ML25329A329

Beaver Valley Power Station, Unit 2 - Initial Operator Licensing Examination Report 05000412/2025301

ML25329A329
Person / Time
Site:	Beaver Valley
Issue date:	11/25/2025
From:	Sarah Elkhiamy NRC/RGN-I/DORS/OB
To:	Blair B Vistra Operations Company
References
50-412/25-301
Download: ML25329A329 (0)

Text

November 25, 2025 Barry Blair Site Vice President Vistra Operations Company, LLC Beaver Valley Power Station P. O. Box 4 - Route 168 Shippingport, PA 15077-0004

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT 2 - INITIAL OPERATOR LICENSING EXAMINATION REPORT 05000412/2025301

Dear Barry Blair:

On October 14, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an examination at the Beaver Valley, Unit 2 facility. The enclosed examination report documents the examination results, which were discussed on November 18, 2025, with Andrew Scott, Director Nuclear Operations, and other members of your staff.

The examination included the evaluation of four applicants for reactor operator licenses, five applicants for instant senior reactor operator licenses, and two applicants for upgrade senior reactor operator licenses. The facility licensee developed the written and operating examinations using NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 12. The license examiners determined ten applicants satisfied the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 55, and the appropriate licenses were issued on November 18, 2025.

The written examination, administered operating test, and documents related to the development and review (outlines, review comments and resolution, etc.) of the examination will be withheld from public disclosure until November 1, 2027. However, because one applicant received a Preliminary Results Letter due to receiving a non-passing grade on the written examination, the applicant was provided a copy of the written examination material. For examination security purposes, your staff should therefore consider the written examination material uncontrolled and exposed to the public No findings were identified during this examination.

B. Blair 2

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Sarah Elkhiamy, Chief Operations Branch Division of Operating Reactor Safety Docket No. 05000412 License No. NPF-73

Enclosure:

Examination Report 05000412/2025301 w/

Attachment:

Supplementary Information cc w/encl: Distribution via ListServ SARAH ELKHIAMY Digitally signed by SARAH ELKHIAMY Date: 2025.11.25 16:10:18 -05'00'

B. Blair 3

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT 2 - INITIAL OPERATOR LICENSING EXAMINATION REPORT 05000412/2025301 DATED NOVEMBER 25, 2025 DISTRIBUTION:

TFish, DORS SElkhiamy, DORS LPinkham, DORS MYoung, DORS NMentzer, DORS NDay, DORS, SRI ANugent, DORS, RI CFragman, DORS, AA RClagg, RI OEDO RidsNrrPMBeaverValley Resource RidsNrrDorlLpl1 Resource R1ORAMAIL Resource DOCUMENT NAME: https://usnrc.sharepoint.com/teams/Region1OperationsBranch/Shared Documents/Examiners/FISH/Exam 25-BV2 Oct 2025/EXAM REPORT.docx ADAMS PKG: ML25024A001 ADAMS ACCESSION NO. ML25329A329 SUNSI Review

Non-Sensitive

Sensitive

Publicly Available

Non-Publicly Available OFFICE RI/DORS/OB RI/DORS/OB NAME TFish/TF SElkhiamy/SE DATE 11/25/2025 11/25/2025 OFFICIAL RECORD COPY

Enclosure EXAMINATION REPORT U.S. NUCLEAR REGULATORY COMMISSION REGION I Docket Number:

05000412 License Number:

NPF-73 Enterprise Identifier: L-2025-OLL-0033 Report Number:

05000412/2025301 Licensee:

Vistra Operations Company, LLC Facility:

Beaver Valley Power Station, Unit 2 Location:

Shippingport, PA Dates:

October 6-10, 2025 (Operating Test Administration)

October 14, 2025 (Written Examination Administration)

October 24, 2025 (Facility Submitted Post-Exam Package)

October 15 - November 7, 2025 (NRC Examination Grading)

November 18, 2025 (Licenses Issued)

Examiners:

T. Fish, Chief Examiner M. Patel, Senior Operations Engineer B. Dyke, Operations Engineer C. Henckel, Operations Engineer Approved By:

Sarah Elkhiamy, Chief Operations Branch Division of Operating Reactor Safety

2 Enclosure

SUMMARY

ER 05000412/2025301; October 6-14, 2025; Beaver Valley Power Station, Unit 2; Initial Operator Licensing Examination Report.

