IR 05000395/2026090
| ML26033A004 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 02/05/2026 |
| From: | Mark Franke NRC/RGN-II/DORS/PB2 |
| To: | Carr E Dominion Energy |
| References | |
| EAF-RII-2025-0235 IR 2026090 | |
| Download: ML26033A004 (0) | |
Text
SUBJECT:
VIRGIL C. SUMMER - NRC INSPECTION REPORT 05000395/2026090; PRELIMINARY WHITE FINDING
Dear Eric S. Carr:
The enclosed inspection report documents a finding with an associated violation that the U.S. Nuclear Regulatory Commission (NRC) has preliminarily determined to be of low (White)
safety significance. The finding involved a failure to properly pre-plan and perform maintenance on the overspeed trip device of the safety-related turbine-driven emergency feedwater (TDEFW)
pump in accordance with written procedures, documented instructions, or drawings appropriate for the circumstances, which resulted in the inoperability of the TDEFW pump. We assessed the significance of the finding using NRCs significance determination process (SDP) and the best available information. The attachment to the inspection report contains a detailed risk evaluation with the basis of our preliminary significance determination. The finding is also an apparent violation of NRC requirements and is being considered for escalated enforcement action in accordance with the Enforcement Policy, which can be found on the NRCs Web site at http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html.
In accordance with NRC Inspection Manual Chapter 0609, we intend to complete our evaluation using the best available information and issue our final determination of safety significance within 90 days of the date of this letter. The NRCs SDP is designed to encourage an open dialogue between your staff and the NRC; however, neither the dialogue nor the written information you provide should affect the timeliness of our final determination.
Before we make a final decision on this matter, we are providing you with an opportunity to (1) attend a Regulatory Conference where you can present to the NRC your perspective on the facts and assumptions the NRC used to arrive at the finding and assess its significance, or (2) submit your position on the finding to the NRC in writing. If you request a Regulatory Conference, it should be held within 40 days of receipt of this letter, and we encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective. The focus of the Regulatory Conference is to discuss the significance of the finding and not necessarily the root cause(s) or corrective action(s) associated with the finding. If a Regulatory Conference is held, it will be open for public observation. If you decide to submit only a written response, such submittal should be sent to the NRC within 40 days of your receipt of this letter. If you decline to request a Regulatory Conference or to submit a written response, you relinquish your right to appeal the final SDP February 5, 2026 determination, in that by not doing either, you fail to meet the appeal requirements stated in the Prerequisite and Limitation sections of Attachment 2 of NRC Inspection Manual Chapter 0609.
If you choose to send a response, it should be clearly marked as a "Response to Apparent Violation; (EAF-RII-2025-0235)" and should include for the apparent violation: (1) the reason for the apparent violation or, if contested, the basis for disputing the apparent violation; (2) the corrective steps that have been taken and the results achieved; (3) the corrective steps that will be taken; and (4) the date when full compliance will be achieved. Your response should be submitted under oath or affirmation and may reference or include previously docketed correspondence, if the correspondence adequately addresses the required response.
Additionally, your response should be sent to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Center, Washington, DC 20555-0001 with a copy to Matthew Fannon, Chief, Projects Branch 2, U.S. Nuclear Regulatory Commission, Region 2, 245 Peachtree Center Avenue N.E, Suite 1200, Atlanta, GA 30303-1200 within 40 days of the date of this letter. If an adequate response is not received within the time specified or an extension of time has not been granted by the NRC, the NRC will proceed with its enforcement decision or schedule a Regulatory Conference.
Please contact Matthew Fannon at 404-997-4547 and in writing within 10 days of the issue date of this letter to notify the NRC of your intentions. If we have not heard from you within 10 days, we will continue with our significance determination and enforcement decision. The final resolution of this matter will be conveyed in separate correspondence.
Because the NRC has not made a final determination in this matter, no Notice of Violation is being issued for this inspection finding at this time. In addition, please be advised that the number and characterization of the apparent violation described in the enclosed inspection report may change as a result of further NRC review.
