IR 05000445/2025301
| ML25259A041 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/16/2025 |
| From: | Heather Gepford NRC/RGN-IV/DORS/OB |
| To: | Peters K Vistra Operations Company |
| References | |
| 50-445/25-301, 50-446/25-301 50-445/OL-25, 50-446/OL-25 | |
| Download: ML25259A041 (1) | |
Text
September 16, 2025
SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 - NRC EXAMINATION REPORT 05000445/2025301 AND 05000446/2025301
Dear Mr. Peters:
On September 10, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an initial operator license examination at Comanche Peak Nuclear Power Plant, Units 1 and 2. The enclosed report documents the examination results and licensing decisions. A technical debrief was conducted on July 29, 2025, with Bobby Simpson, Nuclear Training Director, and other members of your staff. A telephonic exit meeting was conducted on September 10, 2025, with Mr. H. Schill, Nuclear Plant General Manager, who was provided the NRC licensing decisions.
The examination included the evaluation of ten applicants for reactor operator licenses, nine applicants for instant senior reactor operator licenses, and three applicants for upgrade senior reactor operator licenses. The license examiners determined that twenty of the twenty-two applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued. There were two post-examination comments submitted by your staff. Enclosure 1 contains details of this report and Enclosure 2 summarizes post-examination comment resolution.
No findings were identified during this examination. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Heather J. Gepford, Ph.D., Chief Operations Branch Division of Operating Reactor Safety Docket Nos. 50-445; 50-446 License Nos. NPF-87; NPF-89 Enclosures:
1.
Examination Report 05000445/2025301 and 05000446/2025301 2.
NRC Post-Examination Comment Resolution Electronic distribution via LISTSERV Signed by Gepford, Heather J.
on 09/16/25
ML25259A041 SUNSI Review:
ADAMS:
Non-Publicly Available Non-Sensitive Keyword:
By: JCK Yes No Publicly Available Sensitive NRR-079 OFFICE SOE:DORS:OB SOE:DORS:OB OE:DORS:OB SOE:DORS:OB NAME JKirkland KClayton RWilliams CHarrington SIGNATURE
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DATE 09/16/25 09/16/25 09/16/25 09/16/25 OFFICE SOE:DORS:OB SOE:DORS:OB C:DORS:OB NAME DYou TFarina HGepford SIGNATURE
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DATE 09/16/25 09/16/25 09/16/25
Enclosure U.S. NUCLEAR REGULATORY COMMISSION Examination Report Docket Numbers:
05000445, 05000446 License Numbers:
NPF-87, NPF-89 Report Numbers:
05000445/2025301 and 05000446/2025301 Enterprise Identifier:
L-2025-OLL-0045 Licensee:
Vistra Operations Company, LLC Facility:
Comanche Peak Nuclear Power Plant, Units 1 and 2 Location:
Glen Rose, Texas Inspection Dates:
July 21, 2025, to September 10, 2025 Inspectors:
D. You, Senior Operations Engineer (Chief Examiner)
J. Kirkland, Senior Operations Engineer K. Clayton, Senior Operations Engineer T. Farina, Senior Operations Engineer C. Harrington, Senior Operations Engineer R. Williams, Operations Engineer Approved By:
Heather J. Gepford, Ph.D., Chief Operations Branch Division of Operating Reactor Safety
SUMMARY Examination Report 05000445/2025301 and 05000446/2025301; July 21, 2025 -
September 10, 2025; Comanche Peak Nuclear Power Plant, Units 1 and 2; Initial Operator Licensing Examination Report The NRC examiners evaluated the competency of ten applicants for reactor operator licenses, nine applicants for instant senior reactor operator licenses, and three applicants for upgrade senior reactor operator licenses at Comanche Peak Nuclear Power Plant, Units 1 and 2.
The licensee developed the examinations using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 12. The written examination was administered by the licensee on August 1, 2025. The NRC examiners administered the operating tests on July 21 - 29, 2025.
The NRC examiners determined that twenty of the twenty-two applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
A.
NRC-Identified and Self-Revealing Findings None.
B.
Licensee-Identified Violations None.
REPORT DETAILS OTHER ACTIVITIES - INITIAL LICENSE EXAM
.1 License Applications a.
Scope The NRC examiners reviewed all license applications submitted to ensure each applicant satisfied relevant license eligibility requirements. The NRC examiners also audited four of the license applications in detail to confirm that they accurately reflected the subject applicants qualifications. This audit focused on the applicants experience and on-the-job training, including control manipulations that provided significant reactivity changes.
b.
Findings No findings were identified.
.2 Examination Development a.
Scope The NRC examiners reviewed integrated examination outlines and draft examinations submitted by the licensee against the requirements of NUREG-1021. The NRC examiners conducted an onsite validation of the operating tests.
b.
Findings The NRC examiners provided outline, draft examination, and post-validation comments to the licensee. The licensee satisfactorily completed comment resolution prior to examination administration.
The NRC examiners determined the written examinations and operating tests initially submitted by the licensee were within the range of acceptability expected for a proposed examination.
.3 Operator Knowledge and Performance a.
Scope On August 1, 2025, the licensee proctored the administration of the written examinations to all applicants. The licensee staff graded the written examinations, analyzed the results, and presented their analysis and post-examination comments to the NRC on August 14, 2025.
The NRC examination team administered the various portions of the operating tests to all applicants from July 21 - 29, 2025.
b.
Findings No findings were identified.
Eighteen applicants passed the written examination and all parts of the operating test.
