IR 05000315/2018001
| ML18123A305 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/02/2018 |
| From: | Kenneth Riemer Division Reactor Projects III |
| To: | Gebbie J Indiana Michigan Power Co |
| References | |
| IR 2018001 | |
| Download: ML18123A305 (26) | |
Text
First initial La UNITED STATES NUCLEAR REGULATORY COMMISSION May 2, 2018
SUBJECT:
DONALD C. COOK NUCLEAR POWER PLANT, UNITS 1 AND 2NRC INTEGRATED INSPECTION REPORT 05000315/2018001 AND 05000316/2018001
Dear Mr. Gebbie:
On March 31, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Donald C. Cook Nuclear Power Plant, Units 1 and 2. On April 3, the NRC inspectors discussed the results of this inspection with yourself 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 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 the licensee 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. These NCVs are described in the subject inspection report.
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 Donald C. Cook Nuclear 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 the Donald C. Cook Nuclear 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 Karla Stoedter Acting for/
Kenneth Riemer, Chief Branch 2 Division of Reactor Projects Docket Nos. 50-315; 50-316 License Nos. DPR-58 and DPR-74 Enclosure:
IR 05000315/2018001; 05000316/2018001 cc: Distribution via ListServ
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring licensees performance by conducting an integrated quarterly inspection at the Donald C. Cook Nuclear Plant, Units 1 and 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. NRC and self-revealed findings, violations, and additional items are summarized in the table below.
List of Findings and Violations Failure of Unit 1 Turbine Driven Auxiliary Feedwater Pump to Reach Rated Speed Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Systems Green P.5, Operating 71111.15 NCV 05000315/2018001-01 Experience Closed A self-revealed finding of very low safety significance with an associated Non-Cited Violation of Technical Specification 5.4 Procedures, occurred on December 21, 2017, when the Unit 1 Turbine-Driven Auxiliary Feedwater Pump failed to reach rated speed during a surveillance.
Procedure 12-MHP-5021-056-008, Turbine-Driven Auxiliary Feedwater Pump Governor Valve Maintenance, was not appropriate for the circumstances in that direction was not given to check that the governor valve could fully open following maintenance on the governor valve.
Operation of Letdown System Leads to Voiding and Subsequent Relief Valve Lift Cornerstone Significance Cross-Cutting Report Section Aspect Initiating Events Green N/A 71111.20 NCV 05000315/2018001-02 Closed The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of Technical Specification 5.4, Procedures, when the licensee failed to maintain a procedure for operating the letdown system. As a result, a water-hammer occurred which caused a safety-related relief valve to lift, which discharged reactor coolant to the Pressurizer Relief Tank until letdown was isolated.
Additional Tracking Items Report Type Issue Number Title Section Status URI 05000315/2017004-02 Unit 1 Letdown System Safety Valve 71111.20 Closed Lift During Preparations for Cooldown LER 05000316/2016-002-00 Emergency Diesel Generators 71153 Closed Declared Inoperable due to a Manufacturing Design Issue LER 05000315/2017-001-00 Unit 1 Turbine Driven Auxiliary 71153 Closed Feedwater Pump Inoperable for Longer Than Allowed by Technical Specifications LER 05000316/2017-001-00 Unit 2 Containment Hydrogen 71153 Closed Skimmer Fan #1 Inoperable Longer than Allowed by Technical Specifications
TABLE OF CONTENTS
PLANT STATUS
INSPECTION SCOPES
............................................................................................................
REACTOR SAFETY
..................................................................................................................
RADIATION SAFETY
................................................................................................................
OTHER ACTIVITIES - BASELINE
............................................................................................
INSPECTION RESULTS
.........................................................................................................
EXIT MEETINGS AND DEBRIEFS
......................................................................................... 16
DOCUMENTS REVIEWED
..................................................................................................... 16
PLANT STATUS
Unit 1 began the inspection period at 100 percent and remained at or near 100 percent for the
entire inspection period.
Unit 2 began the inspection period at 100 percent power. On February 26, 2018, the licensee
started a downpower in preparation for a refueling outage. On March 1, the licensee shut the
unit down for refueling. Unit 2 remained shutdown for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in
effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with
their attached revision histories are located on the public website at http://www.nrc.gov/reading-
rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared
complete when the IP requirements most appropriate to the inspection activity were met
consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection
Program - Operations Phase. The inspectors performed 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
External Flooding (1 Sample)
The inspectors evaluated readiness to cope with external flooding during the week of
February 19, 2018.
71111.04Equipment Alignment
Partial Walkdown (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1) Unit 1 East Control Air;
(2) Spent Fuel Pool Cooling System;
(3) 1AB Emergency Diesel Generator (EDG); and
(4) East Diesel and Electrical Fire Pumps.
71111.05AQFire Protection Annual/Quarterly
Quarterly Inspection (4 Samples)
The inspectors evaluated fire protection program implementation in the following selected
areas:
(1) Unit 2 CD 4KV Switchgear Room;
(2) Unit 2 Switchgear Cable Vault;
(3) Unit 2 Control Room; and
(4) Unit 2 Control Room Cable Vault.
71111.06Flood Protection Measures
Internal Flooding (1 Sample)
The inspectors evaluated internal flooding mitigation protections in the Service Water
Screen House.
71111.11Licensed Operator Requalification Program and Licensed Operator Performance
Operator Requalification (1 Sample)
The inspectors observed and evaluated simulator training on February 6, 2018, and Job
Performance Measures (JPM) on February 7, 2018.
Operator Performance (1 Sample)
The inspectors observed and evaluated cooldown of the reactor on March 1, 2018.
Operator Exams (1 Sample)
The inspectors reviewed and evaluated requalification examination results on
February 27, 2018.
Operator Requalification Program (1 Sample)
The inspectors evaluated the operator requalification program from January 15 through
January 26, 2018.
71111.13Maintenance Risk Assessments and Emergent Work Control (5 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent work
activities:
(1) Elevated risk associated with containment work, during the week of February 12, 2018;
(2) Elevated risk associated with reduced inventory, during the week of March 5, 2018;
(3) Elevated risk associated with Unit 2 dual train Essential Service Water (ESW) outage,
during the week of March 19, 2018;
(4) Emergent work to repair an air start relay on the 1CD EDG; and
(5) Management of Unit 2 Accumulator levels and Technical Requirements Manual (TRM)
compliance following Unit 2 cooldown.
71111.15Operability Determinations and Functionality Assessments (4 Samples)
The inspectors evaluated the following operability determinations and functionality
assessments:
(1) Containment Sump due to debris;
(2) Past operability: Unit 1 Turbine Driven Auxiliary Feedwater Pump (TDAFP) failure to
reach rated speed;
(3) Repairs of 1CD EDG Air Start Check Valves; and
(4) EDG Air Start Check Valves past operability.
71111.19Post Maintenance Testing (4 Samples)
The inspectors evaluated the following post maintenance tests:
(1) 1CD EDG Air Start Relay replacement;
(2) ESW valve 2-WMO-718 following torque adjustments;
(3) 2AB EDG following outage maintenance activities; and
(4) Unit 2 Reactor Trip Breaker Auxiliary Relay replacement.
