ML19221B727
| ML19221B727 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/09/2019 |
| From: | O'Keefe N NRC/RGN-IV/DNMS/NMSB-B |
| To: | Peters K Vistra Operations Company |
| References | |
| IR 2019002 | |
| Download: ML19221B727 (34) | |
See also: IR 05000445/2019002
Text
August 9, 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 -
INTEGRATED INSPECTION REPORT 05000445/2019002 AND
Dear Mr. Peters:
On June 30, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
Comanche Peak Nuclear Power Plant, Units 1 and 2. On July 2, 2019, the NRC inspectors
discussed the results of this inspection with Mr. Tom McCool and other members of your staff.
A telephonic re-exit was conducted on July 31, 2019, with Mr. Tom McCool and other members
of your staff to discuss a change in characterization on one finding. The results of this
inspection are documented in the enclosed report.
Three findings of very low safety significance (Green) are documented in this report. Each of
these findings involved violations of NRC requirements. Additionally, one Severity Level IV
violation without an associated finding is documented in this report. We are treating these
violations as non-cited violations (NCVs) consistent with Section 2.3.2.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 Comanche Peak.
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 Comanche Peak.
K. Peters
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 by CYoung for/
Neil F. O'Keefe Chief
Reactor Projects Branch B
Docket Nos. 05000445 and 05000446
License Nos. NPF-87 and NPF-89
Enclosure:
As stated
cc w/ encl: Distribution via LISTSERV
K. Peters
3
COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 - INTEGRATED
INSPECTION REPORT 05000445/2019002 AND 05000446/2019002 - DATED AUGUST 9,
2019
DISTRIBUTION:
DISTRIBUTION:
SMorris, RA
MShaffer, DRA
AVegel, DRP
MHay, DRP
RLantz, DRS
GMiller, DRS
DCylkowski, RC
DDodson, RIV/OEDO
VDricks, ORA
JWeil, OCA
MOBanion, NRR
AMoreno, RIV/CAO
BMaier, RSLO
RKellar, IPAT
JJosey, DRP
RAlexander, DRP
PJayroe, DRP/IPAT
MHerrera, DRMA
AAthar, DRP
LReyna, DRP
R4Enforcement
DOCUMENT NAME: CP2019002-RP-JEJ
ADAMS ACCESSION NUMBER: ML19221B727
X SUNSI Review
ADAMS:
Non-Publicly Available
X Non-Sensitive
Keyword:
By:BKT
X Yes No
X Publicly Available
Sensitive
OFFICE
SRI/DRP/B
RI/DRP/B
RI/DRP/B
AC/DRS/EB1
C/DRS/EB2
C/DRS/OB
NAME
JJosey
AAthar
RKumana
GGeorge
NTaylor
GWerner
SIGNATURE
/RA/
/RA/
/RA/
/RA/
/RA/
/RA by
COsterholtz
for/
DATE
08/05/2019
07/31/2019
07/31/2019
08/05/2019
08/02/2019
08/01/2019
OFFICE
C:CRS/RCB
C:DNMS/RIB
TL/IPAT
C:DRP/B
NAME
MHaire
GWarnick
RKellar
NOKeefee
SIGNATURE
/RA/
/RA/
/RA/
/RA/
DATE
07/31/2019
08/01/2019
8/01/2019
08/08/2019
FFICIAL RECORD COPY
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Numbers:
05000445 and 05000446
License Numbers:
Report Numbers:
05000445/2019002 and 05000446/2019002
Enterprise Identifier: I-2019-002-0011
Licensee:
VISTRA Operations Company, LLC
Facility:
Comanche Peak Nuclear Power Plant, Units 1 and 2
Location:
Glen Rose, TX 76043
Inspection Dates:
March 17, 2019 to June 30, 2019
Inspectors:
R. Alexander, Senior Project Engineer
I. Anchondo-Lopez, Reactor Inspector
A. Athar, Resident Inspector
B. Baca, Health Physicist
L. Carson, Senior Health Physicist
N. Hernandez, Operations Engineer
J. Josey, Branch Chief
R. Kumana, Senior Resident Inspector
B. Larson, Senior Operations Engineer
Approved By:
Neil F. O'Keefe, Chief
Reactor Projects Branch B
Division of Reactor Projects
2
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees
performance by conducting a integrated 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.
List of Findings and Violations
Failure to Report a Change in Medical Condition of a Licensed Operator
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Not Applicable
NCV 05000446,05000445/2019002-
01
Open
Not Applicable
The NRC identified a Severity Level IV non-cited violation (NCV) of 10 CFR 55.25,
"Incapacitation Because of Disability or Illness," for the licensees failure to notify the NRC
within 30 days of a change in a licensed operators medical condition.
Failure to Evaluate a Change to the Facilities AC Power System
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
None (NPP)
The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion
III, Design Control, associated with the licensees failure to assure that a design change met
regulatory requirements for sharing of systems among units. Specifically, the licensee
performed a change to the facility to allow the inclusion of Unit 1 specific safety-related loads
on common panels XEC 2-1 and XEC 1-1 but failed to verify that there would be no adverse
effect on the performance of safety functions if these unitized loads were powered from
Unit 2.
Inadequate Operability Evaluation of Control Room Air Conditioning Unit X-01
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
[H.11] -
Challenge the
Unknown
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-422.01, Operability Determination and
Functionality Assessment Program. Specifically, on December 15, 2018, control room air
conditioning (CRAC) Unit X-01 tripped due to a low lube oil condition caused by freon
absorption in the oil. Operations personnel subsequently declared CRAC X-01 operable and
placed it back in service without understanding the cause of the trip or establishing a
reasonable expectation for operability of the unit.
3
Failure to Provide Adequate Procedure Results in a Reactor Trip
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
Open/Closed
[H.12] - Avoid
Complacency
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 adequate procedural guidance for operators checking for
buzzing relays. This resulted in a feedwater isolation to steam generator 2-04 and operators
inserting a manual reactor trip. The licensee entered this issue into the corrective action
program as Condition Report CR-2019-001949.
Additional Tracking Items
Type
Issue Number
Title
Report Section
Status
LER
05000446,05000445/2
019-001-00
LER 2019-001-00 for
Comanche Peak Nuclear
Power Plant (CPNPP),
Units 1 and 2, LTOP Power
Operated Relief Valve
(PORV) Setpoint.
Closed
LER
05000446/2019-001-
00
LER 2019-001-00 For
Comanche Peak Nuclear
Power Plant (CPNPP) Unit 2,
Manual Reactor Trip Due to
Feedwater Isolation Valve
Closure.
Closed
4
PLANT STATUS
Unit 1 began this inspection period in coast down at 91 percent rated thermal power. The unit
coasted down until April 20, 2019, when the unit was shut down to commence a refueling
outage. On May 26, 2019, the unit began a reactor startup and reached rated thermal power on
May 31, 2019. The unit remained at or near rated thermal power for the remainder of the
inspection period.
Unit 2 operated at or near rated thermal power for the entire inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in
effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with
their attached revision histories are located on the public website at http://www.nrc.gov/reading-
rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared
complete when the IP requirements most appropriate to the inspection activity were met
consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection
Program - Operations Phase. The inspectors performed plant status activities described in
IMC 2515 Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem
Identification and Resolution. The inspectors reviewed selected procedures and records,
observed activities, and interviewed personnel to assess licensee performance and compliance
with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
External Flooding Sample (IP Section 03.04) (1 Sample)
(1)
The inspectors evaluated readiness to cope with external flooding associated with
Squaw Creek Reservoir level reaching 776.5 feet.
