. The
licensee also created an action to revise the procedure and incorporate the required test
pressure.
Corrective Action References: Action Request 4130828, NRC Question on Last
Performance of 9861.09D006, and AR 4139396, Enhancement of 9861.09 & Data Sheets
SX Boundary Leak Test
Performance Assessment:
Performance Deficiency: The inspectors determined the licensees failure to establish leak
rate test procedures for SX boundary valves 1CC075A and 1CC076A that included provisions
for ensuring the required differential test pressure was met during testing was contrary to
CFR Part 50, Appendix B, Criterion XI, Test Control, and was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Mitigating Systems Cornerstone attribute of procedure
quality, and it adversely affected the cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent undesirable
consequences. Specifically, the failure to establish leak rate test procedures for the SX
boundary valves that ensured the required differential test pressure was met during testing
did not ensure the capability of the SX system or UHS to respond to an initiating event to
prevent undesirable consequences.
Significance: The inspectors determined the finding affected the Mitigating Systems
Cornerstone and assessed its significance using SDP Appendix A, The Significance
Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening
Questions. The finding screened as having very low safety significance (Green) because the
inspectors were able to answer No to all of the associated screening questions. Specifically,
the licensee demonstrated that the most recent leak rate testing of the SX boundary valves
was performed above the minimum required differential test pressure, and the results of the
testing supported operability of the SX system and the UHS.
Cross-cutting Aspect: No cross cutting aspect was assigned to the finding because the
finding was not reflective of current licensee performance. Specifically, the deficient
procedures discussed above had been in place for greater than two years.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part,
that: (1) a test program be established to assure that all testing required to demonstrate that
structures, systems, and components will perform satisfactorily in service is identified and
performed in accordance with written test procedures which incorporate the requirements and
acceptance limits contained in applicable design documents; and (2) test procedures include
provisions for assuring that all prerequisites for the given test have been met, that adequate
test instrumentation is available and used, and that the test is performed under suitable
environmental conditions.
Contrary to the above, as of April 23, 2018, the licensee failed to assure that: (1) testing
required to demonstrate that structures, systems, and components would perform
satisfactorily in service was identified and performed in accordance with written test
procedures which incorporated the requirements and acceptance limits contained in
applicable design documents; and (2) test procedures included provisions for assuring that all
prerequisites for the given test had been met and that the test was performed under suitable
environmental conditions. Specifically, the test procedures for performing leak rate testing
of safety-related SX boundary valves 1CC075A and 1CC076A to demonstrate they would
perform satisfactorily in service did not include provisions to ensure the required differential
test pressure was met during testing.
Disposition: This violation is being treated as an NCV, consistent with Section 2.3.2 of the
71124.05Radiation Monitoring Instrumentation
Licensee Identified Non-Cited Violation IP 71124.05
This violation of very low safety significance was identified by the licensee and was entered
into the licensees CAP. This violation is being treated as an NCV consistent with
Section 2.3.2 of the Enforcement Policy.
Enforcement:
Violation: Technical Specification 5.4.1, states in part, that written procedures/instructions
shall be established, implemented and maintained covering activities specified within
Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, Section 10, Chemical and
Radiochemical Control Procedures, included: calibration of laboratory equipment,
radiochemical requirements for sampling, validity of calibration techniques, and adequacy of
analysis.
Licensee procedure, CY-AA-130-201, Radiochemistry Quality Control, Revision 5.
Step 4.1.2.3 associated with Liquid Scintillation Counters states, in part, calibrate the system
annually when in service, when initially placed in service, whenever repairs or component
replacement could affect the instrument response and as necessary.
Step 4.3.1.5 states, in part, that Initiate Corrective Actions Process and document corrective
measures if recommended acceptance criteria are exceeded.
Step 4.4.1 states, in part, Perform the following requirements: review control charts for
completeness, adverse trends and biases at the frequencies listed below
Contrary to the above, in 2016 and 2017, the licensee failed to implement the Radiochemical
Quality Control activities required by Procedure CY-AA-130-201, Radiochemistry Quality
Control steps 4.1.2.3, 4.3.1.5 and 4.4.1. Specifically, the radiochemical quality control
practices including calibration of the liquid scintillation detectors, establishment of applicable
quality control charts and associated review, and the evaluation of a disagreement within the
inter-laboratory comparison program were not performed annually as required by Procedure
CY-AA-130-201 in 2016 and 2017.
Significance/Severity Level: The inspectors determined the performance deficiency was more
than minor because it adversely affected the Program & Process attribute of the cornerstone
of Radiation SafetyPublic Radiation Safety objective to ensure adequate protection of
public health and safety from exposure to radioactive materials released into the public
domain as a result of routine civilian nuclear reactor operation. The inspectors assessed the
significance of the finding using SDP Appendix D, Public Radiation Safety Significance
Licensee Identified Non-Cited Violation IP 71124.05
Determination Process and concluded the violation was of very low safety significance
(Green) because the finding was associated with the effluent release program but was not
associated with the failure to implement the effluent program nor did public dose exceed 10 CFR 50 Appendix I or 10 CFR 20.1301(e) criteria.
