ML19130A154: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
 
(StriderTol Bot change)
 
(6 intermediate revisions by the same user not shown)
Line 1: Line 1:
#REDIRECT [[IR 05000445/2019001]]
{{Adams
| number = ML19130A154
| issue date = 05/10/2019
| title = NRC Integrated Inspection Report 05000445/2019001 and 05000446/2019001
| author name = Haire M
| author affiliation = NRC/RGN-IV/DRP/RPB-A
| addressee name = Peters K
| addressee affiliation = Vistra Operations Company, LLC
| docket = 05000445, 05000446
| license number = NPF-087, NPF-089
| contact person =
| document report number = IR 2019001
| document type = Inspection Report, Letter
| page count = 41
}}
See also: [[see also::IR 05000445/2019001]]
 
=Text=
{{#Wiki_filter:May 10, 2019
Mr. Ken Peters, Senior Vice President 
  and Chief Nuclear Officer
VISTRA Operations Company, LLC
P.O. Box 1002
Glen Rose, TX 76043
SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 - NRC
INTEGRATED INSPECTION REPORT 05000445/2019001 AND
05000446/2019001
Dear Mr. Peters:
On March 31, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at your Comanche Peak Nuclear Power Plant, Units 1 and 2.  On April 2, 2019, the NRC
inspectors discussed the results of this inspection with Mr. Steven Sewell and other members of
your staff.  The results of this inspection are documented in the enclosed report.
NRC inspectors documented seven findings of very low safety significance (Green) in this
report.  These findings involved violations of NRC requirements.  Additionally, NRC inspectors
documented one Severity Level IV violation with no associated finding.  The NRC is treating
these violations as non-cited violations (NCV) consistent with Section 2.3.2.a of the
Enforcement Policy.
The inspectors also documented a licensee-identified violation which was determined to be of
very low safety significance in this report.  The NRC is treating this violation as a non-cited
violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest the violations or significance or severity of the violations documented in this
inspection report, you should provide a response within 30 days of the date of this inspection
report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: 
Document Control Desk, Washington, DC  20555-0001; with copies to the Regional
Administrator, Region IV; the Director, Office of Enforcement; and the NRC resident inspector at
the Comanche Peak Nuclear Power Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the U.S. Nuclear Regulatory Commission, ATTN:  Document Control Desk,
Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the
NRC resident inspector at the Comanche Peak Nuclear Power Plant.
 
K. Peters
2
2
This letter, its enclosure, and your response (if any) will be made available for public inspection
and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document
Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for
Withholding.
Sincerely,
/RA/
Mark S. Haire, Chief
Project Branch A
Division of Reactor Projects
Docket Nos. 50-445 and 50-446
License Nos. NPF-87 and NPF-89
Enclosure:
Inspection Report 05000445/2019001 
and 05000446/2019001 
 
3
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Number(s): 
05000445 and 05000446
License Number(s):
NPF-87 and NPF-89
Report Number(s):
05000445/2019001 and 05000446/2019001
Enterprise Identifier: I-2019-001-0011
Licensee:
Vistra Operations Company, LLC
Facility:
Comanche Peak Nuclear Power Plant, Units 1 and 2
Location:
Glen Rose, TX 76043
Inspection Dates:
January 1, 2019 to March 31, 2019
Inspectors:
W. Cullum, Reactor Inspector
R. Deese, Senior Reactor Analyst
J. Drake, Senior Reactor Inspector
J. Josey, Senior Resident Inspector
R. Kumana, Resident Inspector
W. Sifre, Senior Reactor Inspector
Approved By:
Mark S. Haire, Chief
Project Branch A
Division of Reactor Projects
 
4
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees
performance by conducting a Quarterly inspection at Comanche Peak Nuclear Power 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. 
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: 71111.18.
List of Findings and Violations
Inadequate Corrective Actions for Failure to Ensure Containment Hatch Closure Capability
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Barrier Integrity
Green
NCV 05000445; 05000446/2019001-01 
Closed
[H.6] - Design
Margins
71111.04
The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, Corrective Actions, associated with the licensees failure to take adequate
corrective actions for an inadequate containment closure procedure.  Specifically, in
December 2017, the NRC identified that the licensee's procedure for emergency closure of
the Unit 1 and 2 containment equipment hatches was inadequate, and the licensee failed to
take adequate actions to correct the issue prior to the next outage.
Failure to Evaluate a Change to the Facility DC Power System
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Not Applicable
NCV 05000445/2019001-02 
Closed
Not Applicable
71111.04
The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.59 for the
licensees failure to obtain a license amendment or perform a written evaluation
demonstrating the basis for not obtaining a license amendment, prior to making a change to
the facility as described in the final safety analysis report.
Failure to Monitor or Demonstrate Control of Performance Under the Maintenance Rule
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000445; 05000446/2019001-03 
Closed
None
71111.12
The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(2), with
three examples, for failure to demonstrate effective control of performance of a maintenance
rule scoped system through appropriate preventive maintenance.
 
5
Failure to Control Hazard Barriers During Maintenance
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000445/2019001-04 
Closed
[H.14] -
Conservative
Bias
71111.13
The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(4) for failure to
implement risk mitigating actions during diesel generator maintenance.
Failure to Follow Procedure When A Degraded Condition Was Identified
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000445; 05000446/2019001-05 
Closed
[H.14] -
Conservative
Bias
71111.15
The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to
follow the requirements of Station Procedure STI-421.01, Initiation of Issue Reports,
Revision 0.  Specifically, station personnel failed to notify the shift manager of an issue with
material storage in the protected area.  This issue required evaluations and compensatory
actions for resolution.
Failure to Perform Safety Evaluations in Accordance with 10 CFR 50.59
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000445; 05000446/2019001-06 
Closed
[H.9] - Training
71111.17T
The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion V,
Instructions, Procedures, and Drawings, (with four examples) in which the licensee failed to
complete 50.59 evaluations as required by station procedures.
Inadequate Maintenance Instructions Result in Loss of Assessment Capability
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Emergency
Preparedness
Green
NCV 05000445; 05000446/2019001-07 
Closed
[H.8] -
Procedure
Adherence
71152
The inspectors reviewed a self-revealed Green, non-citied violation of 10 CFR 50,
Appendix B, Criterion V, "Instruction, Procedures, and Drawings," that occurred due to
inadequate maintenance instructions for safety-related radiation monitors that resulted in a
major loss of assessment capability.
 
6
Failure to Establish Adequate Procedural Guidance for Flushing Lithium at Power
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000446/2019001-08 
Closed
[H.11] -
Challenge the
Unknown
71152
The inspectors reviewed a Green, self-revealed non-cited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the
licensees failure to establish an adequate procedure for flushing lithium from the residual
heat removal system.  This resulted in safety injection accumulators 2-01 and 2-02 discharge
to the safety injection test header causing level drops in both accumulators and
accumulator 2-01 pressure dropped to below the operability limit resulting in an unplanned
component inoperability.
Additional Tracking Items
Type
Issue Number
Title
Report
Section
Status
NOV
05000446/2018011-01 
Failure to Maintain a Quality
Record Complete and Accurate
in All Material Respects
92702
Closed
LER
05000446/2018-001-00 Unit 2 Automatic Reactor Trip
Due to Turbine Trip, on
March 19, 2019
71153
Closed
 
7
PLANT STATUS
Unit 1 began the inspection period at or near rated thermal power.  On February 1, 2019, the
unit was down powered to 64 percent for turbine testing.  The unit was returned to rated thermal
power the same day.  On March 22, 2019, the unit began power coast down to a refueling
outage, ending the inspection period at 92 percent power.
Unit 2 began the inspection period in a refueling outage.  On January 14, 2019, the unit began a
reactor startup.  The unit shut down on January 15, 2019, due to a main turbine primary water
leak.  On January 18, 2019, the unit began a reactor startup and reached rated thermal power
on January 22, 2019.  On March 2, 2019, the unit was tripped due to a failure of a main
feedwater isolation valve.  The unit began a reactor startup the same day and reached rated
thermal power on March 4, 2019.  The unit remained at or near rated thermal power 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 2515, Light-Water Reactor Inspection Program -
Operations Phase.  The inspectors performed plant status activities described in Inspection
Manual Chapter 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
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.03) (1 Sample)
The inspectors evaluated readiness for impending adverse weather conditions for severe
thunderstorms on March 13, 2019.
71111.04 - Equipment Alignment
Partial Walkdown (IP Section 02.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1)
Unit 1, safety injection pump 1-01 while 1-02 was out of service for maintenance on
February 5, 2019
(2)
Unit 2, containment hatches on February 13, 2019
 
8
(3)
Units 1 and 2, common class-1E DC power on March 5, 2019
(4)
Units 1 and 2, seismic monitoring system on March 18, 2019
71111.05Q - Fire Protection
Quarterly Inspection (IP Section 03.01) (5 Samples)
The inspectors evaluated fire protection program implementation in the following selected
areas:
(1)
fire area 2CA, Unit 2 reactor building on January 9, 2019
(2)
fire zones TB201 and TB202, control room emergency lighting battery rooms on
January 14, 2019
(3)
fire zone 1SB2A, Unit 1 safety injection pump 1-01 on March 11, 2019
(4)
fire zone 2SB4, Unit 2 containment spray chemical add tank on March 13, 2019
(5)
fire zone SM157, stairwell in the southeast corner of the safeguards building on
March 26, 2019
71111.06 - Flood Protection Measures
Inspection Activities - Internal Flooding (IP Section 02.02a.) (1 Sample)
The inspectors evaluated internal flooding mitigation protections in the service water intake
structure on March 12, 2019.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)
(2 Samples)
(1)
The inspectors observed and evaluated licensed operator performance in the Control
Room during Unit 2 startup on January 14, 2019.
(2)
The inspectors observed and evaluated licensed operator performance in the Control
Room during Unit 2 startup on January 18, 2019.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
The inspectors observed and evaluated a simulator-based loss of coolant accident scenario
on March 27, 2019.
 
9
71111.12 - Maintenance Effectiveness
Routine Maintenance Effectiveness Inspection (IP Section 02.01) (3 Samples)
The inspectors evaluated the effectiveness of routine maintenance activities associated with
the following equipment and/or safety significant functions:
(1)
common low voltage power distribution failure to align to normal power supply on
February 28, 2019
(2)
Unit 1, battery charger and inverter failures which occurred in June 2018, on 
February 28, 2019
(3)
service air check valve failure during surveillance testing on March 14, 2019
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent work
activities:
(1)
Unit 1, risk mitigating actions during emergency diesel generator 1-01 lube oil fill on
January 17, 2019
(2)
Unit 1, risk mitigating actions while safety injection pump 1-02 was out of service on
February 5, 2019
(3)
Unit 1, risk assessment during sequencer undervoltage replacement on
February 13, 2019
(4)
Units 1 and 2, removal of service water pipe tunnel missile shield CPX-SWMEBB-01
on February 28, 2019
(5)
Units 1 and 2, risk mitigating actions with transformer XST2 unavailable on
March 29, 2019
71111.15 - Operability Determinations and Functionality Assessments
Sample Selection (IP Section 02.01) (5 Samples)
The inspectors evaluated the following operability determinations and functionality
assessments:
(1)
CR-2019-000324, Units 1 and 2, environmental qualification of steam generator
atmospheric relief valves on January 10, 2019
(2)
CR-2019-000456, Units 1 and 2, Electroswitch Part 21 relay issue on
January 14, 2019
 
10
(3)
TR-2019-001119, Units 1 and 2, tornado missile evaluation for equipment storage on
February 13, 2019
(4)
TR-2019-000805, Units 1 and 2, operations support center HVAC sensor failure on
February 14, 2019
(5)
CR-2019-002132, Unit 1, environmental qualification of service water valves with
teflon components on March 12, 2019
71111.17T - Evaluations of Changes, Tests, and Experiments
Sample Selection (IP Section 02.01) (35 Samples)
The inspectors reviewed the following evaluations (items 1 through 8), screenings, and/or
applicability determinations for 10 CFR 50.59 from September 30, 2016, to
January 14, 2019.
(1)
EV-CR-2016-001706-8, Revision1; FDA-2016-000025-01 temporary modification of 
2RC-8054A to repair a leak on pressurizer 01 Pressure Transmitter.
(2)
AEV-CR-2016-005587-9; FDA-2016-000142-01, LDCR SA-2016-013 and
LDC R TR-2016-003, Missile Probability Analysis Revision.
(3)
EV-TR-2017-003173-5 ABN-104, Revision 9; PCN-9 addition of alternate residual
heat removal path and use of safety injection pump for core cooling in Mode 6.
(4)
EV-TR-2017-007959-13; Perform 50.59 Evaluation for FDA-2017-000106-02
Generator Repair Plan and 59SC-2017-000106-02.
(5)
EV-2014-013052-9; Modification to change the isolated phase bus cooling fans start
logic to provide seven out of eight dampers open requirement using digital
equipment.
(6)
EV-CR-2016-003267-10; FDA-2016-000075-01 Unit 1 pressurizer instrument
isolation valves class change (LDCR-SA-2016-010).
(7)
EV-TR-2018-004520-14; Evaluate operator action for isolation of faulted battery
charger from its battery per 50.59 screen EV-TR-2018-004520-13.
(8)
EV-CR-2017-004574-2; 59SC - STA-707-1 50.59 screen for 2RF16 changes to
procedures for reactor vessel head and upper internals lifts.
(9)
EV-TR-2015-006849-4; 59SC - Provide 50.59 SC to support DCP-17-000010 to input
FZ locations of raceways and equipment into GENESIS in support of 
ME-CA-0000-1086 revision.
(10)
EV-TR-2018-004520-10; 59SC - Perform a 50.59 screen for a compensatory
measure to jumper battery cell.
 
11
(11)
EV-CR-2014-003412-18; 59SC - Perform 50.59 applicability determination and
screen for additional plugging for component cooling water heat exchanger 2-01 in
2RF14.
(12)
EV-TR-2018-003799-6; Perform 10CFR50.59 review of minor fuel design changes
documented in NF-TB-16-21. 
(13)
EV-TR-2018-003700-2; Refer to the attached VDRT package which contains the
requested screen and complete VDRT-5608075 package for valve XWT-0634.
(14)
EV-TR-2018-000169-4; 50.59 screen for backseating of 1MS-0357, SG 1-03
blowdown downstream isolation valve.
(15)
EV-TR-2018-000198-1; Maintenance clearance placed for isolation of 1-LG-2706A
may exceed 90 days.
(16)
EV-TR-2018-000199-1; Maintenance clearance placed for diesel generator starting
compressor solenoid 1-SV-3422-1F may exceed 90 days.
(17)
EV-TR-2018-000600-1; Shift manager clearance placed to isolate TBX-CSFLSI-01
seal water injection filter 01.
(18)
EV-CR-2016-008147-3; Compensatory action of installing scaffolding for medium
energy line break (MELB) barrier.
(19)
EV-CR-2017-007829-1; 59SC - Compensatory actions to install temporary equipment
for flow measurement.
(20)
EV-CR-2017-010212-1; 59SC - Shift manager clearance CP17-0913 due to
feedpump deluge valve not resetting.
(21)
EV-CR-2017-012952-28; 59SC - Shift manager clearance to remove fuses 2-
KXA/0746 and 2-KXB/0746.
(22)
EV-CR-2018-004743-2; 59SC - Compensatory action to blow down the receiver once
per shift.
(23)
EV-TR-2016-005840-10; 59SC - VDRT-5575487 Which includes vendor final
acceptance tests for open phase protection equipment for XST1.
(24)
EV-TR-2017-000041-32; 59SC - VDRT-5397434, Fuel transfer system transfer cart
weldment.
(25)
EV-TR-2017-003173-4; 59SC - Review for revision to ABN-104 based on 
EV-TR-2017-003173-3 for loss of residual heat removal events.
(26)
EV-CR-2018-002390-5; 59SC - Changes made under EV-CR-2018-002390-4.
(27)
EV-CR-2018-006758-1; 59SC - Screen for the compensatory action for average
containment temperature.
 
12
(28)
EV-CR-2018-007384-1; 59SC - Perform 50.59 screen changes to procedures 
OPT-612B and OPT-613B.
(29)
EV-CR-2016-007812-1; 59SC - Perform a 10CFR50.59 Review per STA-707 to
update UFSAR Table 9.5-18 to specify tube plugging limit for diesel generator jacket
water coolers for Unit 1 and Unit 2.
(30)
EV-TR-2018-008391-16; 59SC - Perform a 10CFR50.59 Review per STA-707 to plug
tubes in the component cooling water heat exchangers.
(31)
EV-CR-2018-002189-2; 59SC - 50.59 screen for compensatory action to maintain 
2-HV-2334A accumulator pressure above 2100psi.
(32)
EV-CR-2016-008215-20; 59SC - 50.59 review of compensatory measures to isolate
suction and discharge pressure indication on CT and SF pumps; 
ref:  EV-CR-2016-008215-19.
(33)
EV-TR-2016-009344-1; 59SC - Shift Manager Clearance CP16-1381 initiated to
maintain X-PV-3218A isolated following failure of a functional stroke; request a
50.59SC to determine impact on the plant.
(34)
EV-CR-2018-005954-3; 59SC - Seal injection filters housing bolts and potential
excessive torque specification VDRT-5655877.
(35)
EV-TR-2016-010572-2; 59SC - 59SC - Perform a 50.59 screen for hanging shift
manager clearance CP16-1614 on 2-HS-2802A for damage to upper journal bearings
on the motor for Circulating Water Pump Motor 2-03.
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) 
(2 Samples)
The inspectors evaluated the following temporary or permanent modifications:
(1)
Unit 2, pressurizer power operated relief valve accumulator pressure setpoint
modification on February 14, 2019
(2)
bladder addition to safety-related tanks on March 11, 2019
71111.19 - Post Maintenance Testing
Post Maintenance Test Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the following post maintenance tests:
(1)
Unit 2, diesel generator 2-02 following intercooler crack and jacket water repair on
February 12, 2019 
(2)
Unit 2, pressurizer spray valve following actuator rebuild on February 20, 2019
 
13
(3)
Unit 1, diesel generator 1-01 following fuel injector torqueing on March 13, 2019
(4)
Unit 2, residual heat removal pump 2-02 following pump refurbishment on
March 19, 2019
(5)
Unit 2, auxiliary feedwater pump 2-01 following maintenance on March 20, 2019
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated refueling outage 2RF17 activities from January 1, 2019, to
January 18, 2019, completing the sample for the refueling outage which started on
December 8, 2018 (see Inspection Report 05000445/2018004; 05000446/2018004 (ADAMS
Accession No. ML19042A345)).  Specifically, the inspectors completed Inspection
Procedure 71111.20, Sections 03.01.d through e, during this inspection period.
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance tests:
Containment Isolation Valve (ISO) (IP Section 03.01) (1 Sample)
Unit 2, service air containment isolation valve test on March 7, 2019
Surveillance Testing (IP Section 03.01) (1 Sample)
Unit 2, OPT-601B auxiliary feedwater flow control valve accumulator pressure drop test on
March 26, 2019
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
IE01:  Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) (2 Samples)
(1)
Unit 1 from January 2018 through December 2018
(2)
Unit 2 from January 2018 through December 2018
IE03:  Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) 
(2 Samples)
(1)
Unit 1 from January 2018 through December 2018
(2)
Unit 2 from January 2018 through December 2018
 
14
IE04:  Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) 
(2 Samples)
(1)
Unit 1 from January 2018 through December 2018
(2)
Unit 2 from January 2018 through December 2018
71152 - Problem Identification and Resolution
Annual Follow-up of Selected Issues (IP Section 02.03) (2 Samples)
The inspectors reviewed the licensees implementation of its corrective action program
related to the following issues:
(1)
radiation monitor failures due to failure to install a jumper during maintenance on
February 28, 2019
(2)
safety injection accumulator discharge due to inadequate procedure on
March 29, 2019
71153 - Follow-up of Events and Notices of Enforcement Discretion
Event Report (IP Section 03.02) (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 05000446/2018-001-00, "Unit 2 Automatic Reactor Trip Due
to Turbine Trip," on March 19, 2019
The inspectors determined that it was not reasonable to foresee or correct the cause
discussed in the LER; therefore, no performance deficiency was identified.  The inspectors
also concluded that no violation of NRC requirements occurred. 
OTHER ACTIVITIES - TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
92702 - Follow-up on Corrective Actions for Violations And Deviations
Follow-up - Corrective Actions - Violations and Deviations (1 Sample)
On March 28, 2019, the inspectors reviewed the licensees response to
NOV 05000446/2018011-01, "Failure to Maintain a Quality Record Complete and Accurate
in All Material Respects," and determined that the reason for the violation, corrective actions
taken and planned to address recurrence, and the date when full compliance will be
achieved for this violation is adequately addressed and captured on the docket.
 