Impending Severe Weather Sample (IP Section 03.03) (1 Sample)
(1)
The inspectors evaluated readiness for impending adverse weather conditions for
severe thunderstorms on March 17, 2019.
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1)
Unit 2, train B station service water (protected under defense-in-depth strategy-03) on
May 1, 2019
(2)
Unit 1, centrifugal charging pump 1-02 while pump 1-01 was out of service on
June 11, 2019
5
(3)
Unit 2, motor drive auxiliary feedwater pump 2-02 while pump 2-01 was out of service
on June 13, 2019
(4)
Unit 2, reactor coolant system leakage detection instrumentation on June 24, 2019
71111.04S - Equipment Alignment
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated system configurations during a complete walkdown of the
service water intake and ultimate heat sink system on June 24, 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 zones 2SA1B and 2SA2A, Unit 2, safeguards building 773' elevation on May 28,
2019
(2)
fire zone AA21a, auxiliary building elevation 790' CCW HX room and hallway area
outside of room on May 14, 2019
(3)
fire zone AA21b, auxiliary building elevation 810' hallways on June 21, 2019
(4)
fire zone SB2c, auxiliary building elevation 773 on June 25, 2019
(5)
fire zone SB4, auxiliary building 773 elevation on June 26, 2019
71111.07A - Heat Sink Performance
Annual Review (IP Section 02.01) (1 Sample)
The inspectors evaluated readiness and performance of:
(1)
Unit 1, diesel generator 1-01 jacket water heat exchanger
71111.08P - Inservice Inspection Activities (PWR)
PWR Inservice Inspection Activities Sample (IP Section 03.01) (1 Sample)
(1)
The inspectors verified that the reactor coolant system boundary, steam generator
tubes, reactor vessel internals, risk-significant piping system boundaries, and
containment boundary are appropriately monitored for degradation and that repairs
and replacements were appropriately fabricated, examined and accepted by
reviewing the following activities from April 29 to May 3, 2019:
03.01.a - Nondestructive Examination and Welding Activities.
6
The inspectors directly observed or reviewed records of the following
Nondestructive activities:
1. Ultrasonic Examination
a. Weld TBX-2-2501-29, Report Number 1RF20-UT-018, Elbow-to-Pipe
in Residual Heat Removal System
b. Weld TBX-2-2501-34, Report Number 1RF20-UT-028, Tee-to-Pipe in
Residual Heat Removal System
c. Weld TBX-1-1100-3, Report Number RV-ISI 2019, Reactor Vessel
Intermediate to Lower Shell weld
d. Weld TBX-1-1100-4, Report Number RV-ISI 2019, Reactor Vessel
Lower Shell to Bottom Head weld
e. Weld TBX-1-1100-7, Report Number RV-ISI 2019, Reactor
Vessel Upper Shell Longitudinal Weld
f. Weld TBX-1-4100-1, 2, Report Number SE-338-01 , Hot Leg One
Reactor Vessel Nozzle to Safe End and Safe End to Pipe
g. Weld TBX-1-4100-14, 13, Report Number SE-293-01 , Cold Leg One
Reactor Vessel Nozzle to Safe End and Safe End to Pipe
h. Weld TBX-1-4200-14, 13, Report Number SE-247-01 , Cold Leg Two
Reactor Vessel Nozzle to Safe End and Safe End to Pipe
i.
Weld TBX-1-4300-1, 2, Report Number SE-158-01 , Hot Leg Three
Reactor Vessel Nozzle to Safe End and Safe End to Pipe
j.
Weld TBX-1-4200-1, 2, Report Number SE-202-01 , Hot Leg Two
Reactor Vessel Nozzle to Safe End and Safe End to Pipe
k. Weld TBX-1-4300-14, 13, Report Number SE-113-01 , Cold Leg Three
Reactor Vessel Nozzle to Safe End and Safe End to Pipe
l.
Weld TBX-1-4400-14, 13, Report Number SE-67-01 , Cold Leg Four
Reactor Vessel Nozzle to Safe End and Safe End to Pipe
m. Weld TBX-1-4400-1, 2, Report Number SE-158-01 , Hot Leg Four
Reactor Vessel Nozzle to Safe End and Safe End to Pipe
2. Magnetic Particle Examination
a. Weld TBX-2-2301-H4, Report Number 1RF20-MT-002, Pipe Welded
Attachment in Feedwater System
3. Liquid Penetrant Examination
a. Weld TBX-2-2568-H28, Report Number 1RF20-PT002, Pipe Welded
Attachment in Safety Injection System
4. Visual Examination
a. Weld TBX-2-2568-H28, Report Number 1RF20-VT-055, Pipe Dual
Struts in Safety Injection System
03.01.b - Pressurized-Water Reactor Vessel Upper Head Penetration Examination
Activities were not required this outage.
03.01.c - Pressurized-Water Reactor Boric Acid Corrosion Control Activities.
The inspector reviewed five boric acid corrosion evaluations and associated
corrective actions contained in condition report 2019-03234.
7
03.01.d - Pressurized-Water Reactor Steam Generator Tube Examination Activities
were not required this outage.
Identification and Resolution of Problems:
The inspector reviewed 16 notifications that dealt with inservice inspections issues
and found that items were entered into the corrective action program at the
appropriate level and addressed correctly.
71111.11B - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Requalification Program (IP Section 03.04) (1 Sample)
(1)
Biennial Requalification Written Examinations
The inspectors evaluated the quality of the licensed operator biennial requalification
written examination administered on May 29, 2019.
Annual Requalification Operating Tests
The inspectors evaluated the adequacy of the facility licensees annual requalification
operating test.
Administration of an Annual Requalification Operating Test
The inspectors evaluated the effectiveness of the facility licensee in administering
requalification operating tests required by 10 CFR 55.59(a)(2) and that the facility
licensee is effectively evaluating their licensed operators for mastery of training
objectives.
Requalification Examination Security
The inspectors evaluated the ability of the facility licensee to safeguard examination
material, such that the examination is not compromised.
Remedial Training and Re-examinations
The inspectors evaluated the effectiveness of remedial training conducted by the
licensee, and reviewed the adequacy of re-examinations for licensed operators who
did not pass a required requalification examination.
Operator License Conditions
The inspectors evaluated the licensees program for ensuring that licensed operators
meet the conditions of their licenses.
Control Room Simulator
The inspectors evaluated the adequacy of the facility licensees control room
simulator in modeling the actual plant, and for meeting the requirements contained in
8
Problem Identification and Resolution
The inspectors evaluated the licensees ability to identify and resolve problems
associated with licensed operator performance.
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 1 shutdown on April 20, 2019.
(2)
The inspectors observed and evaluated licensed operator performance in the control
room during Unit 1 startup on May 26, 2019.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
(1)
The inspectors observed and evaluated a simulator-based loss of coolant accident
scenario on June 24, 2019.
