ML23192A527
| ML23192A527 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/12/2023 |
| From: | April Nguyen NRC/RGN-III/DORS |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| Shared Package | |
| ML21188A270 | List: |
| References | |
| 50-373/23-301, 50-374/23-301 50-373/OL-23, 50-374/OL-23 | |
| Download: ML23192A527 (1) | |
See also: IR 05000373/2023301
Text
July 12, 2023
David P. Rhoades
Senior Vice President
Constellation Energy Generation, LLC
President and Chief Nuclear Officer (CNO)
Constellation Nuclear
4300 Winfield Road
Warrenville, IL 60555
SUBJECT: LASALLE COUNTY STATION, UNITS 1 AND 2 NRC INITIAL LICENSE
EXAMINATION REPORT 05000373/2023301; 05000374/2023301
Dear David Rhoades:
On June 1, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed the initial operator
licensing examination process for license applicants employed at your LaSalle County Station.
The enclosed report documents the results of those examinations. Preliminary observations
noted during the examination process were discussed on May 09, 2023, with Scott Arebaugh
and other members of your staff. An exit meeting was conducted by teleconference on
June 12, 2023, between John Van Fleet, Site Vice President, of your staff, and
Travis Iskierka-Boggs, Senior Operator Licensing Examiner, to review the proposed final
grading of the written examination for the license applicants. During the telephone conversation,
the NRC resolutions of three post-examination comments submitted by the facility, initially
received by the NRC on June 1, 2023, were discussed.
The NRC examiners administered an initial license examination operating test during the week
of May 01, 2023. The written examination was administered by LaSalle County Station training
department personnel on May 12, 2023. Five Senior Reactor Operator and eight Reactor
Operator applicants were administered license examinations. The results of the examinations
were finalized on June 26, 2023. Nine applicants passed all sections of their respective
examinations. Four applicants were issued Preliminary Results letters. Four applicants were
issued senior operator licenses and five 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 June 1, 2025.
However, since four applicants received a Preliminary Results letter because of a written
examination grade that is less than 80.0 percent, 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.
D. Rhoades 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 Title 10 of the Code of Federal Regulations, Part 2.390, Public
Inspections, Exemptions, Requests for Withholding.
Sincerely,
Signed by Nguyen, April
on 07/12/23
April M. Nguyen, Chief
Operations Branch
Division of Operating Reactor Safety
Docket Nos. 50-373; 50-374
Enclosures:
1. Examination Report
2. Post-Examination Comments,
Evaluation, and Resolutions
3. Simulator Fidelity Report
cc: Distribution via LISTSERV
J. Nugent, Acting Manager
Operations Training,
LaSalle County Station
D. Rhoades 3
Letter to David Rhoades from April M. Nguyen dated July 12, 2023.
SUBJECT: LASALLE COUNTY STATION, UNITS 1 AND 2 NRC INITIAL LICENSE
EXAMINATION REPORT 05000373/2023301; 05000374/2023301
DISTRIBUTION:
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RidsNrrPMLaSalle Resource
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R3-DORS
ADAMS Accession Number: ML23192A527
OFFICE RIII/DORS/OB/CE RIII/DORS/OB:BC
NAME TIskierka-Boggs:gmp ANguyen
DATE 07/11/2023 07/12/2023
OFFICIAL RECORD COPY
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket Nos: 50-373; 50-374
Report No: 05000373/2023301; 05000374/2023301
Enterprise Identifier L-2023-OLL-0017
Licensee: Constellation Energy Generation, LLC
Facility: LaSalle County Station, Units 1 and 2
Location: Marseilles, IL
Dates: May 01, 2023, to June 01, 2023
Examiners: T. Iskierka-Boggs, Senior Operations Engineer, Examiner
G. Roach, Senior Operations Engineer, Examiner
B. Bartlett, Senior Operations Engineer, Examiner
J. Nance, Operations Engineer, Examiner
Approved by: April Nguyen, Chief
Operations Branch
Division of Operating Reactor Safety
Enclosure 1
SUMMARY OF FINDINGS
ER 05000373/2023301 (DORS); 05000374/2023301 (DORS); 05/01/2023-06/26/2023;
Constellation Energy Generation, LLC; LaSalle County Station, Units 1 and 2; Initial License
Examination Report.
