ML24218A184

Examination Report and Cover Letter

ML24218A184
Person / Time
Site:	Vogtle
Issue date:	08/07/2024
From:	Matthew Endress NRC/RGN-II/DRS/OLB
To:	Martino P Southern Nuclear Operating Co
References
IR 2024301
Download: ML24218A184 (1)

Text

Patrick A. Martino, Site Vice President Southern Nuclear Operating Co., Inc.

Vogtle Electric Generating Plant 7825 River Road Waynesboro, GA 30830

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT UNIT 3 & UNIT 4 - NRC OPERATOR LICENSE EXAMINATION REPORT (0520025/2024301 and 05200026/2024301)

Dear Patrick Martino:

During the period May 13 - 17, 2024, the Nuclear Regulatory Commission (NRC) administered operating tests to employees of your company who had applied for licenses to operate the Vogtle Electric Generating Plant, Units 3 & 4. At the conclusion of the operating tests, the examiners discussed preliminary findings related to the operating tests and the written examination submittal with those members of your staff identified in the enclosed report. The written examination was administered by your staff on May 22, 2024.

Seven Reactor Operator (RO) and eight Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One RO applicant passed the operating test, but failed the written examination. There was one post-administration comment concerning the operating test and three post-administration comments concerning the written examination.

These comments, and the NRC resolution of these comments, are summarized in Enclosure 2.

A Simulator Fidelity Report is included in this report as Enclosure 3.

The initial examination submittal was within the range of acceptability expected for a proposed examination. NRC regional management considered the impacts of the post-examination comment resolution on the evaluation that the written examinations met the expected quality standards. All examination changes agreed upon between the NRC and your staff were made according to NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 12.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm.adams.html (the Public Electronic Reading Room).August 7, 2024 P. Martino 2 If you have any questions concerning this letter, please contact me at (404) 997-4718.

Sincerely, Matthew F. Endress, Chief Operations Branch 2 Division of Reactor Safety Docket Nos.: 5200025, 5200026 License Nos.: NPF-91, NPF-92

Enclosures:

1. Report Details

2. Facility Comments and NRC Resolution

3. Simulator Fidelity Report cc: Distribution via Listserv Signed by Endress, Matthew on 08/07/24

ML24218A184 X Non-Sensitive X Publicly Available X SUNSI Review Sensitive Non-Publicly Available

OFFICE RII/DRS RII/DRS RII/DRS

NAME K. Wallace M. Meeks M. Endress DATE 08/06/24 08/06/24 08/07/24

U.S. NUCLEAR REGULATORY COMMISSION

REGION II

Examination Report

Docket No.: 05200025, 05200026

License No.: NPF-91, NPF-92

Report No.: 05200025/2024301 and 05200026/2024301

Enterprise Identifier: L-2024-OLL-0024 and L-2024-OLL-0025

Licensee: Southern Nuclear Company (SNC)

Facility: Vogtle Electric Generating Plant, Unit 3 & 4

Location: Waynesboro, GA

Dates: Operating Test - May 13 - 17, 2024 Written Examination - May 22, 2024

Examiners: M. Meeks, Chief Examiner, Senior Operations Engineer J. Bundy, Senior Operations Engineer M. Donithan, Senior Operations Engineer M. Kennard, Senior Operations Engineer M. Patel, Senior Operations Engineer (Region I)

S. Battenfield, Operations Engineer T. Morrissey, Senior Resident Inspector

Approved by: Matthew F. Endress, Chief Operations Branch 2 Division of Reactor Safety

Enclosure 1

SUMMARY

ER 05200025/2024301, 05200026/2024301; May 13 - 17, 2024 & May 22, 2024; Vogtle Electric Generating Plant Unit 3 & 4; Operator License Examinations.

Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 12, of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." This examination implemented the operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45, as applicable.

Members of the Vogtle Electric Generating Plant staff developed both the operating tests and the written examination. The NRC developed the written examination outlines. The initial operating test, written RO examination, and written SRO examination submittals met the quality guidelines contained in NUREG-1021. NRC regional management considered the impacts of the post-examination comment resolution on the evaluation that the written examinations met the expected quality standards.