71111.12 - Maintenance Effectiveness
Routine Maintenance Effectiveness Inspection (IP Section 02.01) (1 Sample)
The inspectors evaluated the effectiveness of routine maintenance activities associated with
the following equipment and/or safety significant functions:
(1)
Radiation monitors on June 27, 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 associated with insulation removal on the residual
heat removal system on March 18, 2019
(2)
Unit 1, risk mitigating actions associated with Train B EDG system outage window on
April 22, 2019
(3)
Risk mitigating actions associated with spent fuel pool cooling (after Unit 1
fuel offload) on May 1, 2019
(4)
Unit 1, diesel generator 1-02 cylinder liner and piston emergent replacement on
May 15, 2019
(5)
Unit 1, high energy line break damper testing on May 23, 2019
9
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 02.02) (6 Samples)
The inspectors evaluated the following operability determinations and functionality
assessments:
(1)
Units 1 and 2, CRAC X-01 trip on low load on April 8, 2019
(2)
Unit 1, operability of safety systems with power supply panel XEC 2-1 aligned to
Unit 2 on May 9, 2019
(3)
Unit 2 motor drive auxiliary feedwater pump exceeded alert range for vibrations on
May 10, 2019
(4)
Units 1 and 2, potential tornadic missiles due to storage locations on May 28, 2019
(5)
Unit 1, train B EDG exhaust relief valve doghouse insulation worn through on May 30,
2019
(6)
Control room air conditioning calculation did not evaluate heat load from
panel CPX-ECPRCV-13 on June 13, 2019
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)
(3 Samples)
The inspectors evaluated the following temporary or permanent modifications:
(1)
common service water tunnel sump modification on April 15, 2019
(2)
mechanical stress introduction project on May 22, 2019
(3)
power operated relief valve setpoint changes on May 24, 2019
71111.19 - Post-Maintenance Testing
Post Maintenance Test Sample (IP Section 03.01) (7 Samples)
The inspectors evaluated the following post maintenance tests:
(1)
Unit 1, low temperature over pressure protection setpoint change on April 15, 2019
(2)
Unit 1, diesel generator 1-02 following cylinder liner and piston emergent replacement
on April 29, 2019
(3)
Unit 1, replace oil return trap on control room air conditioning unit X-01, work order 5615078 on May 5, 2019
(4)
Unit 1, work order 5505638, replace regulator on valve 1-8825 on May 6, 2019
10
(5)
Unit 1, work order 5754670, replace solid state protection switch S809 on May 9,
2019
(6)
Unit 1, diesel generator 1-01 following turbocharger replacement on May 13, 2019
(7)
Unit 1, work order 5515877, component cooling water supply valve 1-HV-4515-MO
on May 24, 2019
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated 1RF20 activities from March 19 to May 26, 2019.
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance tests:
Containment Isolation Valve Testing (IP Section 03.01) (1 Sample)
(1)
Unit 1, safety injection check valves 1-8956A and 1-8956B on May 23, 2019
Inservice Testing (IP Section 03.01) (1 Sample)
(1)
Unit 1, OPT-503A stroke time test of valve 1-HV-4168 on May 12, 2019
Surveillance Tests (other) (IP Section 03.01) (4 Samples)
(1)
Unit 1, OPT-417A turbine driven auxiliary feedwater pump control panel load shed
test on March 20, 2019
(2)
power operated relief valve testing following setpoint changes on April 10, 2019
(3)
Unit 2, train A RHR surveillance, OPT-203B on May 15, 2019
(4)
Unit 2, train A integrated test surveillance on May 16, 2019
71114.06 - Drill Evaluation
Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01)
(1 Sample)
(1)
The inspectors evaluated an emergency preparedness drill on June 26, 2019
11
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Contamination and Radioactive Material Control (IP Section 02.03) (1 Sample)
The inspectors evaluated licensee processes for monitoring and controlling contamination
and radioactive material.
(1)
The inspectors evaluated licensee processes for monitoring and controlling
contamination and radioactive material. The inspectors verified the following sealed
sources are accounted for and are intact:
Well Source, 2,178 Curies of Cs-137
HP-60193-XSS, 3.3. Curies Am-241(Be)
HP-60265-XSS, 2.0 Curies Pu-238
HP-60250-XSS, 1.06 Curies Cs 137
High Radiation Area and Very High Radiation Area Controls (IP Section 02.05) (1 Sample)
(1)
The inspectors evaluated risk-significant high radiation area and very high radiation
area controls.
Instructions to Workers (IP Section 02.02) (1 Sample)
The inspectors evaluated instructions to workers including radiation work permits used to
access high radiation areas.
(1)
The inspectors evaluated instructions to workers including radiation work permits
used to access high radiation areas:
Radiation work packages
Mechanical Stress Improvement Process (MSIP): RWPs 2019-2604, 2605, 2606, &
2607
Cavity Decontamination: RWP 2019-1607
Westinghouse Refueling & Support: RWPs 2019-1600 & 1601
Scaffolding Support: RWPs 2019-1215 & 1606
Electronic alarming dosimeter alarms
There were no alarms that occurred during the period of this inspection.
12
Labeling of containers
Cavity Decontamination, 55-gallon drum loaded with contaminated mop heads and
radwaste: 5/1/19
Rad Vault Container - 16 loaded with spent radwaste filters; 5/1/19
Unit-1 Volume Control Tank Room stored with seven 55-gallon drums of radwaste:
5/3/19
Radiation Worker Performance and Radiation Protection Technician Proficiency (IP
Section 02.06) (1 Sample)
(1)
The inspectors evaluated radiation worker performance and radiation protection
technician proficiency.
Radiological Hazard Assessment (IP Section 02.01) (1 Sample)
The inspectors evaluated radiological hazards assessments and controls.
(1)
The inspectors evaluated radiological hazards assessments and controls. The
inspectors reviewed the following:
Radiological surveys
M-20190501-40: Unit-1 Reactor Bldg. 854 Pressurizer Compartment 5/01/19
M-20190501-63: Unit-1 Reactor Bldg. 877 Pressurizer Compartment 5/01/19
M-20190421-9: AMS-4 Unit-1 Reactor Bldg. Containment 905 4/21/19
M-20190421-35: Reactor Bldg. 808 Trash rack 4/21/19
Risk significant radiological work activities
MSIP Project: RWPs 2019-2604, 2605, 2606, & 2607
Cavity Decontamination: RWP 2019-1607
Westinghouse Refueling & Support: RWPs 2019-1600 & 1601
Scaffolding Support: RWPs 2019-1215 & 2607
Air sample survey records
M-20190430-20: AMS-4 Unit-1 Containment Hatch 4/30/19
M-20190430-21: AMS-4 Unit-1 Containment Equipment Hatch 4/30/19
13
M-20190428-1: AMS-4 Unit-1 Containment 905 4/27/19
Radiological Hazards Control and Work Coverage (IP Section 02.04) (1 Sample)
The inspectors evaluated in-plant radiological conditions during facility walkdowns and
observation of radiological work activities.
(1)
The inspectors evaluated in-plant radiological conditions during facility walkdowns
and observation of radiological work activities.
Radiological work package for areas with airborne radioactivity
RWP 2019-1607 M-20190430-21: AMS-4 Unit-1 Containment Equipment Hatch
4/30/19
Cavity Decontamination: M-20190428-1: AMS-4 Unit-1 Containment 905 4/27/19
Steam Generator-1 Area HEPA Filter: 5/1/19
Unit-1 RHR-B Pump Room: M-201801225-10 12/25/18
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Engineering Controls (IP Section 02.01) (1 Sample)
The inspectors evaluated airborne controls and radioactive monitoring.