Four NRC examiners evaluated the competency of four applicants for reactor operator licenses, five applicants for instant senior reactor operator licenses, and two applicants for upgrade senior reactor operator licenses. The facility licensee developed the examinations using NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 12. The written examination was administered by the facility on October 14, 2025. NRC examiners administered the operating tests October 6-10, 2025. The NRC examiners determined that ten applicants satisfied the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 55, and the appropriate licenses have been issued.

A.

NRC-Identified and Self-Revealing Findings No findings were identified.

B.

Licensee-Identified Violations None.

3 Enclosure REPORT DETAILS

4.

OTHER ACTIVITIES (OA) 4OA5 Other Activities (Initial Operator License Examination)

.1 License Applications

a. Scope The examiners reviewed all license applications submitted by the licensee to ensure the applications reflected that each applicant satisfied relevant license eligibility requirements. The applications were submitted on NRC Form 398, Personal Qualification Statement, and NRC Form 396, Certification of Medical Examination by Facility Licensee. The examiners also audited two 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 Operator Knowledge and Performance

a. Examination Scope On October 14, 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 to the NRC on October 24, 2025.

The NRC examination team administered various portions of the operating examination October 6-10, 2025. The applicants for a reactor operator license participated in at least two dynamic simulator scenarios, in a control room and facilities walkthrough test consisting of eleven system tasks, and an administrative test consisting of four administrative tasks. The applicants for instant senior reactor operator licenses participated in at least two dynamic simulator scenarios, a control room and facilities walkthrough test consisting of ten system tasks, and an administrative test consisting of five administrative tasks. The applicants for upgrade senior reactor operator licenses participated in at least one dynamic simulator scenario, a control room and facilities walkthrough test consisting of five system tasks, and an administrative test consisting of five administrative tasks.

b. Findings Ten of the eleven applicants passed the operating test and the written examination. For the written examinations, the reactor operator applicants average score was 80.5 percent and ranged from 77.5 to 83.3 percent. The senior operator applicants average score was 89.4 percent and ranged from 84.5 to 92.7 percent. The text of the examination questions may be accessed in the Agencywide Documents Access and

4 Enclosure Management System (ADAMS) system under the accession numbers noted in. In accordance with current NRC policy, the release of the written examination in ADAMS to the public will be delayed for two years.

Chapter ES-403 and Form ES-403-1 of NUREG-1021 require the licensee to analyze the validity of any written examination questions that were missed by half or more of the applicants. The licensee conducted this performance analysis and submitted the analysis to the chief examiner. Included in this analysis were five post-exam comments.

The post-exam comments and Region Is resolution of the comments are in Attachment 2 to this report.

.3 Initial Licensing Examination Development

a. Examination Scope The facility licensee developed the examinations in accordance with NUREG-1021, Revision 12. All licensee facility training and operations staff involved in examination preparation and validation were listed on a security agreement. The NRC conducted an onsite validation of the operating examinations during the week of September 8, 2025.

b. Findings No findings were identified. The examiners determined that the written and operating examinations initially submitted by the licensee were within the range of acceptability expected for a proposed examination.

.4 Simulation Facility Performance

a. Examination Scope The examiners observed simulator performance with regard to plant fidelity during the examination validation and administration.

b. Findings No findings were identified.

.5 Examination Security

a. Examination Scope The examiners reviewed examination security for examination development during both the onsite preparation week and examination administration week for compliance with NUREG-1021, Revision 12, requirements. Plans for simulator security and applicant control were reviewed and discussed with licensee personnel.

b. Findings No findings were identified.

5 Enclosure 4OA6 Meetings, Including Exit The chief examiner presented the examination results to Andrew Scott, Director Nuclear Operations, and other members of the licensee's staff on November 5, 2025. The licensee did not identify any information or materials used during the examination as proprietary.

