IR 05000317/2022002
| ML22220A038 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/08/2022 |
| From: | Brice Bickett NRC/RGN-I/DORS |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| References | |
| IR 2022002 | |
| Download: ML22220A038 (17) | |
Text
August 8, 2022
SUBJECT:
CALVERT CLIFFS NUCLEAR POWER PLANT - INTEGRATED INSPECTION REPORT 05000317/2022002 AND 05000318/2022002
Dear David Rhoades:
On June 30, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Calvert Cliffs Nuclear Power Plant. On July 14, 2022, the NRC inspectors discussed the results of this inspection with Patrick D. Navin, and other members of your staff. The results of this inspection are documented in the enclosed report.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or the significance or severity of the violation 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 I; the Director, Office of Enforcement; and the NRC Resident Inspector at Calvert Cliffs Nuclear Power Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; and the NRC Resident Inspector at Calvert Cliffs Nuclear Power Plant. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Brice A. Bickett, Branch Chief Projects Branch 3 Division of Operating Reactor Safety
Docket Nos. 05000317 and 05000318 License Nos. DPR-53 and DPR-69
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000317 and 05000318
License Numbers:
Report Numbers:
05000317/2022002 and 05000318/2022002
Enterprise Identifier: I-2022-002-0031
Licensee:
Constellation Energy Generation, LLC
Facility:
Calvert Cliffs Nuclear Power Plant
Location:
Lusby, MD
Inspection Dates:
April 1, 2022 to June 30, 2022
Inspectors:
R. Clagg, Senior Resident Inspector
L. Dumont, Senior Reactor Inspector
S. Obadina, Resident Inspector
G. Walbert, Project Engineer
D. Werkheiser, Senior Reactor Analyst
Approved By:
Brice A. Bickett, Branch Chief
Projects Branch 3
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees
performance by conducting an integrated inspection at Calvert Cliffs Nuclear Power Plant, 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 Perform An Adequate Foreign Material Close-Out Inspection Results in a Unit 2
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
Open/Closed
[H.12] - Avoid
Complacency
A self-revealed Green finding and associated non-cited violation of Technical Specification 5.4.1, Procedures, was identified when Constellation failed to perform an adequate foreign
material close-out inspection for the 12 spent fuel cooling pump breaker cubicle as required
by site procedures. Specifically, on November 21, 2021, Constellation failed to verify that the
12 spent fuel cooling pump breaker cubicle was free of foreign material as required by MA-
AA-716-008, Foreign Material Exclusion Program, Revision 16, which resulted in a Unit 2
Additional Tracking Items
Type
Issue Number
Title
Report Section
Status
LER 2022-001-00 for Calvert
Cliffs Nuclear Power Plant,
Unit 2, Automatic Reactor Trip Due to High Reactor
Coolant System Pressure
Closed
LER 2021-004-00, Calvert
Cliffs Nuclear Power Plant,
Unit 2, Manual Reactor Trip
Due to Lowering Steam
Generator Levels
Closed
PLANT STATUS
Unit 1 operated at or near rated thermal power for the entire 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 activities described in IMC 2515,
Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of
IPs. 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
Seasonal Extreme Weather (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated readiness for seasonal extreme weather conditions prior to
the onset of seasonal hot temperatures for the Units 1 and 2 emergency diesel
generators and intake structures on May 12, 2022
71111.04 - Equipment Alignment
Partial Walkdown (IP Section 03.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1)
Unit 1, 12 and 13 saltwater pumps and associated equipment with 11 saltwater pump
out of service for maintenance, April 29, 2022
(2)
Unit 1, 1B emergency diesel generator, 0C diesel generator, and associated
equipment with 1A emergency diesel generator out of service for maintenance,
May 5, 2022
(3)
Unit 1, 11 saltwater air compressor and associated equipment with 12 saltwater air
compressor out of service for maintenance, June 7, 2022
(4)
Unit 2, 21 emergency core cooling system and associated components with 22
emergency core cooling system air cooler out of service for maintenance,
June 14, 2022
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated system configurations during a complete walkdown of the
