IR 05000373/2016301

LaSalle County Station, Units 1 and 2 - NRC Initial License Examination Report 05000373/2016301; 05000374/2016301 (Jls)

ML16357A679
Person / Time
Site:	LaSalle
Issue date:	12/22/2016
From:	Orlikowski R J Operations Branch III
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
IR 2016301
Download: ML16357A679 (34)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE RD. SUIT E 210 LISLE, IL 60532

-4352 December 22, 2016 Mr. Bryan Senior VP, Exelon Generation Company, LLC President and CNO, Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT: LASALLE COUNTY STATION, UNITS 1 AND 2

- NRC INITIAL LICENSE EXAMINATION REPORT 05000373/2016301; 05000374/2016301

Dear Mr. Hanson:

On November 16, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed the initial operator licensing examination process for license applicants employed at your LaSalle County Station, Units 1 and 2. The enclosed report documents the results of those examination s. Preliminary observations noted during the examination process were discussed on November 4, 2016, with Mr. W. Trafton and other members of your staff

. An exit meeting was conducted by telephone on December 1, 2016

, between Mr

. Trafton of your staff, and Mr. Zoia, Chief Operator Licensing Examiner, to review the proposed final grading of the written examination for the license applicants.

During the telephone conversation, NRC resolution s of the station

's post examination comments, initially received by the NRC on November 16, 2016

, were discussed.

The NRC examiners administered an initial license examination operating test during the wee ks of October 17, October 24, and October 31, 2016. The written examination was administered by training department personnel on November 4, 2016

. Thirteen Senior Reactor Operator and eleven Reactor Operator applicants were administered license examinations. The results of the examinations were finalized on December 13, 2016

. Three applicants failed one or more sections of the administered examination and were issued proposed license denial letters. Twenty-one applicants passed all sections of their respective examinations and twelve were issued senior operator licenses and seven were issued operator licenses. In accordance with NRC policy, the licens es for the remaining two applicants are being withheld pending the outcome of any written examination appeal that may be initiated.

The written examination and other related written examination documentation will be withheld from public disclosure for 24 months per your request

. However, if an applicant received a proposed license denial letter

, because of a written examination grade that is less than 80.0 percent, the applicant will be provided a copy of the written examination. For examination security purposes, your staff should consider that written examination uncontrolled and exposed to the public. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC's Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/readi ng-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/

Robert J. Orlikowski, Chief Operations Branch Division of Reactor Safety Docket No s. 05000373; 05000374 License No s. NPF-11; NPF-18 Enclosure s: 1. OL Examination Report 05000373/2016301; 05000374/2016301 2. Simulation Facility Fidelity Report cc: Distribution via LISTSERV J. Lindsey, Training Director

, LaSalle County Station

Enclosure 1 U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket No s: 05000373; 05000374 License N os: NPF-11; NPF-18 Report No:

05000373/2016301; 05000 374/2016301 Licensee: Exelon Generation Company, LLC Facility: LaSalle County Station, Units 1 and 2 Location:

Marseilles, IL Dates: October 17

- November 16, 2016 Inspectors:

C. Zoia, Chief Examiner J. Seymour, Examiner/

Chief Examiner

-in-Training M. Bielby, Examiner Approved by:

R. Orlikowski, Chief Operations Branch Division of Reactor Safety 2

SUMMARY

Examination Report 05000373/2016301; 05000374/2016301; 10/17/2016

- 11/16/2016; Exelon Generation Company, LLC; LaSalle County Station

Units 1 and 2; Initial License Examination Report.

The announced initial operator licensing examination was conducted by regional Nuclear Regulatory Commission examiners in accordance with the guidance of NUREG

-1021, "Operator Licensing Examination Standards for Power Reactors,

" Revision 10, Supplement 1

. Examination Summary Twenty-one of twenty-four applicants passed all sections of their respective examinations

. Twelve applicants were issued senior operator licenses and seven applicants were issued operator licenses

. Three applicants failed one or more sections of the administered examination and were issued proposed license denials. The licenses for the remainin g two applicants are being held and may be issued pending the outcome of any written examination appeal. (Section 4OA5.1).

