IR 05000456/2025301
| ML26029A447 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 01/30/2026 |
| From: | Paul Zurawski NRC/RGN-III/DORS/OB |
| To: | Mudrick C Constellation Energy Generation, Constellation Nuclear |
| Shared Package | |
| ML23205A058 | List: |
| References | |
| 2025301 | |
| Download: ML26029A447 (0) | |
Text
SUBJECT:
BRAIDWOOD STATION - NRC INITIAL LICENSE EXAMINATION REPORT 05000456/2025301 AND 05000457/2025301
Dear Christopher Mudrick:
On December 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed the initial operator licensing examination process for license applicants employed at Braidwood Station.
The enclosed report documents the results of those examinations. Preliminary observations noted during the examination process were discussed on December 8, 2025, with Victoria Ferguson, Senior Manager - Site Training Manager, and other members of your staff. An exit meeting was conducted via Microsoft Teams on January 21, 2026, between Adam Schuerman, Site Vice President, and Gregory Roach, Senior Operations Engineer, to review the proposed final grading of the examination for the license applicants. During the exit meeting, the NRC resolutions of post-examination comments submitted by the facility, initially received by the NRC on December 31, 2025, were discussed.
The NRC examiners administered an initial license examination operating test during the weeks of December 1, 2025, and December 8, 2025. The written examination was administered by Braidwood Station training department personnel on December 11, 2025. Nine senior reactor operator (SRO) and seven reactor operator (RO) applicants were administered license examinations. The results of the examinations were finalized on January 23, 2026. Fourteen applicants passed all sections of their respective examinations and received operator licenses.
Two applicants were issued Preliminary Results Letters. Eight applicants were issued senior operator licenses and six applicants were issued operator licenses.
The as-administered written examination and operating test, as well as documents related to the development and review (outlines, review comments and resolution, etc.) of the examination will be withheld from public disclosure until December 31, 2027. The enclosure contains details of this report.
However, since two applicants received a Preliminary Results letter because of a written examination grade that is less than 80.0%, the applicants were provided copies of the written examination. For examination security purposes, your staff should consider that written examination uncontrolled and exposed to the public.
January 30, 2026 This letter, its enclosures, 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, Part 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Paul J. Zurawski, Chief Operations Branch Division of Operating Reactor Safety Docket Nos. 50-456, 50-457 License Nos. NPF-72, NPF-77 Enclosure:
1. Examination Report 05000456/2025301; 05000457/2025301 2. Post-Examination Comments, Evaluation, and Resolutions 3. Simulator Fidelity Report cc: Distribution via GovDelivery V. Ferguson, Senior Manager - Site Training Manager Signed by Zurawski, Paul on 01/30/26
SUMMARY OF FINDINGS
ER 05000456/2025301; 05000457/2025301; 12/1/2025-12/31/2025; Constellation Nuclear,
Braidwood Station. Initial License Examination Report.
The announced initial operator licensing examination was conducted by regional Nuclear Regulatory Commission (NRC) examiners in accordance with the guidance of NUREG-1021,
Operator Licensing Examination Standards for Power Reactors, Revision 12.
Examination Summary:
Fourteen of 16 applicants passed all sections of their respective examinations. Eight applicants were issued senior operator licenses and six applicants were issued operator licenses.
(Section 4OA5.1).
REPORT DETAILS
4OA5 Other Activities
.1 Initial Licensing Examinations
a. Examination Scope
The NRC examiners and members of the facility licensees staff used the guidance prescribed in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 12, to develop, validate, administer, and grade the written examination and operating test. The written examination outlines were prepared by the NRC staff and were transmitted to the facility licensees staff. Members of the facility licensees staff developed the operating test outlines and developed the written examination and operating test. The NRC examiners validated the proposed examination during the week of October 27, 2025, with the assistance of members of the facility licensees staff. During the on-site validation week, the examiners audited all license applications for accuracy. The NRC examiners administered the operating test, consisting of job performance measures and dynamic simulator scenarios, during the period of December 1, 2025, through December 8, 2025. The facility licensee administered the written examination on December 11, 2025.
On December 31, 2025, the licensee submitted documentation noting that there were three post-examination comments for consideration by the NRC examiners when grading the written examination. The post-examination comments and the NRC resolution for the post-examination comments are provided in Enclosure 2 of this report.
b. Findings
- (1) Written Examination The NRC examiners determined that the written examination, as proposed by the licensee, was within the range of acceptability expected for a proposed examination.
Less than 20% of the proposed examination questions were determined to be unsatisfactory and required modification or replacement.
During validation of the written examination, several questions were modified or replaced. All changes made to the proposed written examination were made in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors, and documented on Form 2.3-5, Written Examination Review Worksheet.
The Form 2.3-5, the written examination outlines, and both the proposed and final written examinations, will be available electronically in the NRC Public Document Room or from the Publicly Available Records component of NRCs Agencywide Documents Access and Management System (ADAMS) on December 31, 2027 (ADAMS Accession Numbers for the following: administrative files ML23205A056, examination outlines ML23205A057, proposed exam ML23205A059, and as-administered exam ML23205A055).
The NRC examiners graded the written examination on January 21, 2026, and conducted a review of each missed question to determine the accuracy and validity of the examination questions. There were five questions associated with operator actions in response to abnormal and emergency events and two questions associated with plant systems where post-examination analysis identified generic weaknesses in that more than 50% of the applicants responded incorrectly to these questions.
- (2) Operating Test The NRC examiners determined that the operating test, as originally proposed by the licensee, was within the range of acceptability expected for a proposed examination.