Unit 1 high-pressure safety injection system on June 28, 2022.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection (IP Section 03.01) (5 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a
walkdown and performing a review to verify program compliance, equipment functionality,
material condition, and operational readiness of the following fire areas:
(1)
Units 1 and 2, cable chases and control room complex, fire areas 20-24,
May 12, 2022
(2)
Units 1 and 2, horizontal chases and 69' elevation electrical rooms, fire areas 35-38,
May 12, 2022
(3)
0C (station blackout) diesel generator building, fire area EDG0C, May 17, 2022
(4)
Unit 2, east and west electrical penetration rooms, fire areas 26-27, May 25, 2022
(5)
Unit 1, east and west electrical penetration rooms, fire areas 32-33, June 1, 2022
Fire Brigade Drill Performance (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated the on-site fire brigade training and performance during an
unannounced fire drill at the machine shop on June 8, 2022.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
(1)
Unit 1, 12 component cooling heat exchanger maintenance and testing,
June 22, 2022
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
(1)
The inspectors observed and evaluated licensed operator performance in the control
room during Unit 1, 'B' train safety injection valve testing on June 8, 2022.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
(1)
The inspectors observed licensee operator training involving a steam leak inside
containment, a loss of auxiliary feedwater common suction, and a loss of offsite
power in the simulator on June 15, 2022.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following
structures, systems, and components (SSCs) remain capable of performing their intended
function:
(1)
Unit 1, AR 4445774, control element assembly dropped during power ascension,
reviewed on May 2, 2022
Quality Control (IP Section 03.02) (1 Sample)
The inspectors evaluated the effectiveness of maintenance and quality control activities to
ensure the following SSC remains capable of performing its intended function:
(1)
Unit 1, WO C93780548, 11A service water heat exchanger relief valve replacement,
May 24, 2022
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management (IP Section 03.01) (7 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the
following planned and emergent work activities to ensure configuration changes and
appropriate work controls were addressed:
(1)
Unit 1, elevated risk condition due to 11 saltwater pump out of service for
replacement, May 10, 2022
(2)
Unit 1, elevated risk condition due to 1A emergency diesel generator out of service
for maintenance, May 12, 2022
(3)
Unit 2, elevated risk condition due to 22 component cooling heat exchanger out of
service for maintenance, May 24, 2022
(4)
Unit 1, elevated risk condition due to 11A service water heat exchanger out of service
for maintenance, May 27, 2022
(5)
Unit 2, risk informed completion time implementation due to channel 'A' reactor
protection system issues, May 31, 2022
(6)
Unit 1, elevated risk condition due to 12 containment air cooler out of service for
maintenance, June 13, 2022
(7)
Unit 1, elevated risk condition due to 1B emergency diesel generator out of service
for maintenance, June 29, 2022
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the
following operability determinations and functionality assessments:
(1)
Unit 1, AR 4493073, U-4000-11, 14 kilovolt transformer, upper left fan blade sheared
preventing rotation, April 19, 2022
(2)
Unit 2, AR 4493548, gap between wall and frame for emergency hatch #2 (high
energy line break barrier), May 24, 2022
(3)
Unit 1, AR 4501988, 12 containment air cooler local control hand switch failed to
operate in the high position, May 27, 2022
(4)
Unit 1, AR 4501259, dissolved gas analyzer for U-4000-13, 4 kilovolt transformer, in
alarm status, May 31, 2022
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1
Sample)
The inspectors evaluated the following temporary or permanent modifications:
(1)
Unit 2, ECP-17-000204, Ovation turbine control megawatt transducer circuit
modification, June 29, 2022
Severe Accident Management Guidelines Update (IP Section 03.03) (1 Sample)
(1)
The inspectors verified the site's severe accident management guidelines were
updated in accordance with the pressurized water reactor generic severe accident
technical guidelines and validated in accordance with Nuclear Energy Institute 14-01,
Emergency Response Procedures and Guidelines for Beyond Design Basis Events
and Severe Accidents, Revision 1, on June 7, 2022.