3

REPORT DETAILS

4OA5 Other Activities

.1 Initial Licensing Examinations

a. Examination Scope

The U.S. Nuclear Regulatory Commission (

NRC) examiners and members of the facility licensee's staff used the guidance prescribed in NUREG

-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 10, to develop, validate, administer, and grade the written examination and operating test. Members of the facility licensee's staff prepared the outline and developed the written examination and operating test. The NRC examiners validated the proposed examination during the week of September 19, 2016, with the assistance of members of the facility licensee's staff. During the on

-site validation week, the examiners audited three license applicatio ns for accuracy. The NRC examiners, with the assistance of members of the facility licensee's staff, administered the operating test, consisting of job performance measures (JPMs) and dynamic simulator scenarios, during the period o f October 17, 2016, through November 3, 2016. The facility licensee administered the written examination on November 4, 2016

.

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 ES-401-9, "Written Examination Review Worksheet." On November 16, 2016

, the licensee submitted documentation noting that there were eight post-examination comment s 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 included with this report. The Form ES 9, the written examination outlines (ES 2 and ES-401-3), 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 NRC's Agencywide Documents Access and Management System (ADAMS) in 24 months.

(ADAMS Accession Numbers ML15274A405 and ML15274A403

). The NRC examiners graded the written examination on November 29, 2016, and conducted a review of each missed question to determine the accuracy and validity of the examination questions.

4 (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.

Changes made to the operating test

, documented in a document titled, "Operating Test Comments,

" as well as the final as administered dynamic simulator scenarios and JPMs, are available electronically in the NRC Public Document Room or from the Publicly Available Records component of NRC's ADAMS. The NRC examiners completed operating test grading on December 13, 2016.

(3) Examination Results Thirteen applicants at the Senior Reactor Operator level and eleven applicants at the Reactor Operator level were administered written examinations and operating tests. Nineteen applicants passed all portions of their examinations and were issued their respective operating licenses on December 13, 2016. Three applicant s failed one or more sections of the administered examination and were issued proposed license denial s. Two applicant s passed all portions of the license examination, but received a written test grade below 82 percent. In accordance with NRC policy, the applicants

' licenses will be withheld until any written examination appeal possibilities by other applicants have been resolved. If the applicant

's grade is still equal to or greater than 80 percent after any appeal resolution, the applicant will be issued an operating license. If the applicant

's grade has declined below 80 percent, the applicant will be issued a proposed license denial letter and offered the opportunity to appeal any questions the applicant feels were graded incorrectly.

.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 Title10 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 licensee

's examination security activities.

b. Findings

No findings were identified.

4OA6 Management Meetings

.1 Debrief The chief examiner presented the examination team's preliminary observations and findings on November 4, 2016

, to W. Trafton, Site Vice

-President

, and other members of the LaSalle County Station Operations and Training Department staff.

5

.2 Exit Meeting

The chief examiner conducted an exit meeting on December 1, 2016, with Mr. W. Trafton, Site Vice President

, by telephone. The NRC

's final disposition of the station's post-examination comments were disclosed and discussed with Mr. Trafton during the telephone discussion. The examiners asked the licensee whether any of the 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 CONTAC

T Licensee

W. Trafton, Site Vice President

H. Vinyard, Plant Manager

G. Ford, Regulatory Assurance Manager

T. Lanc, Regulatory Assurance

J. Keenan, Operations Director

M. Smith, Operations Shift Manager

J. Lindsey, Training Director

D. Wright, Operations Training Manager

D. Fuson, Training Specialist

C. Betken, Operations Instructor

J. Fiesel, Maintenance Director

M. Fakhreddine, Chemistry

B. Roy, Fleet Assessment

U.S. Nuclear Regulatory

Commission

R. Ruiz, Senior Resident Inspector

C. Hunt, Acting Resident Inspector

C. Zoia, Chief Examiner

J. Seymour, Examiner

/ Chief Examiner

-in-Training

M. Bielby, Examiner

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened, Closed, and Discussed

None LIST OF ACRONYMS USED

ADAMS Agencywide Document Access and Management System

JPM Job Performance Measure

LPRM Local Power Range Monitor

NRC U.S. Nuclear Regulatory Commission

PARS Publicly Available Records System

RCMS Rod Control Management System

SWR Simulator Work Request

SIMULATION FACILITY

FIDELITY REPORT Facility Licensee:

LaSalle County Station, Units 1 and 2

Facility Docket No

s: 50-373; 50-374 Operating Tests Administered:

October 17, 2016

, - November 3, 2016

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 no

n-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

Exelon Nuclear Issue #02733595 While performing a simulator reset during NRC exam simulator scenarios, a simulator operator

failed to correctly position an LPRM bypass switch according to the lineup that was required for the scenario guide. This human performance error was not detected by the simulator

operator due to the use of an override function in the simulator software that permitted the simulator to be reset for the next scenario with the switch remaining out of position. As a result, a malfunction inserted during the subsequent scenario did not present itself to the applicants.

The NRC examiners removed the crew of applicants from the simulator and sequestered them while facility staff investigated and corrected the

issue. Following a delay of approximately 45 minutes, the applicants were returned to the simulator and the scenario

was resumed. The scenario was then completed without further simulator issues. SWR #0132672

During a JPM that required synchronizing an emergency diesel generator to a bus, there was one instance in which the associated

output breaker failed to close. The applicant was removed from the simulator and sequestered by an NRC examiner during facility

investigation of the issue. Facility simulator staff were unable to reproduce the issue. Following a brief delay, the applicant was returned to the simulator and the JPM was resumed. The JPM was then completed without further simulator issues.

SWR #0132673

During a JPM that required synchronizing an emergency diesel generator to a bus, there was one instance in which the associated sync selector switch failed to turn on the synchroscope. The applicant was removed from the simulator and sequestered by an NRC examiner during facility investigation of the issue. Facility simulator staff were unable to reproduce the issue. Following a brief delay, the applicant was returned to the simulator and

the JPM was resumed. The JPM was then completed without further simulator issues.

SWR #0132673

During the operating portion of the NRC exam, there were multiple instances of the 1H13

-P601 silence and test buttons sticking. This resulted in applicants

being unable to silence annunciators from this location and necessitated the silencing of annunciators from another location in the simulator.

N/A During multiple simulator scenarios, a persistent RCMS-related annunciator was present. This was not an expected alarm for the conditions established by the scenario guide

s. The presence of this alarm did not interfere with the execution of the simulator scenarios or cause distraction to the applicants. Due to the nature of the underlying simulator issue

, and the potential impacts of repairs on simulator availability, facility simulator staff deferred repair of the issue until after the operating portion of the NRC exam had been completed.

POST EXAM COMMENTS AND RESOLUTIONS

RO Question 18

(Post-Exam Comment #1)

Original Question:

Unit 1 is at 100% power. The NSO starts 1A CD/CB pump and secures 1B CD/CB pump. One minute later, Annunciator

1H13-P601-F402, MSL A/B Radiation

Monitor HI alarms. (1) Is the MSL A/B Radiation

Monitor High

alarm expected or unexpected?

(2) What is the correct operator action, if any? A. (1) Unexpected.

(2) Commence power reduction

per LGP 3-1. B. (1) Unexpected.

(2) Direct all nonessential personal

to stay clear of Turbine Building Elevation

768. C. (1) Expected.

(2) No additional action required.

Monitor parameters

and trends; annunciator

1N62-P600-B502, OFF GAS PRE-TREATMENT

RADIATION

HI may alarm. D. (1) Expected.

(2) No additional action

required.

Monitor parameters

and trends; annunciator

1N62-P600-B304, STATION VENT STACK RAD HI may alarm. Answer: C Applicant

Feedback:

Answers A and C are both correct and supported by procedures

and engineering changes. Additionally, per Exelon standards per OP

-AA-103-102, Section 4.5.5, alarms that are not previously flagged with LOR's reviewed are not expected which eliminate answers (C) and (D). Facility Response:

A review of OP

-AA-103-102 does state that if an alarm is not previously discussed then the alarm should be considered unexpected.