For administrative purposes, this inspection report provides an update to the apparent violation documented in NRC inspection report 05000395/2025003, dated December 16, 2025, and accessible at http://www.nrc.gov/reading-rm/adams.html via ADAMS Accession Number ML25345A337. 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, Mark E. Franke, Director Division of Operating Reactor Safety Docket No. 05000395 License No. NPF-12
Enclosure:
Inspection Report No.05000395/20260090
Inspection Report
Docket Number:
05000395
License Number:
Report Number:
Enterprise Identifier:
I-2026-090-0001
Licensee:
Dominion Energy South Carolina
Facility:
Virgil C. Summer Nuclear Plant
Location:
Jenkinsville, SC
Inspection Dates:
August 19, 2025, to January 28, 2026
Inspectors:
K. Dials, Resident Inspector
M. Read, Senior Resident Inspector
S. Sandal, Senior Reactor Analyst
T. Stephen, Senior Reactor Analyst
Approved By:
Mark E. Franke, Director,
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees
performance by conducting an NRC inspection at Virgil C. Summer Nuclear 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
Inadequate Maintenance Strategy Resulting in Turbine-Driven Emergency Feedwater Pump
Inoperability
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Preliminary White
Open
EAF-RII-2025-0235
None (NPP)
A self-revealed apparent violation of Technical Specification (TS) 6.8.1, Procedures and
Programs, was identified when the licensee failed to implement a preventive maintenance
procedure to ensure the reliability of the overspeed trip (OST) device for the turbine-driven
emergency feedwater (TDEFW) pump, which resulted in the inoperability and unplanned
unavailability of the pump.
Additional Tracking Items
None.
INSPECTION RESULTS
Inadequate Maintenance Strategy Resulting in Turbine-Driven Emergency Feedwater Pump
Inoperability
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Preliminary White
Open
EAF-RII-2025-0235
None (NPP)
A self-revealed apparent violation of TS 6.8.1, Procedures and Programs, was identified
when the licensee failed to implement a preventive maintenance procedure to ensure the
reliability of the OST device for the TDEFW pump, which resulted in the inoperability and
unplanned unavailability of the pump.
Description: On August 19, 2025, during routine surveillance testing of the TDEFW pump, the
turbine tripped during its initial start. During troubleshooting, the licensee identified worn
components on the OST device, resulting in inadequate engagement between the head lever
and tappet nut. The vendor specification for engagement was 0.030 - 0.060 inches, and the
as-found engagement was 0.011 inches, which caused the OST device to actuate at a lower
speed during the turbine start.
Inspectors reviewed the historical performance of the OST device. The OST device
experienced a similar failure in 2005 which was caused by worn areas on the head bracket,
tappet nut, and bent tappet stem. In addition to initial repairs to the stem and replacement of
the tappet nut, the entire OST device was rebuilt with new components in 2006. Since 2006,
the licensee performed overspeed testing during each refueling outage without a test failure
or adjustments. The apparent cause evaluation following the 2005 failure documented, in
part, that the OST device components will continue to be inspected and components
replaced (as required) during the scheduled overhaul [preventive maintenance] of the EFW
turbine (every 4 refueling outages).
Licensee procedure ER-AA-PRS-1010, Preventative Maintenance Task Basis and
Maintenance Strategy, Revision 18, Step 2.3.11 states, Task type and frequency are
dependent on a components criticality, duty cycle, and service condition. Step 5.2.16
defines the time-based task type as scheduled tasks usually performed without knowledge
of whether it is needed except as is estimated from history of wear-out failure mechanism.
Examples are to inspect and replace or clean or adjust at predetermined time intervals. The
licensees maintenance strategy database associates the OST under the TDEFW pump
template (TPP0008) and does not include a specific classification for the OST device. The
turbine maintenance strategy recurring tasks only included testing the overspeed setpoint and
did not include preventive maintenance. Inspectors determined that, contrary to the
procedure, the licensee did not establish a time-based task commensurate with the criticality
of the subcomponent to inspect, clean, adjust, or maintain the OST device since the
replacement in 2006. The failure to perform preventive maintenance allowed the head lever
and tappet nut to degrade outside of the vendor specification until it actuated prematurely
during a pump start.
Inspectors noted that the licensee had entered NRC Information Notice (IN) 2014-03,
Turbine-Driven Auxiliary Feedwater Pump Overspeed Trip Mechanism Issues, into their
corrective action program. The licensee correctly evaluated the IN and noted the details were
already incorporated into procedure MMP-300.015, Turbine Maintenance, Emergency
Feedwater Pump TPP0008, Revision 18D, Section 7.8, but the review failed to identify that
Section 7.8 was not scheduled to be performed in the preventive maintenance schedule.
Corrective Actions: The licensee repaired the OST device and successfully retested the
TDEFW pump.