71111.19 - Post-Maintenance Testing
Post-Maintenance Test (IP Section 03.01) (8 Samples)
The inspectors evaluated the following post-maintenance testing activities to verify system
operability and/or functionality:
(1)
Unit 1, WO C93756168, 1CV5159A, 12A service water heat exchanger saltwater
strainer flushing valve replacement and testing, April 15, 2022
(2)
Unit 1, WO C93702979, 11 saltwater pump replacement, April 28, 2022
(3)
Unit 1, WO C93751562, perform emergency diesel generator-13 on 1A1 and 1A2
emergency diesel generator engines, May 6, 2022
(4)
Unit 1, WO C93783627, perform speed control knob torque check of 11 auxiliary
feedwater turbine governor, May 10, 2022
(5)
Unit 1, WO C93752186, replace paper drive assembly for 1RE5280, containment
atmosphere particulate radiation monitoring system, May 12, 2022
(6)
Unit 1, WO C93649176, 13 saltwater pump motor replacement, May 24, 2022
(7)
Unit 2, WO C93670498, 22 emergency core cooling system air cooler maintenance
and testing, June 15, 2022
(8)
Unit 2, WO C93783966, 21 instrument air dryer maintenance and testing,
June 30, 2022
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance testing activities to verify system operability
and/or functionality:
Surveillance Tests (other) (IP Section 03.01) (2 Samples)
(1)
Unit 2, STP-O-008A(SA)-2, "Test of 2A DG and 4kV Bus 21 Undervoltage," Revision
35, April 28, 2022
(2)
Unit 2, STP-O-5A22-2, "22 Auxiliary Feedwater Pump Quarterly Surveillance Test,"
Revision 9, May 5, 2022
Inservice Testing (IP Section 03.01) (1 Sample)
(1)
Unit 1, WO C93782745, STP-O-65C3-1, "14 CAC SRW Inlet, 1CV1592, Quarterly
Operability Test," Revision 2, May 12, 2022
71114.06 - Drill Evaluation
Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1
Sample)
(1)
The inspectors observed and evaluated the conduct of an Emergency Preparedness
drill involving the failure of a containment sump motor operator valve with an
earthquake which causes flooding from the 12 saltwater header rupture and service
water pump room, and the loss of the 14 safety-related bus with the loss of cooling
accident on hot leg resulting in a General Emergency declaration on May 10, 2022.
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
MS05: Safety System Functional Failures (IP Section 02.04) (2 Samples)
(1)
Unit 1, April 1, 2021 through March 31, 2022
(2)
Unit 2, April 1, 2021 through March 31, 2022
71152S - Semiannual Trend Problem Identification and Resolution
Semiannual Trend Review (Section 03.02) (1 Sample)
(1)
The inspectors reviewed the licensees corrective action program for potential
adverse trends that might be indicative of a more significant safety issue.
71153 - Follow Up 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)
LER 05000318/2021-004-00, Manual Reactor Trip Due to Lowering Steam Generator
Levels, (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML22018A025). The inspection conclusions associated with this LER
are documented in this report under Inspection Results.
(2)
LER 05000318/2022-001-00, Automatic Reactor Trip Due to High Reactor Coolant
System Pressure (ADAMS Accession No. ML22063A703). The inspectors determined
that the cause of the condition described in the LER was not reasonably within the
licensees ability to be foreseen and corrected and, therefore, was not reasonably
preventable. No performance deficiency nor violation of NRC requirements was
identified.
INSPECTION RESULTS
Observation: Semiannual Trend Review
71152S
The inspectors identified an adverse trend associated with degraded reactor coolant system
Tcold resistance temperature detectors for the Unit 1 and 2 reactor protection systems.
Constellation entered the following issues into their corrective action program in order to be
appropriately addressed.
- On December 8, 2021, as documented in AR 4465414, following surveillance testing
Constellation identified that the Tcold temperature element for channel 'A' of the
reactor protection system on Unit 2 was found to have failed low. Further investigation
identified that the temperature element was degraded. The temperature element was
replaced during a forced outage.
- On April 26, 2022, as documented in AR 4495809, control room operators received an
unexpected alarm and determined that the Tcold indication for reactor protection
system channel 'A' was erratic. Constellation replaced the temperature transmitter and
power supply, and restored channel 'A' of reactor protection system on Unit 2 to
service after satisfactory testing.
- On May 23, 2022, as documented in AR 4501587, Constellation identified that the
Unit 2 reactor protection system channel 'A' Tcold indication was rising unexpectedly.
Constellation performed troubleshooting and determined that the temperature element
needed to be replaced or repaired. The reactor protection system is designed such
that two temperature elements provide indication, and the highest temperature is
selected. Constellation removed the degraded temperature element from service and
has actions to repair or replace the temperature element during the next refueling
outage. The second temperature element is in service and providing appropriate
indication.