15
INSPECTION RESULTS
Inadequate Corrective Actions for Failure to Ensure Containment Hatch Closure Capability
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Barrier Integrity
Green
NCV 05000445; 05000446/2019001-
01 
Closed 
[H.6] - Design
Margins
71111.04
The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, Corrective Actions, associated with the licensees failure to take adequate
corrective actions for an inadequate containment closure procedure.  Specifically, in
December 2017, the NRC identified that the licensee's procedure for emergency closure of
the Units 1 and 2 containment equipment hatches was inadequate and the licensee failed to
take adequate actions to correct the issue prior to the next outage.
Description:  In Inspection Report 2017-004, the NRC documented a non-cited violation for an
inadequate procedure, STI 600.01, "Protecting Plant Equipment and Sensitive Equipment
Controls."  This procedure contained instructions for emergency closure of the containment
equipment hatch during times when the hatch was open, but the ability to close containment
was required.  The inspectors observed that the bolting pattern and required torque that were
identified in the supporting engineering calculation were not incorporated into the procedure. 
The licensees technical evaluation required four bolts spaced 90 degrees apart and torqued
to 30 percent preload values.  The procedure did not require bolts to be evenly spaced and
only required the bolts to be snug tight, a licensee term implying full effort on the tool being
used.  The licensee entered this into their corrective action program.  Subsequently, the
licensee performed an evaluation to justify alternate bolt spacing patterns and revised the
procedure to include adequate bolting patterns.  However, in their evaluation the licensee
stated that no torque requirement existed, and the requirement was only to hold the hatch in
place.
The inspectors observed the containment hatch closure training during Refueling
Outage 2RF17.  The inspectors observed that the bolt patterns used conformed to the revised
procedure and evaluation, but that the hatch operators did not appear to apply any torque to
the bolts.  When the inspectors asked about the bolts, the operators believed that there was
no requirement to apply any torque beyond that needed to hold the hatch in place.
The inspectors determined that by not applying any type of torque to the bolts, the licensee
was not verifying that the containment equipment hatch could be sealed.  A seal is necessary
to ensure that a release of fission product radioactivity within containment will be restricted
from escaping to the environment in the event of a loss of decay heat removal event when the
reactor coolant system was open to the atmosphere.
 
The licensee performed another evaluation and concluded that the minimum torque required
to ensure a seal with four bolts was 144 ft-lbf.  The licensee conducted additional training with
all hatch operators on the requirement to ensure a seal on the hatch.  They also conducted a
demonstration with the assigned operators and concluded that the average operator applying
full effort would achieve greater than 150 ft-lbf.
 
 
16
Corrective Action(s):  The licensee trained the operators on the requirement to ensure the
bolts were adequately torqued and verified through demonstration that the operators could
apply enough torque to ensure the hatch would be sealed.
 
Corrective Action Reference(s):  CR-2018-008300, CR-2019-002533
Performance Assessment:
 
Performance Deficiency:  The inability to assure containment closure during a postulated loss
of decay heat removal or fuel handling accident was a condition adverse to quality.  The
failure to correct a condition adverse to quality is a performance deficiency.
 
Screening:  The inspectors determined the performance deficiency was more than minor
because it was associated with the SSC and barrier performance attribute of the Barrier
Integrity Cornerstone.  It adversely affected the cornerstone objective to provide reasonable
assurance that physical design barriers (fuel cladding, reactor coolant system, and
containment) protect the public from radionuclide releases caused by accidents or events
because the finding represented a loss of reasonable assurance of the ability to close the
containment equipment hatch.  Specifically, the failure to assure that personnel would
adequately torque the bolts on the hatch sufficient to establish a seal would, in an actual
event, result in a loss of the containment barrier.
 
Significance:  The inspectors assessed the significance of the finding using Appendix H,
Containment Integrity SDP.  Using Inspection Manual Chapter 0609, Attachment 04, Initial
Characterization of Findings, dated October 7, 2016, the inspectors determined the finding
was associated with the Barrier Integrity cornerstone.  Using Inspection Manual
Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination
Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 4, Barrier
Integrity Screening Questions, the inspectors determined the finding degraded the ability to
close or isolate containment and required evaluation under Inspection Manual Chapter 0609,
Appendix H, Containment Integrity Significance Determination Process, dated
February 25, 2019.  Using the Large Early Release Frequency (LERF) type screening
process, the inspectors determined the finding was a Type B LERF finding because the
finding did not affect core damage frequency.  The inspectors used 
Table 7.3, Phase 1 Screening - Type B Findings at Shutdown, and determined that a 
Phase 2 estimate was required because the containment equipment hatch affected
containment isolation, which is a system important to LERF.  The inspectors used Table 7.4,
Phase 2 Risk Significance - Type B Findings at Shutdown, to determine the finding was of
very low safety significance (Green) because it did not meet the threshold for low safety
significance (White) for leakage from containment to the environment being greater than
100 percent containment volume per day through containment penetration seals, isolation
valves, or vent and purge systems.  Specifically, the licensee was able to demonstrate
through calculations that the leakage from the containment hatch being closed, but not
sealed, would be no more than 30 percent of the containment volume per day.
Cross-cutting Aspect:  H.6 - Design Margins:  The organization operates and maintains
equipment within design margins.  Margins are carefully guarded and changed only through a
systematic and rigorous process.  Special attention is placed on maintaining fission product
barriers, defense-in-depth, and safety-related equipment.  Specifically, the licensee
incorrectly assumed that a seal on the containment hatch was not required at the onset of an
 
17
accident and that the increased pressure in containment during an accident could be credited
for making a seal on the hatch.
Enforcement:
 
Violation:  10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requires in part
that conditions adverse to quality are promptly identified and corrected.  Contrary to the
above, from December 2017 to December 2018, the licensee failed to promptly correct a
condition adverse to quality.  Specifically, the licensee failed to implement adequate
corrective actions for an inadequate procedure for emergency containment closure to ensure
the containment was sealed, an activity affecting quality.
Enforcement Action:  This violation is being treated as a Non-Cited Violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Evaluate a Change to the Facility DC Power System
Cornerstone
SL-IV
Cross-cutting
Aspect
Report
Section
Not Applicable
NCV 05000445/2019001-02 
Closed 
Not
Applicable
71111.04
The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.59 for the
licensees failure to obtain a license amendment or perform a written evaluation
demonstrating the basis for not obtaining a license amendment prior to making a change to
the facility as described in the final safety analysis report.
Description:  The inspectors reviewed the plant configuration of two common Class 1E DC
power panels that can be powered from either the Unit 1 or Unit 2 Class 1E DC busses.  The
inspectors found that the licensee has shared systems for both units that receive power from
these panels.  The panels also have Unit 1 safety-related systems powered from the panels. 
The inspectors noted that shared systems must meet the requirements of 10 CFR Part 50,
Appendix A, Criterion 5, which states, in part, that structures, systems, and components
important to safety shall not be shared among nuclear power units unless it can be shown
that such sharing will not significantly impair their ability to perform their safety functions. 
The inspectors questioned whether the inclusion of Unit 1 safety-related equipment on the
common panels constituted acceptable sharing of systems between units.
Upon further review, the inspectors determined that the licensee originally committed to
Regulatory Guide (RG) 1.81 to demonstrate compliance with Criterion 5.  The licensee, in its
Final Safety Analysis Report (FSAR), stated that the DC power sources and electric
distribution systems were not shared between the two units, and that safety-related loads
shared between both units are powered from common 125 VDC panels.  The NRC in its
safety evaluation report concluded that the design as described in the FSAR, with shared
systems being powered from the common panels but no unit-specific safety-related systems
powered from the common panels, was acceptable.
In January 2000, the licensee discovered that they had unit-specific safety-related systems
from both Units 1 and 2 on the common panels in addition to the previously evaluated shared
systems, contrary to what was described in their FSAR.  The licensee entered this design
control issue into the corrective action program.  In 2002, the licensee modified the Unit 2
systems to align them to Unit 2 power supplies, but left the Unit 1 systems on the common
panels.  The licensee then revised the FSAR to state that they did not comply with RG 1.81,
 
18
but that the existing configuration of Unit 1 systems was an acceptable exception.  The
inspectors determined that powering Unit 1 systems from the Unit 2 DC power supply and
distribution system constituted a system being shared among units, and that the licensee had
not demonstrated compliance with Criterion 5 for these systems while the panels supplying
Unit 1 systems were powered from Unit 2.  At the time of the inspection, the common panels
were aligned to Unit 1.
The inspectors determined that the inclusion of Unit 1 systems on panels that shared DC
power systems was a change to the facility as described in the FSAR.  The inspectors also
determined that the licensee made the change without performing a written evaluation
demonstrating that a license amendment would not be required.  This impeded the ability of
the agency to perform its regulatory function, requiring disposition using traditional
enforcement. 
Corrective Action(s):  The licensee entered this violation into their corrective action program.
Corrective Action Reference(s):  CR-2019-001711
Performance Assessment:  The inspectors determined this violation was associated with a
minor performance deficiency.
Enforcement: 
The ROPs significance determination process does not specifically consider the regulatory
process impact in its assessment of licensee performance.  Therefore, it is necessary to
address this violation which impedes the NRCs ability to regulate using traditional
enforcement to adequately deter non-compliance. 
Severity:  The violation was determined to be Severity Level IV using section 6.1 of the NRC
Enforcement Policy, dated May 15, 2018, because it was a violation of 10 CFR 50.59, but did
not have a consequence evaluated by the significance determination process as having
low-to-moderate or greater safety significance.
Violation:  Title 10 CFR 50.59 requires, in part, that if the licensee makes changes to the
facility as described in the FSAR without obtaining a license amendment, they must maintain
a written evaluation which provides the basis for determining that the change does not require
a licensee amendment.  Contrary to the above, in April 2002, the licensee made a change to
the facility as described in the FSAR without obtaining a license amendment, but did not
maintain a written evaluation which provides the basis for determining that the change does
not require a licensee amendment.
Enforcement Action:  This violation is being treated as a Non-Cited Violation, consistent with
Section 2.3.2 of the Enforcement Policy.
 
19
Failure to Monitor or Demonstrate Control of Performance Under the Maintenance Rule
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000445; 05000446/2019001-03 
Closed 
None
71111.12
The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(2), with three
examples, for failure to demonstrate effective control of performance of a maintenance rule
scoped system through appropriate preventive maintenance.
Description:  The inspectors identified three examples where the performance of systems,
structures, and components (SSCs) that were subject to the maintenance rule, was not
monitored or demonstrated to be effectively controlled through appropriate preventive
maintenance.
The first example is a violation of 10 CFR 50.65(a)(2) for failure to monitor performance or
demonstrate effective control of performance for the Class 1E battery chargers.  The
inspectors identified a failure of the 1ED1-1 battery charger to successfully perform a
maintenance rule function.  The battery chargers provide DC power to the class 1E DC
busses from the Class 1E AC busses.  The vital bus inverters rely on effective control of
DC voltage ripple on the battery charger output to allow synchronization with class 1E AC
power prior to being placed online.  The licensee incorporated a limit of 2 percent voltage
ripple into the design basis document for the DC system.  However, the licensee did not
perform any testing or preventive maintenance to ensure output voltage ripple remained
within limits.  As a result, the DC output voltage ripple of the 1ED1-1 battery charger
exceeded acceptable voltage ripple at some point in its service life, ultimately resulting in a
failure of the supported inverter to return to service on June 5, 2018.
The licensee determined that the excessive ripple was caused by a failure of a component in
the battery charger, the X-302 printed circuit board (PCB).  The PCB had last been replaced
in December 2016 and was scheduled for a 10-year replacement frequency.  Subsequent to
that replacement, the licensee documented multiple occurrences where the inverters
supported by that charger did not synchronize correctly.  The licensee had generated work
orders to troubleshoot the inverter but had not completed them prior to the June 2018 failure. 
Following this failure, the licensee performed an evaluation of the event for their maintenance
rule program.  The licensee evaluated the failure as not being a maintenance rule failure
because the battery charger functions, as written, did not describe providing power to the DC
busses.  The inspectors concluded that the function to provide power to the DC busses was a
maintenance rule function and that the June 2018 failure was a functional failure. 
Furthermore, because the failure could have been prevented by either performing preventive
maintenance on the battery charger or by completing the troubleshooting work orders, the
failure was maintenance preventable.  The June 2018 failure exceeded the established
performance criteria, indicating performance was not being effectively controlled, but the
licensee did not monitor performance or set goals.  The licensee entered this issue into the
corrective action program.
The second example is a violation of 10 CFR 50.65(a)(2) for failure to monitor performance or
demonstrate effective control of performance for the common low voltage AC power system. 
The inspectors identified a failure of the common 120 VAC power system to provide Class 1E
power to certain important to safety components that are shared between Units 1 and 2.  The
common panels provide power to shared radiation monitors that require Class 1E power to
function following an accident, which is covered by the maintenance rule under 
 
20
10 CFR 50.65(b)(2)(i).  The panels can be transferred to non-Class-1E power for
maintenance.  Following a planned maintenance activity on Panel XEC1 in October 2016, the
licensee was unable to transfer the panel back to its normal Class 1E source due to a failure
of the transfer switch.  Because the failure represented an inability to receive power from its
Class 1E source, this was a failure to meet its maintenance rule function.  The failure was
maintenance preventable, because the licensee was aware of the potential for these switches
to fail but did not perform preventive maintenance to address the failures.  The licensee
incorrectly concluded that the transfer switch failure was not a maintenance preventable
failure of a maintenance rule function, because the common panels were being monitored
against plant level performance criteria.  The performance of the system cannot be practically
monitored by the use of plant level criteria, because the common low voltage power system
could have unlimited maintenance preventable functional failures without ever meeting the
criteria.  The licensee entered this issue into the corrective action program.
The third example is a violation of 10 CFR 50.65(a)(2) for failure to monitor performance or
demonstrate effective control of performance for the inside reactor containment check
valves 1(2)CA-0016.  Inspectors noted that the performance criteria assigned to the valves
was inadequate and that there had been multiple failures of these valves during testing.
These results should have been classified as repeat maintenance preventable functional
failures and caused the system to be classified as 50.65(a)(1), but the system remained in
50.65(a)(2) status.
 
The inspectors noted that the valves were allowed seven failures in a 24-month monitoring
period.  This was determined to be inadequate because the valves were tested on a 30month
frequency, so the allowed amount of failures could never be exceeded.  Additionally, the
inspectors determined that the cause of the valves failures was a known issue, but the
licensee had not taken action to correct it.  Specifically, the valves and system piping are
carbon steel and are part of the service air system.  The service air system is neither filtered
nor dried which results in water accumulation in the air system.  Water accumulation in the
system causes general corrosion in the piping, resulting in wear particles that affect the
valves ability to close.  The inspectors determined that the licensee was aware of the failure
mechanism, the cause, and a solution for the issue but had prioritized it as a low priority and
was not considering this when evaluating whether the failures were maintenance preventable. 
The inspectors determined that the failures were maintenance preventable and as such, were
repeat failures, because the licensee had failed to perform the appropriate modifications to
the system.  The licensee entered this issue into the corrective action program.
In all these cases, the inspectors determined that the failure to demonstrate effective control
was caused by incomplete descriptions of the applicable maintenance rule functions, which
had been developed during initial implementation of the maintenance rule in the 1990s.
 
Corrective Action(s):  The licensee entered these three examples into the corrective action
program and is reviewing the systems performance.
 
Corrective Action Reference(s):  CR-2018-007884
Performance Assessment:
 
Performance Deficiency:  The failure to monitor the performance or demonstrate effective
control of performance of systems covered by the maintenance rule is a performance
deficiency.
 
21
Screening:  The inspectors determined the performance deficiency was more than minor
because it was associated with the equipment performance attribute of the Mitigating
Systems Cornerstone. It adversely affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences (i.e., core damage) because the finding represented a reduction in
the reliability and availability of mitigating systems.  Specifically, the failure to monitor the
performance of the battery chargers resulted in multiple instances of decreased reliability of
the system.  The common low voltage power system affected the Emergency Preparedness
Cornerstone, and the containment isolation valves affected the Barrier Integrity Cornerstone,
but the Mitigating Systems Cornerstone was selected as the most significant due to the risk
significance of the battery chargers.
 
Significance:  The inspectors assessed the significance of the finding using Appendix A,
Significance Determination of Reactor Inspection Findings for At - Power Situations.  Using
Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated
October 7, 2016, the inspectors determined the finding was associated with the Mitigating
Systems cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance
Determination Process (SDP) For Findings At-Power, Exhibit 2, Mitigating Systems
Screening Questions, the inspectors determined the finding was of very low safety
significance (Green) because the finding did not represent an actual loss of function of at
least a single train for greater than its technical specification allowed outage time.
Cross-cutting Aspect:  No cross-cutting aspect was assigned to this finding because the
inspectors determined the finding did not reflect present licensee performance. 
Enforcement:
 
Violation:  10 CFR 50.65(a)(1), requires, in part, that the holders of an operating license shall
monitor the performance or condition of structures, systems, or components (SSCs) within the
scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a
manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling
their intended functions.
10 CFR 50.65(a)(2) states, in part, that monitoring as specified in 10 CFR 50.65(a)(1) is not
required where it has been demonstrated that the performance or condition of an SSC is
being effectively controlled through the performance of appropriate preventive maintenance,
such that the SSC remains capable of performing its intended function.
Contrary to the above, as of March 31, 2019, the licensee failed to demonstrate that the
performance of the Class 1E battery chargers, the common 120 VAC power panels, and
containment check valves had been effectively controlled through the performance of
appropriate preventive maintenance and did not monitor against licensee-established goals. 
Specifically, the licensee failed to identify, and properly account for preventive maintenance
preventable functional failures of the battery chargers, the common 120 VAC panels, and
containment check valves occurring from October 2016 to June 2018 which demonstrate that
the performance or condition of these SSCs was not being effectively controlled through the
performance of appropriate preventive maintenance and, as a result, that goal setting and
monitoring was required.
Enforcement Action:  This violation is being treated as an non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
 
22
Failure to Control Hazard Barriers During Maintenance
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000445/2019001-04 
Closed 
[H.8] -
Procedure
Adherence
71111.13
The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)4 for failure to
implement risk mitigating actions during diesel generator maintenance.
Description:  On January 17, 2019, the inspectors observed the licensee performing a
maintenance activity to add lube oil to the Unit 1 emergency diesel generator 1-01 sump.  In
order to perform the maintenance, the licensee placed a hose through the normally shut
door S1-28 from the train A switchgear room to the train A diesel generator room.  The door is
a dogged, two-leaf metal hatch that functions as a barrier for fire, flooding, and medium
energy line break (MELB) events.  Prior to performing the maintenance, the licensee
evaluated the risk of opening the door to allow placement of the hose.  The licensee identified
additional compensatory measures to protect the train A switchgear in an evaluation
documented in Tracking Report (TR) 2019-000001.  The licensee determined that the open
door did not pose a flood risk and implemented appropriate compensatory measures to
mitigate the fire risk.  To address the MELB risk, the licensee determined that the open
doorway of the active leaf of door S1-28 could allow a MELB in the diesel generator room to
impact safety-related transformer T1EB3, which provides 480 VAC power to safety-related
bus 1EB3.  The licensee determined that the transformer would be protected if the workers
maintained door S1-28 open no more than 2 inches, with the door secured to prevent it from
opening further.  The licensee determined that opening the door for normal ingress and
egress was acceptable provided the door was secured after personnel passed through.  The
evaluation was attached to the work order and a copy was present at the job site.
When the workers began the job, they identified safety concerns with the door being secured
while personnel were in the diesel generator room.  They decided to leave the door open,
assuming that it was acceptable as long as personnel were in the immediate area to close it. 
When the inspectors arrived at the work site, they noticed the door open with no one passing
through it and questioned the configuration of the door.  The inspectors then contacted the
control room and the licensee secured the door.
The licensee determined that crediting actions to close the door post event did not adequately
mitigate the risk of a MELB.  As a result of the failure to implement the risk mitigating actions,
the licensee determined that the train A 480 VAC bus 1EB3 was inoperable for
approximately 3 hours due to the potential for a MELB to spray water on the transformer.  The
allowable outage time of the bus per Technical Specification 3.8.9 is 8 hours.  The licensee
determined that the bus did not exceed its allowed outage time due to the hazard barrier
being open.
 