71111.20Refueling and Other Outage Activities (Partial Sample)
The inspectors evaluated Unit 2 refueling outage activities from March 1 through
March 31, 2018. Specifically, the inspectors observed portions of the reactor shutdown and
cooldown, toured containment, reviewed select clearances, and observed foreign material
exclusion practices. Further, the inspectors reviewed other outage-related activities such as
performance of reactor coolant instrumentation, electric power lineups, maintenance of
containment integrity, decay heat removal system performance, coolant inventory control
(including during a period of lowered inventory), performance of the spent fuel pool cooling
system, portions of reactor disassembly, reactivity control, and removal of fuel from the core
to the spent fuel pool. Because this outage remained in progress at the end of the
inspection period, this does not yet constitute a complete sample.
As part of this review, the inspectors also closed out Unresolved Item (URI) 2017004-02,
Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown.
71111.22Surveillance Testing
The inspectors evaluated the following surveillance tests:
Routine (3 Samples)
(1) Unit 1 CD monthly load test;
(2) Unit 2 Ice Condenser ice basket as-found weighing; and
(3) Train B Loss of Power/Loss of Cooling Accident testing during Unit 2 refueling outage.
In-service (2 Samples)
(1) Unit 2 Turbine Driven Auxiliary Feedwater (TDAFW) System test; and
(2) Power Operated Relief Valve (PORV) stroke time testing.
Containment Isolation Valve (1 Sample)
(1) As-left local leak rate test on 2-XCR-101.
71114.06Drill Evaluation
Emergency Planning Drill (1 Sample)
The inspectors evaluated an Emergency Planning training drill involving loss of fission
product barriers on February 13, 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.
71124.02Occupational As Low As Reasonably Achievable Planning and Controls
Implementation of As Low As Reasonably Achievable and Radiological Work Controls
(1 Sample)
The inspectors reviewed ALARA practices and radiological work controls by reviewing the
following activities:
(1) RWP-171102; U1C28 Reactor Reassembly Activities; Revision 0;
(2) RWP-171105; U1C28 Reactor Baffle Bolt Inspection and Repair Activities to Include
Lower Internal Movements; Revision 4;
(3) RWP-171142; U1C28 Containment Install. Modify and Remove Scaffold; Revision 0;
(4) RWP 182105; Unit 2 C24 Reactor Baffle Bolt Inspection and Repair Activities to Include
Lower Internal Movements; Revision 2;
(5) RWP-182142; U2C24 Containment Install, Modify and Remove Scaffold ; Revision 0;
and
(6) RWP-182105; U2C24 Reactor Baffle Bolt Inspection and Repair Activities to Include
Lower Internal Movements; Revision 1.
Radiation Worker Performance (1 Sample)
The inspectors evaluated radiation worker and radiation protection technician performance.
OTHER ACTIVITIES - BASELINE
71151Performance Indicator Verification (10 Samples)
The inspectors verified licensee performance indicators submittals listed below
(1) IE01: Unplanned Scrams per 7000 Critical Hours - 2 Samples
(01/01/2017 - 12/31/2017);
(2) IE03: Unplanned Power Changes per 7000 Critical Hours - 2 Samples
(01/01/2017 - 12/31/2017);
(3) IE04: Unplanned Scrams with Complications (USwC) - 2 Samples
(01/01/2017 - 12/31/2017);
(4) MS05: Safety System Functional Failures (SSFFs) - 2 Samples
(01/01/2017 - 12/31/2017); and
(5) BI01: Reactor Coolant System (RCS) Specific Activity - 2 Samples
(01/01/2017 - 12/31/2017).
71152Problem Identification and Resolution
Annual Follow-Up of Selected Issues (1 Sample)
(1) Review of URI 05000315/2017004-02, Unit 1 Letdown System Safety Valve Lift during
Preparations for Cooldown.
71153Follow-Up of Events and Notices of Enforcement Discretion
Licensee Event Reports (3 Samples)
The inspectors evaluated the following licensee event reports which can be accessed at
https://lersearch.inl.gov/LERSearchCriteria.aspx:
(1) Retraction of Licensee Event Report (LER) 05000316/2016-002-00, Emergency
Diesel Generators Declared Inoperable due to a Manufacturing Design Issue, via
licensee letter AEP-NRC-2017-32. This issue was the subject of a previously
documented finding of very low safety significance and Non-Cited Violation (NCV)
(05000315/2017001-03; 05000316/2017001-03, Failure to Control Nonconforming
Delivery Valve Holders on Emergency Diesel Generators). The inspectors did not
identify any issues associated with the retraction.
(2) Licensee Event Report (LER) 05000315/2017-001-00, Unit 1 Turbine Driven Auxiliary
Feedwater Pump Inoperable for Longer Than Allowed by Technical Specifications
(TSs). This issue is the subject of a finding with an associated NCV described in
section 71111.15 of this inspection report. The inspectors reviewed the LER and
determined there were no further findings or violations. The LER is closed. The
inspectors documented one finding and associated violation related to this LER as05000315/2018001-01.
(3) Licensee Event Report (LER) 05000316/2017-001-00, Unit 2 Containment Hydrogen
Skimmer fan #1 Inoperable Longer than Allowed by Technical Specifications on
May 19, 2017. The inspectors documented one finding and associated violation related
to this LER as05000316/2017002-02. The inspectors reviewed the LER and
determined there were no further findings or violations. The LER is closed.
INSPECTION RESULTS
71111.15Operability Determinations and Functionality Assessments
Failure of Unit 1 Turbine Driven Auxiliary Feedwater Pump to Reach Rated Speed
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Systems Green P.5, Operating 71111.15
NCV 05000315/2018001-01 Experience
Open/Closed
A self-revealed finding of very low safety significance with an associated NCV of TS 5.4
Procedures, occurred on December 21, 2017, when the Unit 1 Turbine-Driven Auxiliary
Feedwater Pump (TDAFP) failed to reach rated speed during a surveillance. Procedure
2-MHP-021-056-008, TDAFP Governor Valve Maintenance, was not appropriate for the
circumstances in that direction was not given to check that the governor valve could fully open
following maintenance on the governor valve.
Description:
On December 21, 2017, the licensee performed quarterly surveillance procedure
1-OHP-4030-156-017T, Turbine Driven Auxiliary Feedwater System Test. When started,
the pump reached a speed of approximately 3500 RPM and then very slowly started to
increase. Normally, the pump would quickly come up to the required speed
of 4330-4370 RP
- M. After approximately one hour, pump speed had only reached
4000 RPM. Operator attempts to raise the speed were unsuccessful. The licensee secured
the pump and declared it inoperable for the speed issue and started to investigate the failure.
Initial troubleshooting by the licensee revealed binding in some parts of the governor valve
linkage. Those issues were corrected, but the pump exhibited the same type of behavior as
before when it was restarted. The licensee then replaced the governor, but that did not
correct the problem. Further investigation revealed that a threaded rod making up part of the
linkage appeared a little long compared to the Unit 2 pump and previous pictures of the Unit 1
pump. When the licensee adjusted this, they discovered the speed responded accordingly.
The licensee adjusted the length to set the required speed, ran several surveillances to test
the machine, and declared the pump operable. In the subsequent apparent cause evaluation,
the licensee discovered that while adjusting the rod resolved the problem, the main issue was
that the maintenance procedure did not direct validation of governor valve travel following
completion of linkage setup per Electric Power Research Institute (EPRI) guidelines. This
would help detect issues throughout the linkage between the governor and governor valve, if
they existed, following reassembly of the linkage. The governor valve just had maintenance
performed on it during the refueling outage which had concluded about one month prior. The
apparent cause analysis determined the governor valve could not fully open due to an error in
the linkage setup following that maintenance, which is why the machine did not reach full
speed during the quarterly surveillance test.