The NRC administered the operating tests during the period May 13 - 17, 2024. Members of the Vogtle Electric Generating Plant training staff administered the written examination on May 22, 2024. Seven Reactor Operator (RO) and eight Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One RO applicant passed the operating test, but failed the written examination. Thirteen applicants were issued licenses commensurate with the level of examination administered; however, licensing actions for two SRO applicants who were granted deferrals have been delayed, pending receipt of additional information.

There was one post-examination comment concerning the operating test, and three post-examination comments concerning the written examinations.

No findings were identified.

2 REPORT DETAILS

4. OTHER ACTIVITIES

4OA5 Operator Licensing Examinations

a. Inspection Scope

The NRC evaluated the submitted operating test by combining the scenario events and JPMs in order to determine the percentage of submitted test items that required replacement or significant modification. The NRC also evaluated the submitted written examination questions (RO and SRO questions considered separately) in order to determine the percentage of submitted questions that required replacement or significant modification, or that clearly did not conform with the intent of the approved knowledge and ability (K/A) statement. Any questions that were deleted during the grading process, or for which the answer key had to be changed, were also included in the count of unacceptable questions. The percentage of submitted test items that were unacceptable was compared to the acceptance criteria of NUREG-1021, Operator Licensing Standards for Power Reactors.

The NRC reviewed the licensees examination security measures while preparing and administering the examinations in order to ensure compliance with 10 CFR §55.49, Integrity of examinations and tests.

The NRC performed an audit of license applications during the preparatory site visit in order to confirm that they accurately reflected the subject applicants qualifications in accordance with NUREG-1021.

The NRC administered the operating tests during the period May 13 - 17, 2024. The NRC examiners evaluated eight Reactor Operator (RO) and eight Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. Members of the Vogtle Electric Generating Plant training staff administered the written examination on May 22, 2024. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Vogtle Electric Generating Plant, Units 3 & 4, met the requirements specified in 10 CFR Part 55, Operators Licenses.

During the examination development, validation, and administration processes, the NRC examiners identified several minor procedural enhancements to the facility licensee.

After exam administration was completed, the facility licensee entered these procedural enhancements in their corrective action program; reference CR #11083949, CR#11083951, CR#11083952, CR#11083967, CR#11083971, CR#11083972, and CR#11083974 (all CRs dated 06/11/2024).

The NRC evaluated the performance and fidelity of the simulation facility during the preparation and conduct of the operating tests. One issue related to simulator performance is documented in Enclosure 3.

b. Findings

No findings were identified.

Enclosure 2 The NRC developed the written examination sample plan outline. Members of the Vogtle Electric Generating Plant, Units 3 & 4, training staff developed both the operating tests and the written examination. All examination material was developed in accordance with the guidelines contained in Revision 12, of NUREG-1021. The NRC examination team reviewed the proposed examination. Examination changes agreed upon between the NRC and the licensee were made per NUREG-1021 and incorporated into the final version of the examination materials.

The NRC determined, using NUREG-1021, that the licensees initial examination submittal was within the range of acceptability expected for a proposed examination.

NRC regional management considered the impacts of the post-examination comment resolution on the evaluation that the written examinations met the expected quality standards.

Copies of all individual examination reports were sent to the facility Training Manager for evaluation of weaknesses and determination of appropriate remedial training.

The licensee submitted one post-examination comment concerning the operating test, and three post-examination comments concerning the written examination. A copy of the full text of the post-examination comments may be accessed in the ADAMS system at ADAMS Accession Number ML24213A081. A copy of the final written examination and answer key, with all changes incorporated, may be accessed not earlier than December 22, 2026, in the ADAMS system (ADAMS Accession Number(s)

ML24213A085 and ML24214A002). Note that these written examinations have had proprietary information redacted, as specified by the Vogtle Electric Generating Plant Unit 3 and 4 design and construction vendor.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On May 17, 2024, the NRC examination team discussed generic issues associated with the operating test with P. Martino, Site Vice President, and other members of the Vogtle Electric Generating Plant (Unit 3 and 4) staff. The examiners asked the licensee if any of the examination material was proprietary. The information that the licensee identified as proprietary was handled in a manner consistent with NRC and licensee guidelines for this type of information. On July 19, 2024, the NRC examination team conducted a final exit meeting with A. Nichols, Examination Development Lead, and other members of the Vogtle Electric Generating Plant (Unit 3 and 4) staff to discuss the examination results and provide the licensing details.