Corrective Action(s):  The licensee restored the barrier and entered the issue into the
corrective action program.
 
Corrective Action Reference(s):  CR-2019-000672
 
23
Performance Assessment:
 
Performance Deficiency:  The failure to implement planned risk mitigating actions was a
performance deficiency.
 
Screening:  The inspectors determined the performance deficiency was more than minor
because it was associated with the Configuration Control attribute of the Mitigating Systems
cornerstone. It adversely affected the cornerstone objective to ensure the availability,
reliability, and capability of systems that respond to initiating events to prevent undesirable
consequences (i.e., core damage) because the finding represented a loss of control of
barriers required to ensure the availability of AC power.  Specifically, the failure to maintain
the door in a nearly closed position exposed a Class 1E 480 VAC bus to failure during a
MELB event, resulting in an electrical distribution train being inoperable for several hours.
 
Significance:  The inspectors assessed the significance of the finding using Appendix K,
Maintenance Risk Assessment and Risk Management SDP.  Using Inspection Manual
Chapter 0609, Attachment 04, Initial Characterization of Findings, dated October 7, 2016,
the inspectors determined the finding was associated with the Mitigating Systems
cornerstone.  Using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk
Assessment and Risk Management Significance Determination Process, the inspectors
determined the finding was associated with risk mitigating actions (RMAs) only.  The
inspectors used Flowcharts 1 and 2 to determine that the finding required a determination of
the incremental core damage probability due to the failure to implement RMAs.
A risk analyst performed a bounding analysis of incremental core damage probability
assuming that bus 1EB3 was unavailable along with the train A emergency diesel generator
for the entire exposure time when adequate RMAs were not in place.  This estimate was
bounding because it assumes bus 1EB3 always failed during the exposure time and does not
incorporate the probabilistic occurrences of fire, flooding, line break, and other events could
have rendered bus 1EB3 unavailable, which would result in a lower estimate of incremental
core damage probability.  The resulting bounding estimate in the incremental core damage
probability was 8.1E-8.  The inspectors determined that the finding was of very low safety
significance (Green) because the incremental core damage probability was less than 1E-6
and the finding did not affect the large early release probability.
Cross-cutting Aspect:  H.14 - Conservative Bias: Individuals use decision making-practices
that emphasize prudent choices over those that are simply allowable.  A proposed action is
determined to be safe in order to proceed, rather than unsafe in order to stop.  Specifically,
the licensee personnel assumed that the controls were not necessary without stopping work
and discussing with their supervisor, and did not implement prescribed risk mitigating actions.
Enforcement:
 
Violation:  10 CFR 50.65(a)(4) requires, in part, that the licensee assess and manage the
increase in risk that may result from maintenance activities.  Contrary to the above, on
January 17, 2019, the licensee failed to manage the increase in risk resulting from a
maintenance activity.  Specifically, the licensee did not implement planned risk mitigating
actions that were identified as necessary by the risk assessment.
Enforcement Action:  This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
 
24
Failure to Follow Procedure When A Degraded Condition Was Identified
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000445; 05000446/2019001-05 
Closed 
[H.14] -
Conservative
Bias
71111.15
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to
follow the requirements of Station Procedure STI-421.01, Initiation of Issue Reports,
Revision 0.  Specifically, station personnel failed to notify the shift manager of an issue with
material storage in the protected area. This issue required evaluations and compensatory
actions for resolution.
Description:  On January 31, 2019, inspectors identified that the licensee had allowed
material to be stored in a temporary laydown area inside of the protected area.  Inspectors
noted that several items appeared to be susceptible to being picked up by tornado driven
winds, so the inspectors inquired as to how these items had been evaluated for their current
storage area.  The licensee initiated TR-2019-001119 to capture the inspectors questions. 
As part of TR-2019-001119 the licensee determined that the materials in question had not
been evaluated for its current storage location.  An action was assigned to engineering to
evaluate the materials in question (AI-TR-2019-001119-1).  Engineering completed their
evaluation on February 4, 2019, and engineering management approved the evaluation on
February 6, 2019.  The evaluation determined that there were materials in the laydown area
that were susceptible to being lifted by tornadic winds, and they needed to be strapped down
in such a way as to increase their weights to a point where they were no longer susceptible.
Inspectors reviewed AI-TR-2019-001119-1 on February 14, 2019.  During their review they
determined that the identified condition required an operability review because of the potential
to be in an unanalyzed condition with respect to tornado driven missiles.  However, inspectors
noted that an operability review was not performed because the issue had not been reported
to the control room by engineering upon discovery on February 4, 2019, as required by
Station Procedure STI-421.01, Initiation of Issue Reports, Revision 0, Section 6.1. 
Additionally, there was no guidance or actions in place to adequately strap down the material
to ensure that it did not pose a risk to plant equipment. 
Inspectors informed the licensee of their observations.  The licensee reviewed the issue and
determined that the condition did require an operability review and compensatory actions to
address it pending further review.
 
Corrective Action(s):  The licensee performed an operability determination and establish
compensatory measures that established a reasonable expectation of operability pending
development of additional corrective actions.
 
Corrective Action Reference(s):  CR-2019-001119
 
25
Performance Assessment:
 
Performance Deficiency:  The licensees failure to follow the requirements of
Procedure STI-421.01 when a degraded condition was identified was a performance
deficiency.
 
Screening:  The inspectors determined the performance deficiency was more than minor
because it was associated with the Protection Against External Factors attribute of the
Mitigating Systems cornerstone.  It affected the cornerstone objective to ensure availability,
reliability, and capability of systems that respond to initiating events to prevent undesirable
consequences.  Specifically, the storage of materials without proper evaluations resulted in
the introduction of new and unanalyzed tornadic missiles.
 
Significance:  The inspectors assessed the significance of the finding using Appendix A,
Significance Determination of Reactor Inspection Findings for At - Power Situations.  Using
Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding
was of very low safety significance (Green) because:  (1) it was not a design deficiency; (2) it
did not represent a loss of system and/or function; (3) it did not represent an actual loss of
function of at least a single train for longer than its technical specification allowed outage
time; and (4) it did not result in the loss of a high safety significant non-technical specification
train.
Cross-cutting Aspect:  H.14 - Conservative Bias: Individuals use decision making-practices
that emphasize prudent choices over those that are simply allowable.  A proposed action is
determined to be safe in order to proceed, rather than unsafe in order to stop.  Specifically,
engineering failed to use decision making-practices that emphasize prudent choices over
those that are simply allowable.
Enforcement:
 
Violation:  Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, requires, in part, that activities affecting quality shall be prescribed by documented
instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be
accomplished in accordance with these instructions, procedures, and drawings.
 
Contrary to the above, from February 4-27, 2019, an activity affecting quality was not
accomplished in accordance procedures appropriate to the circumstances.  Specifically,
station personnel failed to notify the shift manager of an issue with material storage in the
protected area (as required by Station Procedure STI-421.01, Initiation of Issue Reports)
which required evaluations and compensatory actions for resolution.
Enforcement Action:  This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
 
26
Failure to Perform Safety Evaluations in Accordance with 10 CFR 50.59 
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000445; 05000446/2019001-
06 
Closed 
[H.9] - Training
71111.17T
The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion V,
Instructions, Procedures, and Drawings, (with four examples) in which the licensee failed to
complete 50.59 evaluations as required by station procedures.
Description:  The inspectors identified four examples where the licensee failed to perform
50.59 evaluations as required by procedures and guidance specified in STA-707, 
10 CFR 50.59 and 10 CFR 72.48 Reviews, Revision 21. 
Example 1.  EV-CR-2017-004743-2, Blow Down the 1-01 Instrument Air Receiver
In the screen for the compensatory measure to blow down the 1-01 air receiver once per shift,
question 1 of the screening was, Does the proposed activity involve a change to an SSC that
adversely affects an UFSAR described design function?  The preparer answered the
question No; the explanation provided had the following statements: The activity is a
Compensatory Measure to blow down the 1-01 Instrument Air Receiver once per shift using
1CI-0012 to remove water from the receiver.  The drip trap (CP1- CIMEDT-01) that performs
the automatic drain will be repaired IAW WO 5474911.  This statement indicates that an
automatic function was replaced with a manual function. 
The vendor manual, AP-0293-B, Ingersoll-Rand Compressor Accessories, dated April 1976,
provides the following guidance on page 7 for liquid carryover, It is important that interstage
separators be drained regularly and we are of the opinion that manual drainage at specified
intervals with the fact of drainage logged, is the proper method, particularly at higher
pressures.  Automatic traps, if used, should have a bypass piped for visual observation and
check on trap operation - the check should be made at stated intervals and the results
logged.  Page 12 of the manual provides guidance that drainage of the receiver following the
aftercooler should be drained at least once per shift.
CPNPP 50-59 RM-6, "CPNPP 50.59 Resource Manual," Revision 6, requires that an
evaluation be performed if an automatic function is replaced with a manual action.  The
preparer and reviewer failed to ensure the appropriate Applicability
Determination/screen/evaluation was performed and the corresponding Applicability
Determination/screen/evaluation form was completed in accordance with guidance provided
in CPNPP 50-59 RM-6.  Screening guidance would require this change to be evaluated prior
to changing from an automatic to a manual function.
Example 2.  EV-CR-2018-007384 RCS Pressure Boundary Leakage Test
This document was to perform a 50.59 review for changes to Procedure OPT-612B, RCS
Pressure Boundary Leakage Test for Loop 1 Cold Leg Injection Valves, and
Procedure OPT 613B, RCS Pressure Boundary Leakage Test for Loop 2 Cold Leg Injection
Valves, to allow the performance of reactor coolant system pressure boundary leakage test
for safety injection loops 1 and 2.  The licensee had attempted to perform a flush of the
residual heat removal system while in Mode 1, an evolution normally performed in Modes 3,
 
27
4, or 5. Inadequate procedure changes and review of the planned process resulted in forward
flow through valves 2-8956A and B.  This placed the unit in a 24-hour LCO to complete
Surveillance Requirement 3.4.14 for valves 2-8956 A and B.  Procedures OPT-612B and
OPT-613B needed to be revised to allow performance of this surveillance in Mode 1.  The
activity required component manipulations that isolated one safety injection accumulator and
rendered one train of residual heat removal inoperable in order to perform the leak check.  A
threaded pipe cap was removed and various normally closed valves were opened to allow
connection of the test rig.  The screener and reviewer failed to recognize that these actions
resulted in an "adverse effect" on the plant. 
CPNPP 50-59-RM6, Section 5.2.2, states, in part, changes that have an adverse effect are
required to be evaluated under 10 CFR 50.59 because they have the potential to increase the
likelihood of malfunctions, increase consequences, create new accidents, or otherwise meet
the 10 CFR 50.59 evaluation criteria.
CPNPP 50-59-RM6, Section 5.2.1 states, Items to Consider When Deciding Whether an Item
is a Change to the Facility:  Does the activity decrease the reliability of an SSC design
function, including either functions whose failure would initiate a transient/accident or
functions that are relied upon for mitigation?  Does the activity reduce existing redundancy,
diversity, or defense-in-depth? 
The screener and reviewer failed to recognize that, even though technical specifications allow
operation with one safety injection accumulator isolated and one train of residual heat
removal inoperable, this resulted in a reduction in the existing redundancy, diversity, and
defense-in-depth that required the performance of an evaluation.
Example 3.  Procedure Change to SOP-102B
Section 1 of the screen for the change to SOP-102B, Residual Heat Removal System,
Revision 15, provided the following description in the change justification section: "Modified
Section 5.2 to allow flushing of the RHR System to the RHUT (ref AI-CR-2018-007381-4),
deleted "Intentionally Left Blank" Pages 3&4 of Attachment 4. Re-sequenced Table of
Contents to reflect new page numbering. Added new prerequisite to Section 2.3 to clarify
intent of Section 5.11 and moved 2.3 to previous page."  The technical reviewer answered
yes to the question: If change is editorial, THEN circle or mark "YES."  Editorial changes, as
limited by STA-202, Attachment 8.F, do not require Administrative Review, Technical Review,
NSR, AD, 50.59 Review or 72.48 Review.
The procedure change (in Section 5.2 to allow flushing of the RHR system) actually
manipulated valves in the safety injection system to isolate the safety injection accumulators
based on lessons learned when the licensee originally attempted to flush the residual heat
removal system while in Mode 1.  The licensee had failed to recognize that the initial
conditions assumed in Procedure SOP-102B had the safety injection accumulators isolated. 
In Mode 1, the safety injection accumulators were in service, and the attempted flush of the
residual heat removal system resulted in flow from the accumulators.  The purpose of the
procedure modification was to isolate the safety injection accumulator to allow a partial flush
of the residual heat removal system.  The preparer, reviewer, and technical reviewer all failed
to identify this aspect of the procedure change.  As a result, the adverse effect on the plant, a
reduction in redundancy to the safety injection system, was not identified, and therefore the
required 10 CFR 50.59 evaluation was not performed. 
 
28
Example 4.  EV-2002-002026-01-00 Bladder Equivalency Evaluation
On May 28, 2002, the licensee performed an equivalency evaluation for replacement
diaphragms for the reactor make up water storage tanks, EV-2002-002026-01-00.  In the
evaluation the licensee identified that the new diaphragm was manufactured with  a material
that has a specific gravity greater than 1.0 which will make it heavier than the water in the
tank, and consequently material which tears or breaks off from the diaphragm will sink into
the tank and potentially into the pump suction, which could cause the pump to malfunction. 
The licensee determined that this was an equivalent change by crediting proper maintenance
and inspection to ensure that a failure of the new material does not occur.
Inspectors determined that this was not an equivalent change because the new diaphragm
introduced the potential for a new adverse effect (bladder failure could result in material
sinking and clogging pump suction) and should have been evaluated.  CPNPP 50-59-RM6 , 
Section 5.2.2 states in part, changes that have an adverse effect are required to be evaluated
under 10 CFR 50.59 because they have the potential to increase the likelihood of
malfunctions, increase consequences, create new accidents, or otherwise meet the 10 CFR
50.59 evaluation criteria.
 
Corrective Action(s):  The licensee entered these issues into the corrective action program.
 
Corrective Action Reference(s):  IR-2019-001271, IR-2019-001317, IR-2019-001428, 
IR-2019-001430
Performance Assessment:
 
Performance Deficiency:  The inspectors determined that not conducting required 
10 CFR 50.59 evaluations was a performance deficiency within the licensee's ability to
foresee and correct.  Specifically, the licensee failed to perform 10 CFR 50.59 evaluations for
the compensatory measure for the instrument air system, the procedure change for the
reactor coolant system pressure boundary leakage test for safety injection loops 1 and 2, the
procedure change for the residual heat removal system flush, and replacement diaphragms
for the reactor make up water storage tanks.
 
Screening:  The inspectors determined the performance deficiency was more than minor
because it was associated with the Human Performance attribute of the Mitigating Systems
Cornerstone and adversely impacted the cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent undesirable
consequences. 
 
Significance:  The inspectors assessed the significance of the finding using Appendix A,
Significance Determination of Reactor Inspection Findings for At - Power Situations.  The
inspectors assessed the significance of the finding using Inspection Manual Chapter 0609.04,
and Inspection Manual Chapter 0609, Appendix A, Exhibit 2. The inspectors determined that
this finding was of very low safety significance (Green), because the finding did not represent
a loss of the emergency core cooling system or the instrument air system safety function, did
not result in any loss of function beyond the technical specification-allowed outage time, and
did not result in the loss of any non-technical specification trains that were designated as high
safety-significance in accordance with the licensees maintenance rule program.
 
29
Cross-cutting Aspect:  H.9 - Training:  The organization provides training and ensures
knowledge transfer to maintain a knowledgeable, technically competent workforce and instill
nuclear safety values.  Specifically, the licensee failed to provide training to maintain a
knowledgeable, technically sound workforce and instill nuclear safety values when
implementing the change process.
Enforcement:
 
Violation:  Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, states Activities affecting quality shall be prescribed by documented instructions,
procedures, or drawings, of a type appropriate to the circumstances and shall be
accomplished in accordance with these instructions, procedures, or drawings.  Contrary to the
above, from May 2002, to February 2019, the team identified four examples where the
licensee failed to follow the requirements of Procedure CPNPP 50.59-RM6, "CPNPP 50.59
Resource Manual," Revision 6.  The procedure required a 10CFR 50.59 evaluation to be
performed if an activity reduces existing redundancy, diversity, or defense in depth or if an
automatic function is replaced with a manual action.  Specifically, the licensee implemented
manual compensatory actions when the automatic trap for the instrument air system failed,
made procedure changes that reduced the redundancy, diversity, reliability, and defense-in-
depth of the emergency core cooling systems, and installed new material in the plant with a
different adverse effect without performing 10 CFR 50.59 evaluations as required.
Enforcement Action:  This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Inadequate Maintenance Instructions Result in Loss of Assessment Capability
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Emergency
Preparedness
Green
NCV 05000445; 05000446/2019001-07 
Closed 
[H.8] -
Procedure
Adherence
71152
The inspectors reviewed a self-revealed Green, non-citied violation of 10 CFR 50,
Appendix B, Criterion V, "Instruction, Procedures, and Drawings", that occurred due to
inadequate maintenance instructions for safety-related radiation monitors which resulted in a
major loss of assessment capability.
Description:  On December 5, 2017, the licensee was performing maintenance on the control
room south ventilation intake radiation monitor under Work Order (WO) 5063234 when they
received audible and visible alarms in the control room indicating a loss of multiple radiation
monitors.  The crew evaluated the indications and determined a major loss of assessment
capability occurred due to the unplanned loss of the main steam line radiation monitors for
steam lines 1 and 3, and the station service water (SSW) radiation monitors.  The loss of
these radiation monitors impacted emergency action levels for radiation effluent.  This event
was reported to the NRC as Event Report No. 53105.
The inspectors reviewed the circumstances of this event including the licensees evaluation
and corrective actions.  The licensees radiation monitoring system consists of four
communication loops of 20 to 30 radiation monitors each.  The loops pass inputs via each
successive monitor to the plant computer system, which then provides required indications to
the control room and emergency response facilities (ERFs).  The licensee determined that the
loss of the affected radiation monitors was due to taking the control room south ventilation
 
30
intake radiation monitor out of service without first installing jumpers in the communication
loop to bypass the monitor.  This resulted in a failure of all other monitors in the affected loop
to provide indication to the plant computer system.
The inadequate maintenance resulted in the simultaneous communications failure of
approximately 27 radiation monitors.  In addition to the monitors that met the criteria for the
report, the inspectors noted the following other monitors that affected emergency
classification:
*
Unit 1 main steam line radiation monitors for main steam lines 1 and 3
*
both Unit 1 SSW radiation monitors and all Unit 1 component cooling water radiation
monitors, their credited backup for the SSW monitors
*
the Unit 1 failed fuel monitor
*
all Unit 1 refueling cavity monitors
*
the Unit 1 containment radiation monitors for particulate, iodine, and gaseous activity
*
the fuel building vent exhaust monitor
The licensee implemented compensatory measures for the affected monitors while restoring
them to service.  The main steam line radiation monitors affected the ability to declare a
General Emergency for high steam line radiation, but the licensee determined that a General
Emergency declaration could have been made using other emergency action levels.  The
inspectors did not identify any concerns with the licensees conclusion regarding emergency
classification.
The inspectors determined that the workers did not install the jumpers because WO 5063234
did not contain instructions to install the jumpers.  The licensee had relied on the knowledge
of a few experienced technicians who were aware that the jumpers needed to be installed
prior to removing a monitor from service.  However, the workers performing WO 5063234 on
the control room south ventilation intake radiation monitor on December 5 were not aware of
the need to install jumpers.
The inspectors determined that licensee Procedure STI-606.03, Work Planning, Section 6.2
requires that work packages identify where jumpers need to be installed.  The inspectors
concluded that the work instructions in WO 5063234 were inadequate.  The control room
south ventilation intake radiation monitor is safety-related, and therefore, the work instructions
were quality related instructions.
 