- On June 28, 2022, as documented in AR 4508003, Constellation identified that the
Tcold indication for channel 'C' of the reactor protection system on Unit 1 was
indicating higher than normal. Constellation performed troubleshooting and
determined that the power supply was degraded. Constellation replaced the power
supply and returned channel 'C' of reactor protection system to service after
satisfactory testing. In AR 4465778, Constellation documented a potential cause for
the degraded Tcold indications as loose or degraded connections at the resistance
temperature detectors quick disconnect wiring harnesses and has sent these items
offsite for evaluation.
The inspectors review noted Constellation has an action to document the final determination
of the cause for the degraded resistance temperature detectors due in August of 2022. In
addition, the inspectors noted degraded power supplies have caused some of the Tcold
indication issues. In AR 4508003, Constellation has an action to review the reactor protection
system power supply failure history and revise the preventive maintenance strategy as
appropriate. The inspectors determined these deficiencies demonstrate an adverse trend with
regards to degraded Tcold resistance temperature detectors and Constellation generated
ARs 4465414, 4465778, 44995809, 4501587, and 4508003 to address the deficiencies. The
NRC inspectors did not identify any findings or violations of more than minor significance
during this review.
Failure to Perform An Adequate Foreign Material Close-Out Inspection Results in a Unit 2
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
Open/Closed
[H.12] - Avoid
Complacency
A self-revealed Green finding and associated non-cited violation of Technical Specification 5.4.1, Procedures, was identified when Constellation failed to perform an adequate foreign
material close-out inspection for the 12 spent fuel cooling pump breaker cubicle as required
by site procedures. Specifically, on November 21, 2021, Constellation failed to verify that the
12 spent fuel cooling pump breaker cubicle was free of foreign material as required by MA-
AA-716-008, Foreign Material Exclusion Program, Revision 16, which resulted in a Unit 2
manual reactor trip.
Description: On November 21, 2021, Constellation was in the process of performing post-
maintenance testing on the 12 spent fuel cooling pump and breaker. When starting the pump,
an arc flash event occurred in the breaker. As a result, the upstream breaker (24A 480 volt
bus feeder breaker) tripped open. This caused a loss of bus 2Y10, which is a non-vital 120
volt AC instrument bus. The loss of bus 2Y10 led to the repositioning of Unit 2 condensate
pump mini-flow valve, condensate booster pump mini-flow valve, and moisture separator
reheater drain tank high level dump valve, which created a secondary transient. This resulted
in the 21 steam generator feedwater pump (SGFP) tripping on low-suction pressure. The 23
SGFP started due to the trip of the 21 SGFP. However, the loss of bus 2Y10 also caused a
loss of the control element drive system and created an inability for the station to manually
move control rods, which is needed when using the 23 SGFP following a loss of either 21 or
22 SGFPs. At 10:46 am, without the ability to manually insert control element assemblies to
lower reactor power, Unit 2 was manually tripped from 100 percent power due to lowering
steam generator levels. Constellation manually initiated auxiliary feedwater to maintain level
in the steam generators. From this event, Constellation submitted LER 05000318/2021-004-
00.
The inspectors reviewed Constellation's root cause investigation, AR 4462339, and noted
that foreign material in the 12 spent fuel cooling pump breaker cubicle caused the arc flash
event. The foreign material was identified as insulating breaker cubicle stab covers that are
used to protect plant personnel from energized parts. The root cause investigation stated that
the installation of breaker cubicle stab covers without a tracking mechanism, which were not
removed from the breaker cubicle following maintenance, was the root cause. The inspectors
reviewed the procedures associated with breaker cubicle maintenance and electrical safety.
Specifically, the inspectors reviewed FTE-53A, Westinghouse 480 Volt Load Center Cubicle
Maintenance, Revision 0002, which provides a warning that the stationary main stabs remain
energized with 480 volt AC. In addition, the inspectors reviewed SA-AA-129, Electrical
Safety, Revision 11, which provides direction for installing insulating covers for energized
parts. Specifically, it states, in part, that the licensee should install protective shields when a
breaker is removed from a compartment and energized bus bars are exposed. The inspector
noted that while the procedures provided direction for installing insulating covers, the
procedures did not provide a tracking mechanism or removal instructions for the insulating
breaker cubicle stab covers.