(1)
Installed ventilation systems
CPNPP Control Room Ventilation System
Unit-1 Containment Ventilation System
Unit-1 Primary Plant Ventilation Exhaust Filter X-16
Unit-1 Primary Plant Ventilation Exhaust Filter X-02
Temporary ventilation system setups
Unit-1 Steam Generator Reactor Bldg. 860 HEPA Filter 4/27/19 to 5/2/19
Unit-1 Refueling Floor 860 HEPA Filter 4/27/19 to 5/2/19
Portable or installed monitoring systems
M-20190430-20: AMS-4 Unit-1 Containment Hatch 4/30/19
M-20190430-21: AMS-4 Unit-1 Containment Equipment Hatch 4/30/19
14
M-20190428-1: AMS-4 Unit-1 Containment 905 4/27/19
Self-Contained Breathing Apparatus for Emergency Use (IP Section 02.03) (1 Sample)
The inspectors evaluated self-contained breathing apparatus program implementation.
(1)
Status and surveillance records for SCBAs
Inspected 3 SCBAs located in the CPNPP Control Room 5/2/19
W.O. 5345020 Surveillance: 10 Units 1 & 2 Primary Assembly SCBAs, 7/10/17
W.O. 5345020 Surveillance: 9 Control Room SCBAs, 7/20/17
SCBA fit for on-shift operators
3 Unit-1 Operation/Control Room Staff:
Reactor Operator Qualification
Operator Crew Qualification
Operations Crew Supervisor
SCBA maintenance check
Inspected 3 SCBAs located in the CPNPP Control Room 5/2/19
W.O. 5345020 Surveillance: 10 Units 1 & 2 Primary Assembly SCBAs, 7/10/17
W.O. 5345020 Surveillance: 9 Control Room SCBAs, 7/20/17
Use of Respiratory Protection Devices (IP Section 02.02) (1 Sample)
The inspectors evaluated the licensees use of respiratory protection devices by:
(1)
Evaluations for the use of respiratory protection
b. RWP 2019-1607: 16 Radiation Protection Technicians (Contractor & CPNPP
Staff)
Respiratory protection use during work activities
b. RWP 2019-1607: 16 Radiation Protection Technicians (Contractor & CPNPP
Staff)
Medical fitness for use of respiratory protection devices
b. Nine Senior Radiation Protection Technicians, ERO Qualified through
February 2020
c. Sixteen Radiation Protection Technicians (Contractor & CPNPP Staff)
Observation of donning, doffing and functional test
Four Westinghouse Refuel Contractors
Two CPNPP Decontamination Staff
Three Unit-1 Operation/Control Room Staff
Respiratory protection device evaluation
Six Gen Tex Pure Flows units
15
Six 3M Versa Flows
Eight Full Face Respirators and associated filter cartridges
Four Self-Contained Breathing Apparatus (SCBA)
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (2 Samples)
(1)
Unit 1 from April 1, 2018 through March 31, 2019
(2)
Unit 2 from April 1, 2018 through March 31, 2019
BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)
(1)
Unit 1 from April 1, 2018 through March 31, 2019
(2)
Unit 2 from April 1, 2018 through March 31, 2019
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04) (2 Samples)
(1)
Unit 1 from April 1, 2018 through March 31, 2019
(2)
Unit 2 from April 1, 2018 through March 31, 2019
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
(1)
September 30, 2018 - April 30, 2019
PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual
Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample.
(IP Section 02.16) (1 Sample)
(1)
September 30, 2018 - April 30, 2019
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)
Unit 2 safety bus undervoltage due to loss of 25 kV loop
(2)
Appendix R emergency lighting
Semiannual Trend Review (IP Section 02.02) (1 Sample)
(1)
The inspectors reviewed the licensees corrective action program for trends that might
be indicative of a more significant safety issue. The inspectors review was focused
on operability evaluations performed by operations department personnel during the
period of December 1, 2018 to June 30, 2019.
16
71153 - Followup of Events and Notices of Enforcement Discretion
Event Report (IP Section 03.02) (2 Samples)
The inspectors evaluated the following licensee event reports (LERs):
(1)
Licensee Event Report 05000446/2019-01, "Manual Reactor Trip Due To Feedwater
Isolation Valve Closure," on June 17, 2019 (ADAMS Accession No. ML19127A143).
The circumstances surrounding this LER are documented in report Section Inspection
Results.
(2)
Licensee Event Report 05000445/2019-01, "LTOP Power Operated Relief
Valve (PORV) Setpoint," on June 28, 2019 (ADAMS Accession No. ML19127A079).
The circumstances surrounding this LER are documented in report Section Inspection
Results.
INSPECTION RESULTS
Observation: Semi-Annual Trend Review
The inspectors review focused on inadequate operability and functionality evaluations and
considered the results of daily inspector corrective action program item screenings during the
period of December 1, 2018 to June 30, 2019.
The inspectors reviewed operability evaluations documented in station condition reports.
During their review the inspectors identified a declining trend associated with operator
knowledge when performing operability and functionality evaluations. Specifically, the
inspectors identified eight examples, including one non-cited violation, of inadequate
operability determinations. Examples are:
TR-2018-5446 identified an issue with a temperature controller for the operations
support center, emergency plan equipment. Operations personnel failed to evaluate
functionality of the equipment despite procedure STI-433.01, Maintaining Equipment
Important to Emergency Response, stating that issues with this equipment required
functionality assessments. This issue could have affected habitability of the
operations support center.
CR-2018-008521 identified that control room air conditioning unit X-01 tripped
following a valid surveillance test start demand due to a low lube oil condition caused
by freon absorption in the oil. Operations personnel subsequently declared
CRAC X-01 operable and placed it back in service without understanding the cause of
the trip or establishing a reasonable expectation for operability of the unit. This could
affect the ability of the machine to perform its safety function of being able to start and
run when needed and had caused an actual failure during the surveillance test.
CR-2019-000324 identified a component on steam generator atmospheric relief
valves as not meeting environmental qualification requirements. Operations
personnel incorrectly determined that the atmospheric relief valves were not required
to be environmentally qualified despite the FSAR stating they were required to
17
function in a harsh environment. This could have resulted in equipment not being
able to function during an event.
CR-2019-001628 identified that an atmospheric relief valve was experiencing leakage.
Operations personnel incorrectly determined that the leakage did not affect the reactor
despite the FSAR stating that during accident scenarios the atmospheric relief valves
are assumed to be closed. This appeared to place the plant outside accident
analyses and therefore could have affected the licensee's ability to mitigate an
accident.
CR-2019-001830 and CR-2019-001836 identified both surface and pitting corrosion
on service water piping. Operations personnel evaluated the surface corrosion in the
documented operability evaluation but failed to evaluate the pitting corrosion. This
could have affected the structural integrity of the piping system which could have
affected the systems ability to respond to an event.
CR-2019-003498 noted boric acid on a safety related snubber. Operations personnel
failed to evaluate the ability of the component to function in the identified condition.
CR-2019-003672 identified that the fire protection system had experienced a water
hammer event. Operations personnel failed to evaluate the functionality of the piping
for this type of event.
Upon identification of these issues by the inspectors the licensee appropriately entered these
issues into the station's corrective action program.