Corrective Action(s):  The licensee stopped maintenance, implemented compensatory
measures, and restored the monitors to service.
 
Corrective Action Reference(s):  CR-2019-002535
 
31
Performance Assessment:
 
Performance Deficiency:  The failure to prescribe adequate work instructions for a quality
related activity is a performance deficiency.
 
Screening:  The inspectors determined the performance deficiency was more than minor
because it was associated with the facilities and equipment attribute of the Emergency
Preparedness Cornerstone.  It adversely affected the cornerstone objective to ensure that the
licensee is capable of implementing adequate measures to protect the health and safety of
the public in the event of a radiological emergency.  Specifically, it resulted in the failure of
multiple pieces of equipment credited for maintaining the licensees emergency plan with
respect to emergency planning standard four, which requires a standard emergency
classification and action level scheme to be in use.
 
Significance:  The inspectors assessed the significance of the finding using Appendix B,
Emergency Preparedness SDP.  Using table 5.4-1, Significance Examples
Section 50.47(b)(4), the finding was determined to be of very low safety significance (Green)
because it was not a degraded risk significant planning standard function.  The planning
standard function was not degraded because, although an emergency action level (EAL) was
rendered ineffective such that a General Emergency would not have been declared for a
particular off-normal event, other EALs could have been used to make an appropriate
declaration.
Cross-cutting Aspect:  H.8 - Procedure Adherence: Individuals follow processes, procedures,
and work instructions.  Specifically, individuals did not follow the work planning procedure
when preparing work instructions for maintenance on the radiation monitors.
Enforcement:
 
Violation:  Title 10 CFR 50, Appendix B, Criterion V, "Instruction, Procedures, and Drawings,"
requires in part that activities affecting quality shall be prescribed by documented instructions
of a type appropriate to the circumstances. Contrary to the above, on December 5, 2017, the
licensee failed to prescribe activities affecting quality by documented instructions of a type
appropriate to the circumstances. Specifically, the licensee prescribed maintenance on a
safety-related radiation monitor with instructions that did not identify jumpers required to
maintain the function of the radiation monitoring system.
Enforcement Action:  This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Establish Adequate Procedural Guidance for Flushing Lithium at Power
Cornerstone
Significance
Cross-cutting
Aspect
Report
Section
Mitigating
Systems
Green
NCV 05000446/2019001-08 
Closed
[H.11] -
Challenge the
Unknown
71152
The inspectors reviewed a Green, self-revealing non-cited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the
licensees failure to establish an adequate procedure for flushing lithium from the residual
heat removal system.  This resulted in safety injection Accumulators 2-01 and 2-02 discharge
to the safety injection test header causing level drops in both accumulators, and
 
32
Accumulator 2-01 pressure dropped to below the operability limit resulting in an unplanned
component inoperability.
Description:  On November 2, 2018, with Unit 2 in Mode 1 operations the licensee performed
an evolution to flush lithium from the residual heat removal system.  The licensee used
Station Procedure SOP-102A, Residual Heat Removal System, Revision 20, Section 5.11, to
perform this evolution.  During the flush safety injection Accumulators 2-01 and 2-02 levels
dropped by 6 percent due to the accumulators discharging to the safety injection test header,
and Accumulator 2-01s pressure dropped to below the operability limit resulting in an
unplanned component inoperability.  Operators stopped the activity and restored level and
pressure in the accumulators.  Condition Report CR-2018-007381 was written to capture the
issue in the corrective action program.
During the licensees investigation of the event it was determined that Procedure SOP-102A,
section 5.11, was not the correct procedure for this evolution because it was not intended for
use in the mode of operation.  The licensee identified two causes for why an incorrect
procedure was used; inadequate coordination and incorrect assumptions.  Inadequate
coordination because operations, chemistry and engineering had used an informal selection
process which lacked rigor when selecting a procedure to perform an infrequently performed
task, and this resulted in no further challenge or verifications of the adequacy of this
procedure.  The licensee also identified that the work scheduling process does not require
operations procedures to be reviewed for impact.  Inadequate assumptions because of the
belief by operations, chemistry and engineering that procedure SOP-102A provided
appropriate instructions for the at-power lithium flush. 
Inspectors reviewed the licensees evaluation and concluded that it identified reasonable
causes and adequately addressed the identified causes.
 
Corrective Action(s):  The licensee immediately stopped the activity, refilled and
re-pressurized the safety injection accumulators.  Subsequent corrective actions were to
revise the work control process to require formal reviews for infrequently performed
non-repetitive activities.
 
Corrective Action Reference(s):  CR-2018-007381
Performance Assessment:
 
Performance Deficiency:  The licensees failure to establish an adequate procedure for
flushing lithium from the residual heat removal system was a performance deficiency.
 
Screening:  The inspectors determined the performance deficiency was more than minor
because it was associated with the equipment performance attribute of the Mitigating
Systems Cornerstone.  It adversely affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences.  Specifically, the use of an inadequate procedure for flushing
lithium resulted in an inoperable safety injection accumulator. 
 
Significance:  The inspectors assessed the significance of the finding using Appendix A,
Significance Determination of Reactor Inspection Findings for At - Power Situations.  Using
Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding
was of very low safety significance (Green) because:  (1) it was not a design deficiency; (2) it
did not represent a loss of system and/or function; (3) it did not represent an actual loss of
 
33
function of at least a single train for longer than its technical specification allowed outage
time; and (4) it did not result in the loss of a high safety significant non-technical specification
train.
Cross-cutting Aspect:  H.11 - Challenge the Unknown: Individuals stop when faced with
uncertain conditions. Risks are evaluated and managed before proceeding.  Specifically,
station personnel failed to stop when faced with uncertain conditions and ensure that risks
were evaluated and managed before proceeding.
Enforcement:
 
Violation:  Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings requires, in part, that activities affecting quality shall be prescribed by documented
instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be
accomplished in accordance with these instructions, procedures, and drawings.
Contrary to the above, on November 2, 2018, an activity affecting quality was not prescribed
by documented instructions, procedures, or drawings, of a type appropriate to the
circumstances.  Specifically, Station Procedure SOP-102A, Residual Heat Removal System,
Revision 20, Section 5.11, provided inadequate guidance for flushing lithium from the residual
heat removal system with the reactor in Mode 1 operation.
Enforcement Action:  This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Licensee-Identified Non-Cited Violation
71111.18
This violation of very low safety significance was identified by the licensee and has been
entered into the licensee corrective action program and is being treated as a non-cited
violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation:  10 CFR Part 50, Appendix B, Criterion III, Design Control, requires in part that
measures shall be established to assure that applicable regulatory requirements and the
design basis are correctly translated into specifications, drawings, procedures, and
instructions.  Contrary to the above, from initial construction to December 2018, the licensee
failed to correctly translate the design basis into specifications and procedures.  Specifically,
the licensee failed to ensure the design basis for nitrogen accumulator pressure for the
pressurizer power operated relief valves (PORV) was correctly translated into the
specification for minimum allowable pressure, resulting in a non-conservative low pressure
alarm setpoint.  As a result, for a period of approximately 30 hours, one Unit 1 PORV would
not have been able to cycle for the required number of operations to mitigate an overpressure
event when required.
Significance:  Green.
Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings,
dated October 7, 2016, Inspection Manual Chapter 0609, Appendix G, Shutdown Operations
Significance Determination Process, dated May 9, 2014, and Appendix G Attachment 1,
Phase 1 Initial Screening and Characterization of Findings, Exhibit 4, Barrier Integrity
Screening Questions, the inspectors determined the finding affected the Barrier Integrity
Cornerstone and required a detailed risk evaluation because the finding involved the
unavailability of a PORV during low temperature overpressure (LTOP) operations.
 
34
A senior risk analyst performed a bounding detailed risk evaluation and assumed that the
PORV not being able to cycle the full credited amount of times prevented the PORV from
fulfilling its LTOP system function.  The analyst used the frequency estimate for overpressure
excursion events from NUREG-0933, Resolution of Generic Safety Issues:  Issue 94: 
Additional Low Temperature Overpressure Protection for Light Water Reactors, to estimate
the initiating event frequency.  Other influential assumptions used by the senior reactor
analyst included an exposure time of approximately 30 hours and that the licensee
maintained the availability of a single additional relief valve (with its associated failure rate
estimated from the 2016 data update to NUREG/CR-6928, Industry-Average Performance
for Components and Initiating Events at U.S. Commercial Nuclear Power Plants) with
capability sufficient to mitigate an LTOP event as described in the final safety analysis report.
Using these assumptions, the senior reactor analyst determined that a bounding increase in
core damage frequency for this issue was 8.9E-8 per year and was, therefore, of very low
safety significance (Green).
Corrective Action Reference(s):CR-2018-008757
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
*
On February 8, 2019, the inspector presented the Evaluations of Changes, Tests and
Experiments inspection results to Mr. Tom McCool and other members of the licensee
staff.
*
On February 13, 2019, the inspector presented the Evaluations of Changes, Tests and
Experiments inspection results to Mr. Tim Hope and other members of the licensee staff.
*
On April 2, 2019, the inspector presented the quarterly resident inspector inspection
results to Steven Sewell and other members of the licensee staff.
 
35
DOCUMENTS REVIEWED
71111.04 - Equipment Alignment
Condition Reports
CR-2000-000142
CR-2017-011443
CR-2018-008300
CR-2019-000653
CR-2019-000672
CR-2019-002533
TR-2017-011236
TR-2017-011749
Procedures
Number
Title
Revision
STI-600.01
Protecting Plant Equipment and Sensitive Equipment Controls
1
SOP-605A
125 VDC Switchgear and Distribution Systems, Batteries and
Battery Chargers
12
Drawings 
Number
Title
Revision
E1-0020 Sh. K
125V DC One Line Diagram
CP-24
E1-0020 Sh. L
125V DC One Line Diagram
CP-23
Miscellaneous 
Documents
Number
Title
Revision
or Date
FDA-2000-00142
Final Design Authorization
02
Calculations
Number
Title
Revision
or Date
MM-90-2671
Technical Evaluation
11/28/1990
71111.12 - Maintenance Effectiveness
Condition Reports
CR-2015-008236
CR-2016-000049
CR-2016-007907
CR-2017-000594
CR-2017-0010477
CR-2017-004704
CR-2018-003921
CR-2018-003945
CR-2018-004761
CR-2019-002622
TR-2016-000169
TR-2016-002742
TR-2016-008960
TR-2018-004761
Work Orders
5380904
5517474
5144575
5220567
5331282
5347463
5377428
 
36
Miscellaneous 
Documents
Number
Title
Revision
or Date
DBD-EE-044
DC Power Systems
28
DBD-EE-043
118V AC Uninterruptible Power Supply System
14
71111.13 - Maintenance Risk and Emergent Work
Condition Reports
TR-2019-000001
Work Orders
5692097
5705947
Procedures
Number
Title
Revision
STA-696
Hazard Barrier Controls
3
71111.17T - Evaluations of Changes, Tests and Experiments
Condition Reports
CR-2017-005150
CR-2017-012952
CR-2018-007381
CR-2018-007384
TR-2019-001160
CR-2019-001179
CR-2019-001200
CR-2019-001240
CR-2019-001249
CR-2019-001271
IR-2019-001316
IR-2019-001317
IR-2019-001318
IR-2019-001428
IR-2019-001430
TR-2017-007959
TR-2018-004675
Work Orders
5352698
5510637
5510645
5510646
5510663
5510664
5510665
5510666
5510588
5510605
5510610
5510611
5510615
5510633
5510634
5510635
5510636
5351262
5351266
5351253
5383860
5351257
5351268
5346400
5284599
5435995
391842
3905518
Procedures
Number
Title
Revision
ODA-401
Control of Annunciators, Instruments, and Protective Relays
11
OPT-612B
RCS Pressure Boundary Leakage Test FOR LOOP 1 CL
INJECTION VALVES
3
 
37
Procedures
Number
Title
Revision
OPT-613B
RCS PRESSURE BOUNDARY LEAKAGE TEST FOR LOOP 2 CL
INJECTION VALVES
3
SOP-102B
RESIDUAL HEAT REMOVAL SYSTEM
15
SOP-609A
DIESEL GENERATOR SYSTEM
21
STA-602
TEMPORARY MODIFICATIONS AND TRANSIENT EQUIPMENT
PLACEMENTS
18
STA-707
10CFR50.59 AND 10CFR72.48 REVIEWS
21
STA-738
FIRE PROTECTION SYSTEMS/EQUIPMENT IMPAIRMENTS
7
STI-422.02
COMPENSATORY ACTIONS & TRANSIENT EQUIPMENT
PLACEMENTS
1
STI-707.04
10CFR50.59 AND 10CFR72.48 REVIEWS APPLICABILITY
DETERMINATIONS
1
TDM-401B
TURBINE/GENERATOR LIMIT CURVES
5
ABN-104
RESIDUAL HEAT REMOVAL SYSTEM MALFUNCTION
9
ABN-104
RESIDUAL HEAT REMOVAL SYSTEM MALFUNCTION
8
ABN-402
MAIN GENERATOR MALFUNCTION
13
ALM-0031A
ALARM PROCEDURE 1-ALB-3A
8
ALM-0031B
ALARM PROCEDURE 2-ALB-3A
4
TDM-401B
TURBINE/GENERATOR LIMIT CURVES
5
Drawings 
Number
Title
Revision
M2-0235
FLOW DIAGRAM, SPENT FUEL POOL COOLING AND
CLEAN-UP SYSTEM
CP-17
M2-2225
INSTRUMENTATION AND CONTROL DIAGRAM, FIRE
DETECTION/PROTECTION SYSTEM CHANNELS 4100,
4102, 4103, 4111
CP-2
COMANCHE 004
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC
CONTROL PANEL CP1/2-EPIBHX-01P
CP-3
COMANCHE 015
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL DAMPER
CONTROL PANEL
CP-1
COMANCHE 006
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC
CONTROL PANEL CP1/2-EPIBHX-01P
CP-3
COMANCHE 008
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC
CONTROL PANEL CP1/2-EPIBHX-01P
CP-2
COMANCHE 010
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC
CONTROL PANEL CP1/2-EPIBHX-01P
CP-2
 
38
Drawings 
Number
Title
Revision
COMANCHE 012
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL AHUA/AHUB
FAN STRTER PANELS CP1/2-EPIBMC-01 AND CP1/2-
EPIBMC-02
CP-3
COMANCHE 014
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL
CP-2
COMANCHE 011
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC
CONTROL PANEL CP1/2-EPIBHX-01P
CP-2
COMANCHE 013A UNIT 1 AND UNIT 2 ISOPHASE BUS CONTROL INTERNAL
WIRING DIAGRAM
CP-3
2323-A1-0507
PRIMARY PLANT AUXILIARY ELECTRICAL AND CONTROL
BUILDING FLOOR PLAN
CP-1
COMANCHE 002
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL INTERIOR
PANEL LAYOUT
CP-2
COMANCHE 003
UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC
CONTROL PANEL CP1/2-EPIBHX-01P
CP-2
COMANCHE 015A UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL DAMPER
CONTROL PANEL
CP-1
COMANCHE 015B UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL DAMPER
CONTROL PANEL
CP-1
M1-0260
FLOW DIAGRAM -  RESIDUAL HEAT REMOVAL SYSTEM
CP-37
M1-0261
FLOW DIAGRAM - SAFETY INJECTION SYSTEM SHEET 1
0F 5
CP-24
M1-0216
FLOW DIAGRAM - COMPRESSED AIR SYSTEM
CP-45
M1-0250
FLOW DIAGRAM - REACTOR COOLANT SYSTEM
CP-34
M1-2300
INSTRUMENTATION AND CONTROL DIAGRAM,
VENTILATION - CONTAINMENT, CHANNEL 5400/5403
CP-7
Miscellaneous 
Documents
Number
Title
Revision
or Date
EVAL-2018-007
CPNPP Nuclear Oversight Audit Report - CONFIGURATION
& DESIGN CONTROL
08/16/2018
DBD-ME-013
Design Basis Document - Containment Isolation System
25
RIR-22946OCR
Receipt Inspection Report
10/06/1983
CP-201700626
Comanche Peak Nuclear Power Plant, Docket Nos. 50-445
and 50-446 and 72-74, 10CFR50.59 Evaluation Summary
Report 020, 10CFR72.48 Evaluation Summary Report 005,
and Commitment Material Change Evaluation Report 014
12/05/2017
DBD-ME-014-02
Design Basis Document - Generator and Exciter System
21
 
39
Vendor
Documents
Number
Title
Revision
or Date
CP-201600573
EVALUATION OF COMANCHE PEAK UNIT 1 CLASS 2 TO
CLASS 1VALVE UPGRADES
05/31/2016
CP1/CP2-
EPIBHX-01E/01F
Damper Position Monitor
08/16/2016
CT-27331
MISSILE PROBABILITY ANALYSIS METHODOLOGY
FOR LUMINANT GENERATION COMPANY LLC,
COMANCHE PEAK UNITS 1 & 2 WITH SIEMENS
RETROFIT TURBINES
8
VDRT-5472306
Unit 2 Generator Stator Damage - Monitoring Installation
Plan
07/21/2017
WPT-18067
Transmittal of LTR-SEE-17-189, Flow Evaluation of Forced
Forward Flow through the Residual Heat Removal Pumps at
Comanche Peak Units 1 & 2
10/03/2017
Calculations
Number
Title
Revision
MEB-391
Minimum Allowable Service Water Flow to Diesel Generators
5
ME-CA-0229-2188 Component Cooling Water Heater Exchanger Fowling Water
Analysis
8
71111.18 - Plant Modifications
Condition Reports
CR-2018-008757
Work Orders
5435249
5689179
Modifications 
Number
Title
Revision
FDA-2018-000119-01
Final Design Authorization
Calculations 
Number
Title
Revision
ME-CA-0000-3342
Air Accumulator Check Valve Leakage - Decay Rate,
Pressure, Time
3
 
40
71152 - Identification and Resolution of Problems
Condition Reports
CR-2017-013243
CR-2018-003808
CR-2019-002535
Work Orders
5540984
5063234
Procedures
Number
Title
Revision
STI-606.03
Work Planning
3
Miscellaneous
Documents
Number
Title
Revision
DBD-EE-023
Radiation Monitoring System
23
 
 
  ML19130A154
SUNSI Review
Complete
By:  RDA
ADAMS
Yes  No
Publicly Available
Non-Publicly Available
Non-Sensitive
Sensitive
Keyword: 
NRC-002
OFFICE
SRI/DRP/A
RI/DRP/A
DRS/EB1
DRS/EB2
DRS/OB
DRS/IPAT
NAME
JJosey
RKumana
VGaddy
GPick
GWerner
RKellar
SIGNATURE
/RA/
/RA/
/RA/
/RA/
/RA/ CCO for
/RA/
DATE
05/07/19
05/03/19
05/02/19
05/08/19
05/03/19
05/06/19
OFFICE
DRS/RCB
DNMS/RIB
SPE:DRP/A
BC/DRP/A
NAME
NMakris
GWarnick
RAlexander
MHaire
SIGNATURE
/RA/
/RA/
/RA/
/RA/
DATE
05/02/19
05/07/19
05/02/19
5/10/2019
}}

Latest revision as of 02:32, 5 January 2025

NRC Integrated Inspection Report 05000445/2019001 and 05000446/2019001
ML19130A154
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 05/10/2019
From: Mark Haire
NRC/RGN-IV/DRP/RPB-A
To: Peters K
Vistra Operations Company
References
IR 2019001
Download: ML19130A154 (41)
v • d • e


See also: IR 05000445/2019001

Text

May 10, 2019

Mr. Ken Peters, Senior Vice President

and Chief Nuclear Officer

VISTRA Operations Company, LLC

P.O. Box 1002

Glen Rose, TX 76043

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 - NRC

INTEGRATED INSPECTION REPORT 05000445/2019001 AND

05000446/2019001

Dear Mr. Peters:

On March 31, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection

at your Comanche Peak Nuclear Power Plant, Units 1 and 2. On April 2, 2019, the NRC

inspectors discussed the results of this inspection with Mr. Steven Sewell and other members of

your staff. The results of this inspection are documented in the enclosed report.