2 KEY POINTS OF CONTACT

Licensee personnel

M. Brummitt, Plant Manager G. Crosby, Operations Training Coordinator A. Ferguson, Simulator Coordinator W. Garrett, Licensing Manager J. Hartman, Shift Operations Manager C. Howard, Examinations Developer S. Leighty, Regulatory Affairs Manager P. Martino, Site Vice President T. Mays, Initial Licensed Training (ILT) Lead Instructor A. McRae, ILT Class Coordinator A. Nichols, Examinations Development Lead J. Overstreet, Training Director C. Parkes, Operations Services Manager/Operations Facility Representative A. Schwartz, Operations Training Manager

3 FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS

A complete text of the licensees post-examination comments can be found in ADAMS under Accession Number ML24213A081.

Item

NRC Scenario 3, Event 5

Facility Licensee Comment:

Purpose:

Crews identified various TS [Technical Specification] CONDITIONS based on their evaluation of a scenario event. NRC has requested we clarify required TS entry conditions for situations that existed, for post-exam comment and to ensure the exam key matches the as-run scenario, specifically related to evaluation of LCO [Limiting Conditions for Operation] 3.7.6, CONDITIONS A, B, and C.

Summary: During NRC scenario 3 event 5, VBS-D244 loses control power and fails closed. VBS-D244 causes VES-PDT004A/B to lower and, within 10 minutes, results in an automatic actuation signal of MCR [Main Control Room] Isolation and Air Supply Initiation (VES). When the automatic MCR Isolation actuation signal is generated VES-V005A/B fail to open as expected. VES-V005A/B will NOT open until operators perform manual MCR Isolation actuation from the PDSP.

No other VBS/VWS components are failed and work accordingly, except VBS-D244 and VES-V005A/B as described above. Based on crew timing and actions, there were two situations encountered:

1. Crew manually actuates MCR Isolation from the PDSP (prior to receiving automatic MCR Isolation actuation signal).

2. Crew receives automatic MCR Isolation actuation signal (prior to manually actuating MCR Isolation from the PDSP) and subsequently performs the manual actuation

Manual and automatic MCR Isolation actuation sends a signal to the following components:

• Closes VBS-V186/V187

• Closes VBS-V188/V189

• Closes VBS-V190/V191

• Opens VES-V005A/B

• Opens VES-V022A/B

• Stage 1 load shed occurs

• Timer for Stage 2 load shed starts (180 minute timer)

LCO 3.7.6 evaluations:

a. CONDITION A:

a. The initiating event (VBS-D244 failure) does NOT require entry into LCO 3.7.6, Condition A. VBS-D244 is NOT one of the dampers or valves covered under this specification or its associated surveillances.

b. If VES-V005A/B fail to open on an automatic actuation signal, then LCO 3.7.6, CONDITION A entry is required.

4

b. CONDITION C: Whenever VES-TE058A/B exceed 75°F, when in the MODES of APPLICABILITY, then LCO 3.7.6, CONDITION C entry is required.

c. CONDITION B: LCO 3.7.6, CONDITION B is NOT required to be entered because load shed occurred as expected in both situations. VES-V005A/B are the only components that failed to actuate.

NOTE: There is a frozen reference folder for ILT-6. The following versions of material are used:

• AP-LT-I-PP-VBS, Version 6.1

• AP-LT-I-PP-VES, Version 5.5

• 3-AOP-501, Version 1.0

• 3-PMS-SOP-001, Attachment 17, Version K=0.10

• SV3-VBS-M6-007, Revision 8 (APP-VBS-M6-007, Revision 10)

• 3-SV3-PMS-J3-366, Revision 9

• 3-GEN-OTS-17-003, Version D=0.3

• 3-VES-OTS-17-001, Version B=0.1

• VEGP 3&4 TECH_SPEC_COL APP A:

o Amendment 108 (Unit 3) - pages 3.7.6-1,2,3 o Amendment 107 (Unit 4) - pages 3.7.6-1,2,3 o Amendment 183 (Unit 3) - pages 3.7.6-4,5 o Amendment 181 (Unit 4) - pages 3.7.6-4,5

• VEGP 3&4 TS BASES:

o B3.7.6-1 through 15, Revision 67

• VEGP 3&4 UFSAR, Revision 12.2

NRC Resolution

The licensees recommendation was accepted.