The announced initial operator licensing examination was conducted by regional U.S. Nuclear
Regulatory Commission (NRC) examiners in accordance with the guidance of NUREG-1021,
Operator Licensing Examination Standards for Power Reactors, Revision 12.
Examination Summary:
Nine of thirteen applicants passed all sections of their respective examinations. Four applicants
were issued senior operator licenses and five applicants were issued operator licenses. Four
applicants were issued Preliminary Results letters for failure of one section of the administered
examination. (Section 4OA5.1).
2
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 March 27, 2023, with the assistance of members of the
facility licensees staff. During the on-site validation week, the examiners audited two
reactor operators and two senior reactor operators license applications for accuracy. The
NRC examiners, with the assistance of members of the facility licensees staff,
administered the operating test, consisting of job performance measures (JPMs) and
dynamic simulator scenarios, during the period of May 01, 2023, through May 05, 2023.
The facility licensee administered the written examination on May 12, 2023.
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 percent of the proposed examination questions were determined to be
unsatisfactory and required modification or replacement.
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 June 1, 2025, (ADAMS Accession Numbers
ML21188A269, ML21188A265, ML21188A267, ML21188A274 respectively).
On June 1, 2023, 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 to this report.
The NRC examiners graded the written examination on June 26, 2023, and conducted a
review of each missed question to determine the accuracy and validity of the
examination questions.
3
(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 percent of the proposed operating test portion of the examination was
determined to be unsatisfactory and required modification or replacement.
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 NRC's ADAMS
on June 1, 2025, (ADAMS Accession Numbers ML21188A269, ML21188A265,
ML21188A267, ML21188A274 respectively).
The NRC examiners completed operating test grading on June 26, 2023.
(3) Examination Results
Five applicants at the Senior Reactor Operator (SRO) level and eight applicants at the
Reactor Operator (RO) level were administered written examinations and operating
tests. Nine applicants passed all portions of their examinations and were issued their
respective operating licenses on June 26, 2023. Four applicants were issued Preliminary
Results letters for failure of the written examination overall section of the administered
examination.
.2 Examination Security
a. Scope
The NRC examiners reviewed and observed the licensee's 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
4
4OA6 Meetings
.1 Debrief
The chief examiner presented the examination team's preliminary observations and
findings on May 09, 2023, to Scott Arebaugh, Manager Operations Training, and other
members of the LaSalle County Station Operations and Training Department staff.
.2 Exit Meeting
The chief examiner conducted an exit meeting on June 12, 2023, with John Van Fleet,
Site Vice President, by teleconference. The NRCs final disposition of the stations
grading of the written examination and post-examination comments were disclosed and
discussed during the telephone. 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
5
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
J. Van Fleet, Site Vice President
C. Smith, Plant Manager
D. Blair (Acting), Director Site Operations
J. Nugent (Acting), Manager Operations Training
D. Mearhoff, Regulatory Assurance Manager
J. Messina, Shift Operations Superintendent
K. Heuser, Operations Training, Initial License Training Lead
J. Glass, Initial License Training, Examination Author
U.S. Nuclear Regulatory Commission
T. Iskierka-Boggs, Chief Examiner
G. Roach, Senior Examiner
B. Bartlett, Senior Examiner
J. Nance, Examiner
J. Benjamin, Senior Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
None
LIST OF ACRONYMS USED
ADAMS Agencywide Document Access and Management System
NRC U.S. Nuclear Regulatory Commission
Attachment
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
RO Question 2
Unit 2 is in MODE 4 with all loads on Unit 2 SAT.
Unit 1 is at 100 percent RTP when a transient occurs.
Unit 1 loss of Auxiliary Power (UAT and SAT)
Unit 1 steam line breaks causing Drywell pressure to rise to 5 psig
Unit 1 power is cross-tied from Unit 2 SAT to 141Y and 142Y
What will the effect be if the 2412 breaker (SAT feed to 241Y) trips open?