The inspectors reviewed WO C93680525 for the breaker cubicle maintenance and noted that
it included documentation for the foreign material close-out inspection and also identified the
work as a foreign material exclusion zone 2. The inspectors reviewed the MA-AA-716-008-F-
01, Work Package Device and Close-out Form, Revision 1, which requires that a final close-
out inspection be performed, where the individual verifies that the system/component is free
of foreign material prior to final system closure. This form was completed in the work package
indicating that the foreign material close-out inspection had been performed. The inspectors
also reviewed MA-AA-716-008, Revision 16, which provided a definition for foreign material
exclusion zone 2, which is a zone that is established in situations where a final visual internal
inspection is possible prior to system closure. The inspectors also reviewed step 14.1.6, of
MA-AA-716-008, Revision 16, which states, in part, that the licensee shall verify that the
system internals, components, and parts being installed are free of foreign materials prior to
reassembly. The inspectors determined that a foreign material close-out inspection should
have identified the insulating breaker cubicle stab covers. The inspectors concluded that the
licensee performed an inadequate foreign material close-out inspection, which resulted in a
Unit 2 manual reactor trip.
Corrective Actions: Constellation revised procedure FTE-53A to prohibit the use of stab
covers during breaker cubicle inspections. Constellation removed all stab covers from the
shop and verified that supply chain does not have any future orders to purchase additional
stab covers.
Corrective Action References: AR 4462339
Performance Assessment:
Performance Deficiency: Constellation's failure to perform an adequate foreign material
close-out inspection was a performance deficiency. WO C93680525 classified the work as
foreign material exclusion zone 2, which requires the use of licensee procedure, MA-AA-716-
008, Foreign Material Exclusion Program, Revision 16. Specifically, step 14.1.6, states, in
part, that the lead worker shall verify that the system internals, components and parts, being
installed are free of foreign materials prior to reassembly. Contrary to this, Constellation failed
to verify that the 12 spent fuel cooling pump breaker cubicle was free of foreign material,
specifically, the insulating breaker cubicle stab covers in the breaker cubicle. As a result, an
arc flash event occurred in the 12 spent fuel cooling pump breaker which subsequently
required a Unit 2 manual reactor trip.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Human Performance attribute of the Initiating Events
cornerstone and adversely 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. Specifically, the failure to perform an adequate foreign material close-out
inspection of the 12 spent fuel cooling pump breaker cubicle resulted in an arc flash event
inside the breaker and subsequently required a Unit 2 manual reactor trip.
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The
inspectors screened this finding for safety significance and determined that a detailed risk
evaluation was required. Specifically, the finding caused a reactor trip and loss of mitigating
equipment relied upon to transition the plant from a reactor trip to a stable shutdown
condition.
The performance deficiency caused the loss of the 480 volt load center 24A, motor control
centers 125/204R, and 120 volt instrument bus 2BUS2Y10. This resulted in a low-suction
pressure trip of the 21 SGFP due to the loss-of-power failed-open position of minimum flow
bypass valves and loss of manual reactor rod control (both due to loss of 2BUS2Y10),
necessitating a manual reactor trip by the operators. The subsequent loss of the 22 SGFP on
high-discharge pressure after the reactor trip was not considered related to the performance
deficiency. This initiating event was evaluated as a transient/reactor trip. However, the senior
reactor analysts also evaluated the event as a loss of main feedwater to ensure a bounding
sensitivity review was performed.
The regional senior reactor analysts used the Systems Analysis Programs for Hands-On
Evaluation, Revision 8.2.6, and the Standardized Plant Analysis Risk Model for Calvert Cliffs
Unit 2, Model Version 8.64, to conduct an initiating event analysis. This included
consideration of the effect on systems, structures, and components that were impacted by the
performance deficiency.