During examination development, scenario 3 event 5 was proposed as a component failure that did not result in TS entry. Per the scenario design, it was expected that the applicant teams would diagnose a loss of control power to VBS-D244, which would result in a loss of MCR air conditioning. The applicants were then expected to actuate MCR isolation from the PDSP in accordance with AOP-501, Loss of Main Control Room Air Conditioning. During the onsite validation/examination prep week, the licensed operator team performing the validation of scenario 3 performed event 5 without identifying any applicable TS.

However, during the actual examination administration, several of the applicant teams identified that MCR temperatures exceeded 75 degrees F; and, depending on the timing of the operating teams actions in conjunction with the integrated plant response, some applicant teams identified the failure of VES-V005A/B to open on the automatic actuation signal, some teams did not, and for some teams the failure of VES-V005A/B to automatically open did not become manifest.

During the scenario runtime and post-scenario follow-up questions, the various Senior Reactor Operator (SRO) applicants identified a variety of entries into TS LCO 3.7.6., Main Control Room Emergency Habitability System (VES). Based on these entries, the NRC examiners and the facility licensee determined a post-examination comment would be warranted to formally resolve the correct TS entries for this event, if any

5 Based on the NRC examiners evaluation, the examiners agree with the facility licensee on the applicable TS entries for this event.

In summary, first, if MCR temperature exceeded 75 degrees F, then entry into LCO 3.7.6 CONDITION C was required. LCO 3.7.6 CONDITION C was entered based upon Thermal mass of one or more required heat sink(s) not within limit(s), and had REQUIRED ACTIONS C.1 Restore required heat sink air temperatures to within limit(s) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND C.2. Restore thermal mass of required heat sink(s) to within limit(s) within 5 days.

If the applicant team had an opportunity to identify the failure of VES-V005A/B to open on the automatic MCR isolation signal, then entry into LCO 3.7.6 CONDITION A was required. LCO 3.7.6 CONDITION A was entered based upon One valve or damper inoperable, and had REQUIRED ACTION A.1 to Restore valve or damper to OPERABLE status in a 7 day COMPLETION TIME.

Based on the failures presented in this scenario event, entry into LCO 3.7.6 CONDITION B, One PMS Division inoperable in one or more MCR load shed panel(s), was not required under any of the circumstances that occurred during the examination administration.

Based on this resolution, all of the SRO applicants who performed Scenario 3 in the SRO watch position (Shift Supervisor at Vogtle Unit 3&4) were graded per the above guidance when evaluating the required TS entries for Event 5.

6 Item

Question 36, K/A ADS A3.01

Facility Licensee Comment:

[N.B. The facility licensee applicant began by listing the text of Question 36 and the initial keyed answer of A]

Facility Recommendation

Recommend changing the key from A PXS-V123A to B PXS-V120A.

Reason

BOTH PXS-V123A & PXS-V120A receive a signal to open at the same time, the difference in TDs will result in PXS-V120A FIRST going from closed to open.

Justification

[facility licensee included a table here showing time/lower IRWST level/signals/open signals]

At 1003:00 IRWST Lower NR level is 25.5% and still lowering.

As given in the stem, automatic ADS Stage 4 actuation is also generated at time 1003:00.

ADS Stage 4 Actuation sends a signal to open PXS-V123A. (PMS-J3-369, PMS-J3-371, PMS-J3-513)

ADS Stage 4 Actuation with IRWST Lower NR Level below Low-3 set point (25.7%) results in a signal generated to open PXS-V120A. (PMS-J3-369/372, PMS-J3-374, PMS-J3-512)

Opening of PXS-V123A is delayed 29 seconds. (PMS-J3-513)

Opening of PXS-V120A is delayed 14 seconds. (PMS-J3-512)

With both valves receiving a signal to open at the same time, the difference in TDs will result in PXS-V120A FIRST going from closed to open.