Less than 20% of the proposed operating test portion of the examination was determined to be unsatisfactory and required modification or replacement.
During the validation of the operating test, several job performance measures (JPMs)were modified or replaced, and some modifications were made to the dynamic simulator scenarios. Changes made to the operating test portion of the examination were made in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors, and documented on Form 2.3-3, Operating Test Review Worksheet. The Form 2.3-3, the operating test outlines, and both the proposed and final as-administered dynamic simulator scenarios and JPMs, will be available electronically in the NRC Public Document Room or from the Publicly Available Records component of NRCs ADAMS on December 31, 2027 (ADAMS Accession Numbers for the following: administrative files ML23205A056, examination outlines ML23205A057, proposed exam ML23205A059, and as-administered exam ML23205A055).
The NRC examiners completed operating test grading on January 22, 2026.
Post-examination analysis revealed generic weaknesses in applicant performance in application of the Technical Specifications, the ability to determine shutdown risk, and performance of Shift Emergency Director actions by senior reactor operator (SRO)applicants.
- (3) Examination Results Nine SRO and seven RO applicants were administered examinations. Eight SRO applicants were issued senior operator licenses, and six RO applicants were issued operator licenses.
.2 Examination Security
a. Scope
The NRC examiners reviewed and observed the licensees implementation of examination security requirements during the examination validation and administration to assure compliance with Title 10 of the Code of Federal Regulations, Section 55.49, Integrity of Examinations and Tests. The examiners used the guidelines provided in NUREG 1021, Operator Licensing Examination Standards for Power Reactors, to determine acceptability of the licensees examination security activities.
b. Findings
None.
4OA6 Meetings
.1 Debrief
The chief examiner presented the examination teams preliminary observations and findings on December 8, 2025, to Victoria Ferguson, Senior Manager - Site Training Manager, and other members of the Braidwood Stations Operations and Training Departments staff.
.2
Exit Meeting
The chief examiner conducted an exit meeting on January 21, 2026, with Adam Schuerman, Site Vice President, via Microsoft Teams. The NRCs final disposition of the stations grading of the written examination and post-examination comments were disclosed and discussed during the Microsoft Teams meeting. The chief examiner asked the licensee whether any of the retained submitted material used to develop or administer the examination should be considered proprietary. No proprietary or sensitive information was identified during the examination or debrief/exit meetings.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- A. Schuerman, Site Vice President
- V. Ferguson, Manager, Senior Manager - Site Training
- K. Heuser, Operations Training Manager
- S. Wiscons, Regulatory Exam Author
- M. Tubergan, Facility Operations Representative
U.S. Nuclear Regulatory Commission
- J. Robbins, Senior Resident Inspector
- P. Smagacz, Resident Inspector
- G. Roach, Chief Examiner
- B. Jebbia, Examiner
- J. Robb, Examiner
- C. Hunt, Limited Examiner
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
None.
LIST OF ACRONYMS USED
Agency-Wide Document Access and Management System
Corrective Action Program
Loss of Coolant Accident
NRC
U.S. Nuclear Regulatory Commission
Reactor Coolant Pump
Reactor Operator
Senior Reactor Operator
Simulator Work Request
Facility/Applicant Comments and NRC Resolutions
Question #3
Unit 1 was at 100% power.
An RCS LOCA occurred followed by a reactor trip and SI.
The crew is performing 1BwEP-0, REACTOR TRIP OR SI.
RCS pressure is BELOW RCP trip criteria, and the RO is monitoring ECCS flow to
confirm the trip criteria.
1FI-917, High Head SI flow indicator indicates 0 GPM, and the indicator is suspected to
be stuck at 0 GP
- M.
As an alternative to 1FI-917Property "Contact" (as page type) with input value "M.</br></br>As an alternative to 1FI-917" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., the LOWEST total SI pump discharge flow at which the RO can
confirm RCP trip criteria is > GPM.
A.
100
B.
200
C.
500
D.
1000
Answer
B
Answer Explanation
A - Plausible: Because the criteria for RCP tripping for CV pump flow is 100 gpm per 1BwEP-0.
B - Correct: Per 1BwEP-0 OAS, trip the RCPs when BOTH RCS pressure is <1425 psig
AND 1FI-917 (CV pump flow) is >100 gpm or SI pump flow is >200 gpm.
C - Plausible: Because 500 gpm is the minimum AF pump flow prior to SG levels being
adequate for a heat sink per 1BwEP-0.
D - Plausible: Because the criteria for adequate ECCS RH pump discharge flow is 1000 gpm
per 1BwEP-0.
Question Development:
During examination outline development, Question 3 was selected as Knowledge and Ability
(K/A) 011 EA1.13, to be a Tier 1 question. Questions formulated from this tier are generally
expected to test an applicants knowledge of how to safely operate the plant during emergency
and abnormal conditions.
Question 3 was developed by Byron Generating Station training staff for use on the 2025 Byron
NRC Initial License Written Examination. The Braidwood exam author took this question from
the Byron exam bank to be used as a bank question on the 2025 Braidwood NRC Initial License
Written Examination. There were no technical or psychometric errors identified by the NRC
Chief Examiners regarding this question on either the Byron or Braidwood examination reviews.
Knowledge and Ability (K/A) Statement: (011 EA1.13) Ability to operate and/or monitor the
following as they apply to (EPE 11) LARGE-Break LOCA (CFR: 41.7 / 45.5 / 45.6): emergency
core cooling system (ECCS); Importance Rating 4.3.