However, the machine had passed 1-OHP-4030-156-017T at the conclusion of the refueling
outage as a post-maintenance test in November of 2017. Additionally, during troubleshooting
after the failure, the machine was also able to pass 1-OHP-5030-156-017TV, TDAFP Trip
and Throttle Valve Test. In their apparent cause evaluation, the licensee explored why the
machine could pass these tests and not the quarterly test that resulted in the failure. When
running 1-OHP-4030-156-017T during the outage, steam pressure was higher due to being
in Mode 3. In the 1-OHP-5030-156-017TV test, a different test loop in the system is used,
so the required horsepower to achieve rated speed was reduced. Because of the different
system conditions, the pump was able to pass those surveillances without the governor valve
full open, but not the test on December 21. Adjustment of the threaded rod permitted
additional travel of the governor valve thus allowing more steam admission to the turbine.
The licensee assessed extent of condition and determined that prior to the outage for Unit 1
and during the current operating cycle for Unit 2, the TDAFPs had passed the quarterly
surveillance. Therefore, this issue only impacted Unit 1 TDAFP operability following the
maintenance window.
Corrective Action: The licensee adjusted the linkage to permit full travel of the governor
valve. As part of the apparent cause evaluation, the licensee will assess various
maintenance procedures for the necessary changes needed to ensure full governor valve
travel following maintenance.
Corrective Action Reference: AR-2017-13059
Performance Assessment:
Performance Deficiency: Safety-related procedure 12-MHP-5021-056-008 was not
appropriate to the circumstances in that it allowed reassembly of the governor valve linkage in
a manner that precluded full governor valve travel. The procedure did not direct a check of
governor valve travel. Had the governor valve been able to fully open as designed, the
machine would have been able to pass all surveillances and be considered operable.
Screening: The inspectors determined the performance deficiency was more than minor
because it adversely affected the Procedure Quality attribute of the Mitigating Systems
cornerstone. Specifically, lack of the necessary detail in the procedure resulted in the Unit 1
TDAFP being inoperable until discovered during a routine surveillance approximately one
month after work and testing had been performed on it.
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix A, Exhibit 2. Per question A.3., the inspectors determined a detailed risk evaluation
was required because the Auxiliary Feedwater (AFW) system was required by TS and a loss
of function for greater than the TS allowed-outage time existed for the turbine-driven train of
- W. The train was inoperable from November 25, 2017, until December 23, 2017, which is
beyond the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by T
- S. Specifically, while the TDAFP could fulfill some of its
PRA functions for the required mission times, it could not be relied upon for all the functions;
namely, functions involving a cooldown to Mode 4 to satisfy certain accident analysis
assumptions.
An NRC Senior Reactor Analyst modeled the impact of the performance deficiency by setting
the basic event for the TDAFP failing to run at True in the DC Cook SPAR model. This was
conservative from a risk assessment perspective because although the pump could be relied
upon in some accident sequences (as described above), for the assessment it was assumed
no functions could be performed. The fire results from the licensees PRA model were then
summed with the NRCs internal results. An exposure time factor (1 month/12 months) was
applied to the result, since this was the time the pump was determined to be inoperable. The
delta-CDF result was 7E-8 (Green). Evaluation of Large Early Release Frequency was not
required because the results were below the 1E-7 threshold. Therefore, based on the
detailed risk analysis, the finding was determined to be of very low safety significance
(Green).
Cross-cutting Aspect: The finding had a cross-cutting aspect in the Operating Experience
component of the Problem Identification and Resolution cross-cutting area, which states that
the licensee will systematically and effectively collect, evaluate, and implement relevant
internal and external operating experience in a timely manner. Specifically, EPRI guidance
for TDAFPs was not incorporated in station procedures which would have precluded the
issue. In light of previous station issues with procedure quality and questions on operability
related to the TDAFPs, the station did an assessment of EPRI guidance and researched other
factors to ensure procedures were adequate for the TDAFPs in 2014 and 2015. The
particular guidance for the governor valve was not incorporated. (P.5)
Enforcement:
Violation: TS 5.4, Procedures, states, in part, that the applicable procedures recommended
in Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements (Operation), are
established, implemented, and maintained. Regulatory Guide 1.33 states, in part, that
maintenance that could affect the performance of safety-related equipment should be
performed in accordance with written procedures appropriate to the circumstances.
Contrary to the above, on November 25, 2017, safety-related procedure
2-MHP-5021-056-008 was not appropriate to the circumstances in that it allowed
reassembly of the governor valve linkage in a manner that precluded full governor valve
travel. In addition, the procedure did not direct a check of governor valve travel.
Disposition: This violation is being treated as an NCV, consistent with Section 2.3.2 of the
71111.20Refueling and Other Outage Activities
Operation of Letdown System Leads to Voiding and Subsequent Relief Valve Lift
Cornerstone Significance Cross-Cutting Report Section
Aspect
Initiating Events Green N/A 71111.20
NCV 05000315 /2018001-02
Open/Closed
The inspectors identified a Green finding and associated NCV of TS 5.4, Procedures, when
the licensee failed to maintain a procedure for operating the letdown system. As a result, a
water-hammer occurred which caused a safety-related relief valve to lift and discharge reactor
coolant to the Pressurizer Relief Tank (PRT).
Description:
On September 13, 2017, Unit 1 had just been shutdown for the refueling outage and
operators were preparing to start the cooldown. The operators lowered charging system flow
in anticipation of reducing letdown flow, which would involve isolation of a 75 gpm letdown
flow orifice. Less than five minutes later, as planned, the operator reduced letdown flow by
isolating the 75 gpm orifice. Less than a minute after that, while trying to raise system
pressure, operators received indications that the safety-related relief valve downstream of the
regenerative heat exchanger in the letdown system had lifted and was discharging reactor
coolant to the PR
- T. After pressure stabilized in the system, operators noted indications that
the relief valve had not fully reseated. Because there is no direct measure of flow past the
relief valve, operators calculated the continued leakage to the PRT to be approximately
gpm. This resulted in the operators declaring an Unusual Event, as the leakage was
above the 10 gpm threshold. The operators isolated the letdown system, which stopped the
continued loss of reactor coolant. The licensee established a multi-discipline team to review
the event and determine actions needed to safely restore letdown and develop a better
understanding of the conditions leading to the relief lifting. The inspectors also reviewed plant
data to validate the licensees conclusion. During this review, the inspectors recognized that
thermodynamic conditions in the letdown system had reached saturation, and the pressure
spike likely occurred from a steam bubble collapse causing a water-hammer. The inspectors
informed the licensee of their concern. In response, the licensee performed a walkdown of
the affected piping to ensure the piping had not been damaged. Prior to continuing the
cooldown, the licensee manually cycled and reseated the relief valve, which allowed letdown
to be restored. Additionally, a pressure transducer for a control valve in the system was
calibrated, as it was initially suspected as being a potential cause for the pressure transient.
Subsequent to these actions, the licensee successfully performed and completed the
cooldown.
After the event, licensee personnel reviewed the data and recalculated the leak rate into the
PRT. This reanalysis determined that the leak rate had not exceeded the threshold for the
declaration of an Unusual Event. As a result, the licensee retracted the event notification.