Further information

As mentioned in PMS-SOP-001, Version K=0.10, PXS-V123A opens upon ADS Stage 4 actuation signal, with 29 second TD. [portion of procedure PMS-SOP-001 not included here]

As mentioned in PMS-SOP-001, Version K=0.10, PXS-V120A opens upon ADS Stage 4 actuation signal, coincident with LRWST NR level Low-3 (25.7%), with 14 second TD. [portion of procedure PMS-SOP-001 not included here]

PXS-V123A (IRWST Injection valve) opens upon receipt of ADS Stage 4 actuation signal.

7 PXS-V120A (IRWST Containment Recirculation Isolation valve) opens upon ADS Stage 4 actuation coincident with IRWST Lower NR level Low-3.

[portion of plant logic diagram not included here]

PXS-V123A operates based on the following controls (PXS-M3C-100)

PXS-V120A operates based on the following controls (PXS-M3C-100)

[portions of procedure PXS-M3C-100 not included here]

(PXS-V123A) ADS Stage 4 actuation signal generates signal to IRWST injection valve (PMS-J3-371):

[portion of plant logic diagram not included here]

The signal from PMS-J3-371 then goes to PMS-J3-513 to open PXS-V123A, after TD (29 seconds):

[portion of plant logic diagram not included here]

(PXS-V120A) PMS-J3-372 identifies Low IRWST NR level as 25.7%

Low IRWST NR level is sent to PMS-J3-374, and goes through AND gate with ADS Stage 4 actuation signal, to generate signal for Containment Recirculation Isolation valve:

[portion of plant logic diagram not included here]

(PXS-V120A continued) PMS-J3-512 shows that PXS-V120A opens after TD (14 seconds):

[portion of plant logic diagram not included here]

PXS-V123A Time Delay is 29 seconds and PXS-V120A Time Delay is 14 seconds.

Since BOTH valves receive a signal to open at the same time, PXS-V120A will go from closed to open FIRST, 15 seconds earlier than PXS-V123A.

NRC Resolution

The licensees recommendation was accepted.

During written exam administration on May 22, 2024, there were no applicant questions pertaining to RO exam Question 36.

During examination development, this question was presented as a modified bank question. The answer to the initial question was valve PXS-V123A. As a result of the modifications made to the bank question, the rate of IRWST level lowering was increased from 0.2% per minute (original question) to 0.5% per minute (as-administered question), and the time that the automatic ADS Stage 4 actuation was generated was changed from 1 minute later (original question) to 3 minutes

8 later (as-administered question). These modifications inadvertently changed the correct answer from valve PXS-V123A, which was the original keyed answer, to PXS-V120A, as discussed below.

Based on the given conditions, at a hypothetical time of 10:00:00, IRWST level is 27.0% and lowering at a constant rate of 0.5% level per minute. At time 10:03:00, an automatic ADS Stage 4 actuation occurs. Based on IRWST level rate change, at time 10:03:00, IRWST level was 25.5%

and lowering.

Valve PXS-123A received an automatic open signal based on automatic ADS Stage 4 at time 10:03:00 with a 29 second delay. Therefore PXS-123A will reposition to open beginning at 10:03:29.

Valve PXS-120A received an automatic open signal based on ADS Stage 4 actuation coincident with IRWST level below the Low-3 setpoint of 25.7%, and therefore received an automatic open signal at time 10:03:00 with a 14 second delay. Therefore PXS-120A will reposition to open beginning at 10:03:14; therefore, PXS-120A will be the FIRST valve to reposition from closed to open, which was the required knowledge elicited from the question statement.

Therefore, the newly presented technical information by the facility licensee warranted a change in the answer key from A to B. All applicants were graded on the written examination with the correct answer of Question 36 as B.