Corrective Action References: Condition Report 1298757
Performance Assessment:
Performance Deficiency: The failure to properly pre-plan and perform maintenance that can
affect the performance of safety-related equipment in accordance with written procedures,
documented instructions, or drawings appropriate for the circumstances, was a performance
deficiency. Specifically, the licensees failure to pre-plan and perform maintenance on the
OST device of the safety-related TDEFW pump in accordance with written procedures was a
performance deficiency that was reasonably within their ability to foresee and correct.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Equipment Performance attribute of the Mitigating
Systems cornerstone and adversely affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, the TDEFW pump tripped during testing, rendering
the pump inoperable.
Significance: The inspectors assessed the significance of the finding using IMC 0609,
Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., The Significance Determination Process (SDP) for Findings At-Power. The
affected cornerstone was Mitigating Systems, as determined by IMC 0609, Attachment 4,
Initial Characterization of Findings. The inspectors screened the performance deficiency
using Exhibit 2 of Appendix A and determined a detailed risk evaluation was required
because the degraded condition represented a loss of the Probabilistic Risk Assessment
(PRA) function of one train of a multi-train TS system for greater than its TS allowed outage
time.
A Region II Senior Reactor Analyst performed a detailed risk evaluation. The finding was
preliminarily determined to be of low safety significance (White). The preliminary risk estimate
was obtained by performing a conditional failure analysis of the TDEFW pump using a 73-day
exposure period. The dominant sequences were associated with control building fire initiating
events accompanied by loss of power to the 7.2kV buses and transfer of the control room
when TDEFW pump function was not available. See Attachment, TURBINE-DRIVEN
EMERGENCY FEEDWATER DETAILED RISK EVALAUTION, for a summary of the
preliminary risk determination analysis.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to
this finding because the inspectors determined the finding did not reflect present licensee
performance.
Enforcement:
Violation: Technical Specification 6.8.1, Procedures and Programs, requires, in part, that
written procedures shall be established, implemented, and maintained covering the activities
referenced in the applicable procedures recommended in Appendix A of Regulatory Guide
1.33, Quality Assurance Program Requirements (Operation), Revision 2. Section 9 of
Regulatory Guide 1.33, Procedures for Performing Maintenance, requires, in part, 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 appropriate for the circumstances.
Contrary to the above, since 2006, the licensee failed to pre-plan or perform preventive
maintenance on the Unit 1 TDEFW pump using written procedures, documented instructions,
or drawings appropriate for the circumstances, which affected the performance of this safety-
related equipment. Specifically, the licensee did not establish preventive maintenance tasks
for the Unit 1 TDEFW pump's OST device since its replacement in 2006. Consequently,
certain mechanical components of the OST device degraded over time, leading to the
inoperability of the TDEFW pump.
Enforcement Action: This violation is being treated as an apparent violation pending a final
significance (enforcement) determination.
ATTACHMENT: TURBINE-DRIVEN EMERGENCY FEEDWATER DETAILED RISK
EVALAUTION
OVERALL RISK SUMMARY
The V.C. Summer turbine-driven emergency feedwater (TDEFW) pump was rendered
inoperable due to tripping on an invalid overspeed condition caused by wear on the trip
mechanisms tappet nut. A risk evaluation using a 73-day exposure period estimated an
increase in core damage frequency (delta-CDF) of 1.40E-06/year (consistent with a White
finding). Fire sequences were strongly dominant and made up approximately 94 percent of the
estimated risk increase. The dominant sequences included control building fire scenarios
accompanied by loss of power to the 7.2kV buses and transfer of the control room when
TDEFW pump function was not available.
EXPOSURE TIME
The exposure time began on June 9, 2025, when the TDEFW pump successfully completed its
final surveillance test before being returned to service. At the completion of the surveillance test
the overspeed trip mechanism was manually exercised in accordance with pump reset
procedures. The licensee manually tripped and reset the overspeed mechanism for steam valve
surveillance testing on July 18, 2025. Although the overspeed trip mechanism was exercised on
July 18th, this activity did not involve starting the pump or introducing steam flow. Exercising the
mechanism alone did not demonstrate the pumps ability to accelerate to operating speed
without a spurious trip. Therefore, this manipulation was not considered a successful functional
test of the pumps safety function under dynamic start conditions. The failed surveillance due to
an invalid overspeed occurred on August 19, 2025. Because wear of the tappet nut would not
be expected to occur (and worsen the condition) while the pump was in standby, the analyst
concluded that the failure mechanism was demand-based and not time-based. The exposure
period ended following the completion of repairs which included rotating the tappet nut and
cleaning the mechanism on August 21, 2025. Accordingly, the full value of exposure time T
(73 days including repair time) was used for the analysis consistent with the guidance discussed
in Section 2.3 of Volume 1 of the Risk Assessment Standardization Project (RASP) handbook
(ADAMS ML17348A149).