Common DG Output Breaker __(1)__ will close and Unit Tie Breakers will __(2)__.
A. (1) 1413, DG feed to 141Y
(2) open
B. (1) 2413, DG feed to 241Y
(2) open
C. (1) 1413, DG feed to 141Y
(2) remain closed
D. (1) 2413, DG feed to 241Y
(2) remain closed
Answer: A
References provided to NRC:
LOA-AP-101 UNIT 1, AC POWER SYSTEM ABNORMAL rev 62
LOS-DG-109 UNIT 1, INTEGRATED DIVISION I RESPONSE TIME SURVEILLANCE
rev 29
Electrical Schematics 1E-0-4412AA, 1E-1-4005AK, 1E-1-4005AK, 1E-1-4005AR,
Questions asked during administration of the NRC written examination
Applicant Comment:
Answers A and C are both correct for this question. The second part of the answer asks if the
Unit Tie Breakers will open or remain closed. Unit Tie Breakers should have a designator to
specify which divisions unit tie breakers the question is referring to. In this questions division 1
unit tie breakers will open. However, division 2 unit tie breakers will remain closed.
Enclosure 2
Facility Position on Applicant Comment:
This question was originally keyed as B. However, during the exam it was discovered that it was
a typo and the correct answer was A.
The intent of the question was that the candidate was to answer the question with respect to the
division 1 unit breakers since that was the division with an undervoltage. However, if the
candidate assumed the question was asking about the division 2 unit tie breakers, those
breakers would remain closed since there was no undervoltage sensed on 142Y/242Y.
The question stem does not specify which unit tie breaker division is being referred to.
Since the question stem can be read as referring to either unit tie breaker division, both answer
choices A and C are correct, and do not conflict with one another.
The first part of the question has only 1 correct answer. The common DG output breaker 1413
DG feed to 141Y will remain closed since unit 1 has an undervoltage and a Loss of Coolant
Accident (LOCA) signal.
The facility recommendation is that both A and C are correct answers.
NRC Evaluation/Resolution:
No questions were asked pertaining to this question during the administration of the NRC
written examination.
The NRC staff reviewed the provided electrical schematics provided as a reference;
LOA-AP-101 UNIT 1, AC POWER SYSTEM ABNORMAL rev 62; and LOS-DG-109 UNIT 1,
INTEGRATED DIVISION I RESPONSE TIME SURVEILLANCE rev 29. The NRC staff also
reviewed the additional information provided regarding system operation and response
expected during the conditions experienced in the stem of the question.
In RO Question 2, the question stem states that the Unit 2 SAT feed breakers for division 1
(ACB 2412) and division 2 (2422) are closed by the statement, Unit 2 is in MODE 4 with all
loads on the Unit 2 SAT. The question stem further states that the Unit 1 SAT feed breakers for
division 1 (ACB 1412) and division 2 (ACB 1422) are open by the statement, Unit 1 is cross tied
from Unit 2 SAT to 141Y and 142Y. The first part of the question asks, What will the effect be if
the 2412 breaker (SAT feed to 241Y) trips open? When the 2412 breaker trips open, both the
SAT feed breakers for division 1 (ACB 1412 and ACB 2412) are open causing division 1 unit tie
breakers (ACB 1414 and ACB 2414) to trip open. Since division 2 still has breaker 2422
(SAT feed to 242Y) closed, the division 2 unit tie breakers (ACB 1424 and ACB 2424) will
remain closed.
As pointed out through the candidates comments to the NRC, the second part of the question
states, Common DG Output Breaker __(1)__ will close and Unit Tie Breakers will __(2)__, and
does not indicate which division unit tie breakers are referred to in this question. Facility
feedback stated that the intent was to question the applicants on division 1 unit tie breakers
which would make the keyed answer correct. However, as the question does not specify the
division, this does not satisfy the facilitys intent of the question.
2
NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 12,
ES-1.2, Guidelines for Taking NRC Examinations, Section B.8 states, in part, 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, Therefore, the applicants were instructed to not make assumptions when answering
written examination questions.