Two applicants were excused from the operating test and passed their written examinations. One reactor operator applicant did not pass the written examination, and one instant senior reactor operator applicant did not pass the written examination. The final examinations and post-examination analysis and comments may be accessed in the ADAMS system under the accession numbers noted in the attachment.
Post-examination analysis revealed two generic weaknesses associated with applicant performance on the written examination. The applicants displayed weaknesses associated with interaction between instrument air system malfunction affecting residual heat removal system and main generator MVAR changes affecting reactor power. One generic weakness associated with applicant performance on the operating test was identified. The applicants had issues identifying the correct malfunctioning component during a main feedwater speed control failure. These weaknesses were captured in the licensees corrective action program as tracking reports TR-2025-004546 and TR-2025-004858. Copies of all individual examination reports were sent to the facility training manager for evaluation and determination of appropriate remedial training.
.4 Simulation Facility Performance a.
Scope The NRC examiners observed simulator performance with regard to plant fidelity during examination validation and administration. No issues were identified.
b.
Findings No findings were identified.
.5 Examination Security a.
Scope The NRC examiners reviewed examination security for examination development during both the onsite preparation week and examination administration week for compliance with 10 CFR 55.49 and NUREG-1021. Plans for simulator security and applicant control were reviewed and discussed with licensee personnel.
Findings No findings were identified.
EXIT MEETINGS AND DEBRIEFS Exit Meeting Summary A technical debrief was conducted between the chief examiner and Bobby Simpson, Nuclear Training Director, and other members of the staff on July 29, 2025. A telephonic exit was conducted on September 10, 2025, between D. You, Chief Examiner, and Mr. H. Schill, Nuclear Plant General Manager.
The licensee did not identify any information or materials used during the examination as proprietary.
ADAMS DOCUMENTS REFERENCED Accession No. ML25254A084 - FINAL WRITTEN EXAMS Accession No. ML25254A097 - FINAL OPERATING TEST Accession No. ML25254A128 - POST-EXAMINATION ANALYSIS-COMMENTS
Enclosure 2 NRC Resolution to the Comanche Peak Nuclear Power Plant Post-Examination Comments A complete text of the licensee's post-examination analysis and comments can be found in ADAMS under Accession Number ML25254A128.
RO Question # 36
LICENSEE COMMENT: The facility contends that the question contains a psychometric flaw (subset issue) that results in two correct answers. On a normal stop of a circulating water pump, the discharge valve FIRST closes 55 degrees from the fully open position, the pump then stops, and the discharge valve then fully closes. Answer A states, discharge valve FIRST closes 55 degrees from the fully open position, and then the pump stops which is true. Answer C states, pump FIRST stops, and then the discharge valve FULLY closes which is also true.
NRC RESOLUTION: The NRC disagrees with the licensees recommendation. Based on a review of the question stem and the choices given, we do not believe there is a subset issue.
The applicant is presented with a question that is asking about the correct sequence of events upon the normal stop of a Circulating Water Pump. Fundamentally, the applicant is presented with three distinct events: discharge valve rotating 55 degrees from the open position, pump stopping, and discharge valve fully closing (the correct sequence of events). Additionally, they are tasked with determining which of these presented events happen FIRST (the word is bolded and capitalized). While the pump does stop before the discharge valve fully closes, this is not the first event to occur. The first event that occurs is the discharge valve rotating closed 55 degrees from the open position. Consequently, the only correct answer is choice A since it lists the discharge valve rotating closed 55 degrees from the fully open position as the first event.
RO QUESTION #46
LICENSEE COMMENT: The facility contends that the question has two correct answers (both C and D). The second part is a basis question where the original key stated, extend the time before reaching bleed and feed criteria is the correct answer. However, this question stem wording does not specifically bound the second part to the bases statement in FRH-0.1B which states: RCP operation results in heat addition to the RCS water. By tripping the RCPs, the effectiveness of the remaining water inventory in the SGs is extended, which extends the time at which the operator action to initiate bleed and feed must occur.
The facility also states that the Westinghouse Owners Group (WOG) procedure FR-H.1 says:
Delaying the loss of secondary heat sink is not the only reason for tripping RCPs. RCPs running can also reduce the effectiveness of bleed and feed. RCP heat input to the RCS will result in increased steam generation hindering the depressurization of the RCS during bleed and feed.
The higher pressure produced by the RCP operation will reduce SI flow and increase inventory lost through the PORVs. Therefore, RCPs should be tripped if AFW flow cannot be established immediately after entering this guideline. The facility then concludes that all RCPs are stopped during performance of FRH-0.1B to minimize heat input into the RCS which will prevent over pressurizing. This would make D a correct answer as well.
NRC RESOLUTION: The NRC disagrees that there are two correct answers. Reactor Coolant Pump (RCP) status does not prevent overpressurizing the RCS.
According to the WOG Emergency Response Guideline, FR-H.1, Response to Loss of Secondary Heat Sink, there is no mention of overpressure protection needed by securing the RCPs. Per this document, the step for securing all RCPs (page 45) only mentions that stopping all RCPs will result in an interim plant transient on RCS pressure and temperature as natural circulation flow conditions are established in the RCS. A pressure transient does not equate to an overpressure condition.
Westinghouse Owners Group for FR-H.1, Section 2.5 (page 31), describes an analysis of three cases during a loss of heat sink where RCPs were either: (1) not tripped throughout the transient, (2) tripped at the same time the reactor tripped, and (3) tripped 5 minutes after the reactor trip. In all three cases the analysis concluded that the pressurizer PORVs would open.
Securing the RCPs only delays the time it takes for PORV opening setpoint to be reached. This means that tripping the reactor coolant pump does not prevent overpressurization, it merely delays the time it takes for RCS pressure to rise.