RISK ANALYSIS/CONSIDERATIONS
1.
The TDEFW failure was modeled as fail-to-start (FTS) due to a spurious overspeed trip of
the pump during its start sequence.
2.
The evaluation considered TDEFW pump recovery credit for a normal attempt to relatch the
overspeed trip mechanism immediately following the initial FTS.
3.
An additional attempt to recover the pump was also considered for station blackout (SBO)
sequences using a beyond design basis procedure (BDMG-4.0, Manual Operation of
Turbine-Driven Emergency Feedwater Pump (Includes Abnormal Operation), Revision 2) to
disconnect the governor valve mechanical linkage and take manual local control of the trip
and throttle valve (removing the impact due to the worn tappet nut).
4.
Flexible Coping (FLEX) mitigating strategies and equipment were credited in the analysis
using 24-hour PRA mission time. FLEX equipment reliability was modeled using information
contained in PWROG-18042-NP, Revision 1, FLEX Equipment Data Collection and
Analysis (ADAMS ML22123A259).
Systems Analysis Program for Hands-On Integrated Reliability Evaluations (SAPHIRE) software
Version 8.2.12 and V.C. Summer Standardized Plant Analysis Risk (SPAR) model Version 8.82
were used for the evaluation.
1.
The SPAR model was modified to account for the capability of powering the A or B train
7.2 kV engineered safety feature (ESF) bus from the 13.8 kV normally energized section of
the Parr hydro unit switchyard. This power supply is an underground feed via transformer
XTF-5052 that has the capability of providing power to either ESF bus and does not require
manning or start-up of the Parr hydro unit. No credit was provided in the analysis for start-up
and operation of the Parr hydro unit as an alternate source of electrical power. In addition,
the SPAR model was also modified to account for the ability of transformer XTF-31 and
XTF-4 to supply offsite power to the ESF buses. Two Human Error Probability (HEP) basic
events were created to model operator actions to recover offsite power supplies. The first
HEP (ACP-XHE-XM-PARRHYDRO) estimated the failure to manually align the 13.8 kV
offsite source from the normally energized portion of the Parr hydro switchyard to either the
A or B train 7.2 kV ESF bus. The second HEP (ACP-XHE-XA-ALT1DA(B)) estimated the
failure to manually align either XTF-31 to the B and A train ESF buses or XTF-4 to the A
and B train ESF buses.
2.
Following consultation with Idaho National Labs (INL), the SPAR TDEFW fault tree logic
(EFW-TDP-XPP8) was modified to include basic events representing relatch and restart of
the TDEFW pump and implementation of operator procedures to disconnect the governor
valve mechanical linkage and take manual local control of the trip and throttle valve.
3.
The TDEFW pump relatch failure likelihood (EFW-TDP-XPP-8-FRESTART) was estimated
using a constrained non-informed Bayesian update based on one failed attempt to start the
pump in two demands. The current data for the update was selected based on the operators
successfully re-starting the TDEFW pump following its trip on August 19, 2025. The
Bayesian update yielded an estimate failure likelihood of 2.14E-01 which was used in the
evaluation.
4.
The TDEFW pump failure likelihood following implementation of procedures to disconnect
the governor valve mechanical linkage (EFW-TDP-XPP-8-FRESTARTX2) was estimated
using an informed prior Bayesian update based on one failed attempt to start the pump in
one demand. The current data for this update was based on the one demand and one
failure associated with the performance deficiency with any other demands and failures
removed due to the use of the licensee procedure to remove the overspeed trip from use.
The Bayesian update yielded an estimate failure likelihood of 8.25E-03 which was used in
the evaluation.
5.
The failure likelihood for operator actions (EFW-TDP-XHE-RELATCH) to relatch the TDEFW
pump was estimated using IDHEAS-ECA Version 1.3 and guidance discussed in NRC
Research Information Letter (RIL) 2024-17, Integrated Human Event Analysis for Event and
Condition Assessment (IDHEAS-ECA) Evaluations of SPAR Model Human Failure Events,
dated January 2025 (ML24352A019). The analyst compared the failure likelihood with
SPAR-H estimation techniques which yielded a value of 4.4E-03. The IDHEAS-ECA
evaluation was considered best available information due to its increased capability of
assessing cognition influencing factors and timing uncertainty distributions. IDHEAS-ECA
yielded a failure likelihood of 3.86E-03 which was used in the analysis.