Corrective Action References: Action Request 4114699, Chemistry Quality Control Self-
Assessment Deficiencies, AR 4108505, Liquid Scintillation Counters Quench Curve
Deficiency, and AR 4107332, Analytical Intra-Lab Failures during Self-Assessment
4085425.
71152Problem Identification and Resolution
71152 -
Observation and Minor Performance DeficiencyCorrective Action Program Semiannual
Implementation and Impact on Timeliness of Operability Determinations Trend
Review
During the inspection quarter, the inspectors reviewed a significant number of licensee CAP
documents to assess the following performance attributes:
- complete, accurate, and timely documentation of the identified problem in the CAP;
- evaluation and timely disposition of operability and reportability issues;
- consideration of extent of condition and cause, generic implications, common cause,
and previous occurrences;
- classification and prioritization of the problems resolution commensurate with the
safety significance; and
- identification of negative trends associated with human or equipment performance that
can potentially impact nuclear safety.
Minor Performance Deficiency: The inspectors determined that issues which could impact the
operability of TS-related equipment were generally entered into the CAP in a timely manner.
However, operability determinations were not always performed within the timeframes
established in Section 4.1 of Procedure OP-AA-108-115, Operability Determinations
(CM-1), because some issue reports were not directly routed to the operating shift crew for
review. The CAP software program used by the licensee included a standard set of questions
which were normally answered by the individual entering the issue into the CAP. Depending
on the answers to the questions, the CAP document routing could automatically bypass the
operating shift crew for review.
Screening: This issue screened as minor because all the questions associated with a minor
issue found in IMC 0612, Appendix B, were answered No. The inspectors did not identify
any instance where the failure to perform a timely operability determination had a significant
consequence on licensed activities. However, the inspectors discussed the vulnerability
between the CAP and the operability determination process with the licensee. The licensee
implemented a standing order to require a shift review by the operating crew of condition
reports not directly routed to the shift. In addition, the licensee is trending the number of
condition reports which are returned by the Station Ownership Committee to the shift for
review to determine whether further actions are warranted.
Enforcement: The inspectors did not identify a violation of regulatory requirements associated
with this minor finding because the procedure the licensee failed to follow was a self-imposed
standard.
Observation and Minor Violation - Failure to Enter Discrepant Test 71152Annual
Findings Into the Corrective Action Program Sample Review
The inspectors reviewed AR 4082490, Reactor SCRAM from Trip of 1AP07EJ. The
inspectors selected this sample for review due to the safety significance of the Division 1
and 2 safety-related transformers, which is the subject of the AR. This review focused on
actions associated with newly installed Divisions 1 and 2 4160V to 480V transformers.
As appropriate, the inspectors verified the following attributes during their review of the
licensee's corrective actions for the above condition report and other related condition reports:
- classification and prioritization of the resolution of the problem commensurate with
safety significance; and
- completion of corrective actions in a timely manner commensurate with the safety
significance of the issue.
The inspectors discussed the corrective actions and associated evaluations with licensee
personnel.
As a result of this review the inspectors identified the following minor violation:
Minor Violation: The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion II,
Quality Assurance Program, for the failure to follow procedures associated with the CA
- P.
Specifically, on May 10, 2018, the licensee identified discrepant results while testing
safety-related transformers 0AP06E2 and 1AP12E2 but failed to enter this issue into the CAP
in accordance with PI-AA-120, Issue Identification and Screening Process, Revision 8,
Step 4.3.4, until prompted by the inspectors. Instead, the licensee evaluated the discrepant
results within the work order and found them to be acceptable. The licensee generated
AR 4137994, Insulation Power Factor Results For 0AP06E & 1AP12E, dated May 15, 2018,
after being challenged by the inspectors regarding the need to enter the discrepant test
results into the CAP.
Screening: This issue screened as minor because all the questions associated with a minor
issue found in IMC 0612, Appendix B, were answered No. The failure to document the
discrepant values in the CAP did not adversely impact the safety-related transformers.
Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion II, constitutes a
minor violation that is not subject to enforcement action in accordance with the NRCs
Observation and Minor Violation - Failure to Identify and Correct a 71152 - Annual
Condition Adverse to Quality Sample Review
The inspectors reviewed AR 4116223, Blown Fuses during CPS 9080.23 8.4 for Fast
Transfers. The inspectors selected this sample for review due to repetitive fuse failures
within the safety-related Division 3 NUS Modules dating back to 2013.
As appropriate, the inspectors verified the following attributes during their review:
- complete and accurate identification of the problem in a timely manner commensurate
with its safety significance and ease of discovery;
- consideration of the extent of condition, generic implications, common cause, and
previous occurrences;
- evaluation and disposition of operability/functionality/reportability issues;
- classification and prioritization of the resolution of the problem commensurate with
safety significance;
- identification of corrective actions, which were appropriately focused to correct the
problem; and
- completion of corrective actions in a timely manner commensurate with the safety
significance of the issue.