ATTACHMENT 1: SUPPLEMENTARY INFORMATION ATTACHMENT 2: CONTESTED QUESTIONS AND NRC REGION I RESPONSE

A1-1 Enclosure ATTACHMENT 1: SUPPLEMENTARY INFORMATION KEY POINTS OF CONTACT Licensee Personnel S. DeVault, Exam Support A. Kirsten, Exam Support A. Jaques, Exam Support ITEMS OPENED, CLOSED, AND DISCUSSED NONE ADAMS DOCUMENTS REFERENCED Accession No. ML25317A638 - FINAL-Written Exam (Note: In accordance with current NRC policy, the release of this examination in ADAMS to the public will be delayed for two years.)

Accession No. ML25317A637 - FINAL-Operating Exam (Note: In accordance with current NRC policy, the release of this examination in ADAMS to the public will be delayed for two years.)

A2-1 Enclosure ATTACHMENT 2: CONTESTED QUESTIONS AND NRC REGION I RESPONSE Beaver Valley Unit 2 requested the Region review comments for five questions on the Unit 2 Initial Operator Licensing written exam. The facilitys comments and the Regions associated responses are described below.

Reactor Operator Question 7

7.

Given the following:

The Plant is in MODE 5.

Overpressure Protection System (OPPS) is in service in accordance with 2OM-52.4.R.1.F, Station Shutdown from 100% to MODE 5.

NOTE: The included electrical circuitry drawings for the Pressurizer Power Operated Relief Valves reflect the CURRENT Plant status.

Subsequently, 2RCS-PT402, Rx Clnt System Wide Range Pressure, fails HIGH.

Based on the drawings provided, what is the expected system response to this failure?

A. Only 2RCS*PCV455C is open.

B. Only 2RCS*PCV456 is open.

C. Both 2RCS*PCV455C and 2RCS*PCV456 are open.

D. Neither 2RCS*PCV455C nor 2RCS*PCV456 is open.

Correct Answer: D Facility Request:

The facility requests question #7 be removed from the test.

Facility Justification:

The question stem states that OPPS was placed in service in accordance with Station Shutdown procedure, 2OM-52.4.R.1.F.

• Section K, Step 1.a; Verify 2MSPs-6.68-I and (6.69-I), Reactor Overpressurization PORV PCV 455C (456) Setpoint Functional Tests. These tests verify that the OPPS System is in the correct operating condition.

• The MSPs include a functional test verifying all switches and relay contacts function properly.

• Satisfactory performance of the MSPs is required to declare OPPS in service.

A2-2 Enclosure The electrical diagram provided to the students re"ects that OPPS was not placed into service properly and would have failed the MSP.

This does not comply with NUREG-1021 guidance on question development and introduces unnecessary confusion into the question from con"icting information provided in the stem versus the reference provided, making it impossible for the student to answer the question.

Region 1 Response:

Region 1 agrees with the facilitys comment: The description in the question stem directly conflicts with the status of OPPS shown in the reference provided to the students.

==

Conclusion:==

The question has been removed from the exam and the answer key revised accordingly.

A2-3 Enclosure Reactor Operator Question 11

11. Given the following:

A Loss of ALL AC Power occurred.

The Control Room is performing actions of ECA-0.0, Loss of All AC Emergency Power.

An Extended Loss of AC Power (ELAP) was declared by the Unit Supervisor.

Which of the following describes the design capacity of the 125VDC Station batteries during a Station blackout, and the purpose of declaring an ELAP during ECA-0.0?

1)

The design capacity of the 125VDC Station batteries is ________.

2)

The purpose of declaring an ELAP is to ________.

A.

1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

2) increase monitoring and reduce loads on both trains of batteries B.

1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

2) isolate one train of batteries, then monitor and reduce loads on the other train C.

1) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

2) increase monitoring and reduce loads on both trains of batteries D.

1) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

2) isolate one train of batteries, then monitor and reduce loads on the other train Correct Answer: B Facility Request:

Accept two answers as correct: B as the original correct answer and also Distractor A.

Facility Justification:

Regarding Part 2 of the question, without a required timeframe referenced in the stem of the question, either answer A or B are correct for the stated condition.