6.
The failure likelihood for operator actions (EFW-TDP-XHE-ABNORMAL) to implement
procedures to disconnect the governor valve mechanical linkage, restart the TDEFW pump
and control steam generator (SG) water level was also estimated using IDHEAS-ECA
Version 1.3 and the guidance described in RIL 2024-17. Results were compared with
SPAR-H, which yielded a value of 1.3E-01. The IDHEAS-ECA evaluation was considered
best available information due to its increased capability of assessing cognition influencing
factors and timing uncertainty distributions. IDHEAS-ECA yielded a failure likelihood of
3.95E-02 which was used in the analysis.
7.
Credit for implementation of procedures to recover the TDEFW pump under SBO plant
conditions were implemented in the SPAR model using event tree post-processing rules.
8.
The SPAR model fragilities for the 7.2kV ESF boards were updated with more recent plant-
specific values. Additionally, BIN-4 and BIN-5 seismic events removed from consideration
due to the baseline conditional core damage probability for earthquakes of that magnitude
approaching 1.0.
9.
The following event sequences were used in the evaluation:
Fire
Internal Events
Seismic
Tornado
LARGE EARLY RELEASE FREQUENCY IMPACT
The finding was evaluated in accordance with IMC 0609, Appendix H, Containment Integrity
Significance Determination Process, as a Type A finding. Although the estimated change in
core damage frequency (delta-CDF) was greater than 1E-07/year, the dominant accident
sequences did not involve SG tube rupture or interfacing system loss of coolant accidents.
Therefore, the issue associated with the TDEFW pump would not be expected to be a
significant contributor to an increase in large early release frequency (delta-LERF) risk. Delta-
CDF was determined to be the risk metric of interest for this evaluation.
CALCULATIONS
SAPHIRE condition assessments were performed using the SAPHIRE Events and Conditions
Assessment (ECA) module and setting the TDEFW FTS basic event to True.
Best Estimate:
Model
Event
Sequence
delta-CDP
(1 Year)
delta-CDP
(73 Days)
Fire
6.55E-06*
1.31E-06*
Internal Events
3.36E-07
6.72E-08
Seismic
9.21E-08
1.84E-08
High Wind
2.14E-08
4.28E-09
Internal Flood
9.09E-11
1.82E-11
Tornado
6.33E-12
1.27E-12
Total
1.40E-06
- Fire risk was estimated from V.C. Summer Fire PRA sensitivity data by adjusting estimated
Incremental Core Damage Probability (ICDP) using IDHEAS ECA HEP values over the 73-day
exposure period (reference sensitivity 2).
Internal Events dominant cutsets included reactor trip accompanied by failure of both motor-
driven emergency feedwater (EFW) pumps and failure to initiate feed and bleed.
Fire dominant contributors using the V.C. Summer Fire PRA model included scenarios with loss
of power to the Class 1E 7.2 kV buses accompanied by loss of TDEFW pump function and
relocation of control room function.
Sensitivity 1 - No TDEFW Pump Recovery Credit
To evaluate the sensitivity of analysis results with respect to credit for implementation of SBO
procedures to disconnect the governor valve mechanical linkage, restart and control the
TDEFW pump, the SPAR model was adjusted to remove recovery credit.
Model
Event
Sequence
No Recovery Credit
delta-CDP (73 Days)
Fire
9.20E-06*
Internal Events
1.79E-07
Seismic
2.52E-07
High Wind
1.97E-08
Internal Flood
5.32E-11
Tornado
7.04E-11
Total
9.65E-06
- Fire risk was estimated from V.C. Summer Fire PRA sensitivity data by adjusting estimated
ICDP assuming the HEP values were failed over the 73-day exposure period (reference
sensitivity 2).
Sensitivity 2 - Recovery Human Error Probability
The sensitivity of analysis results with respect to HEPs for implementation of procedures to
disconnect the governor valve mechanical linkage, restart the TDEFW pump and throttle flow to
prevent SG overfill were evaluated. IDHEAS ECA HEP results were compared with values
derived from the Electric Power Research Institute (EPRI) human reliability analysis (HRA)
calculator to determine the overall impact on analysis results for Fire sequences (the most
significant contributor to risk estimates).
The V.C. Summer PRA utilized two separate basic events for the alternative start of the TDEFW
pump per the beyond design basis procedure (BDMG-4) and control of SG level following the
pump start. The detailed risk evaluation utilized one combined HEP event with two critical tasks
for both the alternative start of the TDEFW pump (critical task 1) and control of the SG level
(critical task 2). In the ECA analysis of external events, the combined HEP event was used, and
credit was given for SBO sequences.