Description: While reviewing the historical ARs associated with the NUS fuse failures, the
inspectors discovered licensee information indicating the NUS fuse failures were likely caused
by voltage/current transients within the upstream, safety-related 480V to 120V regulating
transformer. The purpose of the transformer was to regulate voltage and current to the
downstream components including the NUS modules. However, degradation in the
transformers ability to regulate voltage and current levels could create a condition where the
voltage and current levels exceeded the NUS fuse rating causing fuse failure. The licensee
documented the potential transformer degradation issue on September 20, 2013, in
AR 1561455, Division 3, Group 1 Instruments Found De-energized during CPS 9080.23,
Specifically, the licensee stated, The most probable cause of the failure of the NUS modules
was the transient voltage overshoot of the regulating transformer causing the transient
protection varistors on the five NUS modules to actuate, drawing a near fault current until the
individual and line feed fuses blew. Station procedure PI-AA-125, Corrective Action
Program, defined equipment failure as, damage to or degradation of a system, structure or
component that may cause or contribute to the event. Based on the information documented
in AR 1561455, the licensee identified transient voltage overshoots in the 480V to 120V
regulating transformer, which was a degraded condition causing the NUS modules to fail. Per
the licensee definition this would constitute an equipment failure. No further action was taken
to identify and correct the regulating transformer degradation until the transformer failed on
March 18, 2018, impacting multiple pieces of safety-related Division 3 equipment.
Minor Violation: Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires
conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations,
defective material and equipment, and nonconformances are promptly identified and
corrected. Contrary to this requirement, on September 20, 2013, the licensee identified a
failure of the 480V to 120V regulating transformer, which manifested itself as a voltage
overshoot causing the failure of the NUS modules, but failed to take actions to correct the
condition. On March 18, 2018, the regulating transformer subsequently degraded further
causing it to fail in a manner that tripped the upstream breaker and impacted additional pieces
of safety-related Division 3 equipment.
Screening: This issue screened as minor because all the questions associated with a minor
issue found in IMC 0612, Appendix B, were answered No. Specifically, the inspectors
determined that although the transformer failure affected Division 3 equipment, the failure
would not have impacted the Division 3 equipments ability to respond to a DBE or the
capability to shut down the reactor and maintain it in a safe shutdown condition.
Enforcement: The failure to comply with 10 CFR 50, Appendix B, Criterion XVI, constitutes a
minor violation that is not subject to enforcement action in accordance with the NRCs
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 10, 2018, the inspector presented the quarterly integrated inspection results to
Mr.
- T. Stoner, and other members of the licensee staff.
- On May 11, 2018, the inspectors presented the radiation protection program inspection
results to Mr.
- T. Stoner, and other members of the licensee staff.
- On May 4, 2018, the inspector presented the in-service inspection activities results to
Mr.
- T. Krawcyk, and other members of the licensee staff.
THIRD PARTY REVIEWS
The inspectors and branch chief reviewed the most recent Institute of Nuclear Power
Operations reports.
DOCUMENTS REVIEWED
71111.01Adverse Weather Protection
- CPS 3501.01; High Voltage Auxiliary Power System; Revision 28d
- CPS 3505.01C001; Generator Backfeed Checklist; Revision 10a
- CPS 3501.01E001; High Voltage Auxiliary Power System Electrical Lineup; Revision 14
- CPS 3502.01; 480 VAC Distribution; Revision 10a
- CPS 3502.01E001; 480 VAC Distribution Electrical Lineup; Revision 14