Commencing in referenced procedure, ECA-0.0, declaring the ELAP in Step 19 RNO and transitioning to FSG-4, you are directed to perform FSA-10 in Step 4 and FSA-11 in Step 5.

Attachment

• The initiating CAUTION in FSA-10 states; Isolating One Dc Train Within Two Hours, And Performing Load Shed On Both Dc Trains Should Be Completed Within Three Hours For Battery Preservation.

• The initiating CAUTION in FSA-11 states Isolating One Dc Train Within Two Hours, And Performing Load Shed On Both Dc Trains Should Be Completed Within Three Hours For Battery Preservation.

• Therefore, both answers A & B are correct in accordance with Station Attachments FSA-11.

Furthermore:

Answer A:

• Referring to ECA-0.0, Step 19, the RNO directs the declaration of ELAP and implementation of FSG-4.

• FSG-4 is prefaced with a Caution to load shed on both DC trains.

Answer B:

• Referencing ECA-0.0, the NOTE that precedes the declaration of ELAP in step 19, cites that Battery Coping Calculations assumes one train of batteries is isolated within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Aligning with both applicable Station procedures and without a required time component in the question stem, the wording of the FSG-4 caution makes load shed on both DC trains a correct action for ELAP coping strategy (answer A), as well as the ECA-0.0 isolation of one train of batteries. (answer B)

Region 1 Response:

Comment not accepted.

As indicated by the justification provided by the Station, the Note that precedes the declaration of an ELAP in step 19 of ECA-0.0 cites that the Battery Coping Calculations assume one train of batteries is isolated within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the in service battery train is load shed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Beaver Valley Unit 2 lesson plans include the following objective: 3SQS-53.3, Revision 5 Objective 4. Explain from memory the basis for ALL Cautions and Notes, IAW BVPS-EOP Executive Volume. Distractor A does not include the necessary procedural action of isolating a battery train; therefore, this answer choice is incorrect.

==

Conclusion:==

No answer key change is warranted.

A2-2 Enclosure Reactor Operator Question 45

45.

2GSS-PCV201 and 2GSS-PCV202, operating in series, reduces Main Steam pressure to approximately ________ as gland sealing steam for the high-pressure rotor glands.

NOTE:

2GSS-PCV201, Main Steam to Gland Steam pressure control valve 2GSS-PCV202, Gland Steam to HP Turbine pressure control valve A. 1 PSIG B. 40 PSIG C. 125 PSIG D. 150 PSIG Correct Answer: A Facility Request:

Remove Question 45 from the test Facility Justification:

Per NUREG-1021 (ES-1.2 Section B) questions should be answered based on actual plant operation, procedures and references.

In accordance with current plant conditions and operational guidance, gland sealing steam is no longer supplied by the main steam system.

Normal system arrangement is to have Auxiliary Steam supplying Gland Steam through 2GSS-42, bypassing 2GSS-PCV201.

This is done via 2OM-52.4.R.2.S, Part L, Step C.3, closing upstream 2GSS-MOV201 isolating 2GSS-PCV201 from Main Steam.

Due to this "ow path, Main Steam pressure is not reduced across those PCVs as the question suggests.

With Main Steam currently isolated from gland sealing steam, the correct answer would be 0 psig and therefore no answers are correct.

Region 1 Response:

Comment accepted. Beaver Valley Unit 2 does not operate Gland Sealing Steam as described in the question stem and is therefore an invalid question.

A2-3 Enclosure

==

Conclusion:==

The question has been removed from the exam and the answer key revised accordingly.

A2-4 Enclosure Reactor Operator Question 66

66.

Given the following:

The Plant is at 100% power.

Which of the following conditions or events (considered individually) will require Technical Specification action(s) to be performed within one hour or less?

A.

RWST borated water temperature drops to 50°F.

B.

One Containment Pressure Transmitter fails to zero.

C.

RWST borated water volume drops to 840,200 gallons.

D.

BOTH Train A - Phase B (CIB) manual Control Switches are declared inoperable.

Correct Answer: C Facility Request:

Remove question 66 from the test.