The two EPRI HRA calculator HEPs and distribution values were used for the sensitivity. The
EPRI HEP values were summed to be consistent with the combination of those two critical tasks
into a single HEP using the IDHEAS application. The EPRI HRA calculator distribution
information was then utilized to estimate the change in Fire risk as a function of HEP values
between the 5th and 95th percentile over a 73-day exposure period.
The analyst noted that the EPRI HRA calculator yielded lower end Fire risk estimates around
1.03E-06 while IDHEAS ECA results were higher at 1.31E-06. The sensitivity included
consideration of SPAR model sequences in addition to Fire.
Model
Event
Sequence
delta-CDP (73 Days)
delta-CDP (73 Days)
Fire
1.03E-06
1.31E-06*
Internal Events
6.26E-08
6.72E-08
Seismic
1.25E-08
1.84E-08
High Wind
3.56E-09
4.28E-09
Internal Flood
1.82E-11
1.82E-11
Tornado
2.04E-13
1.27E-12
Total
1.11E-06
1.40E-06
- Fire risk was estimated from V.C. Summer Fire PRA sensitivity data by adjusting estimated
ICDP using IDHEAS ECA HEP values over the 73-day exposure period.
Sensitivity 3 - SPAR Model Estimated Fire Risk
SPAR model ECA condition assessments for fire sequences were performed with a 73-day
exposure time both with and without credit for alternate sources of AC power to the 7.2kV buses
from the 13.8kV, 115kV and 230kV transformers. Although these sources of power could be
available to mitigate failure of the TDEFW pump, there was uncertainty regarding whether the
SPAR model would correctly account for necessary cables that could be damaged during
specific fire scenarios. This sensitivity established an upper (no credit for alternate AC sources)
and lower (full credit for alternate AC sources) bound for SPAR model fire risk. The SPAR
model produced results at 5.80E-07/year when this offsite power credit was applied and 1.78E-
05/year when it was not applied. Therefore, SPAR model fire risk was highly sensitive to the
application of this credit. The analyst noted that the V.C. Summer Fire PRA model yielded an
estimated delta-CDP value of 1.08E-06 for a 73-day exposure period which was within the
SPAR models upper and lower bounds of fire risk. Additionally, the analyst noted that the SPAR
model fire information had been incorporated into the model more than 10 years ago. Given the
SPAR model uncertainties associated with cable fire damage and the age of the information
used to develop the SPAR fire model, the analyst concluded that the V.C. Summer peer-
reviewed NFPA-0805 Fire PRA model could be considered best available information for the
estimation of fire risk.
Model
Event
Sequence
SPAR lower bound
FIRE Sequences
delta-CDP
SPAR upper bound
FIRE Sequences
delta-CDP
Fire
5.80E-07
1.78E-05
Internal Events
6.72E-08
6.72E-08
Seismic
1.84E-08
1.84E-08
High Wind
4.28E-09
4.28E-09
Internal Flood
1.82E-11
1.82E-11
Tornado
1.27E-12
1.27E-12
TOTAL
6.70E-07
1.79E-05
Sensitivity 4 - 34-Day Exposure Time Based on July 18, 2025, OST Manipulation
The last successful operational test of the pump prior to its failure occurred on June 9, 2025. On
July 18, 2025, the licensee performed a quarterly steam valve test. The test required the
TDEFW pump to be in a trip condition to prevent it from starting during the steam valve testing.
The licensee manually tripped the overspeed mechanism prior to the steam valve test and reset
it afterwards. The overspeed trip function was not tested and the TDEFW pump was not started
on July 18, 2025.
The local manual manipulation of the OST mechanism on July 18, 2025, introduced potential
uncertainty regarding the condition exposure period. However, this manipulation did not
constitute a functional test of the overspeed trip mechanism under operating conditions because
the pump was never started. The failure mode, spurious trip during dynamic start, could not be
evaluated by simply tripping and resetting the mechanism without subjecting it to actual turbine
acceleration and load. When the OST mechanism is reset and returned to its standby condition,
stacking of dimensional tolerances in the OST mechanism can result in small variabilities in how
the mechanism is physically latched from operation to operation (i.e., the precise physical
contact points between the tappet nut and the head lever when latched).
The sensitivity was performed to demonstrate the impact on the overall analysis if the exposure
period was reduced from 73 days to 34 days in length (the time between July 18, 2025, and
return to service of the TDEFW pump on August 21, 2025).