Licensee personnel
developed several ideas for the cause of the transient but various
organizations disagreed as to what the primary cause was. The inspectors discussed their
concerns regarding the lack of a systematic assessment of the condition with licensee
management. In the fourth quarter of 2017, the inspectors opened an Unresolved Item
(05000315/2017004-02) after they reviewed licensee documentation in the Corrective Action
Program (CAP) regarding this event and could not determine what, if any, performance
deficiencies existed.
After discussing the issue with the inspectors, the licensee performed an Apparent Cause
Evaluation (ACE) on the event in the current inspection period. The inspectors reviewed the
ACE and their own data they had collected. The inspectors determined that a water-hammer
event occurred in the system based on the temperatures and pressures that existed in the
system when operators isolated the 75 gpm orifice. After isolation, pressure dropped (as
expected). At that time, available data indicated temperature was above the point of
saturation for the lowered pressure because of the reduction in charging flow performed by
the operators right before the orifice was isolated. This led to steam bubble formation, and
when pressure was adjusted by the operators, the bubbles collapsed, causing a pressure
spike which lifted the relief valve. Subsequent inspection of the valve later in the outage
revealed severe damage to the internals, which was likely the reason the valve did not reseat
immediately once pressure was relieved. The inspectors determined safety-related
procedure 1-OHP-4021-003-001, Letdown, Charging, and Seal Water Operation did not
contain guidance on how to avoid saturation in the letdown system. The inspectors
determined the issue was NRC-identified based on the inspector observations which led to a
better understanding of the cause (via the ACE), and because of the early NRC observations
which led the licensee to inspect the piping following the water-hammer.
Corrective Action(s): The licensee isolated the letdown system to stop the continued loss of
coolant. Then, the relief valve was exercised to allow it to reseat. The letdown system was
then restored to allow continued cooldown of the plant. After inspector questioning on how
the event occurred, the licensee performed an ACE which recommended changes to
procedures and pre-job briefs to address the potential for saturation conditions in the system.
Corrective Action Reference: AR-2018-0634
Performance Assessment:
Performance Deficiency: The inspectors identified that the licensee failed to maintain a
procedure for operation of the letdown system that included precautions and limitations
appropriate for a quality procedure as described in RG 1.33, Quality Assurance Program
Requirements (Operation). Specifically, precautions and limitations associated with potential
saturation conditions in the system were not included. The licensee is committed to RG 1.33
via TS 5.4, Procedures, and operation of the letdown system is described as a
safety-related process per RG 1.33. Further, the section of piping which experienced the
pressure transient was safety-related.
Screening: The inspectors determined the performance deficiency was more than minor
because it adversely affected the Procedure Quality attribute of the Initiating Events
cornerstone. Specifically, a water-hammer occurred in the letdown system which resulted in
damage to a safety-related relief valve, loss of primary coolant, and declaration of an Unusual
Event (later retracted).
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix A, Exhibit 1. For question A.1., the inspectors determined the finding could not
result in exceeding the leak rate for a small Loss Of Cooling Accident (LOCA) because the
physical configuration of the system and procedural controls limit the maximum letdown flow
to be within the capacity of the coolant charging pumps. For question A.2., the finding could
not affect systems used to mitigate LOCAs because while the charging pumps are within the
system, they are sufficiently isolated from the affected portion of the system by other
components. Further, during an accident, the pump suctions would swap to a different
source, which would isolate them from any leaks upstream. Based on these screening
criteria, the finding screened to Green, or very-low safety significance.
Cross-cutting Aspect: No cross-cutting aspect was assigned to this finding, because the
inspectors determined the finding did not reflect present licensee performance. Specifically,
this procedure was in use for several years and the inspectors identified no recent
opportunities nor operating experience that would have helped the licensee identify the
procedural issue earlier. Additionally, the inspectors had no significant concerns with the
operator response to this event.
Enforcement:
Violation: TS 5.4, Procedures, states, in part, that the applicable procedures recommended
in RG 1.33, Quality Assurance Program Requirements (Operation), are established,
implemented, and maintained. Contrary to this requirement, safety-related procedure
1-OHP-4021-003-001 was not maintained. The sites Quality Assurance Program
Description commits the site to ANSI Standard N18.7-1976/ANS-3.2, Administrative
Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants, with
regard to standards for plant operational procedures. The ANSI N18.7-1976/ANS-3.2
standard requires establishment of precautions to alert the individual performing a task to
important measures used to protect equipment and personnel. Further, R.G. 1.33 stipulates
that operation of the letdown system is a safety-related activity requiring operating
procedures.
Contrary to the above, on September 13, 2017, the procedure did not address the
precautions to preclude water-hammer in the letdown system via appropriate steps or
precautions/limitations. This resulted in operation of the letdown system that caused a water-
hammer and subsequent leakage of reactor coolant through a partially open relief valve on
September 13, 2017. The leakage was stopped on the same day when operators isolated the
letdown system.
Disposition: This violation is being treated as an NCV, consistent with Section 2.3.2 of the
The disposition of this finding and associated violation closes URI 05000315/2017004-02.
71152Problem Identification and Resolution
ObservationReview of URI 05000315/2017004-02, Unit 1 Letdown System IP 71152
Safety Valve Lift During Preparations for Cooldown.
The inspectors performed a review of the licensees follow-up regarding the lifting of a relief
valve in the letdown system on Unit 1 during the Fall 2017 refueling outage. Details of the
event are provided in the above-mentioned URI and in the finding which resulted from
inspector review of the issue described below.
After the plant had been stabilized following isolation of the letdown system, the inspectors
noted that pressure/temperature conditions in the letdown system may have exceeded the
saturation point. Given that, and the fact that the relief valve at lifted, the inspectors
discussed the possibility of a water-hammer event having occurred in the letdown system. As
a result, the licensee performed walkdowns of accessible piping to verify no damage had
occurred. In terms of operator response, the inspectors did not identify any issues with how
the operators controlled the plant following the lift of the relief valve.
Regarding licensee evaluation of the event in the CAP, the inspectors provided several
observations to the licensee. While the licensee did appropriately identify the issue and
document it in the CAP in a timely manner, the licensees resolution did not clearly identify the
cause of the event. An AR existed for the fact the relief valve lifted, another AR documented
saturation conditions may have existed, and another AR documented potential
abnormal/erratic operation of a control valve in the system related to the event. In reviewing
the ARs and in discussions with different licensee departments involved in reviewing the
event, the inspectors were unable to determine what, if any, performance deficiencies existed.
Differing opinions existed between licensee departments. This resulted in the opening of the
above-mentioned UR
- I. Additionally, the inspectors were unable to determine if any of the
corrective actions developed by the licensee were appropriate.
For this sample, the inspectors continued reviewing available plant computer data, current
and previous revisions of the operating procedure for the letdown system, and had continuing
discussions with operations personnel. Additionally, the inspectors reviewed the CAP for
internal operating experience regarding similar events that may have happened. The
inspectors identified a similar event on Unit 2 from 2004, but through further discussions with
licensee personnel and review of additional plant computer data, the inspectors concluded
that the 2004 event had occurred for reasons other than what have been described in the
finding below. After discussing their observations with licensee personnel, the station elected
to perform an Apparent Cause Evaluation (ACE) to ensure a systematic review was done.
The inspectors reviewed the ACE and identified the finding discussed under Inspection
Procedure 71111.20.
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 April 3, 2018, the inspectors presented the inspection results to Mr.
- J. Gebbie, and other
members of the licensee staff. The licensee acknowledged the issues presented;
- On January 26, 2018, the inspectors presented the biennial licensed operator requalification
training (LORT) program inspection results to Mr.