Also a recent plant modification per EC 364641

to relocate the hydrogen injection

points in the condensate header has been successful at reducing the frequency of the MSL Rad Monitor Hl alarms following CD/CB pump lineup changes. The question provides the

candidate with having to determine if the alarm was an expected alarm based on the definition from OP

-AA-103-102 without clarifying information in the stem. Without this information and the engineering change reducing the frequency of the alarm then the correct answer is no longer valid since the

1st part of the two parts states EXPECTED.

Since both C and D state EXPECTED, they can be considered incorrect per procedural process.

In addition

, the student could make the assumption that the alarm was caused by a fuel failure, due to the engineering change, and therefore the alarm would be unexpected and require a power reduction as stated in answer

A.

STATION RECOMMENDATION:

There is NO correct answer

References:

OP-AA-103-102, Watch

-Standing Practices, Revision 16

(reference withheld from public disclosure due to proprietary content)

EC 364639, Mitigate MSL Rad Monitor Spikes

, Revision 2

This reference states, in part, that:

LaSalle Station, in the mid 1990's, installed plant modifications to inject hydrogen into the reactor feedwater, the Hydrogen Water Chemistry (HW) system.

The purpose of the

modifications was to protect the Reactor Internals and Reactor Recirculation Piping by

reduction

of Intergranular Stress Corrosion Cracking (IGSCC).

This modification (EC 364639) will relocate the current point of hydrogen injection into line

1CD07A-30" suction piping, from between 1CBO1

PB (1B) and 1CBO1PC (1C) Condensate

Booster (CB) pumps to a point in line 1CD07A

-30" upstream of all four Condensate Booster

pumps. This relocation of hydrogen injection will reduce the MSL radiation spikes by

providing

a more uniform mixing of hydrogen in the Condensate Booster system.

EC 364641, Mitigate MSL Rad Monitor Spikes

, Revision 0

This reference states, in part, that:

LaSalle Station, in the mid 1990's, installed plant modifications to inject hydrogen into the

reactor feedwater, the Hydrogen Water Chemistry (HW) system.

The purpose of the

modifications was to protect the Reactor Internals and Reactor Recirculation Piping

by reduction

of Intergranular Stress Corrosion Cracking (IGSCC).

This modification (EC 364641) will relocate the current point of hydrogen injection into line

2CD07 A-30" suction piping, from between 2CB01PB (2B) and 2CB01PC (2C) Condensate

Booster (CB) pumps to a point in line 2CD07 A

-30" upstream of all four Condensate Booster

pumps. This relocation of hydrogen injection will reduce the MSL radiation spikes by

providing a more uniform mixing of hydrogen in the Condensate Booster system.

LOR-2H13-P601-F402, MSL A/B Radiation Monitor Downscale/INOP/Hl, Revision 3

This reference states in section C.2 that:

During CP changes and CD/CB Pump swaps MSL Rad monitor Hl alarms have

spuriously annunciated in the past when HWC is online.

This phenomenon is an actual

radiation level change induced by N

-16 production, which is a normal by

-product of H2

gas injection into the reactor. The suspect cause is a release of H2 gas within the

CD/CB piping from a pocketed location.

When this finite amount of gas reaches the

reactor, it results in the formation of N

-16 and is detected as a spike on the MSL and OG

Pretreatment Rad monitors.

LOR-1N62-P600-B502, Off Gas Pre-Treatment Radiation Monitor High Radiation, Revision

This reference states in section C that: During CP changes and CD/CB pump swaps, OG Pretreatment Rad Monitor

Hl alarms have spuriously annunciated in the past when HWC is on-line. This phenomenon is an actual radiation level change induced by N-16 production, which is a normal by product

of H2 gas injection into the reactor.

The suspect cause is a release of H2 gas within the CD/CB piping from a pocketed location.

