ML16314C026
| ML16314C026 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 11/08/2016 |
| From: | Jeremy Groom NRC/RGN-IV/DRP/RPB-A |
| To: | Peters K TEX Operations Company |
| JEREMY GROOM | |
| References | |
| IR 2016003 | |
| Download: ML16314C026 (29) | |
See also: IR 05000445/2016003
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION IV
1600 E. LAMAR BLVD.
ARLINGTON, TX 76011-4511
November 8, 2016
Mr. Ken Peters, Senior Vice President
and Chief Nuclear Officer
TEX Operations Company LLC
P.O. Box 1002
Glen Rose, TX 76043
SUBJECT: COMANCHE PEAK NUCLEAR POWER PLANT - NRC INTEGRATED
INSPECTION REPORT 05000445/2016003 and 05000446/2016003
Dear Mr. Peters:
On September 30, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Comanche Peak Nuclear Power Plant, Units 1 and 2. On September 29,
2016, the NRC inspectors discussed the results of this inspection with Mr. S. Sewell, Senior
Director of Engineering and Regulatory Affairs, and other members of your staff. Inspectors
documented the results of this inspection in the enclosed inspection report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
All of these findings involved violations of NRC requirements.
If you contest the violations or significance of these NCVs, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with
copies to the Regional Administrator, Region IV; the Director, Office of Enforcement,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident
inspector at the Comanche Peak Nuclear Power Plant, Units 1 and 2.
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 Regional Administrator, Region IV; and the NRC resident inspector at the
Comanche Peak Nuclear Power Plant, Units 1 and 2.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public
Inspections, Exemptions, Requests for Withholding, a copy of this letter, its enclosure, and your
response (if any) will be available electronically for public inspection in the NRCs Public
Document Room or from the Publicly Available Records (PARS) component of the NRC's
K. Peters -2-
Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible
from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic
Reading Room).
Sincerely,
/RA/
Jeremy R. Groom, Branch 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 05000445/2016003 and
w/ Attachment: Supplemental Information
cc w/ encl: Electronic Distribution
SUNSI Review ADAMS Non-Sensitive Publicly Available Keyword:
By: JRG Yes No Sensitive Non-Publicly Available NRC-002
OFFICE SRI:DRP/A RI:DRP/A SPE:DRP/A BC:EB1 BC:EB2 BC:OB BC:PSB2
NAME JJosey RKumana RAlexander TFarnholtz GWerner VGaddy HGepford
SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ /RA/
DATE 10/21/16 10/24/16 10/19/16 10/19/16 10/25/16 10/20/16 10/20/16
OFFICE TL-IPAT BC:DRP/A
NAME THipschman JGroom
SIGNATURE /RA/ /RA/
DATE 10/19/16 11/8/16
Letter to Ken Peters from Jeremy Groom dated November 8, 2016
SUBJECT: COMANCHE PEAK NUCLEAR POWER PLANT-NRC INTEGRATED
INSPECTION REPORT 05000445/2016003 and 05000446/2016003
DISTRIBUTION:
Regional Administrator (Kriss.Kennedy@nrc.gov)
Deputy Regional Administrator (Scott.Morris@nrc.gov)
DRP Director (Troy.Pruett@nrc.gov)
DRP Deputy Director (Ryan.Lantz@nrc.gov)
DRS Director (Anton.Vegel@nrc.gov)
DRS Deputy Director (Jeff.Clark@nrc.gov)
Senior Resident Inspector (Jeffrey.Josey@nrc.gov)
Resident Inspector (Rayomand.Kumana@nrc.gov)
Administrative Assistant (VACANT)
Branch Chief, DRP/A (Jeremy.Groom@nrc.gov)
Senior Project Engineer, DRP/A (Ryan.Alexander@nrc.gov)
Project Engineer, DRP/A (Thomas.Sullivan@nrc.gov)
Project Engineer, DRP/A (Mathew.Kirk@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Project Manager (Margaret.Watford@nrc.gov)
Team Leader, DRS/IPAT (Thomas.Hipschman@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
ACES (R4Enforcement.Resource@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
RIV Congressional Affairs Officer (Angel.Moreno@nrc.gov)
RIV/ETA: OEDO (Jeremy.Bowen@nrc.gov)
ROPreports
Electronic Distribution for Comanche Peak Nuclear Power Plant
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000445, 05000446
Report: 05000445/2016003 and 05000446/2016003
Licensee: TEX Operations Company, LLC
Facility: Comanche Peak Nuclear Power Plant, Units 1 and 2
Location: 6322 N. FM-56, Glen Rose, Texas
Dates: July 1 through September 30, 2016
Inspectors: J. Josey, Senior Resident Inspector
R. Kumana, Resident Inspector
W. Cullum, Reactor Inspector
Approved Jeremy R. Groom
By: Chief, Project Branch A
Division of Reactor Projects
A-1 Attachment
SUMMARY
IR 05000445/2016003 and 05000446/2016003; 07/01/2016 - 09/30/2016; Comanche Peak
NPP, Units 1 and 2; Maintenance Effectiveness, Problem Identification and Resolution
The inspection activities described in this report were performed between July 1, 2016, through
September 30, 2016, by the resident inspectors at the Comanche Peak Nuclear Power Plant
and an inspector from the NRCs Region IV office. Two findings of very low safety significance
(Green) are documented in this report. Both of these findings involved a violation of NRC
requirements. The significance of inspection findings is indicated by their color (Green, White,
Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance
Determination Process. Their cross-cutting aspects are determined using Inspection Manual
Chapter 0310, Aspects within the Cross-Cutting Areas. Violations of NRC requirements are
dispositioned in accordance with the NRC Enforcement Policy. The NRCs program for
overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process.