While the inspectors determined that most of these issues were minor in nature, inspectors
determined that they were the result of operator knowledge gaps with respect to the facility's
licensing basis documented in the FSAR, which is a primary part of the process for assessing
operability and functionality. The inspectors determined that these issues were indicative of a
declining trend associated with operator knowledge when performing operability evaluations.
Failure to Report a Change in Medical Condition of a Licensed Operator
Cornerstone
Severity
Cross-Cutting
Aspect
Report
Section
Not
Applicable
NCV 05000446,05000445/2019002-01
Open
Not
Applicable
The NRC identified a Severity Level IV non-cited violation (NCV) of 10 CFR 55.25,
"Incapacitation Because of Disability or Illness," for the licensees failure to notify the NRC
within 30 days of a change in a licensed operators medical condition.
Description:
On August 8, 2018, during a biennial license physical examination for a senior reactor
operator, the designated medical examiner (DME) determined that the individual required a
prescription for corrective lenses; therefore, a condition was required to be added to the
individuals license. The DME further determined that a no solo condition was also required
for a worsening health condition. At that time, the individual was undergoing evaluation and
treatment by his primary care physician. The facility licensee failed to notify the NRC within
30 days of learning of the disability or illness. The inspectors identified this failure to inform
18
the NRC on March 19, 2019, during a sampling review of medical records for individual
license holders.
The inspectors verified, through interviews, that the individual had been wearing the
prescription lenses as prescribed, and that at no time was the individual in solo operation of
the controls, as verified by station log reviews and interviews with various station personnel.
The inspectors also noted that the individual had committed no errors while on shift or been
involved in any operational mishaps or near misses.
Corrective Actions: On March 28, 2019, Comanche Peak Nuclear Power Plant submitted a
letter to the NRC with a revised NRC Form 396, Certification of Medical Examination by
Facility Licensee, that reflected two new license conditions: one for wearing the prescription
lenses and one prohibiting solo operation of the controls. The prescription lens and no solo
conditions were required for fitness-for-duty reasons in accordance with the American
National Standards Institute (ANSI)-3.4-1983, Medical Certification And Monitoring Of
Personnel Requiring Operator Licenses For Nuclear Power Plants,
Corrective Action References: Condition Report CR-2019-002428
Performance Assessment: The inspectors determined this violation was associated with a
minor performance deficiency.
The performance deficiency is minor because a licensing decision was not made due to the
absence of this medical information, and the individual did follow the required restrictions for
the diagnosed medical conditions.
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 inspectors determined the violation to be a Severity Level IV violation similar to
Example 6.4.d.1.a in the NRC Enforcement Policy. Specifically, the licensee non-willfully
failed to inform the NRC of a change in an operators medical condition, which did not
contribute to the NRC making an incorrect regulatory decision. In addition, the individual
adhered to the ANSI 3.4 requirements for the changes in his medical condition.
Violation: Title 10 CFR 55.25 requires, in part, that if a licensed senior operator develops a
permanent physical condition that causes the licensee to fail to meet the requirements of
10 CFR 55.21, the facility shall notify the Commission within 30 days of learning of the
diagnosis. For conditions where a license condition is required, the facility licensee must
provide medical certification on NRC Form 396, "Certification of Medical Examination by
Facility Licensee." Contrary to the above from August 8, 2018 to March 28, 2019, the
licensee failed to notify the Commission, within 30 days of learning the diagnosis, of a change
in medical condition of a licensed senior operator that developed a permanent physical
condition that caused the licensed senior operator to fail to meet the requirements of
10 CFR 55.21. Specifically, on August 8, 2018, during a biennial license physical for the
senior reactor operator, the DME determined that the individual required conditions on his
license for prescription corrective lenses and a no solo condition, but the NRC did not
receive the revised NRC Form 396 until March 28, 2019.
19
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 Facilities AC Power System
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
None (NPP)
The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion
III, Design Control, associated with the licensees failure to assure that a design change met
regulatory requirements for sharing of systems among units. Specifically, the licensee
performed a change to the facility to allow the inclusion of Unit 1 specific safety-related loads
on common panels XEC 2-1 and XEC 1-1 but failed to verify that there would be no adverse
effect on the performance of safety functions if these unitized loads were powered from
Unit 2.
Description:
While performing status walkdowns on April 1, 2019, the inspectors noted that the licensee
had shifted power supplied to AC buses XEC 2-1 and XEC 1-1 from Unit 1 power to Unit 2
power. Because of previous issues identified with common power buses
(NCV 05000445/2019001-02, Failure to Evaluate a Change to the Facility DC Power
System) the inspectors asked the licensee if there were any Unit 1 specific loads supplied by
these panels, and whether these loads were operable while powered from Unit 2.
The licensee informed the inspectors that there were Unit 1 loads on buses XEC 2-1 and
XEC 1-1 and that since the swap had been done in accordance with an approved procedure,
SOP-607A, 118 VAC Distribution System and Inverters, and the busses were being power
by a safety-related 1E power supply from the other unit, all loads were operable.
Inspectors were concerned that this configuration did not appear consistent with the facility
design requirements and current licensing bases. Of note, 10 CFR 50, Appendix A, General
Design Criterion 5, Sharing of structures, systems, and components, states 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, including, in the event of an accident in one unit, an orderly shutdown
and cooldown of the remaining units.
Inspectors reviewed NUREG-0797, Safety Evaluation Report related to the operation of
Comanche Peak Steam Electric Station, Units 1 and 2 Docket Nos. 50-445 and 50-446,
Supplement 22 and the facilities FSAR and identified the following with regard to sharing
safety systems:
The site was licensed with a commitment to Regulatory Guide (RG) 1.81, Shared
Emergency and Shutdown Electric Systems for Multi-Unit Nuclear Power Plants,
without exceptions.
RG 1.81 contains three regulatory Positions.
o C.1 required that DC systems in multi-unit nuclear power plants should not be
shared.
20
o C.2 covered plants which were under construction before June 1,1973 and
stated that these plants would be reviewed by the NRC on an individual case
basis. This position gave seven criteria that should be satisfied by the
applicant.
o C.3 stated that for multi-unit plants for which a construction permit was issued
on or after June 1, 1973, should have separate and independent onsite
emergency and shutdown electric systems.
The facility submitted their construction permit application for Unit 1 after June 1,
1973.
NUREG-0797, Section 8.3, states that there is no sharing of emergency power
sources between units, which is in accordance with RG 1.81.
FSAR Section 8.3.1.1.9, Sharing of Equipment between Two Units, states, in part,
that nuclear-safety-related loads associated with each unit are powered exclusively
from Class 1E systems of that particular unit and nuclear-safety-related loads
common to both units are powered from Class 1E MCCs and distribution panels which
have supplies from each unit. The inspectors noted that this was no longer true
because there were unit-specific loads on shared panels XEC1-1 and XEC 2-1.
Furthermore, the inspectors noted that in January 2000, the licensee had previously
discovered that they had unit-specific safety-related loads from both Units 1 and 2 on
common panels XEC 1-1 and XEC 2-1 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 as Condition Report CR-2000-000142. As
corrective actions the licensee removed the Unit 2 loads from these common panels, changed
the normal power supply of XEC 2-1 from Unit 2 to Unit 1, left the Unit 1 loads on both of
these common panels, revised the description of their commitment to RG 1.81, and
performed a 10 CFR 50.59 screening for this change.