When this finite amount of gas reaches the reactor, it results in the formation of N-16 and is detected as a spike on the MSL and OG Pretreatment Rad Monitors.

NRC Final Resolution:

The NRC reviewed the aforementioned material related to this question. The NRC agreed with the facility position concerning how this question should be dispositioned. It was noted that t

he LOR procedures still discuss a

possible high radiation alarm condition.

It was also noted that

an engineering change was made to prevent such high radiation alarms.

The definition of 'expected alarm' as it relates to the question was determined to add

further ambiguity.

Based upon these considerations,

the NRC concluded that no correct answer

exists for this question. Therefore, RO question #18 has been deleted from the exam.

RO Question 27

(Post-Exam Comment #2)

Original Question:

Unit 1 is in a LOCA (3) Drywell pressure

is 10 psig and rising slowly (4) Drywell and Suppression

Chamber Hydrogen

is 1% (5) Drywell and Suppression

Chamber Oxygen is 2% What action is required?

The Hydrogen Recombiner

must be-.. A. STOPPED manually B. STARTED manually C. Verified to have AUTO-TRIPPED D. Verified to have AUTO-STARTED Answer: B Applicant

Feedback:

The comment made

was that drywell pressure of 10

psig and rising slowly does not adequately describe the time to reach 15.3

psig which is the point at which the Hydrogen Recombiner will trip. The Hydrogen Recombiner should be stopped manually prior to reaching 15.3 psig which makes (A)

also correct.

Facility Response:

LGA-003 lesson plan states

if LGA-003 is entered then parallel execution is also required because of the symptomatic approach to emergency response precludes the prioritization of any one action path since independence for initiating events and transients must be maintained.

Therefore

, the Hydrogen leg of LGA

-003 is entered because the stem of the question states that

Hydrogen is 1%, which leads to entering LGA

-011 and starting the Hydrogen Recombiner.

There is procedure guidance in LGA

-HG-101 to shutdown the Hydrogen Recombiner when drywell pressure exceeds 15.3 psig, but the Hydrogen Recombiner does not have an auto start feature, and therefore would not be running to require STOPPING Manually.

The LGA-HG-101 indicates that the Hydrogen Recombiner will trip at 15.3 psig of drywell pressure and direct the operator to use containment sprays to reduce drywell pressure before restarting the Hydrogen Recombiner.

STATION RECOMMENDATION: ACCEPT ONLY (B) as the correct answer

References:

LGA-HG-101, Operation of the Hydrogen Recombiner as a Mixing System, Revision 1

This reference states in section E.1.b that

This reference states in section F.2 that:

NRC Final Resolution:

The NRC reviewed the aforementioned material related to this question. The NRC agreed with the facility position concerning how this question should be dispositioned. The stem of

the question is clear that drywell pressure is currently 10 psig with a slowly rising trend.

Per LGA-HG-101, the recombiner is manually stopped after 15.3 psig is exceeded, which also corresponds to the automatic trip setpoint. Based upon these considerations, the NRC concluded that there should be no change to key for RO question #27.

RO Question 40

(Post-Exam Comment #3)

Original Question:

RCIC is operating

in the PRESSURE CONTROL MODE with

the RCIC Pump Discharge Flow Controller

in AUTO set to 600 GPM. Which of the following

set of RCIC system control

manipulations

would result in the FASTEST RATE of RISE in Suppression Pool

water temperature?

Throttle 1E51-F022, Full Flow Test Upstream

Valve,

(1) , in order to

(2) . A. (1) Open Maximize pump flowrate B. (1) Closed Maximize pump flowrate C. (1) Open Maximize pump discharge

pressure D. (1) Closed Maximize pump discharge

pressure Answer: D Applicant

Feedback:

Closing the 1E51

-F022 causes the turbine to spin faster and output more heat to the suppression pool.

Flow will remain at the maximum of 600 GPM. In automatic 600 GPM

is the maximum and flow will go no higher which makes (B) also correct.