Cornerstone: Initiating Events
- Green. The inspectors identified a non-cited violation of 10 CFR 50.65(a)(4), Requirements
for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees
failure to adequately manage the increase in risk associated with the potential for a loss of
decay heat removal during refueling outages. Specifically, the licensee implemented a risk
management action that did not reduce the risk, but instead called for placing a safety
injection pump in service during periods where this action is prohibited by plants technical
specifications for low temperature over pressure protection. The inspectors determined this
was an ineffective risk management action because the use of a safety injection pump
during low pressure and temperature conditions would place the plant in an unanalyzed
condition, resulting in an increase in risk. As an immediate corrective action, the licensee
initiated Condition Report CR-2015-009109 to evaluate appropriate risk management
actions. This finding was entered into the licensees corrective action program as Condition
Report CR-2015-009109.
The failure to manage the increase in risk associated with the potential for a loss of decay
heat removal during refueling activities is a performance deficiency. The performance
deficiency was more than minor, and therefore a finding, because it was associated with the
procedure quality attribute of the Initiating Events Cornerstone and affected the cornerstone
objective to limit the likelihood of events that upset plant stability and challenge critical safety
functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance
Determination Process, dated May 19, 2005, Flowchart 1, Assessment of Risk Deficit, the
inspectors determined the need to calculate the risk deficit to determine the significance of
this issue. A senior reactor analyst performed a bounding qualitative assessment and
determined the incremental core damage probability deficit was less than 1E-6 and the
incremental large early release probability deficit was less than 1E-7, based on the
availability of additional equipment to mitigate the loss of decay heat removal. In
accordance with Flowchart 1 in Appendix K, because incremental core damage probability
deficit was less than 1E-6 and incremental large early release probability deficit was less
than 1E-7, the finding screened as having very low safety significance (Green). The finding
has a human performance cross-cutting aspect associated with bases for decisions, in that,
the licensee failed to ensure that operations leadership adequately communicate potential
A-2
problems with the risk management action to start a safety injection pump when in a mode
of applicability for low temperature over pressure protection [H.10]. (Section 4OA2)
Cornerstone: Mitigating Systems
- Green. The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2), Requirements
for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, the
licensee failed to demonstrate that the performance of the Unit 2 auxiliary feedwater check
valves was being effectively controlled through the performance of appropriate preventive
maintenance. The licensees failure to perform appropriate maintenance resulted in several
failures of the check valves. The licensee entered this issue into corrective action program
as CR-2016-008312.
The licensees failure to effectively monitor the performance of maintenance rule scoped
equipment in accordance with 10 CFR 50.65(a)(2) was a performance deficiency. The
performance deficiency was more than minor, and therefore a finding, because it was
associated with the equipment performance attribute of the Mitigating Systems Cornerstone
and affected the cornerstone objective to ensure availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences. Specifically,
the licensee failed to demonstrate that the performance of the Unit 2 auxiliary feedwater
check valves was being effectively controlled through the performance of appropriate
preventive maintenance which resulted in failures of the valves. Using Inspection Manual
Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for
Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of
very low safety significance (Green) because the finding (1) was not a deficiency affecting
the design and qualification of a mitigating structure, system, or component, and did not
result in a loss of operability or functionality, (2) did not represent a loss of system and/or
function, (3) did not represent an actual loss of function of at least a single train for longer
than its allowed outage time, or two separate safety systems out-of-service for longer than
their technical specification allowed outage time, and (4) did not represent an actual loss of
function of one or more non-technical specification trains of equipment designated as high
safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensees maintenance
rule program. A cross-cutting aspect was not assigned to this finding because the
performance deficiency occurred in 1996, and therefore, is not indicative of current licensee
performance. (Section 1R12)
Licensee-Identified Violations
None
A-3
PLANT STATUS
Unit 1 and Unit 2 began the inspection period at approximately 100 percent power and operated
at that power level for the entire inspection period.
REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01)
.1 Summer Readiness for Offsite and Alternate AC Power Systems
a. Inspection Scope
On July 20, 2016, the inspectors completed an inspection of the stations off-site and
alternate-ac power systems. The inspectors inspected the material condition of these
systems, including transformers and other switchyard equipment to verify that plant
features and procedures were appropriate for operation and continued availability of off-
site and alternate-ac power systems. The inspectors reviewed outstanding work orders
and open condition reports for these systems. The inspectors walked down the
switchyard to observe the material condition of equipment providing off-site power
sources. The inspectors verified that the licensees procedures included appropriate
measures to monitor and maintain availability and reliability of the off-site and alternate-
ac power systems.
These activities constituted one sample of summer readiness of off-site and alternate-ac
power systems, as defined in Inspection Procedure 71111.01.
b. Findings
No findings were identified.
1R04 Equipment Alignment (71111.04)
.1 Partial Walk-Down
a. Inspection Scope
The inspectors performed partial system walk-downs of the following risk-significant
systems:
- August 4, 2016, Unit 2, turbine driven and motor driven auxiliary feedwater
pumps
- August 23, 2016, Unit 1, train A 125 VDC distribution system
- September 20, 2016, Units 1 and 2, fire protection piping in the service water
intake structure
A-4
The inspectors reviewed the licensees procedures and system design information to
determine the correct lineup for the systems. They visually verified that critical portions
of the systems or trains were correctly aligned for the existing plant configuration.
These activities constituted three partial system walk-down samples as defined in
Inspection Procedure 71111.04.
b. Findings
No findings were identified.