The inspectors reviewed 50.59 screening document 59SC-2000-000142-02-02 and noted that
as justification for this change the licensee stated, in part;
that the design is controlled such that loads of only Unit 1 are fed from
a common panel and the normal source of power for the panel is
Unit 1the unit specific loads of only one unit are fed from a common
panel and the common panel normal power source is the same unit,
therefore, under normal operation, the interaction between the units for
maintenance and test operation will be no different than what is required
for a common panel. The time when such common panel is aligned to
the units other than the one whose specific loads it feeds will be limited,
any additional interaction needed for maintenance and test activities will
be limited only.
Based on this justification, the licensee concluded that the requirements of RG 1.81,
Regulatory Position C2 were met. Specifically, the sharing of the system is limited between
two units only. A single failure at the system level, due to redundancy for common systems
being maintained the same as for each unit, shall not preclude capability to automatically
supply minimum ESF loads in any one unit and safely shutdown the other unit assuming a
21
loss of off-site power.
The inspectors determined that the licensee inappropriately applied Regulatory Position C.2.
Specifically:
RG 1.81 Regulatory Position C.2 was not applicable to Comanche Peak because their
construction permit application was submitted after June 1, 1973. Also, this position
included the statement that the NRC would review the proposed design for plants
using this set of standards and did not contain an allowance for self-approval.
Comanche Peak was originally evaluated for compliance against regulatory
position C.3, and the licensee failed to properly change their commitment through the
commitment change process.
The licensee failed to document an analysis that demonstrated that such sharing will
not significantly impair their ability to perform their safety functions, including, in the
event of an accident in one unit, an orderly shutdown and cooldown of the remaining
units.
The unit specific loads on the common panels could result in new failure modes for previously
evaluated accidents. For example, non-1E buses 1EB1-2 and 1EB1-3 have shunt trip
isolation devices that receive a signal from a contacts powered from XEC 1-1. The
licensing basis credits these shunt trips as isolation devices between the non-1E and 1E
buses. During a safety injection actuation on Unit 1 with XEC 1-1 being powered from Unit 2
a single failure (loss of power to this bus) would result in the shunt trip devices not actuating
and leaving the non-1E buses aligned to the Unit 1 emergency diesel generators. This
configuration could impact the emergency diesel generators capability to provide adequate
power for safe shutdown. Based on this the inspectors determined that the inclusion of Unit 1
specific safety related loads on common panels XEC 1-1 and XEC 2-1 was a change to the
facility as described in the FSAR. The inspectors also determined that the licensee had failed
to ensure that this change was subject to design control measures commensurate with those
applied to the original design. Specifically, the NRC had reviewed and approved the stations
shared safety-related electrical bus schemes, and in this instance the site had authorized this
change without any review to determine if there were adverse interactions.
The licensee entered this violation into their corrective action program. The licensee
performed an analysis, EV-CR-2019-003684-2, to evaluate the effects on Unit 1 with cross
powered safety loads. This analysis determined that there were potential detrimental effects
associated with cross connecting Unit 1 safety loads to Unit 2 and recommended that these
loads be powered from Unit 1 pending further review.
Corrective Actions: The licensee entered this violation into their corrective action program.
The licensee performed an analysis, EV-CR-2019-003684-2, to evaluate the effects on Unit 1
with cross powered safety loads. This analysis determined that there were potential
detrimental effects associated with cross connecting Unit 1 safety loads to Unit 2 and
recommended that these loads be powered from Unit 1 pending further review.
Corrective Action References: CR-2019-003684
Performance Assessment:
22
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Design Control attribute of the mitigating systems
cornerstone and 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).
Significance: The inspectors assessed the significance of the finding using Appendix A, The
Significance Determination Process (SDP) for Findings At-Power.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to
this finding because the inspectors determined the finding did not reflect present licensee
performance.
Enforcement:
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.
Inadequate Operability Evaluation of Control Room Air Conditioning Unit X-01
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
[H.11] -
Challenge the
Unknown
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-422.01, Operability Determination and
Functionality Assessment Program. Specifically, on December 15, 2018, control room air
conditioning (CRAC) Unit X-01 tripped due to a low lube oil condition caused by freon
absorption in the oil. Operations personnel subsequently declared CRAC X-01 operable and
placed it back in service without understanding the cause of the trip or establishing a
reasonable expectation for operability of the unit.
Description: On December 15, 2018, during the performance of integrated testing CRAC
Unit X-01 tripped due to low lube oil pressure. Operators declared X-01 inoperable and
initiated Condition Report (CR) 2018-008521 to capture this issue in the station's corrective
action program. Operations also requested an evaluation from engineering for the identified
condition.
Engineering documented an evaluation in EV-CR-2018-008521-1 later in the day on
December 15, 2018. This evaluation concluded that CRAC X-01 had tripped due to
refrigerant migration into the oil, and went on to state that the condition did not need to be
corrected because under normal system loading the velocity of the refrigerant traveling
23
through the system and the piping arrangements returns the oil back to the compressors
crankcase (removing the potential for the low pressure condition at the suction of the
compressor). Operators then declared CRAC X-01 operable.
The inspectors subsequently reviewed CR-2018-008521 and the engineering evaluation
documented in EV-CR-2018-008521-1. During their review inspectors questioned the
adequacy of the engineering evaluation and the operability decision with respect to the cause
of the trip being refrigerant migration. Specifically, the safety function of the CRAC units is to
start and run when needed to maintain the control room environment during accidents, and
refrigerant migration occurs when the unit is secured and is caused by differential pressure
between the oil in the compressors crankcase and refrigerant. Inspectors noted that the
CRACs are not continuously in operation and each unit typically runs one CRAC for two
weeks with the other in standby for that period of time, and CRAC X-01 had been secured for
approximately two weeks prior to integrated testing.
The inspectors reviewed Procedure STI-422.01, Operability Determination and Functionality
Assessment Program, and noted that the following definitions are provided for operability
declaration in section 4.15, presumption of operability in section 4.17, and reasonable
expectation of operability in Section 4.18. These definitions establish that: an operability
declaration is a decision made by an SRO on the operating shift crew that there is reasonable
expectation that an SSC can perform its specified safety function; the presumption of
operability is based on the concept that technical specifications are organized and
implemented on the presumption that systems are operable and without information to the
contrary, it is reasonable to assume that once a system or component is established as
operable, it will remain operable with surveillances providing that assurance; and a
reasonable expectation of operability is the expectation that the evidence collected supports
determining that a TS SSC is capable of performing its specified safety function.
Section 6.2.G states, in part, that operability determinations should include the effect or
potential effect of the degraded or non-conforming condition on the affected SSCs ability to
perform the specified safety functions, and whether there is a reasonable expectation of
operability, including the basis for the determination.
The inspectors determined that the licensee had failed to evaluate the potential effect of
refrigerant migration on CRAC X-01 with respect to its ability to perform its specified safety
function and establish a reasonable expectation of operability. Specifically, the licensee
evaluated the ability of the system to continue running with the condition present but failed to
evaluate the ability to successfully start and reach a stable running condition, despite the fact
that the evaluation was triggered by a failure to successfully start. The inspectors informed
the licensee of their concerns and the licensee initiated CR-2019-005092 to capture this issue
in the stations corrective action program.