9 Item

Question 61, K/A WGSA4.03

Applicant Comment

[N.B. The facility licensee began by listing the text of Question 61 and the initial keyed answer of A]

Facility Recommendation

Recommend changing the key from A 1003:00 / can to B 1003:00 / can NOT.

Reason The second half of the question tests, based on the current plant conditions, whether the valve (WGS-V051) can be bypassed using PLS controls from the MCR.

Based on the current plant conditions, the design of the system, and facility standards and expectations:

• (Unclear stem where information provided could reasonably result in applicant misunderstanding the intent) There is no physical bypass flowpath around WGS-V051

• (Newly discovered technical information that supports a change) There is no procedure guidance to direct bypassing High radioactivity signal for WGS-V051.

• (Unclear stem where information provided could reasonably result in applicant misunderstanding the intent) Opening a flow path, with a known High rad condition and no procedure guidance, results in non-conservative action that goes against Operations Personnel Responsibilities, Standards and Expectations

Taking into consideration the unclear stem and the newly discovered technical information, facility proposes that the answer key for the second half question should be changed to can NOT.

Justification

1. Meaning of word bypass

• The question specifically tests whether the valve (WGS-V051) can be bypassed, not signals associated with radiation, pressure, or fan flow can be bypassed.

• The exam developers intended the question to test knowledge of whether I&C signals associated with radiation, pressure or fan flow could be bypassed and subsequent action be taken to operate controls on WGS-V051 to manually override these signals and open WGS-V051.

• The Oxford English Dictionary defines bypass as a secondary channel, pipe, or connection to allow a flow when the main one is closed or blocked.

• Merriam Webster defines bypass as a channel carrying a fluid around a part and back to the main stream.

• The term bypass is commonly used in the commercial nuclear power industry for valves around the main flowpath (ref. dictionaries above - Feed Reg Bypass, Pressurizer Spray Bypass, Turbine Bypass, et al).

• As written, the question is actually testing whether-or-not a bypass flowpath around WGS-V051 can be operated from the MCR to restore Waste Gas flow based on the given conditions.

10

2. Procedure guidance

• In accordance with NMP-OS-007, all Operations personnel are responsible to:

-Ensure the health and safety of the public through safe and efficient operation of the plant in accordance with procedures, nuclear standards and expectations, and regulatory requirements

• Procedure guidance does NOT exist to direct action for bypassing WGS-V051 with the given conditions

• Based on the current plant conditions, and in accordance with WGS-ARP-001-037, WGS would have to be removed from service using WGS-SOP-001, Attachment 8 and can NOT be bypassed from the MCR with the current conditions.

3. Safety and health of the public / standards and expectations

• In accordance with NMP-OS-007, all Operations personnel are responsible to:

-Demonstrate and promote a conservative approach by having nuclear, radiological, and personnel safety as top priorities

• Facility expects licensed operators to take conservative action to place the plant in a safe condition that protects the safety of the health and public and would not proceed in the face of uncertainty to bypass WGS-V051 with the current conditions

• No procedure guidance exists for bypassing high radiation signals associated with WGS-V051 and no context is provided why bypassing I&C signals would be safe and prudent action

NRC Resolution

The facility licensees recommendation was not accepted.

During written exam administration on May 22, 2024, there were no questions asked by the applicant class related to written examination Question 61.

The facilitys contention that the use of the term bypassed in the question stem was confusing as could be interpreted as a physical bypass valve or bypass line around the WGS-V051 valve was not accepted; specifically, the question stem asked if WGS-V051 [can or can NOT] be bypassed using PLS controls from the MCR, not if there was a physical bypass line or bypass valve around the WGS-V051 valve in the plant. Moreover, the PLS control window for the WGS-V051 valve contains a poke (virtual button) specifically for a bypass of WGS-V051.