Question Administration:
NUREG-1021, ES-4.3, item B.2.i, identifies that for applicants taking the RO written
examination, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is to be allotted. The RO applicant contesting this question took 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
and 57 minutes to complete the written examination.
- No applicants asked a question or requested clarification for Question 3 during
administration of the Byron and Braidwood written examinations.
- Thirteen (of 16) applicants submitted answer choice B (the keyed answer) on the Braidwood
examination and seven (of eight) applicants submitted answer choice B (the keyed answer)
on the Byron examination as correct during examination administration (83% pass rate).
- One RO applicant, in a single post-examination comment, proposed an answer key
change for Question 3. The applicant contesting the question chose answer C for
Question 3. The remaining three applicants (in total between the two exams) who answered
Question 3 incorrectly all chose answer A.
Applicant Comment:
In order to confirm RCP trip criteria is met at least one SI pump must indicate greater than
200 gpm. At a TOTAL SI pump discharge flow of 200 gpm neither individual SI pump discharge
flow would indicate greater than 200 gpm and the RO would NOT be able to confirm RCP trip
criteria.
Applicant Technical Justification for Comment:
Westinghouse Owners Group ERG Generic Issue report for RCP TRIP/RESTART includes
discussions of plant conditions where RCP trips would be required. The requirements to trip
RCPs for various accident conditions are incorporated into Braidwoods emergency response
procedure network.
From RCP TRIP/RESTART:
The report clarifies further:
In the Optimal Recovery Guidelines (ORGs), the RCPs are not tripped unless this 2-part
criterion is satisfied. It cannot be emphasized too strongly that a fundamental condition
which must be satisfied for RCP trip during an emergency condition is that at least one
high pressure SI pump be in operation and capable of delivering flow to the RCS. If this
fundamental condition is not met, the RCPs should not be tripped regardless of whether or
not the plant parameters indicate that a trip setpoint has been reached. Analysis has shown
that if the SI system is not in operation, the RCPs can be operated to provide core heat
removal.
1BwEP-0, REACTOR TRIP OR SAFETY INJECTION, Step 20 is CHECK IF RCPS
SHOULD BE STOPPED:
This includes the plant-specific conditions that correspond to the RCP Trip Criteria of
RCP TRIP/RESTART. Selected plant parameters reaching critical setpoints is met when RCS
pressure is less than 1425 psig. Successful operation of the safety injection system is met
when at least one pump is injecting from among the four high-pressure-safety-injection pumps:
1CV01PA, 1CV01PB, 1SI01PA, and 1SI01P
- B. 1FT-917, CENT CHG PUMP FLOW
TRANSMITTER, is downstream of where the CV pumps discharges combine therefore the first
bullet point of step 20.b checks 1CV01PA and 1CV01PB with a single meter. 1FT-0918,
SAFETY INJECTION PUMP 1A DSCH FLOW TRANSMITTER, and 1FT-922, SAFETY
INJECTION PUMP 1B DSCH FLOW TRANSMITTER, are each upstream of where the
SI pumps discharges combine therefore the second bullet point of step 20.b checks 1SI01PA
and 1SI01PB with two meters. With 1FI-917 indicating 0 gpm and suspected to be stuck (per
the stem) the only way to confirm successful operation of the high head safety injection system
is using SI pump discharge flow indications, 1FI-918 and/or 1FI-922, on 1PM06J.
The exam question was written with the understanding that GREATER THAN 200 GPM is a
requirement for TOTAL SI pump discharge flow. Technical information has been newly
discovered that reveals this was an erroneous understanding. The requirement is NOT greater
than 200 gpm total SI pump discharge flow, but instead is a check that at least one SI pump is
successfully injecting into the core. The indication of successful operation of an SI pump is
indication of flow on its individual flow meter.
Design analysis BYR99-010/BRW-99-0017-I provides documentation for the bases of setpoints
selected for use in Emergency Operating Procedures. Section 7.17.7 pertains to setpoint S.8
and is included here in its entirety:
A (referenced in the Basis section of 7.17.7) includes the following note about the
SI Pump Discharge Flow meters, 1(2)FI-0918 and 1(2)FI-0922:
There is a division very near zero (between 0 and 200 gpm). It is not labeled. It is very
difficult to distinguish between this mark and zero.
Section 2.5 (referenced in the Basis section of 7.17.7) includes the following statement:
For a number of the flow parameters, the operator is directed to check for flow or
determine if there is flow from the... pumps. The Emergency Response Guidelines (ERGs),
Footnotes, and the associated Background Information (Reference 5.1) define these values
as the minimum flow that indicates injection into the system. Therefore, these values
represent a just-in-range setpoint to verify a minimum flow. The singular approach taken to
arrive at the bases for these minimum verifiable flow setpoints is to assess a certain
percentage-of-span value that would account for the total channel inaccuracy and also any
noise readings. This value would then be conservatively rounded up to the next, or first,
discernible minor division of the main control board indicator for that parameter.
From 1PM06J:
In order to confirm RCP trip criteria (as required in the stem) the operator must observe a flow
greater than 200 gpm on 1FI-918 and/or 1FI-922 because any indication below that value is not
reliable.
Per NUREG-1021 Rev 12 ES-1.2 section B.8 both SI pumps can be assumed to be running
and providing half of total SI pump discharge flow. With a total SI pump discharge flow barely
greater than 200 gpm (answer choice B.) NEITHER 1FI-918 nor 1FI-922 would indicate
greater than 200 gpm therefore RCP trip criteria could NOT be confirmed.