Corrective Actions: The licensee performed an operability evaluation which established a
reasonable expectation of operability and is evaluating corrective actions.
Corrective Action References: CR-2019-002092
Performance Assessment:
Performance Deficiency: The failure to evaluate the potential effect of refrigerant migration
on CRAC X-01 with respect to its ability to perform its specified safety function and establish
a reasonable expectation of operability was a performance deficiency.
24
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 and 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). Specifically, the licensee evaluated the issue
enough to understand the physical mechanism that caused of the chiller trip but failed to
evaluate the reason this mechanism was present to the degree that it could create a condition
that resulted in an actual trip.
Significance: The inspectors assessed the significance of the finding using Appendix A, The
Significance Determination Process (SDP) for Findings At-Power.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with
uncertain conditions. Risks are evaluated and managed before proceeding. Specifically,
operators failed to stop and use all available information to ensure they fully understood the
issue prior to declaring CRAC X-01 operable.
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.
Procedure STI-422.01, Operability Determination and Functionality Assessment Program, a
procedure that is appropriate to the circumstances of evaluating the operability of
safety-related components, Section 6.2.G requires, in part, that operability determinations
should include the effect or potential effect of the degraded or non-conforming condition on
the affected SSCs ability to perform the specified safety functions, and whether there is a
reasonable expectation of operability, including the basis for the determination.
Contrary to the above, on December 15, 2019, an activity affecting quality was not
accomplished in accordance with procedures appropriate to the circumstances. Specifically,
operations personnel failed to follow the requirements of station procedure STI-422.01,
Operability Determination and Functionality Assessment Program, Section 6.2.G.
Specifically, the licensee failed to evaluate the potential effect of refrigerant migration on
CRAC X-01 with respect to its ability to start with refrigerant in the oil and establish a
reasonable expectation of operability.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Provide Adequate Procedure Results in a Reactor Trip
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
Open/Closed
[H.12] - Avoid
Complacency
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
25
licensees failure to establish adequate procedural guidance for operators checking
for buzzing relays. This resulted in a feedwater isolation to steam generator 2-04 and
operators inserting a manual reactor trip.
Description:
On March 3, 2019 an operator was checking for buzzing relays as required by the Shift
Manager Daily Activities Log. During these checks the operator heard an audible buzzing
sound and attempted to positively identify exactly which relay was buzzing by touching the
suspected relay to check for vibration. When the operator made contact with the relay they
also inadvertently made contact with the relay plunger which resulted in steam generator 2-04
feedwater isolation valve 2-HV-2137 shutting. When valve 2-HV-2137 shut this resulted in a
loss of main feedwater to steam generator 2-04 which caused an alarm for divergent level
and resulted in operators manually tripping the reactor. The licensee entered this issue in the
stations corrective action program as CR-2019-001949.
The licensee performed an Organizational Effectiveness Investigation and determined that
the cause of this event was that the Shift Manager Daily Activities Log did not provide
adequate guidance to operators for how to safely perform buzzing relay checks. The lack of
adequate guidance resulted in the operator using their own judgement based on their
individual risk perception.
The inspectors reviewed the licensee evaluation and determined that the licensees actions
had determined the cause and the proposed corrective actions appeared to be adequate to
address this issue.
Corrective Actions: The licensees corrective actions for this issue were to remove buzzing
relay checks from the operators responsibilities and perform a risk review of all routine
operations activities to ensure that adequate procedural guidance is provided.
Corrective Action References: CR-2019-001949
Performance Assessment:
Performance Deficiency: The failure to provide adequate procedural guidance on how to
perform buzzing relay checks was a performance deficiency.
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. It adversely affected the cornerstone objective of limiting the likelihood of events
that upset plant stability and challenge critical safety functions during shutdown as well as
power operations because the finding resulted in operators manually tripping the reactor.
Specifically, the inadequate procedure resulted in an operator using their own judgement and
inadvertently touching the relay plunger causing a feedwater isolation to steam
generator 2-04.
Significance: The inspectors assessed the significance of the finding using Appendix A, The
Significance Determination Process (SDP) for Findings At-Power.
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the
possibility of mistakes, latent issues, and inherent risk, even while expecting successful
outcomes. Individuals implement appropriate error reduction tools. Specifically, the operator
26
used their own judgement to decide to touch a relay based on their individual risk perception
without planning for mistakes or understanding the risk of the activity.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, states, in part, that activities affecting quality shall be prescribed by documented
instructions, procedures, or drawings, of a type appropriate to the circumstances. Contrary to
the above, on March 3, 2019 the licensee failed to prescribe by documented instructions,
procedures, or drawings appropriate to the circumstances that provided guidance for
operators performing buzzing relay checks. This resulted in a feedwater isolation to steam
generator 2-04 and operators inserting a manual reactor trip.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On July 2, 2019, the inspectors presented the quarterly resident inspector inspection
results to Mr. Tom McCool and other members of the licensee staff.
On May 3, 2019, the inspectors presented the Occupational Safety Cornerstone
Inspection IP 71124.01/03 and PI IP 71151 ORS/PRS Exit Meeting to Mr. T. McCool,
Senior Vice President and other members of the licensee staff.
On May 15, 2019, the inspectors presented the Inservice Inspection Exit Meeting to
Dave Goodwin, Director Site Engineering and other members of the licensee staff.
On June 3, 2019, the inspectors presented the Biennial Requalification Inspection to
Gary Struble, Regulatory Affairs Specialist and other members of the licensee staff.
27
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71111.08P Corrective Action
Documents
CR-YYYY-NNNN
2017-12392, 2017-11619, 2017-12371, 2017-12391,
2017-12859, 2018-00318, 2018-02385, 2018-03983,
2018-05224, 2018-08475, 2018-08535, 2019-03374,
2019-00542, 2019-03234, 2019-00211, 2019-05839.