- CPS 4200.01; Loss of AC Power; Revision 24c
- CPS 4201.01; Loss of DC Power; Revision 8c
71111.04Equipment Alignment
- CPS 3312.01V002; Residual Heat Removal Instrument Valve Lineup; Revision 9a
- CPS 3312.01E001; Residual Heat Removal Electrical Lineup; Revision 17
- CPS 3312.01V001; Residual Heat Removal Valve Lineup; Revision 17c
- CPS 4303.01P023; Cross-Connecting Div 3 DG To Div 1(2) ECCS Electrical Busses;
Revision 2b
- AR 4138790; Division 2 DG Air Receivers Found Isolated During Rounds
- CPS 3506.01V001; Diesel Generator and Support Systems Valve Lineup; Revision 13A
- CPS 3506.01E001; Diesel Generator and Support Systems Electrical Lineup; Revision 18c
- CPS 3506.01V002; Diesel Generator and Support Systems Instrument Valve Lineup;
Revision 11b
- AR 4141038; NRC Has a Question on Completion for Step 8.2.1.5 of 3309.01
- WO 4788132-01; HPCS Instrument Line Fill and Vent
- CPS 9051.02; HPCS Valve Operability Test; Revision 42a
- CPS 3309.01; High Pressure Core Spray (HPCS); Revision 17a
- CPS 9051.05; HPCS Discharge Header Filled and Flow Path Verification; Revision 28c
- CPS 8801.06C001; H22 Panel Mounted Instrument Valve Operation Checklist; Revision 33c
- CPS 3309.01; High Pressure Core Spray (HPCS); Revision 17b
- OP-AA-108-106; Equipment Return to Service; Revision 5
- OP-AA-108-101; Control of Equipment and System Status; Revision 14
- CPS 9051.02; HPCS Valve Operability Test; Revision 42a
71111.05AQFire Protection Annual/Quarterly
- CPS 1893.04M600; 702 Radwaste: Basement (South) Prefire Plan; Revision 5a
- CPS 1893.04M003; Clinton Power Station Pre-Fire Plan Legend; Revision 1
- CPS 1893.04M001; Prefire Plan Cross Index; Revision 3c
- CPS/USAR 3.8 Radwaste Building; Appendix E
- CPS 1893.04M105; 707 Auxiliary: RHR C Pump Room Prefire Plan; Revision 5
- CPS 1893.04; Fire Fighting; Revision 18d
- CPS 3213.01; Fire Detection and Protection; Revision 30a
- AR 1629870; NRC ID Pre Fire Plan Incorrect Door Designation
- OP-AA-201-008; Pre-Fire Plan Manual; Revision 4
- AR 4071504; 1H13-U743 Has a Zone 1 Trouble
- AR 4102235; MCR Halon Panel 1H13-U706 Zone 1 Trouble Locked In
- AR 4135433; NRC IDD: FP Impairment Form Not Found
- AR 4125276; Door 916 Knob Broken
- CC-AA-201; Plant Barrier Impairment Permit 2018-04-005 (C1R18)
- CC-AA-201; Plant Barrier Impairment Permit 2018-04-006 (C1R18)
- CC-AA-201; Plant Barrier Impairment Permit 2018-04-007 (C1R18)
- CC-AA-201; Plant Barrier Impairment Permit 2018-04-004 (C1R18)
- CPS 3213.01P004; Fire Protection Halon Systems; Revision 0b
- CPS 1893.01; Fire Protection Impairment Reporting; Revision 20f
- OP-MW-201-007; Fire Protection System Impairment Control; Revision 7
- CPS 1893.04M364; 800 Control: Main Control Room Prefire Plan; Revision 3a
- CPS 1893.04; Fire Fighting; Revision 18c
- Drawing M05-1102, Sheet Number 001; P&ID Control Room HVAC (VC); Revision V
- Drawing M05-1102, Sheet Number 2; P&ID Control Room HVAC (VC); Revision J
- Drawing M05-1102, Sheet Number 003; General Arrangement Control Bldg. Main Floor Plan
El. 800-0; Revision N
- Drawing M05-1102, Sheet Number 4; P&ID Control Room HVAC (VC); Revision M
71111.07Heat Sink Performance
- WO 1858468; Perform 1E12B001A HX Performance Test IAW 2700.20
- CPS 2700.20; RHR A Heat Exchanger, 1E12B001A Thermal Performance Test Covered by
71111.08In-service Inspection Activities
- C1R17-APR-25; Ultrasonic Examination Summary Sheet for weld 1-RT-36-6A; 05/22/2017
- C1R17-VEN-002; Ultrasonic Examination for component 1-RH-26-2-3; 05/19/2017
- C1R17-MT-004; Magnetic Particle Examination for component 1SD270-SP16-WA;
05/20/2017
- C1R17-MT-001; Magnetic Particle Examination for component RPV-SK; 05/12/2017
- WPS 1-1-GTSM-PWHT; ASME Welding Procedure Specification Record; Revision 2
- ER-AA-335-018; Visual Examination of ASME IWE Class MC and Metallic Liners of IWL
Class CC Components; Revision 13
- EC 619802; Issue Flaw Evaluation for N1B-W-1 Nozzle of RR System Piping; 05/25/2017
- WO 01142657-01; MM 1B21F002: (K) Replace in Event of LLRT Failure; 05/23/2017
- WO 01504438-01; MM 1B21F001: *OC* Replace in the Event of LLRT Failure; 05/23/2017
