ML23192A527

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NRC Initial License Examination Report 05000373/2023301; 05000374/2023301
ML23192A527
Person / Time
Site: LaSalle  Constellation icon.png
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

License Nos. NPF-11; NPF-18

Enclosures:

1. Examination Report

05000373/2023301;

05000374/2023301

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:

Jeffrey Josey

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RidsNrrPMLaSalle Resource

RidsNrrDroIrib Resource

John Giessner

Mohammed Shuaibi

Diana Betancourt-Roldan

Allan Barker

R3-DORS

Ikeda Betts

Colleen Schmidt

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

License Nos: NPF-11; NPF-18

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,

1E-2-4005AR, and 1E-2-4005AK

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