1R05 Fire Protection (71111.05)
.1 Quarterly Inspection
a. Inspection Scope
The inspectors evaluated the licensees fire protection program for operational status
and material condition. The inspectors focused their inspection on four plant areas
important to safety:
- August 5, 2016, Fire area 2SC7, Unit 2 turbine driven auxiliary feedwater pump
room
- September 19, 2016, Fire area SB2a, Unit 1 train A residual heat removal, safety
injection, containment spray pumps rooms
- September 19, 2016, Fire area SE16, Unit 1 Electrical Equipment Room
- September 19, 2016, Fire area 2SE16, Unit 2 Electrical Equipment Room
For each area, the inspectors evaluated the fire plan against defined hazards and
defense-in-depth features in the licensees fire protection program. The inspectors
evaluated control of transient combustibles and ignition sources, fire detection and
suppression systems, manual firefighting equipment and capability, passive fire
protection features, and compensatory measures for degraded conditions.
These activities constituted four quarterly inspection samples, as defined in Inspection
Procedure 71111.05.
b. Findings
No findings were identified.
A-5
.2 Annual Inspection
a. Inspection Scope
On September 20, 2016, the inspectors completed their annual evaluation of the
licensees fire brigade performance. This evaluation included observation of two fire
drills:
- March 22, 2016, Unit 1, announced drill, contaminated waste fire drill, 832 foot
corridor
- June 22, 2016, Unit 2, announced drill, 858 foot elevation valve gallery
During these drills the inspectors evaluated the capability of the fire brigade members,
the leadership ability of the brigade leader, the brigades use of turnout gear and fire-
fighting equipment, and the effectiveness of the fire brigades team operation. The
inspectors also reviewed whether the licensees fire brigade met NRC requirements for
training, dedicated size and membership, and equipment.
These activities constituted one annual inspection sample, as defined in Inspection
Procedure 71111.05.
b. Findings
No findings were identified.
1R06 Flood Protection Measures (71111.06)
a. Inspection Scope
On September 23, 2016, the inspectors completed an inspection of the stations ability to
mitigate flooding due to internal causes. After reviewing the licensees flooding analysis,
the inspectors selected one plant area containing risk-significant structures, systems,
and components that were susceptible to flooding:
- Units 1 and 2, service water intake structure
The inspectors reviewed plant design features and licensee procedures for coping with
internal flooding. The inspectors walked down the selected areas to inspect the design
features, including the material condition of seals, drains, and flood barriers. The
inspectors evaluated whether operator actions credited for flood mitigation could be
successfully accomplished.
These activities constituted completion of one flood protection measures sample as
defined in Inspection Procedure 71111.06.
b. Findings
No findings were identified.
A-6
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
(71111.11)
.1 Review of Licensed Operator Requalification
a. Inspection Scope
On September 27, 2016, the inspectors observed a portion of an annual requalification
test for licensed operators. The inspectors assessed the performance of the operators
and the evaluators critique of their performance.
These activities constituted completion of one quarterly licensed operator requalification
program sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
.2 Review of Licensed Operator Performance
a. Inspection Scope
Inspectors observed the performance of on-shift licensed operators in the plants main
control room. At the time of the observations, the plant was in a period of heightened
activity or risk due to testing being performed on reactor protection and response to
unusual plant conditions. The inspectors observed the operators performance of the
following activities:
- July 13, 2016, Unit 2, Observation during slave relay testing
- August 8, 2016, Unit 2, Observation of operators response to heater drain pump
seal water low pressure alarm
- September 26, 2016, Unit 1, Observation of reactor trip breaker testing
In addition, the inspectors assessed the operators adherence to plant procedures,
including conduct of operations procedure and other operations department policies.
These activities constituted completion of one quarterly licensed operator performance
sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope
The inspectors reviewed two instances of degraded performance or condition of safety-
related structures, systems, and components (SSCs):
A-7
- August 20, 2016, Unit 2, main feedwater system split flow bypass check valves
- September 23, 2016, Unit 1, pressurizer heater group C blown fuse
The inspectors reviewed the extent of condition of possible common cause SSC failures
and evaluated the adequacy of the licensees corrective actions. The inspectors
reviewed the licensees work practices to evaluate whether these may have played a
role in the degradation of the SSCs. The inspectors assessed the licensees
characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance
Rule), and verified that the licensee was appropriately tracking degraded performance
and conditions in accordance with the Maintenance Rule.
These activities constituted completion of two maintenance effectiveness samples, as
defined in Inspection Procedure 71111.12.
b. Findings
Introduction. The inspectors identified a Green, non-cited violation of 10 CFR
50.65(a)(2), Requirements for monitoring the effectiveness of maintenance at nuclear
power plants. Specifically, the licensee failed to demonstrate that the performance of
the Unit 2 auxiliary feedwater check valves was being effectively controlled through the
performance of appropriate preventive maintenance.
Description. On November 11, 2015, the licensee conducted in-service testing on
feedwater check valve 2FW-0191, one of four steam generator split flow bypass check
valves. During the test, check valve 2FW-0191 failed to meet the sites acceptance
criteria indicating the valve failed to seat. The licensee stopped the test and initiated
Condition Report CR-2015-10961 to document the test failure.
Subsequently, the system engineer performed a maintenance rule functional failure
review of this issue. This review determined that the failure of valve 2FW-0191 to seat
was not a maintenance rule functional failure and the function would remain in (a)(2)
status. Inspectors questioned this assessment because one of the scoped functions of
this feedwater check valve is to shut to prevent bypassing flow from the steam
generators. During discussions with the licensee, the inspectors determined that system
engineer was only evaluating the split flow check valves performance against the main
feedwater systems criteria to provide feedwater to the steam generator, and not against
the criteria related to the valves ability to shut to prevent bypassing flow from the steam
generators. Inspectors also determined that the licensee was not performing
preventative maintenance on the check valves to ensure their ability to close and seat
properly.