The facility licensee further contended that there was no procedure guidance for bypassing WGS-V051. However, the examiners noted that the question stem was not tied to any procedure, and instead of asking if a procedure step requires the WGS-V051 valve to be bypassed (procedure knowledge), the question stem is eliciting knowledge of whether or not the WGS-V051 could be bypassed using PLS controls from the MCR (systems knowledge of interlocks and bypasses). Specific to the issue of whether or not knowledge of WGS-V051 bypasses was proceduralized, note that procedure 3-WGS-SOP-001, Gaseous Radwaste System, Attachment 1, PURGE WGS FOR STARTUP OR MAINTENANCE, contains Precaution and Limitation 2.0.4: Ovation allows the following Closure signals to be bypassed for WGS-V051: RIRCA-017-High, WLS-MP-03A/B OFF or VAS Loss of Flow Signals. WGS-PT014 Low-1 closure and WGS-PT014 Low-2 trip may not be bypassed.

11 The exact same language is contained in 3-WGS-SOP-001 Attachment 2, PLACE WGS IN STANDBY, Precaution and Limitation 2.0.6. The exact same language was also presented to the applicants as training material in lesson plan AP-LT-I-PP-WGS, Gaseous Radioactive Waste System (WGS), specifically on slides 34-39.

Based on the conditions in the stem, the applicants were presented with a high radiation condition on WGS-RY017 and a high pressure on WGS-PT014 at time 10001:00. Then, at time 10:05:00, VAS Loss of Flow signals were received on VAS-FICA-012A/B and VAS-FICA-013A/B. In accordance with the precaution and limitation mentioned above, therefore, the interlocked signals for valve WGS-V051 were high radiation (which can be bypassed), VAS Loss of Flow (which can be bypassed), and high pressure (not low pressure) on PT-014.

Therefore, the examiners determined that the only technically accurate answer for Question 61 remained A.

All applicants were graded on the written examination with the correct answer of Question 61 remaining unchanged as A.

12 Item

Question 62, K/A RMSA2.05

Facility Licensee Comment

[N.B. The facility licensee began by listing the text of Question 62 and the initial keyed answer of A]

Facility Recommendation

In addition to answer A, facility recommends accepting answer B close PSS-V011A as an additional correct answer.

Reason

The question is asking what the next appropriate action is for a high radiation condition in the Primary Sampling System.

The question states that PSS-RY050 (RCS Sample Process Rad) alarm is in.

Upon receipt of this alarm, in accordance with NMP-OS-007-001, Section 4.11 (Alarm Response) the crew is expected to perform the Annunciator Response Procedure (ARP) following receipt and communication of the alarm.

In accordance with B-ADM-OPS-002 (Alarm Response and Status Control) Figure 1(Alarm Response Flowchart) crew is required to take appropriate action for the cause of any alarm using Figure 1 (see further information)

The appropriate ARP for PSS-RY-050 is PSS-ARP-001-009 (see further information).

• Step 2 of the ARP is to ensure PSS-V011A is CLOSED.

• Step 6 of the ARP is to perform AOP-113 (Reactor Coolant System High Activity).

ARP steps are required to be performed in sequential order (ref. NMP-AP-003).

The question states that PSS-V011A (Liquid Sample Line A CIV AOV) is OPEN and asks what the next required action is.

Both A. Isolate Letdown and B. close PSS-V011A should be considered correct answers because ARP actions are required to be performed in order. PSS-V011A could be closed prior to implementing AOP-113. Isolating letdown is also an appropriate action for these conditions.

Justification

PSS-RY050 (RCS Sample Process Rad) alarm is an entry condition for AOP-113, and justification that the ARP would be performed prior to AOP-113 entry.

NMP-AP-005-004 states In the case of a single component failure, ARP actions may address the condition more directly than AOP actions. In such cases, it is acceptable for the SS to direct

13 the ARP actions that provide simple, component specific direction to correct the malfunction and then reference applicable AOP(s) after ARP actions are complete.

The appropriate course of action is to perform sequential steps in the ARP to close PSS-V011A in step 2 then implement AOP-113 in step 6. Because PSS-V011A is open and an active ARP step provides direction to close it, answer B should also be accepted as correct.

[here the facility licensee attached the ARP for PSS-RY050 (procedure 3-PSS-ARP-001-009) and a flowchart from procedure B-ADM-OPS-002, neither included here]

NRC Resolution

The facility licensees recommendation was not accepted.

During written exam administration on May 22, 2024, there were no questions asked by the applicant class related to written examination Question 62.