71111.11Licensed Operator Requalification Program and Licensed Operator Performance
- CPS 3005.01; Unit Power Changes; Revision 43b
- CPS 3001.01; Preparation for Startup & Approach to Critical; Revision 28a
- CPS 3002.01; Heatup and Pressurization; Revision 32d
- CPS 3004.01; Turbine Startup and Generator Synchronization; Revision 33g
71111.13Maintenance Risk Assessments and Emergent Work Control
- WC-AA-104-F-01; Risk Screening/Mitigation Plan; Revision 0
- CPS 3310.01; Reactor Core Isolation Cooling (RI); Revision 30b
- WO 4738538-01; 9384.01R20 OP ERAT SVC Protective Relays Functional Test
- AR 4144410; Incorrect Communication with NDO Report
- CPS 9384.01; ERAT SVC Protective Relays Functional Test; Revision 7
71111.15Operability Determinations and Functionality Assessments
- Drawing E03-IP662; Internal-External Wiring Diagram NSPS Div. 2 Cabinet IH13-P662;
Revision E
- Drawing E02-IRH99; Schematic Diagram Residual Heat Removal System (RH) Residual Heat
Removal System (NSPS) (IEI2-IO50); Revision P
- Drawing 10-5105-D70; System Power Wiring P630; Revision 2
- Drawing E02-1AN99; Schematic Diagram Annunciator System (AN) NSSS Annunciator
System; Revision H
- Drawing 914E 891; Resid Heat Removal Sys RHR #5; Revision 3
- Drawing E03-IP662; Internal-External Wiring Diagram NSPS Div. 2 Cabinet 1H13-P662;
Revision B
- Drawing E02-IRH99; Schematic Diagram Residual Heat Removal System (RH) Residual Heat
Removal System (NSPS) (IEI1-IO5O); Revision R
- AR 4121864; Unexpected MCR Annun 5065-4B E12-F042B in Manual Override
- AR 4083875; MCR Annunciator 5065-4B E12-F042B in Manual Override
- AR 4140689; DG 1A Tripped During 9080.01
- AR 4141981; Revise DG Operating Procedure to Provide Loading Guidance
- EC 624113; Evaluate RHR A Min Flow Wall Thinning
71111.18Plant Modifications
- EC 624166; Evaluation of the Impact to 1E12F064B During Field Work to Replace Portions of
1RH19BA and 1RH19BB; Revision 0
- Drawing M05-1075, Sheet Number 002; P&ID Residual Heat Removal (RH); Revision AN
- Drawing M05-1075, Sheet Number 003; P&ID Residual Heat Removal (RH); Revision AI
71111.19Post Maintenance Testing
- AR 4126058; Documentation Deficiency Identified
- AR 4121240; Containment 828 Personnel Airlock Interlock Failure
- AR 4135906; Valve Stem Broken for 1E22-F012
- OP-AA-103-105; Limitorque Motor-operated and Chainwheel Operated Valve Operations;
Revision 5
- CPS 8219.01C001; Personnel Airlock Maintenance Checklist; Revision 16
- CPS 8120.37; Valve Packing Installation; Revision 1
- CPS 9080.01; Diesel Generator 1A OperabilityManual and Quick Start Operability;
Revision 55f
- CPS 9861.05D003; HPCS Water Leak Rate Test Data Sheet; Revision 24a
- CPS 3309.01; High Pressure Core Spray; Revision 17b
- CPS 9381.01; MOV Thermal Overload Bypass Verification; Revision 38
- CPS 9861.05; Water Local Leak Rate Testing; Revision 26b
- CPS 9051.02D001; HPCS Valve Operability Data Sheet; Revision 37c
- CPS 9051.01; HPCS Valve Operability Test; Revision 42a
- CPS 9843.02; Operational Pressure of Class 1, 2 and 3 Systems; Revision 44a
- CPS 9843.02D001; Generic Class 1, 2 and 3 Operational Pressure Test Data Sheet;
Revision 44
- CPS 9015.03; SLC, SDV Monthly Valve Verification; Revision 25c
- CPS 9015.01; Standby Liquid Control System Operability; Revision 41e
- CPS 9159.01; Reactor Coolant System Leakage Test; Revision 11c
- WO 4780634; 1VH01CA SX Pump Room A
- WO 4766924-01; Containment 828 Personnel Airlock Interlock Failure
- WO 4766924-02; Containment 828 Personnel Airlock Interlock Failure
- WO 4776677-01; 9080.01 DG 1A Operability
- WO 4782517; Op Leak Check and Stroke 1E22F012
- WO 4780267; PMT: OPS Stroke 1B33-F060B FCV B Verify Operation
- WO 4765528; 9015.01 Op SLC Pump Operability
- WO 47660657; 1C41F004B Replace Parts After Firing per 9015.02
- M05-1074; P&ID High Pressure Core Spray; Sheet 1; Revision AH
- Exelon Nuclear Issue 4126058; Documentation Deficiency Identified; 03/30/2018
- CPS 8219.01; Personnel Airlock Maintenance; Revision 19
- AR 4122636; Unusual Noise Coming From Airlock Mechanism
71111.20Refueling and Other Outage Activities