The inspectors subsequently reviewed the last test data for all four of the steam
generator split flow bypass check valves. In this review the inspectors noted that in
2011 valve 2FW-0192 had failed to meet the established acceptance criteria, yet the
failure was not noted as a functional failure. Additionally, in 2012, valves 2FW-0191,
2FW-0192, and 2FW-0193 all failed to meet the established acceptance criteria, and
again the failures were not noted as functional failures.
The inspectors noted that 10 CFR 50.65(a)(2) requires, in part, that monitoring as
specified in 10 CFR 50.65(a)(1) is not required where it has been demonstrated that the
A-8
performance of a system is being effectively controlled through the performance of
appropriate preventive maintenance, such that the system remains capable of
performing its intended function. Based on their review, the inspectors determined that
the licensee failed to demonstrate that the performance of the Unit 2 feedwater check
valves was being effectively controlled. Specifically, the licensee was not performing
preventative maintenance on the check valves, resulting in the valves failing to close on
multiple occasions during testing.
The inspectors informed the licensee of the concerns and the licensee initiated condition
report CR-2016-008312 to capture this issue in the stations corrective action program.
The licensee recognized that they were not correctly monitoring the function of these
check valves. Specifically, the licensee determined that monitoring the check valves
only as part of the main feedwater system was not adequate since the systems
performance criteria is to provide feedwater to the steam generators, and the check
valves function is to close to prevent bypass flow. The licensee subsequently performed
a review to determine if other safety-related check valves were also not being monitored
correctly. Based on this review the licensee determined that there were 841 safety-
related check valves (of which 230 were classified as run to failure) that were not being
monitored against their scoped criteria. To correct this issue, the licensee created a new
monitoring function for safety related check valves which monitors the close function,
and moved the equipment to 10 CFR 50.65(a)(1) monitoring requirements because they
determined that they were not able to demonstrate that the performance of the check
valves was being effectively controlled.
Analysis. The licensees failure to effectively monitor the performance of maintenance
rule scoped equipment in accordance with 10 CFR 50.65(a)(2) was a performance
deficiency. The performance deficiency was more than minor, and therefore a finding,
because it was associated with the equipment performance attribute of the Mitigating
Systems Cornerstone and affected the cornerstone objective to ensure availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, the licensee failed to demonstrate that the
performance of the Unit 2 auxiliary feedwater check valves was being effectively
controlled through the performance of appropriate preventive maintenance which
resulted in failures of the valves. Using Inspection Manual Chapter (IMC) 0609,
Appendix A, The Significance Determination Process (SDP) for Findings At-Power,
dated June 19, 2012, inspectors determined that this finding was of very low safety
significance (Green) because the finding (1) was not a deficiency affecting the design
and qualification of a mitigating structure, system, or component, and did not result in a
loss of operability or functionality, (2) did not represent a loss of system and/or function,
(3) did not represent an actual loss of function of at least a single train for longer than its
allowed outage time, or two separate safety systems out-of-service for longer than their
technical specification allowed outage time, and (4) did not represent an actual loss of
function of one or more non-technical specification trains of equipment designated as
high safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensees
maintenance rule program. A cross-cutting aspect was not assigned to this finding
because the performance deficiency occurred in 1996 when the steam generator split
flow bypass check valve was initially scoped under the Maintenance Rule, and therefore,
is not indicative of current licensee performance.
Enforcement. Title 10 CFR 50.65(a)(1) requires, in part, that holders of an operating
license shall monitor the performance of systems and components against licensee
A-9
established goals, in a manner sufficient to provide reasonable assurance that such
structures, systems, and components are capable of fulfilling their intended safety
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 of a
system is being effectively controlled through the performance of appropriate preventive
maintenance, such that the system remains capable of performing its intended function.
Contrary to the above, from initial maintenance rule scoping in 1996 to September 2016,
the licensee did not monitor the performance of the Unit 2 auxiliary feedwater system
check valves against licensee-established goals in a manner sufficient to provide
reasonable assurance that the check valves were capable of fulfilling their intended
safety functions, and the licensee did not demonstrate that the performance of check
valves was being effectively controlled through the performance of appropriate
preventive maintenance, such that the system remained capable of performing its
intended function. In response to this issue the licensee created a new monitoring
function for safety related check valves, and moved the equipment to 10 CFR
50.65(a)(1) monitoring requirements pending further review. Since this violation was of
very low safety significance (Green) and has been entered into the corrective action
program as Condition Report CR-2016-008312, this violation is being treated as a non-
cited violation consistent with Section 2.3.2 of the NRC Enforcement Policy.
(NCV 05000446/2016003-01, Failure to Adequately Monitor Feedwater System Check
Valve Performance)
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
On July 7, 2016, the inspectors reviewed a risk assessment and the risk management
actions taken by the licensee in response to elevated risk associated with performing an
oil sample on spent fuel pool pump X-01.
The inspectors verified that this risk assessment was performed timely and in
accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant
procedures. The inspectors reviewed the accuracy and completeness of the licensees
risk assessment and verified that the licensee implemented appropriate risk
management actions based on the result of the assessment.
The inspectors also observed portions of three emergent work activities that had the
potential to affect the functional capability of mitigating systems:
- August 18, 2016, Unit 2, Steam generator blowdown isolation valve 2-HV-2399
elastomer replacement
- September 1, 2016, Units 1 and 2, unanalyzed condition associated with the
turbine driven auxiliary feedwater pumps
- September 16, 2016, Unit 2, loop A safety chiller emergent maintenance
The inspectors verified that the licensee appropriately developed and followed a work
plan for these activities. The inspectors verified that the licensee took precautions to
minimize the impact of the work activities on unaffected SSCs.