Of the flow rates listed as answer choices 500 gpm is the LOWEST total SI pump
discharge flow at which the RO could confirm RCP trip criteria because both meters would
indicate greater than 200 gpm.
Facility Position on Applicant Comment:
In accordance with NUREG-1021 ES-4.4, Section C.3.c, newly discovered technical information
supports changing the key to accept C., 500, as the correct answer.
NRC Resolution:
Region III evaluated this contention in accordance with NUREG-1021, ES-4.4, paragraph
C.3.c. Specifically, to determine if the following types of errors were adequately identified and
justified for effecting an answer key change.
- Did Question 3 lack necessary information upon which to answer or contain an
unclear stem that confused the applicants?
Neither the applicants contention nor facilitys assessment claimed that Question 3 lacked
any necessary information upon which to base an answer. No applicants asked questions
or requested clarification during exam administration regarding Question 3.
Region III concluded that Question 3 did not possess an unclear or confusing stem.
(SRO) job requirements?
Neither the applicants contention nor facilitys assessment claimed incorrect license level
or lack of job link as a basis for their respective contentions.
Region III concluded that Question 3 did not test the wrong license level and that the
tested subject matter was linked to job requirements. Specifically, K/A 011 EA1.13.
- Did Question 3 contain unintended typographical errors?
Neither the applicants contention nor facilitys assessment claimed unintended
typographical errors as a basis for their respective contentions.
Region III concluded that there were no unintended typographical errors in Question 3.
- Did either the applicant or facility contention introduce newly discovered technical
information which supports an answer key change?
The applicants contention and the facilitys assessment have determined that new
technical information was identified which supports changing the examination answer key
so that answer choice C becomes the only correct answer. Specifically, that with both
Safety Injection (SI) pumps operating at approximately the same discharge conditions,
when their discharge flow instruments come on scale at 200 gpm as seen on 1FI-918
and 1FI-922, their total flow will be greater than 400 gpm; therefore, 500 gpm or answer
choice C would be the lowest value above this minimum total flow value.
Region III disagrees that new technical information has been provided concerning SI flow
following a loss of coolant accident (LOCA) with regards to determining whether reactor
coolant pump (RCP) trip criteria is met. Region III agrees with the applicant with regards to
their assessment of reactor pressure and charging pump flow instrument 1FI-917. In
addition, Region III agrees that to ensure RCP trip criteria is met when performing step 20
of 1BwEP-0, REACTOR TRIP OR SAFETY INJECTION or at any time when performing
this procedure since this criteria is listed on the procedures Operator Action Summary
page, with conditions given in the stem, 1FI-918 or 1FI-922 must be reading greater than
200 gpm. Region III agrees that both SI pumps would receive an automatic start signal on
low reactor pressure and would be operating under similar (not necessarily identical)
conditions. Region III also agrees that the value of greater than 200 gpm that is required on
1FI-918 or 1FI-922 is because it is the lowest readable/reliable flow value for these
instruments. Region III disagrees with the applicant on how the lowest total SI flow to meet
the RCP trip criteria is determined. In the circumstance of 1FI-918 reading slightly above
the 200 gpm line and 1FI-922 reading slightly below 200 gpm which is entirely plausible as
the SI pumps will be operating with similar not necessarily identical discharge conditions,
the RCP trip criteria would be met in conjunction with the other conditions described in the
question stem. Specifically, SI flow as read on 1FI-918 would be reading greater than
200 gpm. 1FI-922, even though there is a small deflection on the meter, is not readable
and therefore, no flow value can be ascertained for this SI pump. For this example the total
SI flow would be that which was read on 1FI-918 (i.e., greater than 200 gpm). In addition, it
is important to note that this question is a Tier 1 question which, NUREG 1021 describes as
a question in which the applicant is tested on their knowledge of operator actions
associated with abnormal and emergency operating procedures. Step 20 of BwEP-0 along
with the Operator Action Summary page for 1BwEP-0 clearly spell out SI flow of greater
than 200 gpm for the SI pumps contribution to the RCP trip criteria during a LOCA.
Administering an exam question in which 500 gpm was the correct answer for this question
would be unsatisfactory per NUREG 1021 as it would require the applicants to recall flow
instrument minimum readable values from memory which is minutia knowledge and would
have an answer which contradicts information provided in the stations emergency
operating procedures.
In summary, Region III disagrees with the applicants contention and in accordance with
NUREG 1021, ES-4.4, paragraph C.3.c, the answer key for the written examination will
remain unchanged with answer choice B as the ONLY correct answer.
Question #4
Unit 1 is in Mode 6.
The reactor head is resting on the reactor vessel.
1BwGP 100-6, REFUELING OUTAGE, is in progress at step F.19, PREPARE to
remove the Rx Vessel Head and place it in the storage area.
Reactor vessel level is at 399'.
Subsequently the following sequence of events occur,
Reactor vessel level begins to drop.
The running RH pump experiences cavitation conditions.
The crew shuts down the running RH pump.
Which of the following abnormal operating procedures has met entry conditions?
- A.
1BwOA PRI-1Property "Contact" (as page type) with input value "A.</br></br>1BwOA PRI-1" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., EXCESSIVE PRIMARY PLANT LEAKAGE
- B.
1BwOA PRI-10Property "Contact" (as page type) with input value "B.</br></br>1BwOA PRI-10" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., LOSS OF RH COOLING
- C.