NRC inspectors documented seven findings of very low safety significance (Green) in this

report. These findings involved violations of NRC requirements. Additionally, NRC inspectors

documented one Severity Level IV violation with no associated finding. The NRC is treating

these violations as non-cited violations (NCV) consistent with Section 2.3.2.a of the

Enforcement Policy.

The inspectors also documented a licensee-identified violation which was determined to be of

very low safety significance in this report. The NRC is treating this violation as a non-cited

violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violations or significance or severity of the violations documented in this

inspection report, you should provide a response within 30 days of the date of this inspection

report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional

Administrator, Region IV; the Director, Office of Enforcement; and the NRC resident inspector at

the Comanche Peak Nuclear Power Plant.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a

response within 30 days of the date of this inspection report, with the basis for your

disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,

Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the

NRC resident inspector at the Comanche Peak Nuclear Power Plant.

K. Peters

2

2

This letter, its enclosure, and your response (if any) will be made available for public inspection

and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document

Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for

Withholding.

Sincerely,

/RA/

Mark S. Haire, Chief

Project Branch A

Division of Reactor Projects

Docket Nos. 50-445 and 50-446

License Nos. NPF-87 and NPF-89

Enclosure:

Inspection Report 05000445/2019001

and 05000446/2019001

3

U.S. NUCLEAR REGULATORY COMMISSION

Inspection Report

Docket Number(s):

05000445 and 05000446

License Number(s):

NPF-87 and NPF-89

Report Number(s):

05000445/2019001 and 05000446/2019001

Enterprise Identifier: I-2019-001-0011

Licensee:

Vistra Operations Company, LLC

Facility:

Comanche Peak Nuclear Power Plant, Units 1 and 2

Location:

Glen Rose, TX 76043

Inspection Dates:

January 1, 2019 to March 31, 2019

Inspectors:

W. Cullum, Reactor Inspector

R. Deese, Senior Reactor Analyst

J. Drake, Senior Reactor Inspector

J. Josey, Senior Resident Inspector

R. Kumana, Resident Inspector

W. Sifre, Senior Reactor Inspector

Approved By:

Mark S. Haire, Chief

Project Branch A

Division of Reactor Projects

4

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees

performance by conducting a Quarterly inspection at Comanche Peak Nuclear Power 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.

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: 71111.18.

List of Findings and Violations

Inadequate Corrective Actions for Failure to Ensure Containment Hatch Closure Capability

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Barrier Integrity

Green

NCV 05000445;05000446/2019001-01

Closed

[H.6] - Design

Margins

71111.04

The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B,

Criterion XVI, Corrective Actions, associated with the licensees failure to take adequate

corrective actions for an inadequate containment closure procedure. Specifically, in

December 2017, the NRC identified that the licensee's procedure for emergency closure of

the Unit 1 and 2 containment equipment hatches was inadequate, and the licensee failed to

take adequate actions to correct the issue prior to the next outage.

Failure to Evaluate a Change to the Facility DC Power System

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Not Applicable

NCV 05000445/2019001-02

Closed

Not Applicable

71111.04

The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.59 for the

licensees failure to obtain a license amendment or perform a written evaluation

demonstrating the basis for not obtaining a license amendment, prior to making a change to

the facility as described in the final safety analysis report.

Failure to Monitor or Demonstrate Control of Performance Under the Maintenance Rule

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000445;05000446/2019001-03

Closed

None

71111.12

The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(2), with

three examples, for failure to demonstrate effective control of performance of a maintenance

rule scoped system through appropriate preventive maintenance.

5

Failure to Control Hazard Barriers During Maintenance

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000445/2019001-04

Closed

[H.14] -

Conservative

Bias

71111.13

The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(4) for failure to

implement risk mitigating actions during diesel generator maintenance.

Failure to Follow Procedure When A Degraded Condition Was Identified

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000445;05000446/2019001-05

Closed

[H.14] -

Conservative

Bias

71111.15

The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to

follow the requirements of Station Procedure STI-421.01, Initiation of Issue Reports,

Revision 0. Specifically, station personnel failed to notify the shift manager of an issue with

material storage in the protected area. This issue required evaluations and compensatory

actions for resolution.

Failure to Perform Safety Evaluations in Accordance with 10 CFR 50.59

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000445;05000446/2019001-06

Closed

[H.9] - Training

71111.17T

The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, (with four examples) in which the licensee failed to

complete 50.59 evaluations as required by station procedures.

Inadequate Maintenance Instructions Result in Loss of Assessment Capability

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Emergency

Preparedness

Green

NCV 05000445;05000446/2019001-07

Closed

[H.8] -

Procedure

Adherence

71152

The inspectors reviewed a self-revealed Green, non-citied violation of 10 CFR 50,

Appendix B, Criterion V, "Instruction, Procedures, and Drawings," that occurred due to

inadequate maintenance instructions for safety-related radiation monitors that resulted in a

major loss of assessment capability.

6

Failure to Establish Adequate Procedural Guidance for Flushing Lithium at Power

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000446/2019001-08

Closed

[H.11] -

Challenge the

Unknown

71152

The inspectors reviewed a Green, self-revealed non-cited violation of 10 CFR Part 50,

Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the

licensees failure to establish an adequate procedure for flushing lithium from the residual

heat removal system. This resulted in safety injection accumulators 2-01 and 2-02 discharge

to the safety injection test header causing level drops in both accumulators and

accumulator 2-01 pressure dropped to below the operability limit resulting in an unplanned

component inoperability.

Additional Tracking Items

Type

Issue Number

Title

Report

Section

Status

NOV 05000446/2018011-01

Failure to Maintain a Quality

Record Complete and Accurate

in All Material Respects

92702

Closed

LER 05000446/2018-001-00 Unit 2 Automatic Reactor Trip

Due to Turbine Trip, on

March 19, 2019

71153

Closed

7

PLANT STATUS

Unit 1 began the inspection period at or near rated thermal power. On February 1, 2019, the

unit was down powered to 64 percent for turbine testing. The unit was returned to rated thermal

power the same day. On March 22, 2019, the unit began power coast down to a refueling

outage, ending the inspection period at 92 percent power.

Unit 2 began the inspection period in a refueling outage. On January 14, 2019, the unit began a

reactor startup. The unit shut down on January 15, 2019, due to a main turbine primary water

leak. On January 18, 2019, the unit began a reactor startup and reached rated thermal power

on January 22, 2019. On March 2, 2019, the unit was tripped due to a failure of a main

feedwater isolation valve. The unit began a reactor startup the same day and reached rated

thermal power on March 4, 2019. The unit remained at or near rated thermal power 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 2515, Light-Water Reactor Inspection Program -

Operations Phase. The inspectors performed plant status activities described in Inspection

Manual Chapter 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

71111.01 - Adverse Weather Protection

Impending Severe Weather Sample (IP Section 03.03) (1 Sample)

The inspectors evaluated readiness for impending adverse weather conditions for severe

thunderstorms on March 13, 2019.

71111.04 - Equipment Alignment

Partial Walkdown (IP Section 02.01) (4 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following

systems/trains:

(1)

Unit 1, safety injection pump 1-01 while 1-02 was out of service for maintenance on

February 5, 2019

(2)

Unit 2, containment hatches on February 13, 2019

8

(3)

Units 1 and 2, common class-1E DC power on March 5, 2019

(4)

Units 1 and 2, seismic monitoring system on March 18, 2019

71111.05Q - Fire Protection

Quarterly Inspection (IP Section 03.01) (5 Samples)

The inspectors evaluated fire protection program implementation in the following selected

areas:

(1)

fire area 2CA, Unit 2 reactor building on January 9, 2019

(2)

fire zones TB201 and TB202, control room emergency lighting battery rooms on

January 14, 2019

(3)

fire zone 1SB2A, Unit 1 safety injection pump 1-01 on March 11, 2019

(4)

fire zone 2SB4, Unit 2 containment spray chemical add tank on March 13, 2019

(5)

fire zone SM157, stairwell in the southeast corner of the safeguards building on

March 26, 2019

71111.06 - Flood Protection Measures

Inspection Activities - Internal Flooding (IP Section 02.02a.) (1 Sample)

The inspectors evaluated internal flooding mitigation protections in the service water intake

structure on March 12, 2019.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)

(2 Samples)

(1)

The inspectors observed and evaluated licensed operator performance in the Control

Room during Unit 2 startup on January 14, 2019.

(2)

The inspectors observed and evaluated licensed operator performance in the Control

Room during Unit 2 startup on January 18, 2019.

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)

The inspectors observed and evaluated a simulator-based loss of coolant accident scenario

on March 27, 2019.

9

71111.12 - Maintenance Effectiveness

Routine Maintenance Effectiveness Inspection (IP Section 02.01) (3 Samples)

The inspectors evaluated the effectiveness of routine maintenance activities associated with

the following equipment and/or safety significant functions:

(1)

common low voltage power distribution failure to align to normal power supply on

February 28, 2019

(2)

Unit 1, battery charger and inverter failures which occurred in June 2018, on

February 28, 2019

(3)

service air check valve failure during surveillance testing on March 14, 2019

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (5 Samples)

The inspectors evaluated the risk assessments for the following planned and emergent work

activities:

(1)

Unit 1, risk mitigating actions during emergency diesel generator 1-01 lube oil fill on

January 17, 2019

(2)

Unit 1, risk mitigating actions while safety injection pump 1-02 was out of service on

February 5, 2019

(3)

Unit 1, risk assessment during sequencer undervoltage replacement on

February 13, 2019

(4)

Units 1 and 2, removal of service water pipe tunnel missile shield CPX-SWMEBB-01

on February 28, 2019

(5)

Units 1 and 2, risk mitigating actions with transformer XST2 unavailable on

March 29, 2019

71111.15 - Operability Determinations and Functionality Assessments

Sample Selection (IP Section 02.01) (5 Samples)

The inspectors evaluated the following operability determinations and functionality

assessments:

(1)

CR-2019-000324, Units 1 and 2, environmental qualification of steam generator

atmospheric relief valves on January 10, 2019

(2)

CR-2019-000456, Units 1 and 2, Electroswitch Part 21 relay issue on

January 14, 2019

10

(3)

TR-2019-001119, Units 1 and 2, tornado missile evaluation for equipment storage on

February 13, 2019

(4)

TR-2019-000805, Units 1 and 2, operations support center HVAC sensor failure on

February 14, 2019

(5)

CR-2019-002132, Unit 1, environmental qualification of service water valves with

teflon components on March 12, 2019

71111.17T - Evaluations of Changes, Tests, and Experiments

Sample Selection (IP Section 02.01) (35 Samples)

The inspectors reviewed the following evaluations (items 1 through 8), screenings, and/or

applicability determinations for 10 CFR 50.59 from September 30, 2016, to

January 14, 2019.

(1)

EV-CR-2016-001706-8, Revision1; FDA-2016-000025-01 temporary modification of

2RC-8054A to repair a leak on pressurizer 01 Pressure Transmitter.

(2)

AEV-CR-2016-005587-9; FDA-2016-000142-01, LDCR SA-2016-013 and

LDC R TR-2016-003, Missile Probability Analysis Revision.

(3)

EV-TR-2017-003173-5 ABN-104, Revision 9; PCN-9 addition of alternate residual

heat removal path and use of safety injection pump for core cooling in Mode 6.

(4)

EV-TR-2017-007959-13; Perform 50.59 Evaluation for FDA-2017-000106-02

Generator Repair Plan and 59SC-2017-000106-02.

(5)

EV-2014-013052-9; Modification to change the isolated phase bus cooling fans start

logic to provide seven out of eight dampers open requirement using digital

equipment.

(6)

EV-CR-2016-003267-10; FDA-2016-000075-01 Unit 1 pressurizer instrument

isolation valves class change (LDCR-SA-2016-010).

(7)

EV-TR-2018-004520-14; Evaluate operator action for isolation of faulted battery

charger from its battery per 50.59 screen EV-TR-2018-004520-13.

(8)

EV-CR-2017-004574-2; 59SC - STA-707-1 50.59 screen for 2RF16 changes to

procedures for reactor vessel head and upper internals lifts.

(9)

EV-TR-2015-006849-4; 59SC - Provide 50.59 SC to support DCP-17-000010 to input

FZ locations of raceways and equipment into GENESIS in support of

ME-CA-0000-1086 revision.

(10)

EV-TR-2018-004520-10; 59SC - Perform a 50.59 screen for a compensatory

measure to jumper battery cell.

11

(11)

EV-CR-2014-003412-18; 59SC - Perform 50.59 applicability determination and

screen for additional plugging for component cooling water heat exchanger 2-01 in

2RF14.

(12)

EV-TR-2018-003799-6; Perform 10CFR50.59 review of minor fuel design changes

documented in NF-TB-16-21.

(13)

EV-TR-2018-003700-2; Refer to the attached VDRT package which contains the

requested screen and complete VDRT-5608075 package for valve XWT-0634.

(14)

EV-TR-2018-000169-4; 50.59 screen for backseating of 1MS-0357, SG 1-03

blowdown downstream isolation valve.

(15)

EV-TR-2018-000198-1; Maintenance clearance placed for isolation of 1-LG-2706A

may exceed 90 days.

(16)

EV-TR-2018-000199-1; Maintenance clearance placed for diesel generator starting

compressor solenoid 1-SV-3422-1F may exceed 90 days.

(17)

EV-TR-2018-000600-1; Shift manager clearance placed to isolate TBX-CSFLSI-01

seal water injection filter 01.

(18)

EV-CR-2016-008147-3; Compensatory action of installing scaffolding for medium

energy line break (MELB) barrier.

(19)

EV-CR-2017-007829-1; 59SC - Compensatory actions to install temporary equipment

for flow measurement.

(20)

EV-CR-2017-010212-1; 59SC - Shift manager clearance CP17-0913 due to

feedpump deluge valve not resetting.

(21)

EV-CR-2017-012952-28; 59SC - Shift manager clearance to remove fuses 2-

KXA/0746 and 2-KXB/0746.

(22)

EV-CR-2018-004743-2; 59SC - Compensatory action to blow down the receiver once

per shift.

(23)

EV-TR-2016-005840-10; 59SC - VDRT-5575487 Which includes vendor final

acceptance tests for open phase protection equipment for XST1.

(24)

EV-TR-2017-000041-32; 59SC - VDRT-5397434, Fuel transfer system transfer cart

weldment.

(25)

EV-TR-2017-003173-4; 59SC - Review for revision to ABN-104 based on

EV-TR-2017-003173-3 for loss of residual heat removal events.

(26)

EV-CR-2018-002390-5; 59SC - Changes made under EV-CR-2018-002390-4.

(27)

EV-CR-2018-006758-1; 59SC - Screen for the compensatory action for average

containment temperature.

12

(28)

EV-CR-2018-007384-1; 59SC - Perform 50.59 screen changes to procedures

OPT-612B and OPT-613B.

(29)

EV-CR-2016-007812-1; 59SC - Perform a 10CFR50.59 Review per STA-707 to

update UFSAR Table 9.5-18 to specify tube plugging limit for diesel generator jacket

water coolers for Unit 1 and Unit 2.

(30)

EV-TR-2018-008391-16; 59SC - Perform a 10CFR50.59 Review per STA-707 to plug

tubes in the component cooling water heat exchangers.

(31)

EV-CR-2018-002189-2; 59SC - 50.59 screen for compensatory action to maintain

2-HV-2334A accumulator pressure above 2100psi.

(32)

EV-CR-2016-008215-20; 59SC - 50.59 review of compensatory measures to isolate

suction and discharge pressure indication on CT and SF pumps;

ref: EV-CR-2016-008215-19.

(33)

EV-TR-2016-009344-1; 59SC - Shift Manager Clearance CP16-1381 initiated to

maintain X-PV-3218A isolated following failure of a functional stroke; request a

50.59SC to determine impact on the plant.

(34)

EV-CR-2018-005954-3; 59SC - Seal injection filters housing bolts and potential

excessive torque specification VDRT-5655877.

(35)

EV-TR-2016-010572-2; 59SC - 59SC - Perform a 50.59 screen for hanging shift

manager clearance CP16-1614 on 2-HS-2802A for damage to upper journal bearings

on the motor for Circulating Water Pump Motor 2-03.

71111.18 - Plant Modifications

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

(2 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1)

Unit 2, pressurizer power operated relief valve accumulator pressure setpoint

modification on February 14, 2019

(2)

bladder addition to safety-related tanks on March 11, 2019

71111.19 - Post Maintenance Testing

Post Maintenance Test Sample (IP Section 03.01) (5 Samples)

The inspectors evaluated the following post maintenance tests:

(1)

Unit 2, diesel generator 2-02 following intercooler crack and jacket water repair on

February 12, 2019

(2)

Unit 2, pressurizer spray valve following actuator rebuild on February 20, 2019

13

(3)

Unit 1, diesel generator 1-01 following fuel injector torqueing on March 13, 2019

(4)

Unit 2, residual heat removal pump 2-02 following pump refurbishment on

March 19, 2019

(5)

Unit 2, auxiliary feedwater pump 2-01 following maintenance on March 20, 2019

71111.20 - Refueling and Other Outage Activities

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

The inspectors evaluated refueling outage 2RF17 activities from January 1, 2019, to

January 18, 2019, completing the sample for the refueling outage which started on

December 8, 2018 (see Inspection Report 05000445/2018004; 05000446/2018004 (ADAMS

Accession No. ML19042A345)). Specifically, the inspectors completed Inspection

Procedure 71111.20, Sections 03.01.d through e, during this inspection period.