A-10
These activities constituted completion of four maintenance risk assessments and
emergent work control inspection samples, as defined in Inspection Procedure 71111.13.
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments (71111.15)
a. Inspection Scope
The inspectors reviewed seven operability determinations that the licensee performed
for degraded or nonconforming SSCs:
- March 28, 2016, CR-2016-003089, operability determination for control room air
conditioner X-01 partial refrigerant charge
- July 12, 2016, CR-2016-006613, operability determination for diesel generator
2-01 86-2 lockout relay actuation
- August 22, 2016, CR-2016-007251, operability determination for turbine driven
auxiliary feedwater pump 1-01 indicating light socket/bulb melted
- August 24, 2016, CR-2016-007653, operability determination for motor driven
auxiliary feedwater pump room heat up analyses
- August 31, 2016, CR-2016-007840, operability determination for safety injection
pump 2-01 oil leak
- September 8, 2016, CR-2016-008000, operability determination for diesel
generator 2-01 failed KVAR meter
- September 21, 2016, CR-2016-007880, operability determination for auxiliary
feedwater pumps following identification of an unanalyzed condition
The inspectors reviewed the timeliness and technical adequacy of the licensees
evaluations. Where the licensee determined the degraded SSC to be operable the
inspectors verified that the licensees compensatory measures were appropriate to
provide reasonable assurance of operability. The inspectors verified that the licensee
had considered the effect of other degraded conditions on the operability of the
degraded SSC.
These activities constituted completion of seven operability and functionality review
samples, as defined in Inspection Procedure 71111.15.
b. Findings
No findings were identified.
A-11
1R18 Plant Modifications (71111.18)
a. Inspection Scope
On September 15, 2016, the inspectors reviewed a temporary plant modification to
remove sentinel valves from the turbine driven auxiliary feedwater pumps on Unit 1
and 2.
The inspectors verified that the licensee had installed these temporary modifications in
accordance with technically adequate design documents. The inspectors verified that
these modifications did not adversely impact the operability or availability of affected
SSCs. The inspectors reviewed design documentation and plant procedures affected by
the modifications to verify the licensee maintained configuration control.
These activities constituted completion of one sample of temporary modifications, as
defined in Inspection Procedure 71111.18.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed four post-maintenance testing activities that affected risk-
significant SSCs:
- April 5, 2016, Unit 1, offsite power supply breaker 1EA2-1 post maintenance test
- May 25, 2016, Unit 1, service water pump 1-01 replacement
- August 23, 2016, Unit 2, Steam generator 2-03 blowdown isolation valve
2-HV-2399 testing following elastomer replacement
- September 15, 2016, Unit 1 and Unit 2, turbine driven auxiliary feedwater pumps
following temporary modification
The inspectors reviewed licensing and design-basis documents for the SSCs and the
maintenance and post-maintenance test procedures. The inspectors observed the
performance of the post-maintenance tests to verify that the licensee performed the tests
in accordance with approved procedures, satisfied the established acceptance criteria,
and restored the operability of the affected SSCs.
These activities constituted completion of four post-maintenance testing inspection
samples, as defined in Inspection Procedure 71111.19.
A-12
b. Findings
No findings were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors observed four risk-significant surveillance tests and reviewed test results
to verify that these tests adequately demonstrated that the SSCs were capable of
performing their safety functions:
Other surveillance tests:
- May 26, 2016, Unit 1, stroke test of power operated relief valve 1-PCV-456
- August 5, 2016, Unit 2, start and flow test of the turbine driven auxiliary
feedwater pump
- August 23, 2016, Unit 1, stroke test of containment sump pump discharge line
outside-containment isolation valve 1-HV-5157
- September 8, 2016, Unit 2, start test of diesel generator 2-01
The inspectors verified that these tests met technical specification requirements, that the
licensee performed the tests in accordance with their procedures, and that the results of
the test satisfied appropriate acceptance criteria. The inspectors verified that the
licensee restored the operability of the affected SSCs following testing.
These activities constituted completion of four surveillance testing inspection samples,
as defined in Inspection Procedure 71111.22.
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation (71114.06)
.1 Emergency Preparedness Drill Observation
a. Inspection Scope
The inspectors observed an emergency preparedness drill on September 28, 2016, to
verify the adequacy and capability of the licensees assessment of drill performance.
The inspectors reviewed the drill scenario, observed the drill from the simulator and
emergency operations facility, and attended the post-drill critique. The inspectors
verified that the licensees emergency classifications, off-site notifications, and protective
action recommendations were appropriate and timely. The inspectors verified that any
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emergency preparedness weaknesses were appropriately identified by the licensee in
the post-drill critique and entered into the corrective action program for resolution.
These activities constituted completion of one emergency preparedness drill observation
sample, as defined in Inspection Procedure 71114.06.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and
Security
4OA1 Performance Indicator Verification (71151)
.1 Mitigating Systems Performance Index: Emergency AC Power Systems (MS06)
a. Inspection Scope
The inspectors reviewed the licensees mitigating system performance index data for the
period of July 1, 2015 through June 30, 2016 to verify the accuracy and completeness of
the reported data. The inspectors used definitions and guidance contained in Nuclear
Energy Institute Document 99-02, Regulatory Assessment Performance Indicator
Guideline, Revision 7, to determine the accuracy of the reported data.
These activities constituted verification of the mitigating system performance index for
emergency ac power systems for Units 1 and 2, as defined in Inspection
Procedure 71151.
b. Findings
No findings were identified.
.2 Mitigating Systems Performance Index: High Pressure Injection Systems (MS07)
a. Inspection Scope
The inspectors reviewed the licensees mitigating system performance index data for the
period of July 1, 2015 through June 30, 2016 to verify the accuracy and completeness of
the reported data. The inspectors used definitions and guidance contained in Nuclear
Energy Institute Document 99-02, Regulatory Assessment Performance Indicator
Guideline, Revision 7, to determine the accuracy of the reported data.