Facility Justification:

Answer B) LCO 3.3.2 Condition A requires immediate entry into table 3.3.2-1.

Answer C) LCO 3.5.4 Condition B, 1 Hour to restore RWST to operable status Answer D) LCO 3.3.2 Condition A requires immediate entry into table 3.3.2-1.

Answers B, C & D are all conditions that require actions to be performed in 1hr/immediately.

Based on the above information, the facility recommends removal of question 66.

Region 1 Response:

Comment accepted. Three or more answer choices are correct. Therefore, the question is invalid.

==

Conclusion:==

The question has been removed from the exam and the answer key revised accordingly.

A2-5 Enclosure Reactor Operator Question 69

69. Given the following:

The Plant is at 100% power.

A Large Break LOCA occurs.

The Reactor trips and SI is actuated.

All systems function as designed.

The crew enters E-0, Reactor Trip or Safety Injection, and transitions to E-1, Loss of Reactor or Secondary Coolant.

While in E-1, the following VALID annunciator and associated indication are received:

A6-3E, COOLING TOWER PUMP TROUBLE.

2CWS-P21A, A CT Pump, has a BRIGHT WHITE light lit and the RED light NOT lit.

In accordance with the guidance in FLT-OP-1002, Conduct of Operations, how is the RO ATC REQUIRED to respond to this alarm?

A.

Announce the unexpected alarm to the command SRO; perform the ARP actions in conjunction with the execution of the EOPs, without interfering with EOP execution.

B.

There is no requirement to announce the unexpected alarm; perform the ARP actions in conjunction with the execution of the EOPs, without interfering with EOP execution.

C.

There is no requirement to announce the unexpected alarm; leave the alarm flashing so long as other alarms are not masked.

D.

Announce the unexpected alarm by performing a crew update; leave the alarm flashing so long as other alarms are not masked.

Correct Answer: B Facility Request:

Remove Question 69 from the test Facility Justification:

The stem of the question states that a "VALID annunciator and associated indication" are received for the automatic trip of 2CWS-P21A, 'A' CT Pump. This would be an unexpected condition during the given transient. In addition, the tripping of a large electrical load on the 'A' Normal 4KV Bus (which is currently supplying power to the

A2-6 Enclosure

'AE' Bus) would be perceived as "significant" since it could be evidence of "continuing degradation of existing plant conditions." An unexpected Cooling Tower Pump trip is electrical in nature (see 2OM-31.1.D for CT Pump trips), which indicates a potentially more serious issue with the electrical bus and degrading plant conditions. These conditions require a crew update (see references below). Answer D specifies that the announcement is a Crew Update while A, B, and C only use the word announce. This implies a difference between announce and making a crew update, which is required per FLT-OP-1002. For that reason, A, B, C cannot be considered correct answers.

Region 1 Response:

Comment not accepted. The stem of the question states that a large break LOCA occurs and the crew enters E-0, Reactor Trip or Safety Injection and transitions to E-1, Loss of Reactor or Secondary Coolant. With a large break LOCA in progress, the status of the cooling tower pump is no longer relevant to the crew in light of the significant event in progress.

FLT-OP-1002, Conduct of Operations states the following: Section 6.9.4 (Pg. 47) -

Under this policy, the ROs only need to announce the most significant alarms to the Command SRO, while continuing to evaluate other alarms. Significant alarms are those that indicate a continuing degradation of the existing plant conditions, safety equipment failures, those that annunciate the entry into abnormal or emergency procedures, or a higher E-Plan classification. A loss of a cooling tower pump, which is no longer required for the current plant conditions, does not meet the significant threshold as previously described. Furthermore, announcing the pump trip alarm to the command SRO or performing a crew update at this point in the transient would unnecessarily distract the crew from the required, critical actions to keep the core covered. Last, there is no information in the stem to support an assumption that the cooling tower pump trip... could be evidence of "continuing degradation of existing plant conditions. Students are to answer exam questions based on information provided in the stem, and not on assumptions they may inject into the question.

Based on the above, Choice B is correct.

==

Conclusion:==

No answer key change is warranted.