Facility Response:

Station's Response:

The RCIC operating procedure for the pressure control mode per LOP-RI-09 indicates that the parameter controlled when throttling 1E51

-F022 is RCIC discharge pressure and RCIC flowrate will be automatically maintained at the flow controller setpoint when operated in AUTO. The RCIC system lesson plan states LE51

-F022 is throttled (to increase pressure and the turbine (pump) speed is automatically adjusted to achieve the flow specified.

STATION RECOMMENDATION:

ACCEPT ONLY (D) as the correct answer

References:

LOP-RI-09, Operating the Reactor Core Isolation Cooling System

for Pressure Control, Revision 11 This reference states in step E.1.10 that:

E.1.10 THROTTLE 1(2)E51

-F022, RCIC Full Flow Test Upstrm

Valve, as needed to maintain desired Rx pressure and/or cooldown

rate. This reference states in Attachment A (Hardcard

- RCIC Operations) step 3 that:

3. To change Cooldown Rate:

-F022, RCIC PMP TEST TO CY UPSTREAM VLV, as needed,

to control Reactor Pressure.

NRC Final Resolution:

The NRC reviewed the aforementioned material related to this question. The NRC agreed with the facility position concerning how this question should be dispositioned. It was determined that system operation and procedural direction support the answer key. Furthermore, it was determined that the selection of distractor 'B' would require

an incorrect understanding of system operation. Throttling FO

-22 with the controller in "AUTO" would not be expected to cause a flow change; if a max flow rate was wanted

, the flow controller would need to be adjusted. Based upon these considerations, the NRC concluded that there should be no change to key for RO question #40.

RO Question 53

(Post-Exam Comment #4) Original Question:

Unit 1 is operating

at rated power (6) A trip of the 1A Service Water Pump results

in a Service Water low pressure alarm. (7) Shortly thereafter,

1H13-P601-B301, SERV WTR EFFLUENT RAD HI alarms. (8) NO other alarms have been received on 1H13-P601. What is the source

of the rising radiation

levels? A. RBCCW Heat Exchangers

B. TBCCW Heat Exchangers

C. Fuel Pool Cooling Heat Exchangers

D. Primary Containment

Ventilation

Chiller Condensers

Answer: C Applicant

Feedback:

RBCCW is cooled by service

water and with

reduced service water pressure RBCCW can leak into service water.

RBCCW is a potentially contaminated system. The applicant believes that both 'A' and 'C' are correct answers.

Facility Response:

RBCCW is filled with makeup condensate and is chemically treated.

Leakage of contaminated water into the RBCCW system is possible.

The RBCCW system has a process radiation monitor which also alarms on 1H1

3-P601. The stem of the question

indicates that no other alarms have been received on 1H13

-P60L and indicates that the leak is not into the RBCCW system. STATION RECOMMENDATION: ACCEPT ONLY (C) as the correct answer

References:

LOR-1H13-P601-B301, Service Water Effluent Radiation High, Revision 2

This reference states in section B.4 that:

LOR-1H13-P601-B401, Reactor Building Closed Cooling Water Radiation High, Revision 2

This reference states in section B that:

NRC Final Resolution:

The NRC reviewed the aforementioned material related to this question. The NRC agreed with the facility position concerning how this question should be dispositioned. It was determined that the stem conditions of the question clearly rule out the RBCCW activity necessary to make distractor "A" correct. Specifically, no high RBCCW activity alarm is present, and therefore

, a leak from RBCCW will not cause high activity in Service Water.

Based upon these considerations, the NRC concluded that there should be no change to key for RO question #53.

RO Question 58

(Post-Exam Comment #5)

Original Question:

Unit 1 is at 100% power. Leakage flow has become just high enough

to activate the Flow Switch 1FS-FC015 shown

below. Identify the location(s)

where this Flow switch initiates

an alarm. A. In the MCR ONLY on panel 1PM13J B. In the MCR ONLY on panel 1H13-P601. C. In the Turbine Building at the Fuel Pool Cooling

Panel ONLY

D. In the MCR AND in the Turbine Building

at the Fuel Pool Cooling

Panel Answer: B Applicant

Feedback:

The Main Control Room can be identified as the correct from the print, and filling the reactor building floor drain sump will cause an alarm on 1PM13J. The applicant believes that the question requires memorization of all Main Control Room Alarm tiles and that the question should be removed from the exam.