1BwOA REFUEL-2Property "Contact" (as page type) with input value "C.</br></br>1BwOA REFUEL-2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., REFUELING CAVITY OR SPENT FUEL POOL LEVEL LOSS
- D.
1BwOA S/D-2Property "Contact" (as page type) with input value "D.</br></br>1BwOA S/D-2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., SHUTDOWN LOCA
Answer
B
Answer Explanation
A - Plausible: Because 1BwOA PRI-1 is written for RCS leakage. However, it is only
applicable during Modes 1, 2, 3, and 4.
B - Correct: The stem conditions meet entry conditions of 1BwOA PRI-10 of oscillating
RH pump parameters and dropping reactor vessel level.
C - Plausible: Because 1BwOA REFUEL-2 is written for RCS leakage from the refueling
cavity. However, the refueling cavity would not be filled in the stem conditions.
D - Plausible: Because 1BwOA S/D-2 is written for RCS leakage that occurs during either
Mode 3 after the SI accumulators are isolated or Mode 4.
Question Development:
During examination outline development, Question 4 was selected as Knowledge and Ability
(K/A) 025 AK1.04, to be a Tier 1 question. Questions formulated from this tier are generally
expected to test an applicants knowledge of how to safely operate the plant during emergency
and abnormal conditions.
Question 4 was developed by the facility as a new question for use on the 2025 Braidwood NRC
Initial License Written Examination. There were no technical or psychometric errors identified by
the NRC Chief Examiner regarding this question during his review.
Knowledge and Ability (K/A) Statement: (025 AK1.04) Knowledge of the operational implications
and/or cause and effect relationships of the following concepts as they apply to (APE 25) LOSS
OF RESIDUAL Heat Removal System (CFR: 41.8 / 41.10 / 45.3): Loss of inventory while at
reduced inventory; Importance Rating 4.3.
Question Administration:
NUREG-1021, ES-4.3, item B.2.i, identifies that for applicants taking the SRO written
examination, 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> is to be allotted and for applicants taking the RO written examination,
hours is to be allotted. The SRO applicant contesting this question took 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and
minutes and the RO applicant contesting this question took 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 57 minutes to
complete the written examination.
- Three applicants asked a question or requested clarification for Question 4 during
administration of the written examination. The facility proctors determined that based on the
questions asked or clarifications requested, no additional information was to be provided. In
addition, the applicants contesting the grading of this question did not ask a question or
request clarification for Question 4.
- Seven (of 16) applicants submitted answer choice B (the keyed answer) as correct
during examination administration (44% pass rate).
- Two applicants, in a single post-examination comment, proposed an answer key
change for Question 4. The two applicants contesting the question along with four other
applicants chose answer choice C for Question 4. The remaining three applicants who
answered Question 4 incorrectly chose choice D.
Applicant Comment:
The reactor vessel rests immediately below the refueling cavity and their levels are hydraulically
equivalent; therefore, any changes to reactor vessel level would also be indicated as changes to
the refueling cavity level. With the conditions stated in the stem 1BwOA REFUEL-2 would have
met entry conditions; specifically, a rapid drop in refueling cavity level and annunciator 1-6-C3,
REFUELING CAVITY LEVEL HIGH LOW, in alarm.
Applicant Technical Justification for Comment:
Section C.5 of BwAP 340-1, USE OF PROCEDURES FOR OPERATING DEPARTMENT,
describes the use of _BwGP-100 series procedures with C.5.c giving details on the use of
flowcharts:
The stem states that 1BwGP 100-6, REFUELING OUTAGE, is in progress at step F.19 and that
reactor vessel level is at 399'. Per 1BwGP 100-6T1, 1BwGP 100-6 FLOWCHART, step F.12
would already have been performed:
1BwGP 100-6 step F.12:
Step F.12 of BwOP RC-4 establishes installed RCS level indication. This step restores
1LT-RY048 (by opening 1RY095A, 1RY104A, 1RY104B, 1RY095D, and 1RY095E) and
annunciator 1-6-C3 REFUELING CAVITY LVL HIGH/LOW (by landing the lifted-lead at
TB-A-25 and taking PA20JC-AR CS1 to ON).
From BwOP RC-4:
From BwOP RC-4:
From 20E-1-4031RY25:
Annunciator 1-6-C3, REFUELING CAVITY LVL HIGH/LOW, alarms on a Low level of 424' 2".
1LI-RY048, Refuel Cav Level, is a meter on 1PM06J and indicates from 392' to 426'. Several
steps in BwOP RC-4 check the indication prior to reaching 399' in the reactor vessel.
BwOP RC-4 step F.16.c:
BwOP RC-4 step F.31.b:
There is effectively no distinction between indications of refueling cavity level and reactor vessel
level because the reactor vessel is a physical extension of the refueling cavity. This is
demonstrated by the two BwOP RC-4 steps shown above referring to 1LI-RY048 as Refuel
Cav Level (step F.16.c) and Reactor Vessel Refueling Level (step F.31.b).
The initial conditions stated in the stem, i.e., everything prior to Subsequently, combined
with the above discussion of BwOP RC-4 reveals that among the initial indications that would be
present in the control room are (1) Refuel Cav Level, 1LI-RY048, at 399' and (2) annunciator
1-6-C3, REFUELING CAVITY LVL HIGH/LOW, with SER 1552 REFUELING CAVITY LEVEL
LOW.
1BwOA REFUEL-2 lists the following symptoms or entry conditions:
When the Reactor vessel level begins to drop per the stem, Refueling Cavity Level on
1LI-RY048 will drop because the level indication taps off the bottom of the reactor vessel.