- OU-CL-104; Shutdown Safety Management Program; Revision 16
- OU-AA-103; Shutdown Safety Management Program; Revision 18
- OP-AA-108-117; Protected Equipment Program; Revision 5
- EC 624219; Evaluation To Use SF Through An FC Heat Exchanger As An Alternate
Suppression Pool Cooling Method; Revision 0
- CPS 3312.02; Alternate Shutdown Cooling (A-SDC) Methods; Revision 9c
- CPS 3006.01; Unit Shutdown; Revision 45e
- CPS 3006.01; Unit Shutdown; Revision 46
- CPS 3006.01F001; Unit Shutdown Flowchart; Revision 0a
- CPS 3005.01F002; Unit Power Changes Power Decrease Flowchart; Revision 0
- OP-AB-300-1003; Reactivity Maneuver
- AR 4127384; Implementation of License Amendment 216
- AR 4141791; Failure to Document 9000.06 Verification Step
- AR 4136006; 9000.06 D003 Reactor Coolant Data Recording
- AR 4136029; Reactor Coolant Temp Limits
- CPS 3312.03C001; Alternate SDC Temperature Monitoring Checklist; Revision 0a
- CPS 3312.03; RHRShutdown Cooling (SDC) & Fuel Pool Cooling and Assist (FPC&A);
Revision 11d
- AR 4134615; Issues Identified by NRC During Plant Walk down
- Drawing M-1MS24007C; Shear Lug Orientation; Revision E
- Drawing M-1MS32007C; Shear Lug Orientation; Revision G
- Drawing M27-1601-04A; Drywell Piping Plan, Floor El. 737-0; Revision J
- AR 1682117; Cable Walk down Inspections On-Line and During C1R14 Outage
- AR 4135978; NRC ID: 1E12F009 Valve Grease or Oil Coming Out of Gear Case
- AR 4134687; Cable on Top of SRV has Sheath Pulled Away from Connector
- AR 1300661; Crack in Penetration Seal on Bioshield Wall
- AR 4135961; NRC ID: Packing Leak 1VP091A
- AR 4007963; PSUDrywell Walk downs-Packing Leak on 1VP091A
- Drawing M05-1109, Sheet Number 002; P&ID Drywell Chilled Water System (VP);
Revision AA
- AR 2441862; Partial Light Protector Missing During C1M21 DW Close-Out
- AR 4133866; Cracked Light Fixture in Drywell
- AR 4134695; NRC ID: Drywell Cooling Cabinet Wire With Sheath Removed
- AR 4134726; NRC Identified Green Residue on CRD Piping in Drywell
- AR 4135315; NRC ID: RR FCV A Full Closed LS Arm Will Not Actuate
- CPS 4001.01; Drywell RE/RF LD Monitoring System Summary; Revision 12b
- Drawing S27-1354; Containment Building Pedestal Base & Weir Wall Reinforcing Plan &
Details; Revision N
- Drawing A27-1000-038; Containment Building Basement Plan Area 3; Revision C
- AR 4134627; NRC ID: Material Stored Inappropriately on 707 Aux Bldg
71111.22Surveillance Testing
- WO 1708032-01; 9080.21R20 OP DG 1A Integrated Test (Except Section 8.5)
- CPS 9080.21C001; DG 1A LOP Pretest Checklist; Revision 27b
- CPS 9080.21C005; ECCS Initiation/DG Start Pretest Checklist; Revision 5b
- CPS 9080.21C002; DG 1A LOP/ECCS Pretest/Post-Test Checklist; Revision 27
- CPS 9080.21E001; DG 1AECCS Integrated Electrical Lineup; Revision 24a
- CPS 3506.01C005; Diesel Generator Start Log; Revision 1b
- CPS 9080.21D001; DG 1AECCS Integrated Data Sheet; Revision 26c
- CPS 3506.01C001; Diesel Generator 1A Pre-Start Checklist; Revision 16a
- CPS 3506.01D001; Diesel Generator 1A Operating Logs; Revision 5b
- CPS 3506.01C005; Diesel Generator Start Log; Revision 1b
- CPS 9052.04; LPCS/RHR A Discharge Header Filled and Flow Path Verification; Revision 28e
- AR 1087427; Enhancement to 9080.21 ST Reference
- CPS 9080.24; DG 1A Test Mode Override, Load Reject Operability, and Idle Speed Override;
Revision 5d
- AR 1569511; 1VP04CCB Reset Light Not Lit on Start Up
- AR 3988866; VP Chiller A Safety Circuit Light Does Not Energize
- AR 4134225; EOID 1PIDG043 DIV I DG Fuel Inlet Pressure HI (12 Cyl)
- AR 4134300; VC A Chill Water Pump Breaker Appears Tripped in MCR
- CPS 9080.21; Diesel Generator 1AECCS Integrated; Revision 34b
- Drawing E02-1RH99, Sheet Number 011; Schematic Diagram Residual Heat Removal
System (RH) Residual Heat Removal System (NSPS) (1E12-1050); Revision H
- Drawing E02-1RH99, Sheet Numbers 006 and 010; Schematic Diagram Residual Heat
Removal System (RH) Residual Heat Removal System (NSPS) (1E12-1050); Revision L
- Drawing M05-1075; Sheet Number 004; P&ID Residual Heat Removal (RH); Revision AG
- Drawing E02-1RH99, Sheet Number 511; Residual Heat Removal Sys (RH) Pump 1A Min.