Facility Response:

Leakage through 1FS

-FC015 flow switch passes through the 1REL2A

-1 drain line as shown

on the provided drawing from M

-98-L. This continues to

M-91-2 to M-104-1 to M-104-2 to the 1RF02 sump.

lf the water addition to this sump may cause an alarm 1PM13J

-8402 on the 1PML3J for excessive pump

-out time, excessive pump start frequency or tank hi

-hi level.

The stem of the question is discussing the high flow alarm from flow switc

h 1FS-FC015 which will alarm on 1H13-P601. There are two alarms that could result from leakage

flow from the fuel pool. These alarms are on two different panels in the Main Control Room, the 1PM13J and the 1H13-P601. Answers A and B are correct and since they state ONLY then neither answer can

be completely correct, therefore there is no correct answer.

STATION RECOMMENDATION: There is NO correct answer

References:

LOR-1H13-P601-C207, Fuel Pool Cooling System Trouble, Revision 3

This reference states in section A that:

LOR-1PM13J-B402, South Reactor Building Floor Drain Sump Excessive Pump Out Time, Pump Excessive Start Freq., or Hi-Hi Level, Revision 2

This reference states in section A that:

Applicable System Drawings

NRC Final Resolution:

The NRC reviewed the aforementioned material

related to this question.

This question was

modified from the 2013 NRC exam

at the same facility

. The NRC disagreed

with the facility position regarding this question.

The stem of the question asks specifically about where one particular

flow switch causes an alarm; t

he applicable LOR procedure indicates that this

alarm occurs on the 601 panel.

The question is clear

in asking where the flow switch causes the alarm and not what alarms could occur with high leakage. Based upon these considerations, the NRC concluded that there should be no change to key for RO question #58.

SRO Question 86

(Post-Exam Comment #6)

Original Question:

Unit 1 is operating

at 100% power. Unit 2 is in a refueling outage, and

fuel shuffles

are in progress.

There is an inadvertent

Unit 1 Reactor Building

Ventilation

isolation

and SBGT initiation.

The assist NSO

reports that 1VG001,

Inlet Isolation Damper,

opened, reclosed and CANNOT

be manually reopened.

The Unit 2 Supervisor

will direct

(1)

, and Tech Spec 3.6.4.3 requires entry into an LCO to

(2)

. A. (1) Unit 1 SBGT train to be secured

restore SBGT to operable status, ONLY B. (1) Unit 1 SBGT train to be secured

restore SBGT to operable status, AND immediately

suspend core alterations

and movement of irradiated

fuel C. (1) Unit 2 SBGT train to be secured

restore SBGT to operable status, ONLY D. (1) Unit 2 SBGT train to be secured

restore SBGT to operable status, AND immediately

suspend core alterations

and movement of irradiated

fuel Answer: A Applicant

Feedback:

With the VG train initiated inadvertently it is inoperable and therefore LCO 3.6.4.3 Required Actions E.1, E.2, and E.3 are required to be entered.

The applicant believes that 'B' is the correct answer.

Facility Response:

The stem provides a scenario in which a Unit

Reactor Building Ventilation isolation signal occurs, which also causes a SBGT system initiation.

Secondary Containment Isolation

Instrumentation

is required per

Tech Spec 3.3.6.2.

Per LCO 3.0.6,

when a supported system LCO is not met solely due to a support system LCO not being met, the conditions and require

d actions associated with the supported system are not required to be entered. Only the support system required LCO actions are required to be entered.

The question is focused on SBGT train status and LCO 3.6.4.3 based on the given plant conditions.

STATION RECOMMENDATION: ACCEPT ONLY (A) as the correct answer

References:

Technical Specification and Bases 3.6.4.3

These references state, in part, that:

NRC Final Resolution:

The NRC reviewed the aforementioned material

related to this question.