The symptom of Rapid drop in refueling cavity level. would be present on 1LI-RY048, Refuel
Cav Level.
Additionally, step 2.d of BwAR 1-6-C3 would direct the crew to enter 1BwOA REFUEL-2:
Facility Position on Applicant Comment:
In accordance with NUREG-1021 ES-4.4, Section C.3.c, newly discovered technical
information supports changing the key to accept C., 1BwOA REFUEL-2, REFUELING
CAVITY OR SPENT FUEL POOL LEVEL LOSS, as an additional correct answer.
NOTE: A procedure change request (PCRA 4755255-94) has been initiated to clarify
entry conditions for _BwOA REFUEL-2.
NRC Resolution:
Region III evaluated this contention in accordance with NUREG-1021, ES-4.4, paragraph
C.3.c. Specifically, to determine if the following types of errors were adequately identified and
justified for effecting an answer key change.
- Did Question 4 lack necessary information upon which to answer or contain an
unclear stem that confused the applicants?
Neither the applicants contention nor facilitys assessment claimed that Question 4 lacked
any necessary information upon which to base an answer. Three applicants asked
questions or requested clarification during exam administration regarding Question 4. In
each instance, the facility proctors determined that no additional information was needed to
answer the question as written. In review of the questions/clarifications requested, the
Region III Chief Examiner agreed with the facility proctors that no additional information
was required to be provided to the applicants.
Region III concluded that Question 4 did not possess an unclear or confusing stem.
(SRO) job requirements?
Neither the applicants contention nor facilitys assessment claimed incorrect license level
or lack of job link as a basis for their respective contentions.
Region III concluded that Question 4 did not test the wrong license level and that the
tested subject matter was linked to job requirements. Specifically, K/A 025 AK1.04.
- Did Question 4 contain unintended typographical errors?
Neither the applicants contention nor facilitys assessment claimed unintended
typographical errors as a basis for their respective contentions.
Region III concluded that there were no unintended typographical errors in Question 4.
- Did either the applicant or facility contention introduce newly discovered technical
information which supports an answer key change?
The applicants contentions and the facilitys assessment have determined that
new technical information was identified which supports changing the examination
answer key to also accept answer choice C as an additional correct answer. Specifically, it
was noted that based on the information provided in the stem, with the crew at step F.19 of
1BwGP 100-6, REFUELING OUTAGE, the refueling cavity level instrument LT-RY048
would have already been put in service and reading reactor vessel water level which was
previously drained to (399') 1 foot below the reactor vessel flange. As shown in the
technical discussion above, the refueling cavity water level transmitter receives its input
from the same location as the reactor vessel level indicator LT-RY046. Therefore, at this
moment in time with the refueling cavity completely empty as the reactor vessel head is in
place and reactor coolant system level is being maintained 1 foot below the reactor vessel
flange, LT-RY048 would be reading 399' which is reactor vessel level. In addition,
annuciator 1-6-C3, REFUELING CAVITY LVL HIGH/LOW would have been previously
energized and would be in alarm with LT-RY047 also in service and reading 399' which is
below the low level alarm setpoint of 424'2". All of this would be the case before a leak in
the residual heat removal or reactor coolant system occurs causing reactor vessel and as a
result refueling cavity level indication to begin dropping. With all of this in mind, the
applicants and the facility contest that the entry conditions of 1BwOA REFUEL-2,
REFUELING CAVITY OR SPENT FUEL POOL LEVEL LOSS would have also now been
met. Specifically, an unexplained lowering level on the LT-RY048, refueling cavity level
instrument, and 1-6-C3 alarming.
Region III agrees with the technical information provided above and recognizes that during
question development it was not assessed that the refueling cavity level transmitter would
be in service and providing a separate indication of reactor vessel water level. It should be
noted that the symptoms and entry conditions section of 1BwOA REFUEL-2 specifically
states, [T]he following symptoms may [emphasis added] cause entry into this
procedureRapid drop in refueling cavity level and, [T]he following annuciators may
[emphasis added] cause entry into this procedureREFUELING CAVITY LEVEL HIGH
LOW (1-6-C3). The may in both of these statements implies that there is an expectation
that operator judgement will be used to determine if this is truly the correct procedure to
enter based on actual plant conditions. It should also be noted that annunciator 1-6-C3
would have been an expected alarm when its level transmitter was valved in and its
electronic circuit switch was energized even before the level transient occurred. In their
technical discussion, the applicants noted that entering 1BwOA REFUEL-2 was a
subsequent action when responding to 1-6-C3. Since this annunciator would have been
alarming and acknowledged as expected prior to the level transient, it is debatable as to
whether the crew would have been expected to again refer to the associated alarm
response procedure when the leak occurred.
As it has now been conclusively determined that plant conditions listed in the entry
conditions section of 1BwOA REFUEL-2 would have existed at the time of the coolant leak,
and assessing how the question stem was worded, which asked, [W]hich of the following
abnormal operating procedures has met entry conditions?, Region III agrees that the
procedures listed in both answer choices B and C are applicable in refueling mode
(Mode 6) and that they both would have included entry conditions which were indicated at
the time of the leak. Answer choice B which would address the loss of the residual heat
removal pumps reported in the question stem is the most correct answer and is required for
addressing the conditions listed in the stem. Region III has determined that the path to
enter answer choice C exists through plant indications and through the supplemental
actions of the annunctiator response procedure for alarm 1-6-C3. In this instance, these
answers do not conflict with each other in that the question stem simply asks about which
procedures entry conditions are met. Entering into both of these procedures
simultaneously would not provide conflicting guidance to the operators.