- J. Gebbie, and other members of the
licensee staff;
- On February 27, 2018, the inspectors presented the completed 2018 LORT annual
operating test and biennial written examination inspection results to Mr.
- J. Clark, Licensed
Operator Requalification Training Supervisor; and
- On March 23, 2018, the inspectors presented the ultimate radiation protection program
inspection results to Mr.
- J. Gebbie, and other members of the licensee staff.
THIRD PARTY REVIEWS
The inspectors reviewed the Institute of Nuclear Power Operations (INPO) evaluation and
assessment report discussing plant performance through September 2017.
DOCUMENTS REVIEWED
71111.-1 - Adverse Weather Protection
- 12-OHP-4027-FSG-1501; Flooding Response Deployment; Revision 0
- PMP-4027-FSG-003; Flex Program; Revision 7
- U2C24; Protected Area Laydown Map; 02/02/2018
71111.04Equipment Alignment
- 02-OHL-5030-SOM-035; Unit 2 Tours - Unit 2 Outside Tour; Revision 30
- 12-OHP-4021-018-002; Placing in Service and Operating the Spent Fuel Pit Cooling and
Cleanup System; Revision 31
- 1-OHP-4021-064-001; Operation of Plant and Control Air; Revision 49
- 1-OHP-4030-132-027; AB Diesel Generator Operability Test (Train B); Revision 46
- AirCheck Report; 1-TK-39-DN/Control Air North Dry Air Receiver; 03/09/2017
- AirCheck Report; 1-TK-39-DN/Control Air North Dry Air Receiver; 06/05/2017
- AirCheck Report; 1-TK-39-DN/Unit 1 Control Air North Dry Air Receiver; 09/05/2017
- AirCheck Report; 1-TK-39-DS/Control Air South Dry Air Receiver; 03/09/2017
- AirCheck Report; 1-TK-39-DS/Control Air South Dry Air Receiver; 06/05/2017
- AirCheck Report; 1-TK-39-DS/Unit 1 Control Air North Dry Air Receiver; 12/04/2017
- AirCheck Report; 1-TK-39-DS/Unit 1 Control Air North Dry Air Receiver; 12/04/2017
- AirCheck Report; 1-TK-39-DS/Unit 1 Control Air South Dry Air Receiver; 09/05/2017
- AirCheck Report; 2-TK-39-DN/Control Air North Dry Air Receive Unit 2; 12/05/2017
- AirCheck Report; 2-TK-39-DN/Control Air North Dry Air Receiver; 03/09/2017
- AirCheck Report; 2-TK-39-DN/Control Air North Dry Air Receiver; 06/05/2017
- AirCheck Report; 2-TK-39-DN/Unit 2 Control Air North Dry Air Receiver; 09/05/2017
- AirCheck Report; 2-TK-39-DS/Control Air South Dry Air Receiver; 03/09/2017
- AirCheck Report; 2-TK-39-DS/Control Air South Dry Air Receiver; 06/05/2017
- AirCheck Report; 2-TK-39-DS/Unit 2 Control Air South Dry Air Receiver; 09/05/2017
- AirCheck Report; 2-TK-39-DS/Unit Two Control Air South Dry Air Receiver; 12/06/2017
- AR 2016-9447; Boral Coupon Surveillance Missed Acceptance Criteria; 08/18/2016
- AR 2017-2747; Control Air Sample with 1 Particle >40 Microns; 03/29/2017
- AR 2017-3104; Control Air Sample Particulate Found >40 Microns; 03/21/2017
- NET-28047-02,
09/28
- OP-12-5152T-14; Flow Diagram Fire Protection - Water Piping in Pump House Floor
Elevation 598 -0 Units 1 & 2; 03/19/2009
- OP-1-5151A-50; Flow Diagram Emergency Diesel Generator AB Unit No. 1; Revision 50
- OP-1-5151B-6; Flow Diagram Emergency Diesel Generator AB; Revision 61
71111.05AQFire Protection Annual/Quarterly
- 12-FPP-2270-066-011; Attachment 3; Fire Watch Patrols; 02/11/2018
- 12-FPP-2270-066-011; Attachment 3; Fire Watch Patrols;02/12/2018
- AR 2018-1494; Fire Barrier Penetration Seal F6327 Inoperable; 02/14/2018
- AR 2018-1819; Fire Seal W9993; 02/21/2018
- AR 2018-2376; 2-DR-AUX-463 Found Open; 03/05/2005
- AR 2018-3188; Through Hole Identified in Fire Seal Penetration W5065; 03/20/2018
- AR 2018-3450; Unit 1 Control Room Fire Penetration Requires Cap on Conduit; 03/26/2018
- D.C. Cook Fire Safety Analysis (FSA); Revision 2
- Fire Impairment log Report; 02-12-2018
- Fire Pre-Plans Volume 1; Revision 31
- Fire Pre-plans; Volume 1; Revision 31
- GT 2018-2027; Clarify 12-MHP-5021-EMP-003; 02/27/2018
- NFPPM; American Electric Power D. C. Cook Nuclear Plant; Revision 2
71111.06Flood Protection Measures
- Cook Beyond Design Bases Flood Prevention Plan, Revision 1
- MD-12-Flood-019-N, Mitigating Strategies for Flooding, Revision 1
- MD-12-SCRN-001-N, Screenhouse Internal Flood Levels, Revision 3
71111.11Licensed Operator Requalification Program and Licensed Operator Performance
- 2018 D. C. Cook Nuclear Power Plant Licensed Operator Requalification Program Crew B RO
and SRO Biennial Written Examinations; 01/25/2018
- 2-OHP-4021-001-001; Plant Heatup from Cold Shutdown to Hot Standby; Revision 90
- 2-OHP-4021-001-003; Power Reduction; Revision 55
- ADMIN JPM RO-O-E123; Revision 7; 01/04/2018
- ADMIN JPM SR-O-N0010; Revision 2; 01/04/2018
- Annual Operating Examination Crew Simulator Evaluation; B Shift; 01/24/2017
- Annual Operating Examination Individual Simulator Evaluation; B Shift; 01/24/2017
- AR 2016-1308; Operability Determination Process; 02/03/2016
- AR 2016-13086; Delayed Notification of Unit 2 Refueling Water Storage Tank Level Below
35%; 11/01/2016
- AR 2016-14428; Shutdown Risk Near Miss; 12/16/2016
- AR 2016-5643; PMP-0610-OSD-001 Needs to Better Define Startup; 05/04/2016
- AR 2016-7865; Unit 2 Right Moisture Separator Reheater Expansion Joint Failure; 07/06/2016
- AR 2016-9545; Control Room Log Entries for 2-SRA-2900 Not Accurate; 08/16/2016
- AR 2016-9978; Operability Determination for 1-ERS-1400 Not Accurate; 09/02/2016
- AR 2017-11497; Control Room Log Entries Referencing Appendix R; 11/11/2017
- AR 2017-12804; Post Maintenance Test Not Performed in Correct Mode for 1-NRV-153;
2/13/2017
- AR 2017-12898; Plant Heatup Procedure Step was Incorrectly N/Ad; 12/18/2017
- AR 2017-12948; 1-NRV-153 Mistakenly Declared Operable; 12/19/2017
- AR 2017-2196; Action Way Corrective Actions Discrepancies from Apparent Cause
Evaluation in 2015-3236-3; 02/22/2017
- AR 2017-4529; Control Room Log Entry for Unplanned Technical Specification Entry Not
Bold; 05/03/2017
- AR 2018-0676; One Copy of an NRC Exam Simulator Scenario Could Not Be Located;
01/22/2018