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance tests:

Containment Isolation Valve (ISO) (IP Section 03.01) (1 Sample)

Unit 2, service air containment isolation valve test on March 7, 2019

Surveillance Testing (IP Section 03.01) (1 Sample)

Unit 2, OPT-601B auxiliary feedwater flow control valve accumulator pressure drop test on

March 26, 2019

OTHER ACTIVITIES - BASELINE

71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) (2 Samples)

(1)

Unit 1 from January 2018 through December 2018

(2)

Unit 2 from January 2018 through December 2018

IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02)

(2 Samples)

(1)

Unit 1 from January 2018 through December 2018

(2)

Unit 2 from January 2018 through December 2018

14

IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03)

(2 Samples)

(1)

Unit 1 from January 2018 through December 2018

(2)

Unit 2 from January 2018 through December 2018

71152 - Problem Identification and Resolution

Annual Follow-up of Selected Issues (IP Section 02.03) (2 Samples)

The inspectors reviewed the licensees implementation of its corrective action program

related to the following issues:

(1)

radiation monitor failures due to failure to install a jumper during maintenance on

February 28, 2019

(2)

safety injection accumulator discharge due to inadequate procedure on

March 29, 2019

71153 - Follow-up of Events and Notices of Enforcement Discretion

Event Report (IP Section 03.02) (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 05000446/2018-001-00, "Unit 2 Automatic Reactor Trip Due

to Turbine Trip," on March 19, 2019

The inspectors determined that it was not reasonable to foresee or correct the cause

discussed in the LER; therefore, no performance deficiency was identified. The inspectors

also concluded that no violation of NRC requirements occurred.

OTHER ACTIVITIES - TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL

92702 - Follow-up on Corrective Actions for Violations And Deviations

Follow-up - Corrective Actions - Violations and Deviations (1 Sample)

On March 28, 2019, the inspectors reviewed the licensees response to

NOV 05000446/2018011-01, "Failure to Maintain a Quality Record Complete and Accurate

in All Material Respects," and determined that the reason for the violation, corrective actions

taken and planned to address recurrence, and the date when full compliance will be

achieved for this violation is adequately addressed and captured on the docket.

15

INSPECTION RESULTS

Inadequate Corrective Actions for Failure to Ensure Containment Hatch Closure Capability

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Barrier Integrity

Green

NCV 05000445; 05000446/2019001-

01

Closed

[H.6] - Design

Margins

71111.04

The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B,

Criterion XVI, Corrective Actions, associated with the licensees failure to take adequate

corrective actions for an inadequate containment closure procedure. Specifically, in

December 2017, the NRC identified that the licensee's procedure for emergency closure of

the Units 1 and 2 containment equipment hatches was inadequate and the licensee failed to

take adequate actions to correct the issue prior to the next outage.

Description: In Inspection Report 2017-004, the NRC documented a non-cited violation for an

inadequate procedure, STI 600.01, "Protecting Plant Equipment and Sensitive Equipment

Controls." This procedure contained instructions for emergency closure of the containment

equipment hatch during times when the hatch was open, but the ability to close containment

was required. The inspectors observed that the bolting pattern and required torque that were

identified in the supporting engineering calculation were not incorporated into the procedure.

The licensees technical evaluation required four bolts spaced 90 degrees apart and torqued

to 30 percent preload values. The procedure did not require bolts to be evenly spaced and

only required the bolts to be snug tight, a licensee term implying full effort on the tool being

used. The licensee entered this into their corrective action program. Subsequently, the

licensee performed an evaluation to justify alternate bolt spacing patterns and revised the

procedure to include adequate bolting patterns. However, in their evaluation the licensee

stated that no torque requirement existed, and the requirement was only to hold the hatch in

place.

The inspectors observed the containment hatch closure training during Refueling

Outage 2RF17. The inspectors observed that the bolt patterns used conformed to the revised

procedure and evaluation, but that the hatch operators did not appear to apply any torque to

the bolts. When the inspectors asked about the bolts, the operators believed that there was

no requirement to apply any torque beyond that needed to hold the hatch in place.

The inspectors determined that by not applying any type of torque to the bolts, the licensee

was not verifying that the containment equipment hatch could be sealed. A seal is necessary

to ensure that a release of fission product radioactivity within containment will be restricted

from escaping to the environment in the event of a loss of decay heat removal event when the

reactor coolant system was open to the atmosphere.

The licensee performed another evaluation and concluded that the minimum torque required

to ensure a seal with four bolts was 144 ft-lbf. The licensee conducted additional training with

all hatch operators on the requirement to ensure a seal on the hatch. They also conducted a

demonstration with the assigned operators and concluded that the average operator applying

full effort would achieve greater than 150 ft-lbf.

16

Corrective Action(s): The licensee trained the operators on the requirement to ensure the

bolts were adequately torqued and verified through demonstration that the operators could

apply enough torque to ensure the hatch would be sealed.

Corrective Action Reference(s): CR-2018-008300, CR-2019-002533

Performance Assessment:

Performance Deficiency: The inability to assure containment closure during a postulated loss

of decay heat removal or fuel handling accident was a condition adverse to quality. The

failure to correct a condition adverse to quality is a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the SSC and barrier performance attribute of the Barrier

Integrity Cornerstone. It adversely affected the cornerstone objective to provide reasonable

assurance that physical design barriers (fuel cladding, reactor coolant system, and

containment) protect the public from radionuclide releases caused by accidents or events

because the finding represented a loss of reasonable assurance of the ability to close the

containment equipment hatch. Specifically, the failure to assure that personnel would

adequately torque the bolts on the hatch sufficient to establish a seal would, in an actual

event, result in a loss of the containment barrier.

Significance: The inspectors assessed the significance of the finding using Appendix H,

Containment Integrity SDP. Using Inspection Manual Chapter 0609, Attachment 04, Initial

Characterization of Findings, dated October 7, 2016, the inspectors determined the finding

was associated with the Barrier Integrity cornerstone. Using Inspection Manual

Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination

Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 4, Barrier

Integrity Screening Questions, the inspectors determined the finding degraded the ability to

close or isolate containment and required evaluation under Inspection Manual Chapter 0609,

Appendix H, Containment Integrity Significance Determination Process, dated

February 25, 2019. Using the Large Early Release Frequency (LERF) type screening

process, the inspectors determined the finding was a Type B LERF finding because the

finding did not affect core damage frequency. The inspectors used

Table 7.3, Phase 1 Screening - Type B Findings at Shutdown, and determined that a

Phase 2 estimate was required because the containment equipment hatch affected

containment isolation, which is a system important to LERF. The inspectors used Table 7.4,

Phase 2 Risk Significance - Type B Findings at Shutdown, to determine the finding was of

very low safety significance (Green) because it did not meet the threshold for low safety

significance (White) for leakage from containment to the environment being greater than

100 percent containment volume per day through containment penetration seals, isolation

valves, or vent and purge systems. Specifically, the licensee was able to demonstrate

through calculations that the leakage from the containment hatch being closed, but not

sealed, would be no more than 30 percent of the containment volume per day.

Cross-cutting Aspect: H.6 - Design Margins: The organization operates and maintains

equipment within design margins. Margins are carefully guarded and changed only through a

systematic and rigorous process. Special attention is placed on maintaining fission product

barriers, defense-in-depth, and safety-related equipment. Specifically, the licensee

incorrectly assumed that a seal on the containment hatch was not required at the onset of an

17

accident and that the increased pressure in containment during an accident could be credited

for making a seal on the hatch.

Enforcement:

Violation: 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requires in part

that conditions adverse to quality are promptly identified and corrected. Contrary to the

above, from December 2017 to December 2018, the licensee failed to promptly correct a

condition adverse to quality. Specifically, the licensee failed to implement adequate

corrective actions for an inadequate procedure for emergency containment closure to ensure

the containment was sealed, an activity affecting quality.

Enforcement Action: This violation is being treated as a Non-Cited Violation, consistent with

Section 2.3.2 of the Enforcement Policy.

Failure to Evaluate a Change to the Facility DC Power System

Cornerstone

SL-IV

Cross-cutting

Aspect

Report

Section

Not Applicable

NCV 05000445/2019001-02

Closed

Not

Applicable

71111.04

The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.59 for the

licensees failure to obtain a license amendment or perform a written evaluation

demonstrating the basis for not obtaining a license amendment prior to making a change to

the facility as described in the final safety analysis report.

Description: The inspectors reviewed the plant configuration of two common Class 1E DC

power panels that can be powered from either the Unit 1 or Unit 2 Class 1E DC busses. The

inspectors found that the licensee has shared systems for both units that receive power from

these panels. The panels also have Unit 1 safety-related systems powered from the panels.

The inspectors noted that shared systems must meet the requirements of 10 CFR Part 50,

Appendix A, Criterion 5, which states, in part, that structures, systems, and components

important to safety shall not be shared among nuclear power units unless it can be shown

that such sharing will not significantly impair their ability to perform their safety functions.

The inspectors questioned whether the inclusion of Unit 1 safety-related equipment on the

common panels constituted acceptable sharing of systems between units.

Upon further review, the inspectors determined that the licensee originally committed to

Regulatory Guide (RG) 1.81 to demonstrate compliance with Criterion 5. The licensee, in its

Final Safety Analysis Report (FSAR), stated that the DC power sources and electric

distribution systems were not shared between the two units, and that safety-related loads

shared between both units are powered from common 125 VDC panels. The NRC in its

safety evaluation report concluded that the design as described in the FSAR, with shared

systems being powered from the common panels but no unit-specific safety-related systems

powered from the common panels, was acceptable.

In January 2000, the licensee discovered that they had unit-specific safety-related systems

from both Units 1 and 2 on the common panels in addition to the previously evaluated shared

systems, contrary to what was described in their FSAR. The licensee entered this design

control issue into the corrective action program. In 2002, the licensee modified the Unit 2

systems to align them to Unit 2 power supplies, but left the Unit 1 systems on the common

panels. The licensee then revised the FSAR to state that they did not comply with RG 1.81,

18

but that the existing configuration of Unit 1 systems was an acceptable exception. The

inspectors determined that powering Unit 1 systems from the Unit 2 DC power supply and

distribution system constituted a system being shared among units, and that the licensee had

not demonstrated compliance with Criterion 5 for these systems while the panels supplying

Unit 1 systems were powered from Unit 2. At the time of the inspection, the common panels

were aligned to Unit 1.

The inspectors determined that the inclusion of Unit 1 systems on panels that shared DC

power systems was a change to the facility as described in the FSAR. The inspectors also

determined that the licensee made the change without performing a written evaluation

demonstrating that a license amendment would not be required. This impeded the ability of

the agency to perform its regulatory function, requiring disposition using traditional

enforcement.

Corrective Action(s): The licensee entered this violation into their corrective action program.

Corrective Action Reference(s): CR-2019-001711

Performance Assessment: The inspectors determined this violation was associated with a

minor performance deficiency.

Enforcement:

The ROPs significance determination process does not specifically consider the regulatory

process impact in its assessment of licensee performance. Therefore, it is necessary to

address this violation which impedes the NRCs ability to regulate using traditional

enforcement to adequately deter non-compliance.

Severity: The violation was determined to be Severity Level IV using section 6.1 of the NRC

Enforcement Policy, dated May 15, 2018, because it was a violation of 10 CFR 50.59, but did

not have a consequence evaluated by the significance determination process as having

low-to-moderate or greater safety significance.

Violation: Title 10 CFR 50.59 requires, in part, that if the licensee makes changes to the

facility as described in the FSAR without obtaining a license amendment, they must maintain

a written evaluation which provides the basis for determining that the change does not require

a licensee amendment. Contrary to the above, in April 2002, the licensee made a change to

the facility as described in the FSAR without obtaining a license amendment, but did not

maintain a written evaluation which provides the basis for determining that the change does

not require a licensee amendment.

Enforcement Action: This violation is being treated as a Non-Cited Violation, consistent with

Section 2.3.2 of the Enforcement Policy.

19

Failure to Monitor or Demonstrate Control of Performance Under the Maintenance Rule

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000445;05000446/2019001-03

Closed

None

71111.12

The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(2), with three

examples, for failure to demonstrate effective control of performance of a maintenance rule

scoped system through appropriate preventive maintenance.

Description: The inspectors identified three examples where the performance of systems,

structures, and components (SSCs) that were subject to the maintenance rule, was not

monitored or demonstrated to be effectively controlled through appropriate preventive

maintenance.

The first example is a violation of 10 CFR 50.65(a)(2) for failure to monitor performance or

demonstrate effective control of performance for the Class 1E battery chargers. The

inspectors identified a failure of the 1ED1-1 battery charger to successfully perform a

maintenance rule function. The battery chargers provide DC power to the class 1E DC

busses from the Class 1E AC busses. The vital bus inverters rely on effective control of

DC voltage ripple on the battery charger output to allow synchronization with class 1E AC

power prior to being placed online. The licensee incorporated a limit of 2 percent voltage

ripple into the design basis document for the DC system. However, the licensee did not

perform any testing or preventive maintenance to ensure output voltage ripple remained

within limits. As a result, the DC output voltage ripple of the 1ED1-1 battery charger

exceeded acceptable voltage ripple at some point in its service life, ultimately resulting in a

failure of the supported inverter to return to service on June 5, 2018.

The licensee determined that the excessive ripple was caused by a failure of a component in

the battery charger, the X-302 printed circuit board (PCB). The PCB had last been replaced

in December 2016 and was scheduled for a 10-year replacement frequency. Subsequent to

that replacement, the licensee documented multiple occurrences where the inverters

supported by that charger did not synchronize correctly. The licensee had generated work

orders to troubleshoot the inverter but had not completed them prior to the June 2018 failure.

Following this failure, the licensee performed an evaluation of the event for their maintenance

rule program. The licensee evaluated the failure as not being a maintenance rule failure

because the battery charger functions, as written, did not describe providing power to the DC

busses. The inspectors concluded that the function to provide power to the DC busses was a

maintenance rule function and that the June 2018 failure was a functional failure.

Furthermore, because the failure could have been prevented by either performing preventive

maintenance on the battery charger or by completing the troubleshooting work orders, the

failure was maintenance preventable. The June 2018 failure exceeded the established

performance criteria, indicating performance was not being effectively controlled, but the

licensee did not monitor performance or set goals. The licensee entered this issue into the

corrective action program.

The second example is a violation of 10 CFR 50.65(a)(2) for failure to monitor performance or

demonstrate effective control of performance for the common low voltage AC power system.

The inspectors identified a failure of the common 120 VAC power system to provide Class 1E

power to certain important to safety components that are shared between Units 1 and 2. The

common panels provide power to shared radiation monitors that require Class 1E power to

function following an accident, which is covered by the maintenance rule under

20

10 CFR 50.65(b)(2)(i). The panels can be transferred to non-Class-1E power for

maintenance. Following a planned maintenance activity on Panel XEC1 in October 2016, the

licensee was unable to transfer the panel back to its normal Class 1E source due to a failure

of the transfer switch. Because the failure represented an inability to receive power from its

Class 1E source, this was a failure to meet its maintenance rule function. The failure was

maintenance preventable, because the licensee was aware of the potential for these switches

to fail but did not perform preventive maintenance to address the failures. The licensee

incorrectly concluded that the transfer switch failure was not a maintenance preventable

failure of a maintenance rule function, because the common panels were being monitored

against plant level performance criteria. The performance of the system cannot be practically

monitored by the use of plant level criteria, because the common low voltage power system

could have unlimited maintenance preventable functional failures without ever meeting the

criteria. The licensee entered this issue into the corrective action program.

The third example is a violation of 10 CFR 50.65(a)(2) for failure to monitor performance or

demonstrate effective control of performance for the inside reactor containment check

valves 1(2)CA-0016. Inspectors noted that the performance criteria assigned to the valves

was inadequate and that there had been multiple failures of these valves during testing.

These results should have been classified as repeat maintenance preventable functional

failures and caused the system to be classified as 50.65(a)(1), but the system remained in

50.65(a)(2) status.

The inspectors noted that the valves were allowed seven failures in a 24-month monitoring

period. This was determined to be inadequate because the valves were tested on a 30month

frequency, so the allowed amount of failures could never be exceeded. Additionally, the

inspectors determined that the cause of the valves failures was a known issue, but the

licensee had not taken action to correct it. Specifically, the valves and system piping are

carbon steel and are part of the service air system. The service air system is neither filtered

nor dried which results in water accumulation in the air system. Water accumulation in the

system causes general corrosion in the piping, resulting in wear particles that affect the

valves ability to close. The inspectors determined that the licensee was aware of the failure

mechanism, the cause, and a solution for the issue but had prioritized it as a low priority and

was not considering this when evaluating whether the failures were maintenance preventable.

The inspectors determined that the failures were maintenance preventable and as such, were

repeat failures, because the licensee had failed to perform the appropriate modifications to

the system. The licensee entered this issue into the corrective action program.

In all these cases, the inspectors determined that the failure to demonstrate effective control

was caused by incomplete descriptions of the applicable maintenance rule functions, which

had been developed during initial implementation of the maintenance rule in the 1990s.

Corrective Action(s): The licensee entered these three examples into the corrective action

program and is reviewing the systems performance.

Corrective Action Reference(s): CR-2018-007884

Performance Assessment:

Performance Deficiency: The failure to monitor the performance or demonstrate effective

control of performance of systems covered by the maintenance rule is a performance

deficiency.

21

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the equipment performance attribute of the Mitigating

Systems Cornerstone. It adversely affected the cornerstone objective to ensure the

availability, reliability, and capability of systems that respond to initiating events to prevent

undesirable consequences (i.e., core damage) because the finding represented a reduction in

the reliability and availability of mitigating systems. Specifically, the failure to monitor the

performance of the battery chargers resulted in multiple instances of decreased reliability of

the system. The common low voltage power system affected the Emergency Preparedness

Cornerstone, and the containment isolation valves affected the Barrier Integrity Cornerstone,

but the Mitigating Systems Cornerstone was selected as the most significant due to the risk

significance of the battery chargers.

Significance: The inspectors assessed the significance of the finding using Appendix A,

Significance Determination of Reactor Inspection Findings for At - Power Situations. Using

Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated

October 7, 2016, the inspectors determined the finding was associated with the Mitigating

Systems cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance

Determination Process (SDP) For Findings At-Power, Exhibit 2, Mitigating Systems

Screening Questions, the inspectors determined the finding was of very low safety

significance (Green) because the finding did not represent an actual loss of function of at

least a single train for greater than its technical specification allowed outage time.

Cross-cutting Aspect: No cross-cutting aspect was assigned to this finding because the

inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: 10 CFR 50.65(a)(1), requires, in part, that the holders of an operating license shall

monitor the performance or condition of structures, systems, or components (SSCs) within the

scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a

manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling

their intended functions.

10 CFR 50.65(a)(2) states, in part, that monitoring as specified in 10 CFR 50.65(a)(1) is not

required where it has been demonstrated that the performance or condition of an SSC is

being effectively controlled through the performance of appropriate preventive maintenance,

such that the SSC remains capable of performing its intended function.

Contrary to the above, as of March 31, 2019, the licensee failed to demonstrate that the

performance of the Class 1E battery chargers, the common 120 VAC power panels, and

containment check valves had been effectively controlled through the performance of

appropriate preventive maintenance and did not monitor against licensee-established goals.

Specifically, the licensee failed to identify, and properly account for preventive maintenance

preventable functional failures of the battery chargers, the common 120 VAC panels, and

containment check valves occurring from October 2016 to June 2018 which demonstrate that

the performance or condition of these SSCs was not being effectively controlled through the

performance of appropriate preventive maintenance and, as a result, that goal setting and

monitoring was required.

Enforcement Action: This violation is being treated as an non-cited violation, consistent with

Section 2.3.2 of the Enforcement Policy.

22

Failure to Control Hazard Barriers During Maintenance

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000445/2019001-04

Closed

[H.8] -

Procedure

Adherence

71111.13

The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)4 for failure to

implement risk mitigating actions during diesel generator maintenance.