Model
Event
Sequence
delta-CDP
(73 Days)
delta-CDP
(34 Days)
Fire
1.31E-06*
6.10E-07*
Internal Events
6.72E-08
3.13E-08
Seismic
1.84E-08
8.58E-09
High Wind
4.28E-09
1.99E-09
Internal Flood
1.82E-11
8.47E-12
Tornado
1.27E-12
5.90E-13
Total
1.40E-06
6.52E-07
- Fire risk was estimated from V.C. Summer Fire PRA sensitivity data by adjusting estimated
ICDP using IDHEAS ECA HEP values over the 73-day exposure period.
Although the sensitivity demonstrated that exposure period was influential in the overall analysis
results, the analyst noted the following reasons that the uncertainties associated with the
manipulation of the OST mechanism on July 18, 2025, did not warrant use of an exposure
period less than 73 days:
1.
Additional wear of the tappet nut and head lever engagement surfaces did not occur (and
worsen the degradation) while the pump was in a standby state. Therefore, the degradation
mechanism was determined to be demand-based and not time-based. Section 2.3 of
Volume 1 of the RASP handbook stated that the T exposure time determination approach
was appropriate for standby or periodically operated components that failed due to a
degradation mechanism that was not gradually affecting the component during the standby
time.
2.
Although the manual manipulation of the OST mechanism on July 18, 2025, could have
introduced the potential for additional tappet nut wear, the analyst concluded that the wear
due to OST manipulation on July 18 would not be as significant as the prolonged wear over
many demands that occurred prior to the last successful operation of the pump on June 9 as
a result of the performance deficiency. The wear pattern observed on the tappet nut
reflected many operations on the OST mechanism to arrive at that state of spurious
operation (resulting in pump failure) that occurred during a dynamic start of the TDEFW
pump on August 19.
3.
Although the reset of the OST mechanism on July 18, 2025, introduced the potential for
variability in the engagement surfaces of the tappet nut and the head lever due to
dimensional tolerances, the analyst concluded that there was insufficient evidence to
conclude that variability in manipulation-to-manipulation latching tolerances was the cause
of the failure on August 19, 2025. The analyst could not rule out the possibility that the as-
left relatch engagement condition was more favorable on July 18 than June 6, and simply
not favorable enough to overcome the accumulation of tappet nut wear to withstand the
forces associated with the dynamic start of the pump that occurred during its failure on
August 19. The analyst concluded that the wear of the engagement surfaces that existed
prior to July 18 was the most likely cause of the failure on August 19 and that the exposure
period should begin at the last successful dynamic pump start which would be more
representative of the conditions that occurred when the pump failed and consistent with the
exposure time determination method discussed in the RASP handbook.
4.
The evaluation methods and modeling approach utilized in the detailed risk evaluation were
aligned with how the pump responded to actual dynamic start demands on June 9, 2025,
and August 21, 2025, including providing probabilistic credit in the analysis for the ability to
successfully reset the OST mechanism and restart the TDEFW pump without additional
corrective action.
OVERALL RESULTS
SAPHIRE ECA condition assessments were performed by setting the TDEFW pump FTS basic
event to True over an exposure period of 73-days. The V.C. Summer Fire PRA model was
considered best available information for estimating Fire risk contribution and was strongly
dominant in the results. The dominant sequences included control building fire scenarios
accompanied by loss of power to the 7.2kV buses and transfer of the control room when
TDEFW pump function was not available.
The results are summarized below:
EVENT
SEQUENCE
Best
Estimate
delta-CDP
No Recovery
Credit
delta-CDP
Values
delta-CDP
SPAR Fire
Lower/Upper Bound
delta-CDP
34 Day
Exposure
delta-CDP
Fire
1.31E-06*
9.20E-06*
1.03E-06
5.80E-07 to 1.79E-05
6.10E-07*
Internal Events
6.72E-08
1.79E-07
6.26E-08
6.72E-08
3.13E-08
Seismic
1.84E-08
2.52E-07
1.25E-08
1.84E-08
8.58E-09
High Wind
4.28E-09
1.97E-08
3.56E-09
4.28E-09
1.99E-09
Internal Flood
1.82E-11
5.32E-11
1.82E-11
1.82E-11
8.47E-12
Tornado
1.27E-12
7.04E-11
2.04E-13
1.27E-12
5.90E-13
TOTAL
1.40E-06
9.65E-06
1.11E-06
6.70E-07 to 1.79E-05
6.52E-07
- Fire risk was estimated from V.C. Summer Fire PRA sensitivity data by adjusting estimated
ICDP using IDHEAS ECA HEP values over the 73-day exposure period.