- Closed Simulator Deficiency Records List from January 1; 2017 through January 23; 2018
- Donald C. Cook Nuclear Power Plant Design Function Time Critical Operator Actions
- Individual Operator Training Records-B Shift; Requalification Year 41
- OHI-2070; Operations Training and Qualification; Attachment 4; Active License
Watchstanding Records; B Shift Individuals for 3rd and 4th Quarters 2017
- Open Simulator Deficiency Records List; 01/23/2018
- PMP-7030-SFD-001; Safety Function Determination Program; Revision 9
- Remediation Packages for Licensed Operators; (various); Requalification Years 41 and 42
- RQ-S-4205-U2-T1; Period 4205 U2 Training Scenario 1; Simulator Exercise Guide; Revision 0
- SBT Package for Validation of Scenario RQ-E-ANN-21; Revision 2; 03/03/2017
- SBT Package for Validation of Scenario RQ-E-ANN-59; Revision 0; 03/13/2017
- SEG RQ-E-ANN-17; 01/04/2018
- SEG RQ-E-ANN-32; 01/04/2018
- SEG RQ-E-ANN-60; 01/04/2018
- SEG RQ-E-ANN-63; 01/22/2018
- Simulator Exercise Guide (SEG) RQ-E-ANN-6; 01/04/2018
- Simulator JPM RO-O-AEO-A006; Revision 1; 01/04/2018
- Simulator JPM RO-O-AEO-E005; Revision 3; 01/04/2018
- Simulator JPM RO-O-AEO-E281; Revision 2; 01/04/2018
- Simulator JPM RO-O-AEO-N202; Revision 2; 01/04/2018
- Simulator JPM RO-O-E007A-U12; Revision 0; 01/04/2018
- Simulator JPM RO-O-E022A-U12; Revision 0; 01/04/2018
- Simulator JPM RO-O-E107A-U12; Revision 0; 01/04/2018
- Simulator JPM RO-O-E279-U12; Revision 0; 01/04/2018
- TDG-SIM-005; Scenario Based Testing (SBT) Guideline; Revision 5
- Training Oversight Committee Agenda/Minutes; Discipline Committee: Operations Continuing
Training; 04/07/2017
- Training Oversight Committee Agenda/Minutes; Discipline Committee: Operations Continuing
Training; 07/19/2017
- TRP-2070 SIM-001; Simulator Configuration Control; Revision 11
- TRP-2070 SIM-003; Simulator Performance Testing; Revision 7
- TRP-2070-TAP-300-LOR; Licensed Operator Requalification Training Annual Operating
Test and Biennial Written Examination Development; Revision 6
- TRP-2070-TAP-300-OPS; Operations Training Examination and Simulator Exercise Guide
Development; Revision 17
- TRP-2070-TAP-400; SAT Implementation; Data Sheet 7: Remediation; Revision 28
- TRP-2070-TAP-400; SAT Implementation; Revision 28
- TRP-2070-TAP-400-LOR; Licensed Operator Requalification Training Annual Operating
Test and Biennial Written Examination Implementation; Revision 5
- TRP-2070-TAP-400-SEC; Operations Training NRC Exam Security; Revision 7
- TRP-2070-TAP-OPS; Operations Training Implementation; Revision 48
- U1C28 Core Test; 11/16/2017
- U1C28 Steady State Test; 12/19/2017
- U1C28 Transient Test 11-Load Rejection; 11/28/2017
- U1C28 Transient Test 1-Manual Reactor Trip; 11/15/2017
- U1C28 Transient Test 3-Simultaneous Closure of All MSIVs; 11/28/2017
- U1C28 Transient Test 5-Trip of Reactor Coolant Pump #13; 11/15/2017
- U1C28 Transient Test 7-Maximum Power Rate Ramp 100%-75%-100%; 11/15/2017
- U2C23 Core Test; 07/01/2017
- U2C23 Steady State Test; 04/04/2017
- U2C23 Transient Test 10-Stuck Open PORV 2/Hi Head Injection Inhibited; 06/05/2017
- U2C23 Transient Test 1-Manual Reactor Trip; 06/02/2017
- U2C23 Transient Test 2-Simultaneous Trip of Both Main Feed Pumps; 06/02/2017
- U2C23 Transient Test 4-Simultaneous Trip of All RCPs; 06/02/2017
- U2C23 Transient Test 6-Turbine Trip without Reactor Trip; 06/01/2017
- U2C23 Transient Test 8-Maximum Size RCS Break with Loss of Offsite Power; 06/01/2017
71111.13Maintenance Risk Assessments and Emergent Work Control
- PMP-4010-CAC-001; Containment Access Control; Revision 19
- PMP-4100-sdr-001; Plant Shutdown Safety and Risk Management; Revision 46
- PMP-4100-SDR-002; Outage Risk Assessment and Management; Revision 11
- AR 2018-0246; DGCD-PS Incorrect Fuses Installed; 01/08/2018
- WO 55511745-01; 1-42-DGCD, Bench Check and Replace Relay
- 12-IHP-5021-IMP-001; Lead Lifting/Landing and Electrical Jumper/Fuse Installation and
Removal; Revision 15
-
for FLEX Strategies, Revision 74
71111.15Operability Determinations and Functionality Assessments
- 12MHP5021.032.021; Emergency Diesel Engine Air Start Check Valve Maintenance;
Revision 2
- 12-MHP-5021-032-021; Emergency Diesel Engine Air Start Check Valve Maintenance/
Revision 4
- 12-MHP-5021-056-008; Turbine Driven Auxiliary Feed Pump Governor Valve Maintenance;
Revision 20
- 680-41400; Containment Sump Strainer Replacement Large Size Head Loss Test Report;
Revision 0
- AR 2018-1000; Systpect Starting Air Check Valve Leakage on 1CD 4F; 01/31/2018
- AR 2018-1228; Scattered insulation in containment around/below 2-HV-CEQ-2; 02/08/2018
- AR 2018-1337; Procedural Adherence with CAC-001; 02/09/2018
- Commercial Grad Dedication Evaluation 00011132, Drawing for Air Start Check Valve,
Revision 0
- DB-12-AFWS; Auxiliary Feedwater System; Revision 9
- MD-12-SUMP-001-N; Containment Recirculation Sump Function; Revision 1
- OP-1-5120Y-12; Flow Diagram 100# Control Air System Header Diesel Generators 1AB &
1CD Unit #1; 11/30/2015
- OP-1-5151D-72; Flow Diagram Emergency Diesel Generator CD Unit No. 1; 11/16/2015
- PRA-MSPI-Basis; MSPI Data - Auxiliary Feedwater Systems; Revision 13
- PRA-NB-QU; Internal Events Quantification Notebook; Revision 4
- PRA-NB-SC; Success Criteria Notebook; Revision 2
- PRA-NB-SY-AFW; Donald C. Cook Nuclear Plant Units 1 and 2 Auxiliary Feedwater System
Notebook; Revision 6
- PRA-SDP-2018-01; Risk Analysis of Unit 1 TDAFP Failure to Reach Rated Speed; Revision 0
- UFSAR 14.1.9; Loss of Normal Feedwater Flow; Revision 28
- VTD-DRCO-0011; Terry Turbine (A Division of Dresser Rand) Governor Control System,
Revision 0
- VTD-WORT-0001; Worthington Corporation Installation and Operating Instructions for Four
Cycle - Diesel and Dual Fuel Engines, Type SWB-VEE; Revision 8
71111.19Post Maintenance Testing
- 12-IHP-5030-EMP-014; MOV Diagnsotic Testing; Revision 22
- 2-OHP-4030-232-027AB; AB Diesel Generator Operability Test; Revision 48