Description: On January 17, 2019, the inspectors observed the licensee performing a

maintenance activity to add lube oil to the Unit 1 emergency diesel generator 1-01 sump. In

order to perform the maintenance, the licensee placed a hose through the normally shut

door S1-28 from the train A switchgear room to the train A diesel generator room. The door is

a dogged, two-leaf metal hatch that functions as a barrier for fire, flooding, and medium

energy line break (MELB) events. Prior to performing the maintenance, the licensee

evaluated the risk of opening the door to allow placement of the hose. The licensee identified

additional compensatory measures to protect the train A switchgear in an evaluation

documented in Tracking Report (TR) 2019-000001. The licensee determined that the open

door did not pose a flood risk and implemented appropriate compensatory measures to

mitigate the fire risk. To address the MELB risk, the licensee determined that the open

doorway of the active leaf of door S1-28 could allow a MELB in the diesel generator room to

impact safety-related transformer T1EB3, which provides 480 VAC power to safety-related

bus 1EB3. The licensee determined that the transformer would be protected if the workers

maintained door S1-28 open no more than 2 inches, with the door secured to prevent it from

opening further. The licensee determined that opening the door for normal ingress and

egress was acceptable provided the door was secured after personnel passed through. The

evaluation was attached to the work order and a copy was present at the job site.

When the workers began the job, they identified safety concerns with the door being secured

while personnel were in the diesel generator room. They decided to leave the door open,

assuming that it was acceptable as long as personnel were in the immediate area to close it.

When the inspectors arrived at the work site, they noticed the door open with no one passing

through it and questioned the configuration of the door. The inspectors then contacted the

control room and the licensee secured the door.

The licensee determined that crediting actions to close the door post event did not adequately

mitigate the risk of a MELB. As a result of the failure to implement the risk mitigating actions,

the licensee determined that the train A 480 VAC bus 1EB3 was inoperable for

approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> due to the potential for a MELB to spray water on the transformer. The

allowable outage time of the bus per Technical Specification 3.8.9 is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The licensee

determined that the bus did not exceed its allowed outage time due to the hazard barrier

being open.

Corrective Action(s): The licensee restored the barrier and entered the issue into the

corrective action program.

Corrective Action Reference(s): CR-2019-000672

23

Performance Assessment:

Performance Deficiency: The failure to implement planned risk mitigating actions was a

performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the Configuration Control attribute of the Mitigating Systems

cornerstone. It adversely affected the cornerstone objective to ensure the availability,

reliability, and capability of systems that respond to initiating events to prevent undesirable

consequences (i.e., core damage) because the finding represented a loss of control of

barriers required to ensure the availability of AC power. Specifically, the failure to maintain

the door in a nearly closed position exposed a Class 1E 480 VAC bus to failure during a

MELB event, resulting in an electrical distribution train being inoperable for several hours.

Significance: The inspectors assessed the significance of the finding using Appendix K,

Maintenance Risk Assessment and Risk Management SDP. Using Inspection Manual

Chapter 0609, Attachment 04, Initial Characterization of Findings, dated October 7, 2016,

the inspectors determined the finding was associated with the Mitigating Systems

cornerstone. Using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk

Assessment and Risk Management Significance Determination Process, the inspectors

determined the finding was associated with risk mitigating actions (RMAs) only. The

inspectors used Flowcharts 1 and 2 to determine that the finding required a determination of

the incremental core damage probability due to the failure to implement RMAs.

A risk analyst performed a bounding analysis of incremental core damage probability

assuming that bus 1EB3 was unavailable along with the train A emergency diesel generator

for the entire exposure time when adequate RMAs were not in place. This estimate was

bounding because it assumes bus 1EB3 always failed during the exposure time and does not

incorporate the probabilistic occurrences of fire, flooding, line break, and other events could

have rendered bus 1EB3 unavailable, which would result in a lower estimate of incremental

core damage probability. The resulting bounding estimate in the incremental core damage

probability was 8.1E-8. The inspectors determined that the finding was of very low safety

significance (Green) because the incremental core damage probability was less than 1E-6

and the finding did not affect the large early release probability.

Cross-cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices

that emphasize prudent choices over those that are simply allowable. A proposed action is

determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically,

the licensee personnel assumed that the controls were not necessary without stopping work

and discussing with their supervisor, and did not implement prescribed risk mitigating actions.

Enforcement:

Violation: 10 CFR 50.65(a)(4) requires, in part, that the licensee assess and manage the

increase in risk that may result from maintenance activities. Contrary to the above, on

January 17, 2019, the licensee failed to manage the increase in risk resulting from a

maintenance activity. Specifically, the licensee did not implement planned risk mitigating

actions that were identified as necessary by the risk assessment.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with

Section 2.3.2 of the Enforcement Policy.

24

Failure to Follow Procedure When A Degraded Condition Was Identified

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000445;05000446/2019001-05

Closed

[H.14] -

Conservative

Bias

71111.15

The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to

follow the requirements of Station Procedure STI-421.01, Initiation of Issue Reports,

Revision 0. Specifically, station personnel failed to notify the shift manager of an issue with

material storage in the protected area. This issue required evaluations and compensatory

actions for resolution.

Description: On January 31, 2019, inspectors identified that the licensee had allowed

material to be stored in a temporary laydown area inside of the protected area. Inspectors

noted that several items appeared to be susceptible to being picked up by tornado driven

winds, so the inspectors inquired as to how these items had been evaluated for their current

storage area. The licensee initiated TR-2019-001119 to capture the inspectors questions.

As part of TR-2019-001119 the licensee determined that the materials in question had not

been evaluated for its current storage location. An action was assigned to engineering to

evaluate the materials in question (AI-TR-2019-001119-1). Engineering completed their

evaluation on February 4, 2019, and engineering management approved the evaluation on

February 6, 2019. The evaluation determined that there were materials in the laydown area

that were susceptible to being lifted by tornadic winds, and they needed to be strapped down

in such a way as to increase their weights to a point where they were no longer susceptible.

Inspectors reviewed AI-TR-2019-001119-1 on February 14, 2019. During their review they

determined that the identified condition required an operability review because of the potential

to be in an unanalyzed condition with respect to tornado driven missiles. However, inspectors

noted that an operability review was not performed because the issue had not been reported

to the control room by engineering upon discovery on February 4, 2019, as required by

Station Procedure STI-421.01, Initiation of Issue Reports, Revision 0, Section 6.1.

Additionally, there was no guidance or actions in place to adequately strap down the material

to ensure that it did not pose a risk to plant equipment.

Inspectors informed the licensee of their observations. The licensee reviewed the issue and

determined that the condition did require an operability review and compensatory actions to

address it pending further review.

Corrective Action(s): The licensee performed an operability determination and establish

compensatory measures that established a reasonable expectation of operability pending

development of additional corrective actions.

Corrective Action Reference(s): CR-2019-001119

25

Performance Assessment:

Performance Deficiency: The licensees failure to follow the requirements of

Procedure STI-421.01 when a degraded condition was identified was a performance

deficiency.

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the Protection Against External Factors attribute of the

Mitigating Systems cornerstone. It affected the cornerstone objective to ensure availability,

reliability, and capability of systems that respond to initiating events to prevent undesirable

consequences. Specifically, the storage of materials without proper evaluations resulted in

the introduction of new and unanalyzed tornadic missiles.

Significance: The inspectors assessed the significance of the finding using Appendix A,

Significance Determination of Reactor Inspection Findings for At - Power Situations. Using

Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding

was of very low safety significance (Green) because: (1) it was not a design deficiency; (2) it

did not represent a loss of system and/or function; (3) it did not represent an actual loss of

function of at least a single train for longer than its technical specification allowed outage

time; and (4) it did not result in the loss of a high safety significant non-technical specification

train.

Cross-cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices

that emphasize prudent choices over those that are simply allowable. A proposed action is

determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically,

engineering failed to use decision making-practices that emphasize prudent choices over

those that are simply allowable.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and

Drawings, requires, in part, that activities affecting quality shall be prescribed by documented

instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be

accomplished in accordance with these instructions, procedures, and drawings.

Contrary to the above, from February 4-27, 2019, an activity affecting quality was not

accomplished in accordance procedures appropriate to the circumstances. Specifically,

station personnel failed to notify the shift manager of an issue with material storage in the

protected area (as required by Station Procedure STI-421.01, Initiation of Issue Reports)

which required evaluations and compensatory actions for resolution.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with

Section 2.3.2 of the Enforcement Policy.

26

Failure to Perform Safety Evaluations in Accordance with 10 CFR 50.59

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000445; 05000446/2019001-

06

Closed

[H.9] - Training

71111.17T

The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, (with four examples) in which the licensee failed to

complete 50.59 evaluations as required by station procedures.

Description: The inspectors identified four examples where the licensee failed to perform

50.59 evaluations as required by procedures and guidance specified in STA-707,

10 CFR 50.59 and 10 CFR 72.48 Reviews, Revision 21.

Example 1. EV-CR-2017-004743-2, Blow Down the 1-01 Instrument Air Receiver

In the screen for the compensatory measure to blow down the 1-01 air receiver once per shift,

question 1 of the screening was, Does the proposed activity involve a change to an SSC that

adversely affects an UFSAR described design function? The preparer answered the

question No; the explanation provided had the following statements: The activity is a

Compensatory Measure to blow down the 1-01 Instrument Air Receiver once per shift using

1CI-0012 to remove water from the receiver. The drip trap (CP1- CIMEDT-01) that performs

the automatic drain will be repaired IAW WO 5474911. This statement indicates that an

automatic function was replaced with a manual function.

The vendor manual, AP-0293-B, Ingersoll-Rand Compressor Accessories, dated April 1976,

provides the following guidance on page 7 for liquid carryover, It is important that interstage

separators be drained regularly and we are of the opinion that manual drainage at specified

intervals with the fact of drainage logged, is the proper method, particularly at higher

pressures. Automatic traps, if used, should have a bypass piped for visual observation and

check on trap operation - the check should be made at stated intervals and the results

logged. Page 12 of the manual provides guidance that drainage of the receiver following the

aftercooler should be drained at least once per shift.

CPNPP 50-59 RM-6, "CPNPP 50.59 Resource Manual," Revision 6, requires that an

evaluation be performed if an automatic function is replaced with a manual action. The

preparer and reviewer failed to ensure the appropriate Applicability

Determination/screen/evaluation was performed and the corresponding Applicability

Determination/screen/evaluation form was completed in accordance with guidance provided

in CPNPP 50-59 RM-6. Screening guidance would require this change to be evaluated prior

to changing from an automatic to a manual function.

Example 2. EV-CR-2018-007384 RCS Pressure Boundary Leakage Test

This document was to perform a 50.59 review for changes to Procedure OPT-612B, RCS

Pressure Boundary Leakage Test for Loop 1 Cold Leg Injection Valves, and

Procedure OPT 613B, RCS Pressure Boundary Leakage Test for Loop 2 Cold Leg Injection

Valves, to allow the performance of reactor coolant system pressure boundary leakage test

for safety injection loops 1 and 2. The licensee had attempted to perform a flush of the

residual heat removal system while in Mode 1, an evolution normally performed in Modes 3,

27

4, or 5. Inadequate procedure changes and review of the planned process resulted in forward

flow through valves 2-8956A and B. This placed the unit in a 24-hour LCO to complete

Surveillance Requirement 3.4.14 for valves 2-8956 A and B. Procedures OPT-612B and

OPT-613B needed to be revised to allow performance of this surveillance in Mode 1. The

activity required component manipulations that isolated one safety injection accumulator and

rendered one train of residual heat removal inoperable in order to perform the leak check. A

threaded pipe cap was removed and various normally closed valves were opened to allow

connection of the test rig. The screener and reviewer failed to recognize that these actions

resulted in an "adverse effect" on the plant.

CPNPP 50-59-RM6, Section 5.2.2, states, in part, changes that have an adverse effect are

required to be evaluated under 10 CFR 50.59 because they have the potential to increase the

likelihood of malfunctions, increase consequences, create new accidents, or otherwise meet

the 10 CFR 50.59 evaluation criteria.

CPNPP 50-59-RM6, Section 5.2.1 states, Items to Consider When Deciding Whether an Item

is a Change to the Facility: Does the activity decrease the reliability of an SSC design

function, including either functions whose failure would initiate a transient/accident or

functions that are relied upon for mitigation? Does the activity reduce existing redundancy,

diversity, or defense-in-depth?

The screener and reviewer failed to recognize that, even though technical specifications allow

operation with one safety injection accumulator isolated and one train of residual heat

removal inoperable, this resulted in a reduction in the existing redundancy, diversity, and

defense-in-depth that required the performance of an evaluation.

Example 3. Procedure Change to SOP-102B

Section 1 of the screen for the change to SOP-102B, Residual Heat Removal System,

Revision 15, provided the following description in the change justification section: "Modified

Section 5.2 to allow flushing of the RHR System to the RHUT (ref AI-CR-2018-007381-4),

deleted "Intentionally Left Blank" Pages 3&4 of Attachment 4. Re-sequenced Table of

Contents to reflect new page numbering. Added new prerequisite to Section 2.3 to clarify

intent of Section 5.11 and moved 2.3 to previous page." The technical reviewer answered

yes to the question: If change is editorial, THEN circle or mark "YES." Editorial changes, as

limited by STA-202, Attachment 8.F, do not require Administrative Review, Technical Review,

NSR, AD, 50.59 Review or 72.48 Review.

The procedure change (in Section 5.2 to allow flushing of the RHR system) actually

manipulated valves in the safety injection system to isolate the safety injection accumulators

based on lessons learned when the licensee originally attempted to flush the residual heat

removal system while in Mode 1. The licensee had failed to recognize that the initial

conditions assumed in Procedure SOP-102B had the safety injection accumulators isolated.

In Mode 1, the safety injection accumulators were in service, and the attempted flush of the

residual heat removal system resulted in flow from the accumulators. The purpose of the

procedure modification was to isolate the safety injection accumulator to allow a partial flush

of the residual heat removal system. The preparer, reviewer, and technical reviewer all failed

to identify this aspect of the procedure change. As a result, the adverse effect on the plant, a

reduction in redundancy to the safety injection system, was not identified, and therefore the

required 10 CFR 50.59 evaluation was not performed.

28

Example 4. EV-2002-002026-01-00 Bladder Equivalency Evaluation

On May 28, 2002, the licensee performed an equivalency evaluation for replacement

diaphragms for the reactor make up water storage tanks, EV-2002-002026-01-00. In the

evaluation the licensee identified that the new diaphragm was manufactured with a material

that has a specific gravity greater than 1.0 which will make it heavier than the water in the

tank, and consequently material which tears or breaks off from the diaphragm will sink into

the tank and potentially into the pump suction, which could cause the pump to malfunction.

The licensee determined that this was an equivalent change by crediting proper maintenance

and inspection to ensure that a failure of the new material does not occur.

Inspectors determined that this was not an equivalent change because the new diaphragm

introduced the potential for a new adverse effect (bladder failure could result in material

sinking and clogging pump suction) and should have been evaluated. CPNPP 50-59-RM6 ,

Section 5.2.2 states in part, changes that have an adverse effect are required to be evaluated

under 10 CFR 50.59 because they have the potential to increase the likelihood of

malfunctions, increase consequences, create new accidents, or otherwise meet the 10 CFR

50.59 evaluation criteria.

Corrective Action(s): The licensee entered these issues into the corrective action program.

Corrective Action Reference(s): IR-2019-001271, IR-2019-001317, IR-2019-001428,

IR-2019-001430

Performance Assessment:

Performance Deficiency: The inspectors determined that not conducting required

10 CFR 50.59 evaluations was a performance deficiency within the licensee's ability to

foresee and correct. Specifically, the licensee failed to perform 10 CFR 50.59 evaluations for

the compensatory measure for the instrument air system, the procedure change for the

reactor coolant system pressure boundary leakage test for safety injection loops 1 and 2, the

procedure change for the residual heat removal system flush, and replacement diaphragms

for the reactor make up water storage tanks.

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the Human Performance attribute of the Mitigating Systems

Cornerstone and adversely impacted the cornerstone objective of ensuring the availability,

reliability, and capability of systems that respond to initiating events to prevent undesirable

consequences.

Significance: The inspectors assessed the significance of the finding using Appendix A,

Significance Determination of Reactor Inspection Findings for At - Power Situations. The

inspectors assessed the significance of the finding using Inspection Manual Chapter 0609.04,

and Inspection Manual Chapter 0609, Appendix A, Exhibit 2. The inspectors determined that

this finding was of very low safety significance (Green), because the finding did not represent

a loss of the emergency core cooling system or the instrument air system safety function, did

not result in any loss of function beyond the technical specification-allowed outage time, and

did not result in the loss of any non-technical specification trains that were designated as high

safety-significance in accordance with the licensees maintenance rule program.

29

Cross-cutting Aspect: H.9 - Training: The organization provides training and ensures

knowledge transfer to maintain a knowledgeable, technically competent workforce and instill

nuclear safety values. Specifically, the licensee failed to provide training to maintain a

knowledgeable, technically sound workforce and instill nuclear safety values when

implementing the change process.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and

Drawings, states Activities affecting quality shall be prescribed by documented instructions,

procedures, or drawings, of a type appropriate to the circumstances and shall be

accomplished in accordance with these instructions, procedures, or drawings. Contrary to the

above, from May 2002, to February 2019, the team identified four examples where the

licensee failed to follow the requirements of Procedure CPNPP 50.59-RM6, "CPNPP 50.59

Resource Manual," Revision 6. The procedure required a 10CFR 50.59 evaluation to be

performed if an activity reduces existing redundancy, diversity, or defense in depth or if an

automatic function is replaced with a manual action. Specifically, the licensee implemented

manual compensatory actions when the automatic trap for the instrument air system failed,

made procedure changes that reduced the redundancy, diversity, reliability, and defense-in-

depth of the emergency core cooling systems, and installed new material in the plant with a

different adverse effect without performing 10 CFR 50.59 evaluations as required.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with

Section 2.3.2 of the Enforcement Policy.

Inadequate Maintenance Instructions Result in Loss of Assessment Capability

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Emergency

Preparedness

Green

NCV 05000445;05000446/2019001-07

Closed

[H.8] -

Procedure

Adherence

71152

The inspectors reviewed a self-revealed Green, non-citied violation of 10 CFR 50,

Appendix B, Criterion V, "Instruction, Procedures, and Drawings", that occurred due to

inadequate maintenance instructions for safety-related radiation monitors which resulted in a

major loss of assessment capability.

Description: On December 5, 2017, the licensee was performing maintenance on the control

room south ventilation intake radiation monitor under Work Order (WO) 5063234 when they

received audible and visible alarms in the control room indicating a loss of multiple radiation

monitors. The crew evaluated the indications and determined a major loss of assessment

capability occurred due to the unplanned loss of the main steam line radiation monitors for

steam lines 1 and 3, and the station service water (SSW) radiation monitors. The loss of

these radiation monitors impacted emergency action levels for radiation effluent. This event

was reported to the NRC as Event Report No. 53105.

The inspectors reviewed the circumstances of this event including the licensees evaluation

and corrective actions. The licensees radiation monitoring system consists of four

communication loops of 20 to 30 radiation monitors each. The loops pass inputs via each

successive monitor to the plant computer system, which then provides required indications to

the control room and emergency response facilities (ERFs). The licensee determined that the

loss of the affected radiation monitors was due to taking the control room south ventilation

30

intake radiation monitor out of service without first installing jumpers in the communication

loop to bypass the monitor. This resulted in a failure of all other monitors in the affected loop

to provide indication to the plant computer system.

The inadequate maintenance resulted in the simultaneous communications failure of

approximately 27 radiation monitors. In addition to the monitors that met the criteria for the

report, the inspectors noted the following other monitors that affected emergency

classification:

Unit 1 main steam line radiation monitors for main steam lines 1 and 3

both Unit 1 SSW radiation monitors and all Unit 1 component cooling water radiation

monitors, their credited backup for the SSW monitors

the Unit 1 failed fuel monitor

all Unit 1 refueling cavity monitors

the Unit 1 containment radiation monitors for particulate, iodine, and gaseous activity

the fuel building vent exhaust monitor

The licensee implemented compensatory measures for the affected monitors while restoring

them to service. The main steam line radiation monitors affected the ability to declare a

General Emergency for high steam line radiation, but the licensee determined that a General

Emergency declaration could have been made using other emergency action levels. The

inspectors did not identify any concerns with the licensees conclusion regarding emergency

classification.