In accordance with IMC 0308 and Risk Assessment of Operational Events Handbook
guidance, the Significance Determination Process evaluates the risk increase/significance of
the performance deficiency that causes an initiating event by using the incremental
conditional core damage probability estimate:
Incremental conditional core damage probability = Conditional core damage probability -
Baseline core damage probability
The initiating event was set to 1.0, including any inclusive systems, structures, and
component failures, and the conditional core damage probability was calculated by the
Systems Analysis Programs for Hands-On Evaluation. For this finding, the initiating event
transient/reactor trip was set to 1.0. For the loss of main feedwater sensitivity case, event loss
of main feedwater was set to 1.0. The following additional standardized plant analysis risk
model modifications were made:
21 SGFP (MFW-TDP-FR-PUMP21) set to 0.2, as a surrogate for recovery, (21 SGFP
was recovered shortly after instrument bus 2Y10 was cross-tied to 2Y09, which
provided power to the min-flow valves restoring suction pressure)
AC Bus 24A (ACP-BAC-LP-24A) set to TRUE, (bus was restored 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after
event). This also fails motor control center 125 / 204R and associated loads
Instrument Bus 2Y10 (ACP-BAC-LP-2Y10) failure probability was increased by an
order of magnitude, as a surrogate for recovery, (2Y10 was cross-tied to 2Y09 in 7
minutes per the direction of abnormal operating procedure (AOP-07I-2))
After discussions with Idaho National Labs, motor control center 204R fault tree was
modified to account for short term recovery probability to cross-tie to motor control
center 214R via procedure. A Standardized Plant Analysis Risk-H human error
probability was calculated to be 0.14 for high-stress, moderate complexity, and
minimum time to complete. (conditional core damage probability results were
insensitive to variations to this basic event)
(Sensitivity) Alignments of key equipment were set to run TRUE or standby FALSE.
This was performed to realistically assess the important risk contributors due to the
cross-train capability and flexibilities modeled in the Calvert Cliff Unit 2 standardized
plant analysis risk model
The following influential assumptions were used:
Transient/reactor trip best represented the initiating event
21 SGFP loss was directly related to the performance deficiency; 22 SGFP was lost
after the reactor trip due to an unexpected feedwater system response after receiving
its expected reactor trip override and speed runback and is not considered related to
this performance deficiency
Standby electric feedwater pump (23 SGFP), which started once the 21 SGFP was
lost, subsequently tripped as designed, coincident with the reactor trip and hence is
considered to have little to no mitigating impact to the initiating event
Recovery of feedwater and buses 2Y10/125/204R were feasible and justified
Bus 24A was not readily recoverable based on troubleshooting and repairs. The
licensee provided a damage assessment and justification for possible early recovery,
but the senior reactor analysts did not consider this in the risk assessment
Nominal test and maintenance values; A sensitively case was performed with key
equipment alignments set as existed during the event (loss of main feedwater, service
water, and switchgear ventilation)
Senior reactor analysts preliminary risk assessments determined the finding to be risk
significant and identified dominant event trees as medium-break loss of coolant accidents and
anticipated transients without scram. The dominant sequences involved loss of reactor
coolant pump seals, failure to trip reactor coolant pumps, and high-pressure injection failure.
This was similar for both transient/reactor trip and loss of main feedwater (sensitivity) initiating
event assessments. Further review of dominant sequence cut sets indicated this was mainly
the result of motor control center 204R failure, loss of component coolant water, and
emergency switchgear cooling that results in failure of emergency switchgear loads. This
makes the conditional core damage probability results highly sensitive to alternate switchgear
ventilation (as proceduralized in OI-22H) and the failure probability of the portable generators
powering those fans. A review of calculation CA04570, that documents heat-up of the Unit 1
and 2 switchgear rooms on loss of ventilation, indicates that up to 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> is available to
establish alternate ventilation before exceeding operational temperature limits for the affected
emergency switchgear. Therefore, the senior reactor analysts concluded that this supported a
low failure probability of the emergency switchgear loads from a loss of ventilation. Based on
additional discussions with Constellation and considering the last bus was recovered within
7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the event, additional standardized slant analysis risk model adjustments were
made:
Alternate emergency switchgear ventilation power (ACP-MGN-FC-PORTGEN) was
reduced by two orders of magnitude to account for the large number of units tested
and available
Operator action to establish alternate emergency ventilation (HVC-XHE-XM-
RMCOOL) was reduced by an order of magnitude to account for the large available
time to establish the alternate line-up
- Various basic event probability updates based on senior reactor analyst's discussions
with licensee probabilistic risk assessment staff to account for plant recoveries over
the 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the event
The calculated conditional core damage probability for a transient with the above adjustments
was 8.2E-7. The baseline core damage probability was calculated to be 2.1E-7.