Flow Vlv 1E12-F064A Heat Exch. 1A Outlet Vlv 1E12-F068A; Revision N
- Drawing E02-1RS99, Sheet Number 111; Schematic Diagram Remote Shutdown System
(RS) Remote Shutdown System (1061-1050); Revision G
- Drawing M05-1075, Sheet Number 001; P&ID Residual Heat Removal (RH); Revision AY
- Drawing M05-1075, Sheet Number 002; P&ID Residual Heat Removal (RH); Revision AN
- Drawing M05-1075, Sheet Number 003; P&ID Residual Heat Removal (RH); Revision AI
- WO 4644807-01; 9843.01 10 LRT CAT A VLV LRT (1E22-F004) HPCS INJ
- WO 4776677-01; 9080.01 DG 1A Oper
- Drawing M05-1074, Sheet Number 001; P&ID High Pressure Core Spray (HP); Revision AH
- CPS 2761.02; Leak Rate Testing Equipment Operation; Revision 5g
- CPS 9843.01F003; ISI Category A Valve Identified Leakage; Revision 23c
- CPS 9843.01; ISI Category A Valve Leak Rate Test; Revision 36b
- CPS 9843.01D001; Category A Valve Leak Rate Test Via Telltale Drain; Revision 30b
- CPS 9843.01D002; Category A Valve Leak Rate Test Via Flowmeter; Revision 25b
- CPS 9843.01V005; Leak Rate Testing of HPCS Injection; Revision 27d
- WO 4738575-01; 9051.01R22 OP HPCS Pump & Wtr Leg Pump Oper (RCIC Strg Tank)
- Drawing M05-1074, Sheet Number 001; P&ID High Pressure Core Spray (HP); Revision AH
- CPS 1401.09F002; CAT A Instrument Failure Checklist; Revision 1
- AR 4126502; CAT A Failure 1E22R502
- WO 4625538-01; Perform Calibration (8801.01)
- CPS 1887.00; Administration of In-Service Inspection (ISI) and In-Service Testing (IST)
Program Activities; Revision 8b
- CPS 9051.01D001; HPCS Pump & HPCS Water Leg Pump Operability Data Sheet;
Revision 48b
- CPS 9813.01; Control Rod Scram Time Testing; Revision 41d
- WO 1929564-01; 9813.01R20 OP Scram 10% Control Rods Per 9813.01
- CPS 8221.01; Control Rod Drive (CRD) Hydraulic Control Unit (HCU) Maintenance;
Revision 19
- CPS 3304.01; Control Rod Hydraulic & Control (RD); Revision 37a
- CPS 3304.01P001; Clearing HCU Trouble Alarms; Revision 1
- WO 4581915-01; 9813.01R20 OP Scram 10% Control Rods Per 9813.01
- WO 4652986-01; 9813.01R20 OP Scram 10% Control Rods Per 9813.01
- CPS 9813.02; Option B Scram Time Analysis; Revision 0
- CPS 3304.05; Determination of Control Rod Drive System Problems; Revision 0c
- AR 2587367; EOID: Temperature Cycling on Rod 28-33
- AR 3943220; EOID: High Temperature on CRDM 28-33, Read on 1C11R018
- AR 4010192; Received Unexpected Annunciator 5006-1G, CRD HYD HI Temp
- AR 4125314; Unexpected MCR Annunciator 5006-1G CRD HYDR Temp HI on 32-37
- AR 4125681; HCU 04-37 Level Switch Failed 9413.02
- CPS 5006.01; Alarm Panel 5006 AnnunciatorsRow 1; Revision 32e
- CPS 9413.02; Scram Accumulator Instrumentation Channel Calibration; Revision 40e
71124.01Radiological Hazard Assessment and Exposure Controls
- AR 02739124; Pre-NRC Inspection, 71124.01 Radiological Hazard Assessment and Exposure
Controls; 04/12/2018
- AR 04020055; RPID: Emerging Trend in Electronic Dosimeter Alarms; 06/09/2017
- AR 04020066; RPID: Emerging Trend in Electronic Dosimeter Dose Rate Alarms; 06/09/2017
- AR 04079020; RPID: 2017 RP Crew Clock Review; 11/29/2017
- AR 04105834; RPD: Individual Received Accumulated Dose Alarm; 02/20/2018
- AR 04126399; Potential Trend: Dose Alarms Received by Maintenance; 04/13/2018
- AR 04133783; RPID: Dose Rate Alarm; 05/02/2018
- RP-AA-800; Attachment 2; Source Leak Test Record; 06/26/2017
- RP-AA-800; Attachment 2; Source Leak Test Record; 01/03/2018
- RP-AA-800; Control, Inventory, and Leak Testing of Radioactive Sources; Revision 7
- NISP-RP-002; Radiation and Contamination Surveys; Revision 0
- NISP-RP-005; Access Controls for High Radiation Areas; Revision 0
- NISP-RP-007; Control of Radioactive Material; Revision 0
- NISP-RP-010; Radiological Job Coverage; Revision 0
- RP-AA-462; Controls for Radiographic Operations; Revision 14
- RP-AA-4011004; Controls for the Draining and Decon of BWP-PWR Reactor Cavity and
Associated Pits; Revision 4
- RP-AA-401-1002; Radiological Risk Management; Revision 10
- RP-AA-460; Controls for High and Locked High Radiation Areas; Revision 32
- RP-AA-460-002; Additional High Radiation Exposure Control; Revision 3
- RP-AA-500 Radioactive Material (RAM) Control; Revision 18
- Radiation Work Permit and Associated ALARA File; RWP CL-0-18-00701; C1R18
Suppression Pool Diving; Revision 1
- Radiation Work Permit and Associated ALARA File; RWP CL-0-18-00901; C1R18 Reactor