- AR 2018-3247; Air Leak - 2-XRV-221, DG2AB F Bank Start Air Control Valve; 03/22/2018
- AR 2018-3255; 2-OME-150-AB Unloaded Megger Readings Unsatisfactory; 03/22/2018
- DB-12-CTS; Containment Spray system Design Basis Document; Revision 15
- DB-12-EDG; Emergency Diesel Generators Design Basis Document; Revision 10
- DB-12-EDGS; Design Basis Document for the Emergency Diesel Generator Support Systems;
Revision 10
- OP-1-98035; Diesel Generator ICD Control Elementary Diagram; 09/11/2014
- OP-2-5120Y-12; Flow Diagram 100# Control Air System Header Diesel Generators 2AB &
2CD Unit #2; 10/27/2015
- OP-2-5151B-68; Flow Diagram Emergency Diesel Generator AB Unit No. 2; 12/11/2013
- OP-2-98101-36; Turbine Control Sheet No. 1 Elementary Diagram Unit No. 2 CIA #42373;
10/27/2016
- OP-2-98139-3; Turbine Control Sheet No 1 Elementary Diagram; 12/15/2016
- Operating Logs; Unit 2; 03/22/2018
- PS-2-92059-12; Generator Reat Panel GRB Sh. #2 Wiring Diagram; 01/04/2017
- Technical Data Book, Unit 2; 2-Figure 19.8; Revision 49
- VDS-2-WMO-718; AR-2013-10860-19 to Correct Limit Switch Settings, Revision 3
- WO 55420010-01; 2-52X1-RTTB, Replace Relay and Post-Maintenance Test Relay
- WO 55503287-01; 1-DGCD-2301A, Adjust Fuel Limit Start Potentiometer
- WO 55503457; Meggar Cable Associated with 2-QT-106-AB1
- WO 55505562-24; 2-OME-AB; Pre and Post Maintenance Generator Checks
- WO 55509351; 2-WMO-718-ACT; As-Found/Adjust Limit/As-Left Diagnostic
- WO 55509351; Leakby Unit 2 Essential Service Water Outlet valve on the West CTS HX
71111.20Refueling and Other Outage Activities
- 12-MHP-4050-FHP-023; Reactor Vessel Head Removal with Fuel in the Vessel; Revision 1
- 2-OHP-4021-001-003; Power Reduction; Revision 66
- 2-OHP-4021-001-004; Plant Cooldown from Hot Standby to Cold Shutdown; Revision 72
- 2-OHP-4021-002-005; RCs Draining; Revision 48
- 2-OHP-4021-017-002; Placing in Service the Residual Heat Removal System; Revision 28
- 2-OHP-4030-227-037; Refueling Surveillance; Revision 28
- AEP-94-811; Loop to Loop Power Imbalance; 10/18/1994
- AR 2018-1985; Steel braided tie Downs have Loose Wires on Ends; 02/26/2018
- AR 2018-2012; Unit 2 QPTR Exceeded 1.02 on Downpower; 02/28/2018
- AR 2018-2268; Foreign Material Found in Containment; 03/03/2018
- AR 2018-2321; Teletower Extended While Unattended; 03/04/2018
- AR 2018-2383; Worker in Standard Risk FME Area was not FME Ready; 03/06/2018
- AR 2018-2456; Inadequate FME Controls at Reactor Cavity; 03/06/2018
- AR 2018-2468; MTIS did not Control Tooling in FME High Risk Area; 03/07/2018
- AR 2018-2469; FME Monitors did not Identify Material in an FME Zone; 03/07/2018
- AR 2018-2472; High Risk FME Practices U2 Upper Containment; 03/07/2017
- AR 2018-2676; Install Mechanical Jumper to Reduce Waste Water Generation; 03/102018
- AR 2018-2680; Rusting on 2-VFX-4-V2; 03/10/2018
- FO-18-C-072; FME Practices Observed for MTIS Ironworkers for U2C24 Refueling Outage;
03/06/2018
- PM--2010-PRC-003; Procedure and Work Instruction Usage and Adherence; Revision 49
- PMP-2220-0001-0001; Foreign Material Exclusion (FME); Data Sheet 1; Revision 39
- PMP-2220-001-001; Foreign Material Exclusion (FME); Revision 39
- PMP-4100-SDR-001; Plant Shutdown Safety and Risk Management; Revision 46
- R-CW-CIRC-0014; Clearance to Support Dual Train ESW; 03/20/2018
- WO 55495218; FME Zone Requirements
71111.22Surveillance Testing
- 12-EHP-4030-010-262; Ice Condenser Surveillance and Operability Evaluation; Revision 21
- 12-MHP-4030-010-001; Ice Condenser Basket Weighing Surveillance, Revision 22
- 12-MHP-4030-010-002; Ice Condenser Flow Channel Surveillance; Revision 10
- 1-OHP-4030-132-027CD; CD Diesel Generator Operability Test (Train A); Revision 47
- 2-EHP-4030-234-203; Unit 2 LLRT; Revision 26
- 2-OHP-4030-202-060, PZR Power Operated Relief Valve Testing, 02/28/2018
- 2-OHP-4030-202-060. PZR Power Operated Relief Valve Testing; Revision 19
- 2-OHP-4030-232-217B; DG2AB Load Sequencing & ESF Testing, Revision 52
- 2-OHP-4030-256-017T; Turbine Driven Auxiliary Feedwater System Test; 02/01/2018
- AR 2018- 2158; Administrative Error TDB-2-Fig-19-1; Revision 115; 03/01/2018
- AR 2018-2125. 2-NRV-152 Failed Its IST Closed Stroke Time; 03/01/2018
- AR 2018-2247; DG2AB Field Volt Meter is not Indication Correctly; 03/02/2018
- ECP-12-N1-05, Low Temperature Overpressure Protection, LTOP Setpoint Calculation,
Revision 13
- OP-2-5120D-32; Flow Diagram Containment Control Air 85# & 50# Ring Headers Unit 2;
11/18/2016
- Valve Reference Data Sheet, 2-NRV-152, 12/18/2016
71114.06Drill Evaluation
- 2018 1St Quarter Training Drill Evaluation Report, 03/11/2018
- EP Training Drill Scenario Manual; 02/13/2018
- EP Training Drill Scenario Manual; 02/13/2018
- PMP-2080-EPP-101; Emergency Classification; Revision 21
- PMP-2080-EPP-101; Emergency Classification; Revision 21
71124.01Radiological Hazard Assessment and Exposure Controls
- 12-THP-6010-RPP-405; Analysis of Airborne Radioactivity; Revision 23
- AR 2017-10435; Security Tour performed Differently due to High Rad; 10/14/2017
- AR 2017-11844; Unanticipated Self Reading Dosimeter Dose Rate Alarm; 11/19/2017
- AR 2017-8976; Unanticipated SRD Dose rate Alarm on the 2/3 Steam Generator Platform;
09/17/2017
- AR 2017-9220; Individual ED Dose Alarm; 09/21/2017
- AR 2018-2336; Poor Radiation Worker Practices used by ICE Personnel; 03/05/2018
- AR 2018-2475; Poor Radiation worker Practices Observed by NOS; dated 03/07/2018
- AR 2018-3474; CESA Box Container of Equipment Fastener Found Loose while Stored
Outside of the RPAC; 03/25/2018
- AR 2018-3506; RP did not Verify Neutron Dose Rates for Unit 1 CLV Work Near the Window
of the Annulus; dated 03/27/2018
- AR 2018-3626; Potential Industrial Gap in Radiological Contaminated Area Waste Oil Release
Methods; 04/02/2018
- Baffle Bolt Tool Radiation Survey Plan 2018