These activities constituted verification of the mitigating system performance index for
high pressure injection systems for Units 1 and 2, as defined in Inspection
Procedure 71151.
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b. Findings
No findings were identified.
.3 Mitigating Systems Performance Index: Heat Removal Systems (MS08)
a. Inspection Scope
The inspectors reviewed the licensees mitigating system performance index data for the
period of July 1, 2015 through June 30, 2016 to verify the accuracy and completeness of
the reported data. The inspectors used definitions and guidance contained in Nuclear
Energy Institute Document 99-02, Regulatory Assessment Performance Indicator
Guideline, Revision 7, to determine the accuracy of the reported data.
These activities constituted verification of the mitigating system performance index for
heat removal systems for Units 1 and 2, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution (71152)
.1 Routine Review
a. Inspection Scope
Throughout the inspection period, the inspectors performed daily reviews of items
entered into the licensees corrective action program and periodically attended the
licensees condition report screening meetings. The inspectors verified that licensee
personnel were identifying problems at an appropriate threshold and entering these
problems into the corrective action program for resolution. The inspectors verified that
the licensee developed and implemented corrective actions commensurate with the
significance of the problems identified. The inspectors also reviewed the licensees
problem identification and resolution activities during the performance of the other
inspection activities documented in this report.
b. Findings
No findings were identified.
.2 Annual Follow-up of Selected Issues
a. Inspection Scope
The inspectors selected two issues for an in-depth follow-up:
- During refueling outage 2RF15, October 2015, and refueling outage 1RF18,
May 2016, the licensee credited defense in depth contingency plans, risk
assessments with specified risk management actions, for time periods when the
reactor coolant system would be in a loops not filled condition or when shutdown
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cooling would be in a reduced availability condition due to the increase in risk for
the activities.
The inspectors assessed the licensees risk assessments and the specified risk
management actions. The inspectors identified that the licensee failed to
appropriately manage the risk associated with the activities.
- On May 18, 2016, after completion of preventative maintenance on the lube oil
cooler for coolant charging pump 1-01, a service water leak was discovered
coming from the cooler head. Upon disassembly, the licensee discovered
significant pitting on the head for the heat exchanger. The licensee initiated
Condition Report 2016-004868 to evaluate the issue, though an operability
evaluation was not performed at the time because the unit was not in a mode of
applicability for the charging pump. The licensee determined that this condition
had been previously identified in Condition Report CR-2014-001804, and parts
were on order to replace the pitted head. The licensees corrective action was to
apply Loctite #2, a sealant material, to stop the leak, noting that this had
previously been evaluated as acceptable in Condition Report CR-2006-001208.
Upon further review inspectors determined that the evaluation performed in CR-
2006-001208 was a one-time evaluation for use of Loctite #2, and did not
establish a basis for the current use. Therefore, an operability evaluation was
required for the subsequent use of Loctite. The licensee initiated Condition
Reports CR-2016-004936 and CR-2016-006674 to address this issue, and
documented a current operability evaluation for use of the Loctite.
Inspectors determined that this issue was a minor violation of Title 10 CFR Part
50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which
requires, in part, that activities affecting quality shall be accomplished in
accordance with documented instructions, procedures, or drawings, of a type
appropriate to the circumstances. Station Procedure STI-442.01, Operability
Determination and Functionality Assessment Program, is an Appendix B quality
related procedure that is appropriate to the circumstances for evaluating the
operability of safety-related components. Station Procedure STI-442.01 step 6.1,
requires, in part, that when a potential degraded or nonconforming condition is
identified, the shift manager should ensure the operability determination process
is initiated to determine the operability of the structure, system or component.
The inspectors assessed the licensees problem identification threshold, cause
analyses, extent of condition reviews and compensatory actions. The inspectors
verified that the licensee appropriately prioritized the planned corrective actions
and that these actions were adequate to correct the condition.
These activities constituted completion of two annual follow-up sample as defined in
b. Findings
Introduction. The inspectors identified a Green non-cited violation of
10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants, for the licensees failure to adequately manage the increase in
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risk associated with the potential for a loss of decay heat removal during refueling
outages.
Description. During refueling outage 2RF15, October 2015, when the licensee was
setting up for vacuum fill of the reactor coolant system, inspectors reviewed the stations
defense in depth contingency plan 2RF15-01. The inspectors determined that this
contingency plan was a risk assessment with specified risk management actions for
periods when the reactor coolant system would be in a loops not filled condition or
periods of reduced availability of the shutdown cooling system. Inspectors noted that the
contingency plan for these periods of increased risk directed that if residual heat removal
(shutdown cooling) is lost, operators should establish alternate cooling flow path using
Station Procedure ABN-104, Residual Heat Removal System Malfunction, Revision 9,
section 8.
Inspectors reviewed ABN-104, section 8 and noted that it directed operators to start a
safety injection pump in response to a loss of shutdown cooling. Inspectors identified a
concern that the action to start a safety injection pump would occur while in the mode of
applicability for technical specification 3.4.12, Low Temperature Overpressure
Protection System. Technical specification 3.4.12 requires the safety injection pumps
be made incapable of injecting due to concerns of over pressurizing the reactor coolant
system in modes 4, 5, and 6 (the latter only when the reactor vessel head is installed).
The licensee initiated Condition Report CR-2015-009109 to capture the inspectors
concern in the stations corrective action program.
Subsequently, during refueling outage 1RF18, May 2016, inspectors noted that the
licensee again credited a defense in depth contingency plan (1RF18-01) which again
would have operators start a safety injection pump when technical specification 3.4.12
was in effect. During subsequent reviews, the inspectors determined that the licensee
did not have an evaluation for starting a safety injection pump when low temperature
overpressure protection was in effect.