Disassembly / Reassembly Activities; Revision 1
- Radiation Work Permit and Associated ALARA File; RWP CL-0-18-00906; C1R18 Cavity
Decon Activities; Revision 1
- Radiation Work Permit and Associated ALARA File; RWP CL-0-18-00401; C1R18
Radiography; Revision 1
- CL-1-18-00401 Radiography Shot Plan for the B RHR Pump Room; Revision 4
71124.05Radiation Monitoring Instrumentation
- 04107332; Analytical Intra-Laboratory Failures during Self-Assessment 4085425; 02/23/2018
- AR 04108505; Liquid Scintillation Counters Quench Curve Deficiency; 02/26/2018
- AR 04114699; Chemistry Quality Control Self-Assessment Deficiencies; 03/14/2018
- AR 04085425; Self-Assessment; Chemistry Laboratory and Radiochemistry Quality Control;
03/16/2018
- AR 04136221; NRC ID: NRC RP Baseline Inspection Area of Concern; 05/09/2018
- AR 02739160; Self-Assessment; Radiation Monitoring Instrumentation; 02/02/2018
- CY-AA-130-201; Radiochemistry Quality Control; Revision 5
- CY-AA-160-100; Analytical Results; Revision 4
- RP-AA-229; Fastscan Abacos Plus Whole Body Counter Calibration; Revision 3
- RP-AA-700; Controls for Radiation Protection Instrumentation; Revision 5
- RP-AA-700-1209; Calibration of Shepherd Box Irradiators; Revision 1
- RP Instrumentation Source Checks; 05/05/2018
- 10 CFR 61 Data; #L74310; 09/19/2017
- Instrumentation Out of Tolerance Reports; Various Documents
- GM Portable Instrument Calibration; Various Records
- DMC-3000 Calibrations; Various Record
- Gamma Spectroscopy Calibration Records; Various Records
- Liquid Scintillation Counter Calibration and Quench Curve Records; 04/18/2018
- Ion Chamber Portable Instrument Calibrations; Various Records
- MGP Ram Gam Calibration; #0012623; 01/19/2018
- MGP Telepole Calibration; #079468; 04/11/2018
- MGP Amp-100 Calibration; #077593; 09/12/2017
- RadeCo H-809 Calibration; #5582; 02/04/2018
- Eberline AMS 4 Calibration; #076670; 10/25/2017
- Argos 5 Calibration; #1011-289; 02/18/2018
- PM 12 Calibration; #117; 09/18/2017
- SAM-12 Calibration; #150; 02/15/2018
- Fastscan Whole Body Counter Calibration; 01/30/2018
- Shepherd Model 89 Calibrator Annual Calibration Check; 01/03/2018
- 1RIX-CM060 Drywell High Range Monitor Calibrations; June 2017 and January 2018
- Radiochemistry Cross Check Program Results; 2016-2018
71151Performance Indicator Verification
- Radiochemistry Cross Check Program Results; 2016-2018
- Monthly Radiation Protection Performance Indicators and Supporting Documentation;
April 2017 through March 2018
- Monthly MSPI Performance Indicators and Supporting Documentation: April 2017 through
March 2018
71152Problem Identification and Resolution
- AR 4123070; NOS ID: Safety and Housekeeping Issues at Lake Screen House; 04/04/2018
- AR 4123273; NOS ID: Water Leak on the SX Inlet Pump Valve 1SX009A; 04/05/2018
- AR 4123431; RAT SVC Procedure Not Revised After Last Replacement; 04/05/2018
- AR 4123456; ERAT SVC Procedure Not Revised After Last Replacement; 04/05/2018
- AR 4123502; ERAT SVC Battery B Needs to be Replaced Before PM is Due; 04/05/2018
- AR 4125189; Leak on the Gasoline Pump; 04/10/2018
- AR 4126125; Unexpected Annunciator 5012-8B Ground 125 Vdc MCC 1F; 04/12/2018
- AR 4126135; Drawing Error on E02-1DC07 Sheet 1; 04/12/2018
- AR 4126351; Trend in Issue Report Not Routed for Shift Review; 04/13/2018
- AR 4128657; Sec ID: Passport EIN Names & Status Issues With Sec. EDS; 04/19/2018
- AR 4128764; NOS ID: 19 RGPP Containers Stored in the Hallway; 04/19/2018
- AR 4129180; APS As Found: OSB Elevator Inop With Persons Onboard; 04/20/2018
- AR 4129353; EOID 0HC04G Service Building Elevator OOS; 04/21/2018
- AR 4129790; Procedure Enhancement 9432.04B and 9432.04C; 04/23/2018
- AR 4130240; No Risk Screening in Work Order; 04/24/2018
- AR 4130283; Not All Ops Crews Met the Requirements of OP-AA-201-003; 04/24/2018
- AR 4130819; Found That the Grease Fitting Damaged; 04/25/2018
- AR 4130881; Tools Found During VC B Chiller Walkdown; 04/25/2018
- OP-AA-108-115; Operability Determinations (CM-1); Revision 20
- OP-AA-108-115-1002; Supplemental Consideration for On-Shift Immediate Operability
Determinations (CM-1); Revision 3
- OP-AA-111-101; Operating Narrative Logs and Records; Revision 13
- CPS Operations Department Daily Orders; 04/25/2018
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