Drawings
TBX-1-1100
Inservice Inspection Location Isometric
1
Miscellaneous
CP-201900257
Unit 1 - Third Interval ASME Section XI Inservice Inspection
Program Plan
2
Procedures
MRS-SSP-3498
TBX RV Nozzle to Safe End Weld MSIP Field Service
Procedure
0
STA-737
Boric Acid Corrosion Detection and Evaluation
8
STI-737.01
Boric Acid Corrosion Detection and Evaluation
0
TX-ISI-11
Liquid Penetrant Examination for Comanche Peak Nuclear
Power Plant
5
TX-ISI-302
Ultrasonic Examination of Austenitic Piping Welds
5
TX-ISI-70
Magnetic Particle Examination for Comanche Peak Nuclear
Power Plant
14
TX-ISI-8
VT-1 and VT-3 Visual Examination Procedure
11
WDI-PJF-
1321507-EPP-
001
Examination Program Plan (Scan Plan)
1
71111.11B Corrective Action
Documents
CR-YYYY-NNNN
2017-03868, 2017-03964, 2017-04417, 2017-04499,
2017-04572, 2017-04683, 2017-06489, 2017-07890,
2017-08016, 2017-08672, 2017-09069, 2017-09904,
2017-09915, 2018-03312, 2018-03681, 2018-03745,
2018-03796, 2018-04038, 2018-05098, 2018-05130,
2018-05461, 2018-06369, 2018-07381, 2018-07384,
2018-08504, 2019-00555, 2019-01052, 2019-01240
TR-YYYY-NNNN
2017-03029, 2017-03066, 2017-03068, 2017-03083,
2017-03292, 2017-03770, 2017-05956, 2017-06540,
2017-08662, 2017-07003, 2017-07446, 2017-09096,
2017-12278, 2018-06584, 2018-06280, 2018-06291,
2018-06321, 2018-07045, 2019-01040
28
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
LORT Cycle 18-5 Simulator Crews
2018 Licensed Operator Requalification Written Exam
Feb 22,
2019
2018 Licensed Operator Requalification Operating Test
Jan 30, 2019
2017-2018 Requalification Operating Test Sample Plan
Mar 11,
2019
2017-2018 Requalification Written Exam Sample Plan
Mar 11,
2019
RO Remedial Exam #1
Feb 28,
2018
SRO Remedial Exam #1
Feb 28,
2018
License Reactivation - 2017-2019
Simulator Differences List
Jan 28, 2019
Lessons Learned from Design Modifications and Current
Events Training
Sep 25,
2017
MC-
OPD1.F17.IR4
Cycle 17-6 Design Modifications and Current Events
Feb 20,
2018
SOC-18-03
Simulator Oversight Committee Meeting Minutes
Mar 7, 2018
SOC-18-06
Simulator Oversight Committee Meeting Minutes
Jun 28, 2018
SOC-19-02
Simulator Oversight Committee Meeting Minutes
Feb 5, 2019
Procedures
NTG-104
Nuclear Training Guideline 104 - Implementation
21
ODA-102
Operations Department Administrative Manaual section 102
- Conduct of Operations
27
ODA-407
Operations Department Administrative Manual section 407 -
Procedures Use and Adherence
16
OGPD-13
Operations Guideline 13 - Simulator Training Standards
Jun 27, 2017
OGPD-3
Operations Guideline 3 - Operations Standards and
Expectations
Jun 14, 2018
OGPD-6
Operations Guideline 6 - Operations Department
Performance Management, Monitoring, and Improvement
Jan 7, 2019
SAPT-004
Simulator Annual Performance Test 004 - Core Performance
Test
2
SAPT-1
Simulator Annual Performance Test 1 - Steady State Test
2
29
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
SAPT-2
Simulator Annual Performance Test 2 - Transient
Performance Test
2
SAPT-3
Simulator Annual Performance Test 3 - Malfunction Test
2
STA-105
Station Administrative Manual section 105 - Nuclear Training
Department Administration
13
STA-106
Station Procedure 106 - Nuclear Training Records
1
STA-121
Station Procedure 121 - Licensed Operator Physicals and
License Application Process
5
STA-419
Station Procedure 419 - Management Oversight of Training
Programs
15
STA-501
Station Procedure 501 - Nonroutine Reporting
21
STI-214.01
Station Instruction 214.01 - Control of Timed Operator
Actions
1
TRA-204
Training Procedure 204 - Licensed Operator Requalification
Training
18
TRA-206
Training Procedure 206 - Examination Control and
Implementation
0
TRA-207
Training Procedure 207 - Simulator Configuration
Management
0
TRA-208
Training Procedure 208 - Simulator Training
0
TRA-295
Training Procedure 205 - Shift Technical Advisor Training
12
TRI-204.01
Training Instruction 204.01 - Licensed Operator
Requalification Training Processes and Program Reviews
0
TRI-206.02
Training Instruction 206.02 - NRC Requalification Exam
Development Process
0
TRI-206.02
NRC Exam Development Process
0
TRI-206.04
Training Instruction 206.04 - Licensed Operator Requal
Periodic Exams
0
TRI-208.02
Training Instruction 208.02 - Conduct of LORT Simulator
Training
0
Corrective Action
Documents
CR-YYYY-NNNN
2000-000142, 2019-003684
Miscellaneous
000142-02-02
30
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
IR/TR/CR
2019 - 002448, 002331, 001511, 001117, 001547
2018 - 009060, 009021, 009016, 008400, 00840
Corrective Action
Documents
Resulting from
Inspection
IR-2019
003649, 003650, 003681, 003681
Miscellaneous
1RF20
Cavity Recovery Action Plan for MSIP
2/1/19
DIDCP No.
!RF20-02
Contingency Plan: Containment Closure & Maintaining
Common Area Pressure Boundary
2
RPI-305-2
LHRA/VHRA Key Log
April - May
2019
Procedures
RPI-212
Radioactive Source Control
13
RPI-302
Radiation and Contamination Surveys
1
RPI-303
Radiological Air Sampling
0
RPI-304
Radiological Posting and Labeling
3
RPI-305
Access Controls for High Radiation Areas
3
RPI-629
Radiological Risk Management
5
RPI-700
Sealed Source Leak Testing
13
STA-650
General Health Physics Plan
8
STA-660
Control of High Radiation Areas
18
Radiation
Surveys
M-20190421-29
1-155D-G 1RF20 Initial Entry Survey
M-20190421-35
1-154A/D 1RF20 808' Trash Racks
M-20190421-36
1-154A-D 1RF20 Initial Entry Survey
M-20190421-9
1 905 1RF20 Initial Entry
M-20190423-56
1RF20 (1-155H-K) Post CRUD Burst
M-20190430-33
!RF20 Post Decon of Cavity Floor After Reactor Head Set for
MSIP 1-157
Radiation Work
Permits (RWPs)
RWP 2019-1215
Scaffolding in the RCA
RWP 2019-1600
WEC Refueling
RWP 2019-1604
RWP-2019-1607
MSIP Support, Cavity Decon
Self-Assessments EVAL 2018-0006
CPNPP Nuclear Oversight: Work Management/RP
Audit
7/26/18
31
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
IR/TR/CR
2017-009867, 010605, 010692, 010827, 011717, 011762,
012840, 013033, 2018-000842, 001245, 001285, 001859,
003371, 004052, 005228, 006575, 007249, 008342, 008400
2019-000200, 001117, 001421, 001511, 001547, 001622,
001669, 002036, 002128, 002192, 002448
Corrective Action
Documents
Resulting from
Inspection
IR-2019
003680, 003682
Procedures
RPI-115
Alarm Response
10
RPI-303
Radiological Air Sampling
0
RPI-902
Issue and Control of Respiratory Protection
17
RPI-923
Maintenance and Use of Portable HEPA Filter Ventilation
Units
5
SAF-104
Inspection of Respiratory Protection Equipment
(Maintenance & Repair
12
SOP-801A
Containment Ventilation
14
SOP-816
Primary Plant Ventilation Systems
18
STA-652
Radioactive Material Control
21
STA-653
Contamination Control Program
20
STA-802
Control Room Ventilation
13
STI-211.06
Use of Respiratory Protection
1
Self-Assessments EVAL 2018-006
CPNPP Nuclear Oversight: Work Management/RP Audit
7/28/19
71151
Procedures
NGM-710
Management Review Meetings
11/2/18
RPI-629
Radiological Risk Management
5
Calculations
STI-422.01
Operability Determination and Functionality Assessment
Program
23
Corrective Action
Documents
CR-YYYY-NNNN
2018-008521, 2019-000324, 2019-001628, 2019-001830,
2019-001836, 2019-003498, 2019-003672, 2019-005285
TR-YYYY-NNNN
2018-005446
CR-YYYY-NNNN
2019-001949
Miscellaneous
Shift Manager Daily Activities Log