In summary, Region III agrees with the applicants contention and in accordance with
NUREG 1021, ES-4.4, paragraph C.3.c, the answer key for the written examination will be
changed to accept both answer choices B and C as correct.
Question #43
Unit 1 is in Mode 3.
1BwGP 100-2, PLANT STARTUP in progress.
FW flow is being controlled in MANUAL with the FW Reg Bypass valves.
Under these conditions, FW should be controlled with a constant SG (1) program.
If FW flow needs to be adjusted due to a steam flow/feed flow mismatch, careful manipulation of
FW Reg Bypass Valves should be exercised, to avoid (2) problems.
A. (1) mass
(2) shrink/swell
B. (1) mass
(2) water hammer
C. (1) level
(2) shrink/swell
D. (1) level
(2) water hammer
Answer Explanation
A - Plausible: Because the RCS is operated at a constant mass program.
B - Plausible: Because the RCS is operated at a constant mass program. Also, 1BwGP 100-2
precaution D.4.c states Opening the 1FW039A-D could result in a water hammer event if there
is a large DP across the valve. However, this is a concern in the FW tempering line, but not the
main FW or bypass lines which are designed for large DPs across the feed reg valves.
C - Correct: Per 1BwGP 100-2, the SGs are to be maintained at or above program level. The
program is based upon a constant level program to compensate for shrink and swell during
power changes and transients. 1BwGP 100-2, precaution D.4.a states Careful manipulation of
FW Valves during Low Load conditions should be exercised, so as not to OVER/UNDER feed
S/Gs, causing increased SHRINK/SWELL problems and possible Reactor Trip.
D - Plausible: Because 1BwGP 100-2 precaution D.4.c states Opening the 1FW039A-D could
result in a water hammer event if there is a large DP across the valve. However, this is a
concern in the FW tempering line, but not the main FW or bypass lines which are designed for
large DPs across the feed reg valves.
Answer
C
Question Development:
During examination outline development, Question 43 was selected as Knowledge and Ability
(K/A) 059 K5.08, to be a Tier 2 question. Questions formulated from this tier are generally
expected to test an applicants knowledge of how plant systems are designed and function.
Question 43 was developed by the facility as a new question for use on the 2025 Braidwood
NRC Initial License Written Examination. There was a grammatical error in the question stem
identified by the NRC Chief Examiner regarding this question during his review. This error was
corrected prior to exam administration.
Knowledge and Ability (K/A) Statement: (059 K5.08) Knowledge of the operational implications
or cause and effect relationships of the following concepts as they apply to the (SF4S MFW)
MAIN FEEDWATER SYSTEM (CFR: 41.5 / 45.7): Reason for matching steam flow and MFW
flow when recovering from an S/G level transient in manual control; Importance Rating 3.9.
Question Administration:
NUREG-1021, ES-4.3, item B.2.i, identifies that for applicants taking the SRO written
examination, 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> is to be allotted. The SRO applicant contesting this question took
hours and 15 minutes to complete the written examination.
- No applicants asked a question or requested clarification for Question 43 during
administration of the written examination.
- Ten (of 16) applicants submitted answer choice C (the keyed answer) as correct
during examination administration (63% pass rate).
- One SRO applicant, in a single post-examination comment, proposed an answer key
change for Question 43. The applicant contesting the question along with two other
applicants chose answer D for Question 43. Two of the remaining three applicants who
answered Question 43 incorrectly chose answer B and one applicant chose answer A.
Applicant Comment:
With the conditions stated in the stem careful manipulation of FW Reg Bypass Valves
should be exercised to avoid water hammer problems.
Applicant Technical Justification for Comment:
INPO Student Guide 193006, FLUID STATICS AND DYNAMICS, defines water hammer as
a physical shock to the components of a fluid system, liquid, or vapor caused by impact of
high-velocity liquids. It lists common causes of water hammer including quickly closing a
valve and quickly opening a valve. It states that for water hammer prevention, operators
should operate manual system valves slowly.
The potential for water hammer exists in any flow path. With the conditions stated in the stem,
feed flow is being controlled through 1FW510A/1FW520A/1FW530A/1FW540A, (Feed Reg
Bypass Valves or FRBVs). While the plant is in MODE 3 during a plant startup the Feed Reg
Valves are closed due to their large piping size which could risk overfilling a SG given the
limited steam demands. This results in flow being controlled through the FRBVs with no parallel
paths for the piping leg on the SG side of the FRBVs.
With the conditions stated in the stem, the steam flow would be relatively small (as a result of
RCP heat input only) therefore the feed flow demands would be relatively low. With the FRBVs
nearly closed, the risks of water hammer are higher. By fully closing the FRBVs, the operators
would risk a water hammer event by stopping fluid from flowing and causing pressure waves
through the system. Additionally, when the FRBVs are fully closed, careful manipulation is
required when opening them to prevent a water hammer event when re-initiating flow in the line.
There have been several instances of water hammer problems in the feedwater lines
during various conditions.
AR 04065155: One SG FRV went closed from full power. The pressure transient (shown
below) is consistent with indications of water hammer.
AR 01265004 (from Byron, same system design): Water hammer occurred when cycling
2FW002A/B (high-pressure feedwater heater isolation valves) when repressurizing lines that
had partially depressurized due to leakby.
SER 52-85 (Significant Event Report from San Onofre 1): Water hammer significant enough to
cause a severe body-bonnet leak in a 4-inch check valve resulted when Aux Feed pumps
started on a reactor trip without feed.