The inspectors determined that the workers did not install the jumpers because WO 5063234

did not contain instructions to install the jumpers. The licensee had relied on the knowledge

of a few experienced technicians who were aware that the jumpers needed to be installed

prior to removing a monitor from service. However, the workers performing WO 5063234 on

the control room south ventilation intake radiation monitor on December 5 were not aware of

the need to install jumpers.

The inspectors determined that licensee Procedure STI-606.03, Work Planning, Section 6.2

requires that work packages identify where jumpers need to be installed. The inspectors

concluded that the work instructions in WO 5063234 were inadequate. The control room

south ventilation intake radiation monitor is safety-related, and therefore, the work instructions

were quality related instructions.

Corrective Action(s): The licensee stopped maintenance, implemented compensatory

measures, and restored the monitors to service.

Corrective Action Reference(s): CR-2019-002535

31

Performance Assessment:

Performance Deficiency: The failure to prescribe adequate work instructions for a quality

related activity is a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the facilities and equipment attribute of the Emergency

Preparedness Cornerstone. It adversely affected the cornerstone objective to ensure that the

licensee is capable of implementing adequate measures to protect the health and safety of

the public in the event of a radiological emergency. Specifically, it resulted in the failure of

multiple pieces of equipment credited for maintaining the licensees emergency plan with

respect to emergency planning standard four, which requires a standard emergency

classification and action level scheme to be in use.

Significance: The inspectors assessed the significance of the finding using Appendix B,

Emergency Preparedness SDP. Using table 5.4-1, Significance Examples

Section 50.47(b)(4), the finding was determined to be of very low safety significance (Green)

because it was not a degraded risk significant planning standard function. The planning

standard function was not degraded because, although an emergency action level (EAL) was

rendered ineffective such that a General Emergency would not have been declared for a

particular off-normal event, other EALs could have been used to make an appropriate

declaration.

Cross-cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures,

and work instructions. Specifically, individuals did not follow the work planning procedure

when preparing work instructions for maintenance on the radiation monitors.

Enforcement:

Violation: Title 10 CFR 50, Appendix B, Criterion V, "Instruction, Procedures, and Drawings,"

requires in part that activities affecting quality shall be prescribed by documented instructions

of a type appropriate to the circumstances. Contrary to the above, on December 5, 2017, the

licensee failed to prescribe activities affecting quality by documented instructions of a type

appropriate to the circumstances. Specifically, the licensee prescribed maintenance on a

safety-related radiation monitor with instructions that did not identify jumpers required to

maintain the function of the radiation monitoring system.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with

Section 2.3.2 of the Enforcement Policy.

Failure to Establish Adequate Procedural Guidance for Flushing Lithium at Power

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green

NCV 05000446/2019001-08

Closed

[H.11] -

Challenge the

Unknown

71152

The inspectors reviewed a Green, self-revealing non-cited violation of 10 CFR Part 50,

Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the

licensees failure to establish an adequate procedure for flushing lithium from the residual

heat removal system. This resulted in safety injection Accumulators 2-01 and 2-02 discharge

to the safety injection test header causing level drops in both accumulators, and

32

Accumulator 2-01 pressure dropped to below the operability limit resulting in an unplanned

component inoperability.

Description: On November 2, 2018, with Unit 2 in Mode 1 operations the licensee performed

an evolution to flush lithium from the residual heat removal system. The licensee used

Station Procedure SOP-102A, Residual Heat Removal System, Revision 20, Section 5.11, to

perform this evolution. During the flush safety injection Accumulators 2-01 and 2-02 levels

dropped by 6 percent due to the accumulators discharging to the safety injection test header,

and Accumulator 2-01s pressure dropped to below the operability limit resulting in an

unplanned component inoperability. Operators stopped the activity and restored level and

pressure in the accumulators. Condition Report CR-2018-007381 was written to capture the

issue in the corrective action program.

During the licensees investigation of the event it was determined that Procedure SOP-102A,

section 5.11, was not the correct procedure for this evolution because it was not intended for

use in the mode of operation. The licensee identified two causes for why an incorrect

procedure was used; inadequate coordination and incorrect assumptions. Inadequate

coordination because operations, chemistry and engineering had used an informal selection

process which lacked rigor when selecting a procedure to perform an infrequently performed

task, and this resulted in no further challenge or verifications of the adequacy of this

procedure. The licensee also identified that the work scheduling process does not require

operations procedures to be reviewed for impact. Inadequate assumptions because of the

belief by operations, chemistry and engineering that procedure SOP-102A provided

appropriate instructions for the at-power lithium flush.

Inspectors reviewed the licensees evaluation and concluded that it identified reasonable

causes and adequately addressed the identified causes.

Corrective Action(s): The licensee immediately stopped the activity, refilled and

re-pressurized the safety injection accumulators. Subsequent corrective actions were to

revise the work control process to require formal reviews for infrequently performed

non-repetitive activities.

Corrective Action Reference(s): CR-2018-007381

Performance Assessment:

Performance Deficiency: The licensees failure to establish an adequate procedure for

flushing lithium from the residual heat removal system was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the equipment performance attribute of the Mitigating

Systems Cornerstone. It adversely affected the cornerstone objective to ensure the

availability, reliability, and capability of systems that respond to initiating events to prevent

undesirable consequences. Specifically, the use of an inadequate procedure for flushing

lithium resulted in an inoperable safety injection accumulator.

Significance: The inspectors assessed the significance of the finding using Appendix A,

Significance Determination of Reactor Inspection Findings for At - Power Situations. Using

Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding

was of very low safety significance (Green) because: (1) it was not a design deficiency; (2) it

did not represent a loss of system and/or function; (3) it did not represent an actual loss of

33

function of at least a single train for longer than its technical specification allowed outage

time; and (4) it did not result in the loss of a high safety significant non-technical specification

train.

Cross-cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with

uncertain conditions. Risks are evaluated and managed before proceeding. Specifically,

station personnel failed to stop when faced with uncertain conditions and ensure that risks

were evaluated and managed before proceeding.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and

Drawings requires, in part, that activities affecting quality shall be prescribed by documented

instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be

accomplished in accordance with these instructions, procedures, and drawings.

Contrary to the above, on November 2, 2018, an activity affecting quality was not prescribed

by documented instructions, procedures, or drawings, of a type appropriate to the

circumstances. Specifically, Station Procedure SOP-102A, Residual Heat Removal System,

Revision 20, Section 5.11, provided inadequate guidance for flushing lithium from the residual

heat removal system with the reactor in Mode 1 operation.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with

Section 2.3.2 of the Enforcement Policy.

Licensee-Identified Non-Cited Violation

71111.18

This violation of very low safety significance was identified by the licensee and has been

entered into the licensee corrective action program and is being treated as a non-cited

violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires in part that

measures shall be established to assure that applicable regulatory requirements and the

design basis are correctly translated into specifications, drawings, procedures, and

instructions. Contrary to the above, from initial construction to December 2018, the licensee

failed to correctly translate the design basis into specifications and procedures. Specifically,

the licensee failed to ensure the design basis for nitrogen accumulator pressure for the

pressurizer power operated relief valves (PORV) was correctly translated into the

specification for minimum allowable pressure, resulting in a non-conservative low pressure

alarm setpoint. As a result, for a period of approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, one Unit 1 PORV would

not have been able to cycle for the required number of operations to mitigate an overpressure

event when required.

Significance: Green.

Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings,

dated October 7, 2016, Inspection Manual Chapter 0609, Appendix G, Shutdown Operations

Significance Determination Process, dated May 9, 2014, and Appendix G Attachment 1,

Phase 1 Initial Screening and Characterization of Findings, Exhibit 4, Barrier Integrity

Screening Questions, the inspectors determined the finding affected the Barrier Integrity

Cornerstone and required a detailed risk evaluation because the finding involved the

unavailability of a PORV during low temperature overpressure (LTOP) operations.

34

A senior risk analyst performed a bounding detailed risk evaluation and assumed that the

PORV not being able to cycle the full credited amount of times prevented the PORV from

fulfilling its LTOP system function. The analyst used the frequency estimate for overpressure

excursion events from NUREG-0933, Resolution of Generic Safety Issues: Issue 94:

Additional Low Temperature Overpressure Protection for Light Water Reactors, to estimate

the initiating event frequency. Other influential assumptions used by the senior reactor

analyst included an exposure time of approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and that the licensee

maintained the availability of a single additional relief valve (with its associated failure rate

estimated from the 2016 data update to NUREG/CR-6928, Industry-Average Performance

for Components and Initiating Events at U.S. Commercial Nuclear Power Plants) with

capability sufficient to mitigate an LTOP event as described in the final safety analysis report.

Using these assumptions, the senior reactor analyst determined that a bounding increase in

core damage frequency for this issue was 8.9E-8 per year and was, therefore, of very low

safety significance (Green).

Corrective Action Reference(s):CR-2018-008757

EXIT MEETINGS AND DEBRIEFS

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

On February 8, 2019, the inspector presented the Evaluations of Changes, Tests and

Experiments inspection results to Mr. Tom McCool and other members of the licensee

staff.

On February 13, 2019, the inspector presented the Evaluations of Changes, Tests and

Experiments inspection results to Mr. Tim Hope and other members of the licensee staff.

On April 2, 2019, the inspector presented the quarterly resident inspector inspection

results to Steven Sewell and other members of the licensee staff.

35

DOCUMENTS REVIEWED

71111.04 - Equipment Alignment

Condition Reports

CR-2000-000142

CR-2017-011443

CR-2018-008300

CR-2019-000653

CR-2019-000672

CR-2019-002533

TR-2017-011236

TR-2017-011749

Procedures

Number

Title

Revision

STI-600.01

Protecting Plant Equipment and Sensitive Equipment Controls

1

SOP-605A

125 VDC Switchgear and Distribution Systems, Batteries and

Battery Chargers

12

Drawings

Number

Title

Revision

E1-0020 Sh. K

125V DC One Line Diagram

CP-24

E1-0020 Sh. L

125V DC One Line Diagram

CP-23

Miscellaneous

Documents

Number

Title

Revision

or Date

FDA-2000-00142

Final Design Authorization

02

Calculations

Number

Title

Revision

or Date

MM-90-2671

Technical Evaluation

11/28/1990

71111.12 - Maintenance Effectiveness

Condition Reports

CR-2015-008236

CR-2016-000049

CR-2016-007907

CR-2017-000594

CR-2017-0010477

CR-2017-004704

CR-2018-003921

CR-2018-003945

CR-2018-004761

CR-2019-002622

TR-2016-000169

TR-2016-002742

TR-2016-008960

TR-2018-004761

Work Orders

5380904

5517474

5144575

5220567

5331282

5347463

5377428

36

Miscellaneous

Documents

Number

Title

Revision

or Date

DBD-EE-044

DC Power Systems

28

DBD-EE-043

118V AC Uninterruptible Power Supply System

14

71111.13 - Maintenance Risk and Emergent Work

Condition Reports

TR-2019-000001

Work Orders

5692097

5705947

Procedures

Number

Title

Revision

STA-696

Hazard Barrier Controls

3

71111.17T - Evaluations of Changes, Tests and Experiments

Condition Reports

CR-2017-005150

CR-2017-012952

CR-2018-007381

CR-2018-007384

TR-2019-001160

CR-2019-001179

CR-2019-001200

CR-2019-001240

CR-2019-001249

CR-2019-001271

IR-2019-001316

IR-2019-001317

IR-2019-001318

IR-2019-001428

IR-2019-001430

TR-2017-007959

TR-2018-004675

Work Orders

5352698

5510637

5510645

5510646

5510663

5510664

5510665

5510666

5510588

5510605

5510610

5510611

5510615

5510633

5510634

5510635

5510636

5351262

5351266

5351253

5383860

5351257

5351268

5346400

5284599

5435995

391842

3905518

Procedures

Number

Title

Revision

ODA-401

Control of Annunciators, Instruments, and Protective Relays

11

OPT-612B

RCS Pressure Boundary Leakage Test FOR LOOP 1 CL

INJECTION VALVES

3

37

Procedures

Number

Title

Revision

OPT-613B

RCS PRESSURE BOUNDARY LEAKAGE TEST FOR LOOP 2 CL

INJECTION VALVES

3

SOP-102B

RESIDUAL HEAT REMOVAL SYSTEM

15

SOP-609A

DIESEL GENERATOR SYSTEM

21

STA-602

TEMPORARY MODIFICATIONS AND TRANSIENT EQUIPMENT

PLACEMENTS

18

STA-707

10CFR50.59 AND 10CFR72.48 REVIEWS

21

STA-738

FIRE PROTECTION SYSTEMS/EQUIPMENT IMPAIRMENTS

7

STI-422.02

COMPENSATORY ACTIONS & TRANSIENT EQUIPMENT

PLACEMENTS

1

STI-707.04

10CFR50.59 AND 10CFR72.48 REVIEWS APPLICABILITY

DETERMINATIONS

1

TDM-401B

TURBINE/GENERATOR LIMIT CURVES

5

ABN-104

RESIDUAL HEAT REMOVAL SYSTEM MALFUNCTION

9

ABN-104

RESIDUAL HEAT REMOVAL SYSTEM MALFUNCTION

8

ABN-402

MAIN GENERATOR MALFUNCTION

13

ALM-0031A

ALARM PROCEDURE 1-ALB-3A

8

ALM-0031B

ALARM PROCEDURE 2-ALB-3A

4

TDM-401B

TURBINE/GENERATOR LIMIT CURVES

5

Drawings

Number

Title

Revision

M2-0235

FLOW DIAGRAM, SPENT FUEL POOL COOLING AND

CLEAN-UP SYSTEM

CP-17

M2-2225

INSTRUMENTATION AND CONTROL DIAGRAM, FIRE

DETECTION/PROTECTION SYSTEM CHANNELS 4100,

4102, 4103, 4111

CP-2

COMANCHE 004

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC

CONTROL PANEL CP1/2-EPIBHX-01P

CP-3

COMANCHE 015

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL DAMPER

CONTROL PANEL

CP-1

COMANCHE 006

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC

CONTROL PANEL CP1/2-EPIBHX-01P

CP-3

COMANCHE 008

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC

CONTROL PANEL CP1/2-EPIBHX-01P

CP-2

COMANCHE 010

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC

CONTROL PANEL CP1/2-EPIBHX-01P

CP-2

38

Drawings

Number

Title

Revision

COMANCHE 012

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL AHUA/AHUB

FAN STRTER PANELS CP1/2-EPIBMC-01 AND CP1/2-

EPIBMC-02

CP-3

COMANCHE 014

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL

CP-2

COMANCHE 011

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC

CONTROL PANEL CP1/2-EPIBHX-01P

CP-2

COMANCHE 013A UNIT 1 AND UNIT 2 ISOPHASE BUS CONTROL INTERNAL

WIRING DIAGRAM

CP-3

2323-A1-0507

PRIMARY PLANT AUXILIARY ELECTRICAL AND CONTROL

BUILDING FLOOR PLAN

CP-1

COMANCHE 002

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL INTERIOR

PANEL LAYOUT

CP-2

COMANCHE 003

UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL PLC

CONTROL PANEL CP1/2-EPIBHX-01P

CP-2

COMANCHE 015A UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL DAMPER

CONTROL PANEL

CP-1

COMANCHE 015B UNIT 1 & UNIT 2 ISOPHASE BUS CONTROL DAMPER

CONTROL PANEL

CP-1

M1-0260

FLOW DIAGRAM - RESIDUAL HEAT REMOVAL SYSTEM

CP-37

M1-0261

FLOW DIAGRAM - SAFETY INJECTION SYSTEM SHEET 1

0F 5

CP-24

M1-0216

FLOW DIAGRAM - COMPRESSED AIR SYSTEM

CP-45

M1-0250

FLOW DIAGRAM - REACTOR COOLANT SYSTEM

CP-34

M1-2300

INSTRUMENTATION AND CONTROL DIAGRAM,

VENTILATION - CONTAINMENT, CHANNEL 5400/5403

CP-7

Miscellaneous

Documents

Number

Title

Revision

or Date

EVAL-2018-007

CPNPP Nuclear Oversight Audit Report - CONFIGURATION

& DESIGN CONTROL

08/16/2018

DBD-ME-013

Design Basis Document - Containment Isolation System

25

RIR-22946OCR

Receipt Inspection Report

10/06/1983

CP-201700626

Comanche Peak Nuclear Power Plant, Docket Nos. 50-445

and 50-446 and 72-74, 10CFR50.59 Evaluation Summary

Report 020, 10CFR72.48 Evaluation Summary Report 005,

and Commitment Material Change Evaluation Report 014

12/05/2017

DBD-ME-014-02

Design Basis Document - Generator and Exciter System

21

39

Vendor

Documents

Number

Title

Revision

or Date

CP-201600573

EVALUATION OF COMANCHE PEAK UNIT 1 CLASS 2 TO

CLASS 1VALVE UPGRADES

05/31/2016

CP1/CP2-

EPIBHX-01E/01F

Damper Position Monitor

08/16/2016

CT-27331

MISSILE PROBABILITY ANALYSIS METHODOLOGY

FOR LUMINANT GENERATION COMPANY LLC,

COMANCHE PEAK UNITS 1 & 2 WITH SIEMENS

RETROFIT TURBINES

8

VDRT-5472306

Unit 2 Generator Stator Damage - Monitoring Installation

Plan

07/21/2017

WPT-18067

Transmittal of LTR-SEE-17-189, Flow Evaluation of Forced

Forward Flow through the Residual Heat Removal Pumps at

Comanche Peak Units 1 & 2

10/03/2017

Calculations

Number

Title

Revision

MEB-391

Minimum Allowable Service Water Flow to Diesel Generators

5

ME-CA-0229-2188 Component Cooling Water Heater Exchanger Fowling Water

Analysis

8

71111.18 - Plant Modifications

Condition Reports

CR-2018-008757

Work Orders

5435249

5689179

Modifications

Number

Title

Revision

FDA-2018-000119-01

Final Design Authorization

Calculations

Number

Title

Revision

ME-CA-0000-3342

Air Accumulator Check Valve Leakage - Decay Rate,

Pressure, Time

3

40

71152 - Identification and Resolution of Problems

Condition Reports

CR-2017-013243

CR-2018-003808

CR-2019-002535

Work Orders

5540984

5063234

Procedures

Number

Title

Revision

STI-606.03

Work Planning

3

Miscellaneous

Documents

Number

Title

Revision

DBD-EE-023

Radiation Monitoring System

23

ML19130A154

SUNSI Review

Complete

By: RDA

ADAMS

Yes No

Publicly Available

Non-Publicly Available

Non-Sensitive

Sensitive

Keyword:

NRC-002

OFFICE

SRI/DRP/A

RI/DRP/A

DRS/EB1

DRS/EB2

DRS/OB

DRS/IPAT

NAME

JJosey

RKumana

VGaddy

GPick

GWerner

RKellar

SIGNATURE

/RA/

/RA/

/RA/

/RA/

/RA/ CCO for

/RA/

DATE

05/07/19

05/03/19

05/02/19

05/08/19

05/03/19

05/06/19

OFFICE

DRS/RCB

DNMS/RIB

SPE:DRP/A

BC/DRP/A

NAME

NMakris

GWarnick

RAlexander

MHaire

SIGNATURE

/RA/

/RA/

/RA/

/RA/

DATE

05/02/19

05/07/19

05/02/19

5/10/2019

Retrieved from "https://kanterella.com/w/index.php?title=ML19130A154&oldid=4929537"