After reviewing the system response using Piping and Instrumentation Diagrams provided by
the facility, it has been determined that, depending on which division of unit tie breakers are
assumed for this question, two separate and different answers for the division 1 and division 2
unit tie breakers will occur. The verbiage of the question as written asks the candidates to
answer the question pertaining to Unit Tie Breakers and does not specify which division.
Considering this fact, when answering the question to the Unit Tie Breakers, one must answer
the question, as written, and determine a response for all the Unit Tie Breakers since none are
specified.
As the question is written, candidates are required to not make assumptions pertaining to the
question, nor are to assume information is added or detracted from the stem of the question,
and as such must answer the question for the Unit Tie Breakers. Division 1 Unit Tie Breakers
will open in the given conditions of the question, and Division 2 Unit Tie Breakers will remain
closed with the given conditions of the question. There are no answer options that have open
AND remain closed as part of that option.
NUREG-1021, ES-4.4, Section C.3.f, states, If a question is determined to have three or more
correct answers or there is no correct answer, the NRC will delete the question. Therefore, the
NRC staff has concluded that there are no correct answers for this question, and the question
will be deleted from the written examination.
3
RO Question 13
Unit 1 is at 100 percent RTP when the following occurs.
An SRV spuriously opens
Suppression Pool temperature is rising at 1°F/minute
The minimum Limiting Conditions Operation Suppression Pool temperature above which a
scram is required is __(1)__, and the purpose for limiting temperature is to __(2)__?
A. (1) 105°F
(2) ensure all the steam released through the downcomer lines during a LOCA is
quenched
B. (1) 110°F
(2) ensure all the steam released through the downcomer lines during a LOCA is
quenched
C. (1) 105°F
(2) prevent peak primary containment pressures and temperatures exceeding maximum
allowable values during a postulated DBA
D. (1) 110°F
(2) prevent peak primary containment pressures and temperatures exceeding maximum
allowable values during a postulated DBA
Answer: D
References provided to NRC:
Technical Specification (TS) (LCO) 3.6.2.1, Suppression Pool Average Temperature
TS Basis Document for TS LCO 3.6.2.1
TS LCO 3.6.2.2, Suppression Pool Water Level
TS Basis Document for TS LCO 3.6.2.2
Questions asked during administration of the NRC written examination
Facility Comments and Discussion:
Answers B and D are both correct answers. According to the 3.6.2.1 Tech Spec basis, both
complete steam condensation and primary containment peak pressure and temperature led to
the development of the suppression pool average temperature limit. The basis also states that
the suppression pool must quench all the steam released through the downcomer lines during
a loss of coolant accident (LOCA). The LCO statement states that the requirement to have the
reactor shutdown when the average pool temperature reaches 110F is to ensure that the pool
can absorb the steam without exceeding any maximum allowable values during a postulated
DBA or transient. These statements make B a correct answer.
4
NRC Evaluation/Resolution:
No questions were asked pertaining to this question during the administration of the NRC
written examination.
The facility did not provide a contention for the first part of the question. Therefore, the first part
of the question in which minimum LCO Suppression Pool temperature above which a scram is
required of 110°F is correct as originally keyed.
Regarding the second part of the question, the NRC staff reviewed the additional information
provided in TS LCO Basis Document 3.6.2.1, Suppression Pool Average Temperature. The
background section of the TS basis document states, in part, The suppression pool is designed
to absorb the decay heat and sensible heat released during a reactor blowdown from
safety/relief valve discharges or from a loss of coolant accident (LOCA) Additionally, the
background states, in part, The suppression pool must quench all the steam released through
the downcomer lines during a loss of coolant accident (LOCA). This is the essential mitigative
feature of a peak pressure suppression containment that ensures that the peak containment
pressure is maintained below the design value The background section further states,
The technical concerns that lead to the development of suppression pool average temperature
limits are as follows:
a) Complete steam condensation;
b) Primary containment peak pressure and temperature;
c) Condensation oscillation (CO) loads; and
d) Chugging loads.