The examiners acknowledged that there could be valid circumstances where, under actual operations, it could be true that operators could be performing actions in parallel or as directed by the SRO (Shift Supervisor) such that it could be possible that closing PSS-V011A would be a next action. For example, it would be allowable under Vogtle Unit 3&4 conduct-of-operations for the SRO to assign ARP actions to one RO, and AOP actions to another RO, such that it might be possible that closing PSS-V011A would occur next in a given sequence or timeline; in other words, if that timeline were to take place, the examiners would not construe such a sequence of actions as improper actions.

However, there are other considerations that must be considered when analyzing this question.

The two principal considerations both involve the written exam questions link to procedure AOP-113, Reactor Coolant System High Activity. First, the last bulleted condition listed in the question stem at time stamp 1000:00 is that AOP-113 step 1 Announce High RCS Activity Condition is performed. The intention of including this step is to ensure that the applicants are focused on the specific point (i.e., step 2) in the implementation of AOP-113 that the operators were performing. As the facility licensee identified in their contention, it is the normal flow path to first perform the ARP, and then enter and perform any AOP actions as directed by the ARP or as directed by the SRO/Shift Supervisor. Because this is the normal flowpath of procedure usage, it would be an unwarranted assumption on the part of the applicants taking this question to believe that the ARP and AOP were being performed in parallel, just in such a fashion that the ARP action to attempt closure of PSS-V011A was the next sequential action to be performed. Note that closing PSS-V011A is also an action that is included in AOP-113, just at the end of the procedure, not at step 2 (which is the correct answer of isolate letdown).

Furthermore, recall that NUREG-1021 ES-1.2 step B.8 stated, in part, that:

When answering a question, do not make assumptions about conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question. For example, you should not assume that any alarm has activated unless the question so states or the alarm is expected to activate as a result of the conditions that are stated in the question. Similarly, you should assume that no operator actions have

14 been taken, unless the stem of the question or the answer choices specifically state otherwise. [emphasis added] []

Accordingly, the NRC examiners evaluated that the contention that operators could be performing the ARP actions in parallel with the AOP actions, constituted an unwarranted assumption that was not permitted in accordance with NUREG-1021 ES-1.2.

Secondly, the question statement specifically links the required answer/intended elicited knowledge of the answer to the AOP procedure steps, and not the ARP content: In accordance with AOP-113 and considering ONLY the given responses below, the crew is NEXT required to

_______. In other words, irrespective of how an operations team might be sequentially performing the multiple valid actions of the ARP and AOP, the question is specifically directing the applicant to demonstrate knowledge of the AOP.

The examiners evaluated that it was clear that the correct next action, in accordance with AOP-113, was to isolate letdown (step 2); the AOP-113 procedure does direct the operators to close PSS-V011A, but not until operators perform step 6 (2)(a) RNO (which is, in fact, the -last-step in the procedure). Therefore, in accordance with NUREG-1021, the examiners determined that A remained the one and only technically correct answer to written examination question Q62.

All applicants were graded on the written examination with the correct answer of Question 62 remaining unchanged as A.

15 SIMULATOR FIDELITY REPORT

Facility Licensee: Vogtle Electric Generating Plant, Unit 3 and 4 (AP-1000)

Facility Docket No.: 05200025, 05200026

Operating Test Administered: May 13 - 17, 2024

This form is to be used only to report observations. These observations do not constitute audit or inspection findings and, without further verification and review in accordance with Inspection Procedure 71111.11 are not indicative of noncompliance with 10 CFR 55.46. No licensee action is required in response to these observations.

While conducting the simulator portion of the operating test, examiners observed the following:

Item Description

DR 0030368 During the administration of a simulator/control room JPM for the NRC initial licensed operator exam, the Computerized Procedures System (CPS) experienced an unexpected crash, which resulted in a temporary loss of the ability for operators to access plant procedures and the ability for the simulator to run. The simulator support staff were contacted, and were able to re-boot the CPS server and conduct a CPS system check that ensured all DDS components (including the CPS) were operational. Training management and NRC examiners agreed to continue with the simulator JPM, and the remaining applicants were successfully examined without further incident or re-occurrence of the CPS crash.

Enclosure 3