Inspectors determined that the specified risk management action to start a safety
injection pump would restore flow to the core to mitigate the loss of shutdown cooling.
However, the inspectors also determined that the plant is not analyzed for using a safety
injection pump during periods when the reactor coolant system is at low temperatures
requiring low temperature overpressure protection. The proposed use of safety injection
pumps as described in ABN-104, section 8, without analyses for sufficient relief
capability, created the potential for vessel overpressurization and a challenge to the
reactor coolant system barrier. Any challenge to the reactor coolant system barrier
would serve to increase risk. The inspectors also noted that the licensee had several
options to mitigate a potential loss of shutdown cooling that are analyzed during period
where low temperature overpressure protection is required. Specifically, the inspectors
identified that the licensee could start centrifugal charging pumps to restore core flow
following a loss of shutdown cooling. These pumps have slightly less capacity than the
safety inspection pumps which would be bounded by the relief capability required in
technical specification 3.4.12.
Inspectors informed the licensee of the additional concerns and the licensee added them
to Condition Report CR-2015-009109. Inspectors determined that the licensee had not
started a safety injection pump when technical specification 3.4.12 was in effect during
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1RF19 or 2RF18. As corrective actions, the licensee amended Condition Report
CR-2015-009109 to evaluate appropriate risk management actions.
Analyses. The failure to manage the increase in risk associated with the potential for a
loss of decay heat removal during refueling activities is a performance deficiency. The
performance deficiency was more than minor, and therefore a finding, because it was
associated with the procedure quality attribute of the Initiating Events Cornerstone and
affected the cornerstone objective to limit the likelihood of events that upset plant
stability and challenge critical safety functions during shutdown as well as power
operations. Using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk
Assessment and Risk Management Significance Determination Process, dated May 19,
2005, Flowchart 1, Assessment of Risk Deficit, and determined the need to calculate
the risk deficit to determine the significance of this issue. A senior reactor analyst
performed a bounding qualitative assessment, using insights from Inspection Manual
Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process,
and determined the incremental core damage probability deficit was less than 1E-6 and
the incremental large early release probability deficit was less than 1E-7. The influential
assumptions used by the senior reactor analyst included the low exposure time that the
plant is in LTOP conditions, the initiating event frequency associated with a loss of decay
heat removal, available operator mitigation actions that would prevent the use of safety
injection pumps, and the availability of additional equipment to mitigate the loss of decay
heat removal.
In accordance with Flowchart 1 in Appendix K, because incremental core damage
probability deficit was less than 1E-6 and incremental large early release probability
deficit was less than 1E-7, the finding screened as having very low safety significance
(Green). The finding has a human performance cross-cutting aspect associated with
bases for decisions, in that, the licensee failed to ensure that operations leadership
adequately communicate potential problems with the risk management action to start a
safety injection pump when in a mode of applicability for low temperature over pressure
protection [H.10].
Enforcement. Title 10 CFR 50.65(a)(4) requires, in part, that licensees shall assess and
manage the increase in risk that may result from proposed maintenance activities.
Defense in depth contingency plans 2RF15-01 and 1RF18-01 implement pre-planned
risk assessments and specified risk management actions for times during refueling
outages when the reactor coolant system is depressurized and level is lowered.
Contrary to the above, from October 3, 2015, through May 31, 2016, the licensee failed
to manage the increase in risk from proposed maintenance activities. Specifically, the
licensee implemented a risk management action that did not reduce the risk, instead it
called for placing the plant in an unanalyzed condition which could elevate risk. As an
immediate corrective action the licensee initiated Condition Report CR-2015-009109 to
evaluate appropriate risk management actions. Since this violation was of very low
safety significance (Green) and has been entered into the corrective action program as
Condition Report CR-2015-009109, this violation is being treated as a non-cited violation
consistent with Section 2.3.2 of the NRC Enforcement Policy.
(NCV 05000445/2016003-02; 05000446/2016003-02, Failure to Manage Risk During
Refueling Outages)
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4OA5 Other Activities
a. Inspection Scope
The inspectors evaluated the impact of financial conditions on continued safe
performance at Comanche Peak. In that the licensees parent company, Energy Future
Holdings, was under bankruptcy protection/reorganization during the inspection period,
NRC Region IV conducted special reviews of processes at Comanche Peak. The
inspectors evaluated several aspects of the licensees operations to determine whether
the financial condition of the station impacted plant safety. The factors reviewed
included: (1) impact on staffing, (2) corrective maintenance backlog, (3) changes to the
planned maintenance schedule, (4) corrective action program implementation, and
(5) reduction in outage scope, including risk-significant modifications. In particular, the
inspectors verified that licensee personnel continued to identify problems at an
appropriate threshold and enter these problems into the corrective action program for
resolution. The inspectors also verified that the licensee continued to develop and
implement corrective actions commensurate with the significance of the problems
identified.
The special review of processes at Comanche Peak included continuous reviews by the
Resident Inspectors, as well as the specialist-led baseline inspections completed during
the inspection period which are documented previously in this report.
b. Findings
No findings were identified.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On July 7, 2016, the resident inspectors presented the inspection results to Mr. S. Sewell,
Senior Director of Engineering and Regulatory Affairs, and other members of the licensee staff.
The licensee acknowledged the issues presented. The licensee confirmed that any proprietary
information reviewed by the inspectors had been returned or destroyed.