TS LCO 3.6.2.1, ACTION C states that if the CONDITION of Suppression Pool Average
Temperature is > 110°F but 120°F, that the REQUIRED ACTION of C.1 is to Place the
reactor mode switch in the shutdown position, Immediately. The LCO requirement is Average
temperature 110°F with THERMAL POWER 1% Rated Thermal Power. According to the
basis document, this requirement ensure that the plant will be shut down at > 110°F. The pool is
designed to absorb decay heat and sensible heat but could be heated beyond design limits by
the steam generated if the reactor is not shut down.
After a review of the additional provided documents, the NRC staff have determined that the
Suppression Pool is designed to ensure all the steam released through the downcomer lines
during a LOCA is quenched, in addition to, preventing peak primary containment pressures and
temperatures from exceeding maximum allowable values during a postulated design basis
accident as indicated by the Technical Specification LCO and Background documents.
Additionally, not brought up as a contention from the facility, the NRC staff evaluated the
question from the perspective of license level knowledge to determine if the question meets
Reactor Operator (RO) level of knowledge. The question pedigree and references provided that
support the facilitys position that answer options B and D are both correct are both based on
the information in the facilitys Technical Specification Basis documents. NUREG-1021, ES-4.2,
Figure 4.2-2, Screening for Senior Reactor Operator (SRO)-Only Questions Linked to Title 10 of
the Code of Federal Regulations 55.43(b)(2)(TS), provides a flow chart that may be used to
assess whether a question is RO or SRO level of knowledge. As indicated in the graphics
below, this question screened to be an RO level of knowledge question based on the
information presented from the Technical Specification Basis document is being asked about
the information contained above the line in TS LCO 3.6.2.1.
5
Therefore, the NRC staff concluded that both answer options B and D are correct based on
newly discovered technical information that supports a change to the answer key.
6
RO Question 66
A Reactor startup is in progress on Unit 1.
MODE Switch is in STARTUP
NO control rods have been withdrawn
Initial SRM count rates are:
SRM A 30
SRM B 35
SRM C 50
SRM D 45
When is single notch withdrawal of control rods required per LGP-1-1, NORMAL UNIT
STARTUP?
A. RWM is Inoperable
B. SRM C count rate of 200
C. Group 1 control rods all at position 48
D. Criticality NOT achieved prior to the ECP estimated critical position
Answer: C
References provided to NRC:
LGP-1-1, NORMAL UNIT STARTUP
LOP-RM-01, ROD CONTROL MANAGEMENT SYSTEM
Control Rod Move Sheet, Sequence ID: SAr1.0, Unit 2 Cycle 20 L2C20 Startup
Sequence, Effective Date 03/05/2023
Questions asked during administration of the NRC written examination
Applicant Comment 1:
Answers C and D are both correct. The answer key states that D is not correct because it is a
requirement to stop control rod withdrawal completely per LGP-1- step E.2.4. However, answer
D does not match what is in LGP-1-1 and does NOT meet the requirements to stop control rod
withdrawal.
Per LGP-1-1, the operator would continue to withdraw control rods (single notch withdrawal) to
the ECP and up to +1% band AFTER the ECP in order to achieve criticality. The +1% band
contains multiple control rods and single notch withdrawal is required.
Answer D states criticality NOT achieved prior to the ECP. This condition describes a step in
the approved rod sequence where all steps prior to the ECP have been completed and criticality
has not been achieved. The next rod move is the estimated critical position, and is required to
be single notched.
7
Applicant Comment 2:
Both answers C and D are correct and are supported by procedure. The answer key cites step
E.2.4 from LGP-1-1 to justify why D is an incorrect answer. LGP-1-1, step E.2.4 states: IF the
reactor has NOT been declared critical ON or before a current ECP (including a +1% band),
then as necessary STOP all control rod withdrawals. In simpler wording, step E.2.4 means if
youve reached the ECP +1% without having gone critical, then stop withdrawing rods.
Answer D states: Criticality NOT achieved prior to the ECP.