- NSTS; Annual Inventory Reconciliation Report; dated 01/04/2018 and 01/18/2017
- PMP-6010-RPP-001; General Radiation Worker Instructions; Revision 26
- PMP-6010-RPP-003; High, Locked High and Very High Radiation Area Access; Revision 28
- PMP-6010-RPP-006; Radiation Work Permit Program; Revision 25
- PMP-6010-RPP-301; Control of Material in a Radiologically Controlled Area; Revision 32
- PMP-6010-RPP-301; Control of Material in a Radiologically Controlled Area; Liquid/Granular
- RWP 182105; Unit-2 C24 Reactor Baffle Bolt Inspection and Repair Activities to Include
Lower Internal Movements; Revision 2
- RWP-171102; U1C28 Reactor Reassembly Activities; Revision 0
- RWP-171105; U1C28 Reactor Baffle Bolt Inspection and Repair Activities to Include Lower
Internal Movements; Revision 4
- RWP-171142; U1C28 Containment Install. Modify and Remove Scaffold; Revision 0
- RWP-182105; U2C24 Reactor Baffle Bolt Inspection and Repair Activities to Include Lower
Internal Movements; Revision 1
- RWP-182116; U2C24 In-service Test & Inspections in Auxiliary Building and Plant
Radiological Controlled Areas; Revision 0
- RWP-182142; U2C24 Containment Install, Modify and Remove Scaffold ; Revision 0
- Solid Release Form and Isotopic Analysis; Release Numbers; 17-0095; 17-0096; 17-0097; 18-
23; 18-1125; 18-0006;18-0008; 18-0009 through 18-0014
- Unit-1 Report of Core Barrel Moves for LRSS Clevis Bolt Work; Dose Rate Data ED Location
Historical Data; 04/30/2013
71124.02Occupational As Low As Reasonably Achievable Planning and Controls
- 02717473; CREVS Charcoal Test Results did not Meet Acceptance Criteria; 09/16/2018
- ALARA Committee Meeting A-18-22F; Dose Reduction Through Ownership and
accountability; 03/19/2018
- DC COOK U1C28 Refueling Outage ALARA Report; 02/05/2018
- PMP-6010-ALA-001; ALARA Program Review of Plant Work Activities; Revision 33
71151Performance Indicator Verification
- NEI 99-02; Regulatory Assessment Performance Indicator Guideline; Revision 7
- Performance Indicator Data on Quarterly Report; 01/01/2017 - 12/31/2017
- PMP-7110-PIP-001; Reactor Oversight Program Performance Indicators and Monthly
Operating Report Data; Reactor Coolant Specific Activity; Revision 15
71152Problem Identification and Resolution
- 1-OHP-4021-003-001; Letdown, Charging and Seal Water Operation, Revision 69
- 1-OHP-4021-003-001; Letdown, Charging and Seal Water Operation, Revision 68
- 1-OHP-4021-003-001; Letdown, Charging and Seal Water Operation; Revision 72
- AR 00090196; While Shifting 75 Gallon per Minute Orifices on Line the Regenerative Heat
Exchanger Safety Valve Lifted when 2-QRV-161 was Opened Causing the Reactor Coolant
System to Lose Inventory; 04/29/2004
- AR 00090298; The Unit 2 Letdown Safety Valve (2-SV-51) Lifted on 04/29/2004 While Placing
a 75 Gallon per Minute Orifice In-service; 04/30/2004
- AR 2017-8716; 1-SV-51 Leakrate of 12.2.Gallons per Minute After Unit 1 Reactor Trip;
09/13/2017
- AR 2018-0634; Unit 1 Letdown Safety Lifted While Removing Letdown Orifice; 01/21/2018
- AR-2017-8740; 1-SV-51, Possible Point of Saturation During Unusual Event; 09/13/2017
- DB-12-CVCS; Chemical and Volume Control System; Revision 8
- GT-00087671; OHP-4021-003-001, Letdown Charging and Seal Water Operation, Attachment
13, Operation of Normal Letdown, does not Provide Adequate Direction for Certain
Conditions; 02/17/2004
- OP-1-5129-68; Flow Diagram CVCS-Reactor Letdown & Charging Unit No. 1; 05/06/2017
- UFSAR 14.3; Reactor Coolant System Pipe Rupture (Loss of Coolant Accident); Revision 26
- UFSAR 9.2; Chemical and Volume Control System; Revision 28
71153Follow-Up of Events and Notices of Enforcement Discretion
- AEP-NRC-2017-32; Cancellation of Licensee Event Report 316/2016-002-00; Emergency
Diesel Generators Declared Inoperable Due to a Manufacturing Design Issue; 06/13/2017
- LER 05000316/2016-002-00; Emergency Diesel Generators Declared Inoperable Due to a
Manufacturing Design Issue; 02/09/2017
- LER 05000316/2017-001-00; Unit 2 Containment Hydrogen Skimmer Ventilation Fan #1
Inoperable Longer than Allowed by Technical Specifications; Revision 0
- LER 315/2017001-01; Unit 1 Turbine Driven Auxiliary Feedwater Pump Inoperable Longer
than Allowed by Technical Specifications; 02/19/2018
- NUREG-1022; Event Report Guidelines 10 CFR 50.72 and 50.73; Revision 3
First initial La
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE RD. SUITE 210
LISLE, ILLINOIS 60532-4352
May 2, 2018
Mr. Joel P. Gebbie
Senior VP and Chief Nuclear Officer
Indiana Michigan Power Company
Nuclear Generation Group
One Cook Place
Bridgman, MI 49106
SUBJECT: DONALD
- C. COOK NUCLEAR POWER PLANT, UNITS 1 AND 2NRC
INTEGRATED INSPECTION REPORT 05000315/2018001 AND
Dear Mr. Gebbie:
On March 31, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at your Donald
- C. Cook Nuclear Power Plant, Units 1 and 2. On April 3, the NRC inspectors
discussed the results of this inspection with yourself 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 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 the licensee 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. These NCVs are described in the subject inspection report.
If you contest the violations or significance of these NCVs, you should provide a response within
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 Donald C. Cook Nuclear Power Plant.
J. Gebbie -2-
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 the Donald C. Cook
Nuclear 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 Karla Stoedter Acting for/
Kenneth Riemer, Chief
Branch 2
Division of Reactor Projects
Docket Nos. 50-315; 50-316
License Nos. DPR-58 and DPR-74
Enclosure:
IR 05000315/2018001; 05000316/2018001
cc: Distribution via ListServ
J. Gebbie -3-
Letter to Joel Gebbie from Kenneth Riemer dated May 2, 2018
SUBJECT: DONALD
- C. COOK NUCLEAR POWER PLANT, UNITS 1 AND 2NRC
INTEGRATED INSPECTION REPORT 05000315/2018001 AND