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SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
G. Struble, Manager, Operations/Simulator Training
J. Alldredge, Technician, Radiation Protection
T. Curtis, Lead Environmental Technician
S. Darter, Coordinator, Radiation Protection
S. Dixon, Consulting Licensing Analyst/Regulatory Affairs
T. Emery, Technician, Radiological Environmental Monitoring Program
T. Hope, Manager, Regulatory Affairs
B. Knapp, Acting Manager, Radiation Protection
M. Macho, Supervisor, Radiation Protection
S. Peterson, Senior Calibration Laboratory Technician, Radiation Protection
K. Powell, Supervisor, Radiation Protection
M. Syed, Engineer, Systems Engineer
M. Watkins, Lead Technician, Instruments and Controls Maintenance
J. Barnette, Consultant, Licensing Technologist
S. Bartholomew, Analyst, Emergency Preparedness
G. Bryan, Operations Specialist, Emergency Preparedness
K. Faver, Planner, Emergency Preparedness
R. Fishencord, Planner, Emergency Preparedness
J. Hull, Manager, Emergency Preparedness
R. Marquez, Planner, Emergency Preparedness
S. Sewell, Senior Director of Engineering and Regulatory Affairs
D. Volkening, Manager, Nuclear Oversight
T. McCool, Site Vice President
B. Knowles, Radiation Protection Staff
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
Failure to Adequately Monitor Feedwater System Check Valve
Performance (Section 1R12)
05000445/2016003-
Failure to Manage Risk During Refueling Outages (Section
02;05000446/2016003- NCV
4OA2)
02
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LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
Number Title Revision
STA-629 Switchyard Control and Transmission Grid Interface 7
Section 1R04: Equipment Alignment
Condition Reports
CR-2016-007245
Drawings
Number Title Revision
E1-0020 125V DC One Line Diagram CP-20
E1-0021 Common Auxiliary Control Fuel and Turbine Buildings Normal CP-22
480VC MCCs One Line Diagram
Procedures
Number Title Revision
SOP-904 Fire Protection Main Water Supply and Fire Pumps System 16
OPT-215 Class 1E Electrical Systems Operability 15
Section 1R05: Fire Protection
Condition Reports
CR-2016-002654
Drawings
Number Title Revision
E1-2020 Safeguard Building Fire Detection Plan EL 773-0, 790-6 and CP-2
800-6
Procedures
Number Title Revision
SAF-104 Inspection of Respiratory Protection Equipment (Maintenance 11
and Repair)
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Procedures
Number Title Revision
ABN-901 Fire Protection System Alarms or Malfunctions 2
FPI-103A Fire Preplan Instruction Manual, Unit 1 Safeguards Building 4
Elevation 810-6, Rad. Pen. Area & Elec. Equip. Rm
Miscellaneous Documents
Number Title Revision
-- Fire Protection Report 30
Work Orders
4789803
Section 1R06: Flood Protection Measures
Calculations
Number Title Revision
SI-CA-0000-693 Miscellaneous Building - Flooding Analysis 1
Section 1R11: Licensed Operator Requalification Program and Licensed Operator
Performance
Procedures
Number Title Revision
EOP-3.0A Steam Generator Tube Rupture 9
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Condition Reports
CR-2016-007272 CR-2016-000493 CR-2016-007720 CR-2016-007428 CR-2016-007690
Procedures
Number Title Revision
DID XPWR-SFP- SFP Cooling During Non-Refueling Outage Conditions -
01
STI-600.01 Protecting Plant Equipment and Sensitive Equipment Controls 1
MSM-GO-0213 Sway Strut Maintenance 1
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Work Orders
5320735 5210636
Section 1R15: Operability Determinations and Functionality Assessments
Calculations
Number Title Revision
1-EB-302-4 As Built HVAC Calculation - Auxiliary Feedwater Pump Room 5
Unit 1
Condition Reports
CR-2016-003089 CR-2016-007251 CR-2016-007653 CR-2016-007840
Work Orders
5010266
Section 1R18: Plant Modifications
Miscellaneous Documents
Number Title Revision
FDA-2016- Create Temp Mod FDA to Remove the Sentinel Valves on the 00
000123-01-00 Casing of the TDAFW Pump Turbines
Work Orders
5330786 5330788
Section 1R19: Post-Maintenance Testing
Condition Reports
CR-2016-000493 CR-2016-007559 TR-2016-004759 CR-2016-005744 CR-2016-005216
CR-2016-003163
Drawings
Number Title Revision
E1-0031-07 6.9 kV Switchgear Bus 1EA2 Breaker 1EA2-2 Schematic CP-13
Diagram
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Procedures
Number Title Revision
MSM-G0-0213 Sway Strut Maintenance 1
MSM-G0-4004 Baker On-line Motor Testing 5
MSM-C0-7310 Service Water Pump Maintenance 5
SOP-603A 6900 V Switchgear 16
MSE-G0-0020 Relay Calibration 5
Work Orders
5210636 5330786 4297555 5008028 4947477
4986918 5008083 5136434 4913385
Section 1R22: Surveillance Testing
Condition Reports
CR-2016-007588
Drawings
Number Title Revision
M2-0206 Flow Diagram Auxiliary Feedwater System CP-15
Procedures
Number Title Revision
OPT-206B AFW System 22
OPT-503A Cntmt Isol Valves ASME Testing 15
Work Orders
5270846
Section 1EP6: Drill Evaluation
Procedures
Number Title Revision
EPP-121 Re-Entry, Recovery and Closeout 10
EPP-116 Emergency Repair & Damage Control and Immediate Entries 9
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Procedures
Number Title Revision
EPP-109 Duties and Responsibilities of the Emergency Coordinator / 15
Recovery Manager
ABN-907 Acts of Nature 15
Section 4OA2: Problem Identification and Resolution
Condition Reports
CR-2006-001208 CR-2014-001804 CR-2016-004868 CR-2016-004936
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