No TDEFW Pump Recovery Credit - showed a strongly influential impact to estimated risk
when the BDMG procedure for TDEFW pump recovery was not credited.
Recovery Human Error Probability - demonstrated that there was not an influential change in
estimated risk based on the use of the EPRI HRA method versus the use of IDEAS-ECA
application for estimating HEPs.
SPAR Model Estimated Fire Risk - the SPAR model was very sensitive to consideration of the
sources of power to the 7.2kV switchgear for Fire scenarios. The V.C. Summer Fire PRA model
yielded results were within the lower and upper bound established by the sensitivity and was
considered best available information for estimation of fire risk.
34-Day Exposure Based on July 18, 2025, OST Manipulation - the sensitivity demonstrated that
analysis conclusions were sensitive to the exposure time used. The analyst concluded that the
73-day exposure period was more appropriate to the circumstances and available information,
because the last successful dynamic start of the pump occurred on June 9, 2025. The July 18
activity did not verify the pumps ability to avoid premature overspeed trip and therefore cannot
be used as the inception date.
LICENSEE SIGNIFICANCE EVALUATION
The licensee provided two evaluations that assessed the significance of the TDEFW pump
failure.
The initial review focused on whether a performance deficiency existed and how long the plant
was exposed to the condition. The analysis included (in part) the following points:
A performance deficiency occurred due to inadequate preventive maintenance to ensure
pump reliability.
- Because the wear developed gradually and the exact start of failure could not be
determined, an exposure period of 38 days would be appropriate. Specifically, use of the full
exposure period would not align with the observed failure mechanism, which was gradual
wear over time.
The second review estimated the risk impact of the failure using the V.C. Summer Fire PRA
model. Key points (in part) included:
Use of the same 38-day exposure period included in the previous analysis, with an alternate
approach that combined T/2 and T estimation methods and yielded an exposure period of
approximately 54 days.
Opportunities remained in the Fire PRA to reduce conservatism in the future by applying
updated fire modeling guidance and crediting additional detection and suppression
measures.
Human error probabilities influenced results but did not change the overall conclusion.
The analyst noted several considerations:
From Section 2.3 of the RASP handbook:
This exposure time determination approach (Exposure Time = t + Repair Time) is
appropriate for standby or periodically operated components that fail due to a degradation
mechanism that is not gradually affecting the component during the standby time period.
The t period should be considered for the following cases:
- The failure was determined to have occurred when the component was last functionally
operated in a test or unplanned demand.
- The failure mechanism was unknown, and the root cause assessment was not sufficient
or not complete to identify the cause of the failure.
The observed failure mechanism (tappet nut wear) would be more accurately described as
gradual wear of the OST trip mechanism tappet nut resulting from pump operation and not
gradual wear of the surfaces over time (i.e., the degradation did not gradually progress
during the standby period).
- The use of the T/2 method of estimating condition exposure period to address uncertainties
in failure point inception may be appropriate for consideration for conditions that worsen
while the pump is on standby (i.e., like failures caused by accumulation of corrosion over
time), but would not be appropriate for conditions that only progress as a result of
component operation (i.e., wear of the OST mechanism tappet nut).
There was insufficient information to conclude that the July manipulation of the OST device
for steam valve testing (without operation of the TDEFW pump) was the failure inception
point and should be considered the start of the condition exposure time. The manual
manipulation of the OST device did not test its ability to withstand a dynamic start of the
TDEFW pump without a spurious actuation below the overspeed trip setpoint and, therefore,
could not be considered the last functional test of the OST device prior to the August failure.
- There was insufficient information to draw conclusions regarding the impact of engagement
in the contact surfaces of the OST mechanism that resulted from the steam valve testing
(i.e., the OST mechanism engagement could have been either more favorable or less
favorable following reset of the OST at the end of the test and still not adequate to withstand
the rapid introduction of steam into the pump that occurred at the time of failure).
EXTERNAL EVENTS CONSIDERATIONS
Initial internal event risk estimates were greater than 1E-07, therefore all other external event
sequences were evaluated in the risk assessment. Fire event sequences were determined to be
significant contributors to the overall estimated risk.
CONCLUSIONS/RECOMMENDATIONS
The estimated risk increase (delta-CDF) over the nominal case for the inoperability of the
TDEFW pump was 1.40E-06/year, which should be preliminarily considered a finding of low
(White) significance.