Answer D does NOT say ON or before current ECP (including +1% band), it ONLY states prior
to which means that the ECP has not been reached and, under these conditions, notch control
will REQUIRE single notch withdrawal per the guidance of LGP-1-1, step E.2.3.
Facility Position on Applicant Comment:
Single notch withdrawal is required at the ECP estimated critical position as well as +1% ECP
band in accordance with LGP-1-1 step E.2.3 and the approved control rod sequence package.
The control rod sequence states Do not use continuous withdrawal between 00 and 36 unless
allowed by LGP-1-1. LGP-1-1 does not allow continuous withdrawal at this time since the first
rod group is pulled and the generator is offline with all bypass valves closed.
For operational validity, an approved control rod sequence was provided from the last unit 2
outage startup. The estimated ECP was on control rod 14-43 and criticality was achieved after
the ECP on control rod 46-43. All the rods moved on this page were single notched between
00 and 36 per the note on the approved rod sequence which states, Do not use continuous
withdrawal between 00 and 36 unless allowed by LGP-1-1.
The facility recommends both C and D are correct answers.
NRC Evaluation/Resolution:
No questions were asked pertaining to this question during the administration of the NRC
written examination.
The NRC staff reviewed LGP-1-1, NORMAL UNIT STARTUP, and LGP-RM-01, ROD
CONTROL MANAGEMENT SYSTEM. Specifically, steps E.2.3 and E.2.4 of LGP-1-1 were
reviewed. Based on the information provided in step E.2.3, an operating crew would be
expected to DISCONTINUE Notch Out Override between positions 00 and 36 after the given
condition at the beginning of the step is satisfied. This condition states, When the first rod
group has been pulled to position 48, or Source Range Monitor (SRM) count rate on any
operable SRM reaches 8 times the highest initial SRM count rate. The crew would be expected
to discontinue notch out override between positions 00 and 36 until: at least one bypass valve is
open; the generator is online; or specified otherwise by a Qualified Nuclear Engineer (QNE). As
indicated, this makes answer choice C a correct answer.
8
The NRC staff reviewed the provided Control Rod Move Sheet from the most recent Unit 2
startup. On this form, Step 51 for Control Rod ID 14-43, was identified as the ECP for the
startup. All Group 1 rods were moved from position 00 to 48 using steps 1 through 42 thus
demonstrating through recent operating experience that the Group 1 rods would be at position
48 prior to the estimated critical position. Further demonstrating that answer choice C is
correct.
The contentions provided by both candidates and the facility state that LGP-1-1, Step E.2.4,
requires that the reactor has not been declared critical on or before a current ECP including the
+1% band. They further assert that answer choice D states, Criticality NOT achieved prior to
ECP estimated critical position, and that answer choice D does not include the +1% range as
the wording states that criticality is not reached prior to the ECP. The candidates further assert
that single notch withdrawal is required up to the +1% prior to the requirement to stop for QNE
assessment in step E.2.4 of LGP-1-1.
A review of the provided Unit 2 Startup Control Rod Move Sheet demonstrated that single notch
withdrawal of control rods is required following all of Group 2 rods being at position 48 which
occurred at Step 20 of the Control Rod Move Sheet. This occurs prior to withdrawing control
rods close to the estimated critical position as contested by the candidates. Step E.2.4 provides
direction to STOP all control rod withdrawals when requested by QNE or based on criticality
not being reached on or prior to a current ECP. There is no guidance in step E.2.4 stating that
single notch withdrawal is required as asserted by the candidates. The requirement to perform
single notch withdrawals is directed from Step E.2.3 due to Group 2 rods being at position 48 on
Step 20 of the Control Rod Move Sheet.
Therefore, the NRC staff concluded that the original keyed answer choice of C is the only
correct answer for this question.
9
SIMULATION FIDELITY REPORT
Facility Licensee: LaSalle County Station
Facility Docket Nos: 50-373; 50-374
Operating Tests Administered: May 01, 2023, through May 05, 2023
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
Title 10 of the Code of Federal Regulations, 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
None N/A
Enclosure 3