The incremental conditional core damage probability was 8.2E-7 - 2.1E-7, which resulted in a
risk increase 6.1E-7. The dominant sequence is a transient, with loss of reactor coolant pump
seals (from the loss of component cooling water), failure to trip reactor coolant pumps, and
failure of high-pressure injection.
A sensitivity case treating the event as a loss of main feedwater was conducted and resulted
in similar conditional core damage probability results and dominant sequences. The loss of
main feedwater case was sensitive to the 21 SGFP recovery and availability of low-pressure
feedwater. Both transient/reactor trip and loss of main feedwater risk analyses were
insensitive to crediting diverse and flexible mitigation strategies (i.e., FLEX).
The performance deficiency did cause a transient (reactor trip), however an external event
that could cause a transient cannot cause the performance deficiency to be revealed. The
senior reactor analysts validated through discussions with NRC inspectors, review of the
licensees root cause analysis, etc. that the performance deficiency will only have a risk
impact when connecting the affected breaker to the bus. Therefore, the risk can be assessed
looking only at internal events and risk contribution due to fire, flood, tornado, or seismic
events need not be considered.
A large early release frequency assessment was made using the Systems Analysis Programs
for Hands-On Evaluation and IMC 0609, Appendix H, dated March 23, 2020. The Systems
Analysis Programs for Hands-On Evaluation calculated increase in large early release
frequency was <1E-7 based on zero-factor multipliers for dominant sequences. Also, since
Calvert Cliffs Unit 2 is a combustion engineering designed pressurized water reactor a
separate consequential steam generator tube rupture screening was performed in
accordance with IMC 0609, Appendix H, Section 5, aided by NUREG-2195, Consequential
SGTR Analysis for Westinghouse and Combustion Engineering Plants, Appendix L, dated
July 2017. Since all dominant sequence large early release frequency factors were zero, this
assessment screened out having very low safety significance
Constellation also performed an analysis of the performance deficiency using the Calvert
Cliffs Unit 2 full-power internal events application-specific model based on the model of
record and provided it to the analysts for information. The senior reactor analysts reviewed
the analysis, key inputs, and assumptions and determined that their analysis was consistent
with the NRC method. The licensees calculated incremental conditional core damage
probability was 7.83E-7 with similar dominant sequences identified by the senior reactor
analysts evaluation.
In summary, the increase in risk associated with the performance deficiency using the
incremental conditional core damage probability estimate is 6.1E-7 and represents a finding
of very low safety significance (Green). Incremental conditional core damage probability or
delta core damage frequency are the risk metrics for the Significance Determination Process
to evaluate the significance of inspection findings, and their numerical values are consistent
with the risk informed scale and basis detailed in IMC 0308.
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, Constellation
failed to recognize the task of installing insulating breaker cubicle stab covers introduced
foreign material into the breaker cubicle and should have recognized and planned for the
possibility of mistakes.
Enforcement:
Violation: The Renewed Facility Operating License for Calvert Cliffs Power Plant, Unit 2,
Technical Specification 5.4.1 requires, in part, that written procedures shall be established,
implemented, and maintained as covered in Regulatory Guide 1.33, Revision 2, Appendix A,
February 1978, where Section 9a specifies that maintenance that can affect the performance
of safety-related equipment should be properly pre-planned and performed in accordance
with written procedures, documented instructions or drawings. MA-AA-716-008, Foreign
Material Exclusion Program, Revision 16, step 14.1.6 states, in part, that the lead worker
shall verify that the system internals, components, and parts, being installed are free of
foreign materials prior to reassembly.
Contrary to the above, on November 21, 2021, Constellation failed to satisfactorily implement
MA-AA-716-008 for the 12 spent fuel cooling pump and breaker maintenance. Specifically,
Constellation did not perform an adequate foreign material close-out inspection and left the
insulating breaker cubicle stab covers installed in the breaker cubicle. This resulted in an arc
flash event inside the breaker and subsequently required a Unit 2 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 14, 2022, the inspectors presented the integrated inspection results to Patrick D.
Navin, and other members of the licensee staff.