Facility Position on Applicant Comment:
In accordance with NUREG-1021 ES-4.4, Section C.3.c, newly discovered technical information
supports changing the key to accept D., (1) level (2) water hammer, as an additional correct
answer.
NRC Resolution:
Region III evaluated this contention in accordance with NUREG-1021, ES-4.4, paragraph
C.3.c. Specifically, to determine if the following types of errors were adequately identified and
justified for effecting an answer key change.
- Did Question 43 lack necessary information upon which to answer or contain an
unclear stem that confused the applicants?
Neither the applicants contention nor facilitys assessment claimed that Question 43 lacked
any necessary information upon which to base an answer. No applicants asked questions
or requested clarification during exam administration regarding Question 43.
Region III concluded that Question 43 did not possess an unclear or confusing stem.
(SRO) job requirements?
Neither the applicants contention nor facilitys assessment claimed incorrect license level
or lack of job link as a basis for their respective contentions.
Region III concluded that Question 43 did not test the wrong license level and that the
tested subject matter was linked to job requirements. Specifically, K/A 059 K5.08.
- Did Question 43 contain unintended typographical errors?
Neither the applicants contention nor facilitys assessment claimed unintended
typographical errors as a basis for their respective contentions.
Region III concluded that there were no unintended typographical errors in Question 43.
- Did either the applicant or facility contention introduce newly discovered technical
information which supports an answer key change?
The applicants contention and the facilitys assessment have determined that new
technical information was identified which supports changing the examination answer key
to also accept answer choice D as an additional correct answer. Specifically, that valve
operation when restoring flow must be done in a deliberate manner because water hammer
is a possibility in any water system. The additional technical information provided is
associated with the second portion of the question regarding the affects of water hammer
when operating the feedwater system. There is no contention with the first part of the
question concerning level programming.
Region III disagrees that new technical information has been provided concerning water
hammer with regards to the conditions provided in the question stem. Question 43 was
written to test applicant knowledge of feedwater system operations during startup on Unit 1
with 1BwGP 100-2, PLANT STARTUP in progress. The stem informed the applicant that
the plant was in Mode 3 and steam flow/feed flow mismatch and hence steam generator
water level was being controlled manually with the Feedwater Regulating Bypass valves.
The operators are performing step F.26.c of 1BwGP 100-2, which directs them to control
feedwater with the Feedwater Bypass valves in AUTO or MANUAL through the FW009
valves. This would be the normal, preferred lineup for a Unit 1 startup. Precaution D.4.a of
BwGP 100-2 is written to guide operator performance when performing this step and it
specifically states, [C]areful manipulation of FW Valves during Low Load conditions should
be exercised, so as not to OVER/UNDER feed S/Gs, causing increased SHRINK/SWELL
problems and possible Reactor Trip. This precaution statement validates the keyed
answer C as the correct answer for the conditions provided in the question stem.
Regarding the applicants contention that water hammer should also be considered a
correct answer when operating the Feedwater Bypass valves in manual, the design of the
Unit 1 steam generators and feedwater system must be considered. First, the Unit 1 steam
generators are designed to exhibit a lower recirculation ratio; therefore, under low
feedwater flow conditions, as are being experienced in this question, there is a low chance
of water hammer at the feedwater nozzle. With high recirculation flow in the steam
generator and low feedwater flow rates at low powers as would be the case on Unit 2,
water hammer would be a concern. Second, Unit 1 has a different piping configuration with
regards to the feedwater tempering line, which means the FW039 low flow line valves are
not normally required to be opened during startup operations. In fact, the only precaution
concerning water hammer when operating during startup conditions on Unit 1 is associated
with opening of the FW039 valves. Precaution D.4.c states, [O]pening the 1FW039A-D
could result in a waterhammer event if there is a large p across the valve. If the pressure
in the FW header (16" line) is significantly less than the tempering line pressure, raise the
FW header pressure per the applicable steps of BwOP FW-3. It should be also noted the
question stem is specifically asking about precautions associated with operating the
Feedwater Regulating Bypass valves and not the FW039 valves or just the feedwater
system in general. These important differences between Unit 1 and Unit 2 are why Unit 2
has a separate Water Hammer Prevention System which would be in effect at these low
power, low feedwater flow conditions. Unit 1 does not have this extra system because the
configuration and construction differences mentioned above do not require it when
controlling feedwater under low flow conditions with the Feedwater Regulating and
Regulating Bypass valves. If this question had been written on Unit 2, the answer key
would have to consider both answer choices C and D correct, although shrink/swell would
remain the overriding concern when manually operating the Feedwater Regulating Bypass
valves.
In summary, Region III disagrees with the applicants contention and in accordance with
NUREG 1021, ES-4.4, paragraph C.3.c, the answer key for the written examination will
remain unchanged with answer choice C as the ONLY correct answer.
SIMULATION FACILITY FIDELITY REPORT
Facility Licensee:
Braidwood Station
Facility Docket Nos.:
50-456; 50-457
Operating Tests Administered:
December 1, 2025 to December 8, 2025
The following documents observations made by the NRC examination team during the
initial operator license examination. These observations do not constitute audit or inspection
findings and are not, without further verification and review, indicative of non-compliance with
CFR 55.45(b). These observations do not affect NRC certification or approval of the
simulation facility other than to provide information which may be used in future evaluations.
No licensee action is required in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were
observed:
ITEM
DESCRIPTION
Simulator crash during Initial License Training NRC Exam.