The NRC agreed with the facility position concerning how this question should be dispositioned. It was determined that the applicable technical specifications and bases clearly support the keyed answer of the question. Furthermore, it was determined that the technical specification issue raised by the applicant does not form a technically valid basis for the selection of a different correct answe

r. Based upon these considerations, the NRC concluded that there should be no change to key for

SRO question #86

.

SRO Question 94

(Post-Exam Comment #7)

Original Question:

Core Alterations

have been stopped

by the Refuel SRO due to a Refuel Bridge equipment

failure. Who has the authority

to grant permission

to resume fuel

movement?

A. Dedicated

Refueling

NSO B. Operations

Shift Manager C. Qualified

Nuclear Engineer D. Outage Services Director Answer: B Applicant

Feedback:

Fuel moves require an NSO to give permission.

The applicant believes that both 'A' an 'B' are correct answers.

Facility Response:

The question provides a scenario where fuel moves were stopped due to a refueling bridge equipment failure.

The Shift Manager must grant permission to resume fuel movements following refueling equipment repairs and the NSO must give the Fuel Handling personnel

permission to proceed, as stated in LFP

-100-1 "An NSO assigned to CORE ALTERATIONS shall CHECK the Step for correctness. If it is in agreement w

ith the NCTL (Nuclear Component Transfer List) then the NSO gives the Fuel Handling personnel permission to proceed.

The stem of the question does not specifically direct who is being granted permission to resume fuel moves, but merely states who has the authority to grant permission to resume fuel movements.

In this case the

Operations Shift Manager would provide the NSO permission to resume fuel movements and the

NSO gives the Fuel Handling personnel permission to proceed with fuel movements.

STATION RECOMMENDATION: ACCEPT (A)

AND (B) as the correct answer

References:

LFP-100-1, Master Refuel Procedure, Revision

This reference states in section D.3 that:

This reference states in section E.3.8

that: NRC Final Resolution:

The NRC reviewed the

aforementioned material related to this question.

The NRC agreed with the facility position concerning how this question should be dispositioned. During the administration of the written exam, in response to an applicant question

, additional stem information was provided that repairs had been completed. It was determined that the Stem wording of the question and LFP

-100-1 support a possible interpretation that makes distractor 'A' also a correct answer. Based upon these considerations, the NRC concluded that both 'A' and 'B' should be accepted as correct answers for SRO question #94.

SRO Question 81

(Post-Exam Comment #8)

Original Question:

Unit 1 is in Mode 5, and irradiated

fuel assemblies

are being moved within

the Reactor Pressure Vessel (RPV). In accordance

with LCO 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated

Fuel, RPV cavity water

level must be greater than or equal to

(1)

above the RPV flange; otherwise,

movement of irradiated

fuel assemblies

in the RPV must be IMMEDIATELY

SUSPENDED

to maintain sufficient

water level to

(2)

. A. (1) 22 feet (2) ensure 99.5% of the total iodine released from a damaged fuel assembly is

retained in

the water B. (1) 23 feet

(2) ensure 99.5% of the total iodine released from a damaged fuel assembly is

retained in the water

C. (1) 23 feet (2) retain iodine fission product activity in the event of a fuel handling accident,

keeping offsite doses within limits

D. (1) 22 feet (2) retain iodine fission product activity

in the event of a fuel handling accident,

keeping offsite doses within limits

Answer: D Applicant

Feedback:

The answers for both (A) and (D) are included in the Tech Spec Basis for Tech Spec 3.9.6.

Facility Response:

The Tech Spec Bases for 3.9.6 discusses that the LCO is based on a minimum water level

of 22 feet. The analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water, and that

offsite doses are maintained within allowable limits.

The bases also includes a discussion that 23 feet of water allows

a decontamination factor of 200 to be used in accident analysis for iodine which relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the fuel.

The damaged fuel assembly rods are retained in the refueling cavity water.

The discussion in the bases is supporting information and not directly related to the LCO statement.

STATION RECOMMENDATION: ACCEPT ONLY (D) as the correct answer

References:

Technical Specification 3.9.6 and Bases

These references state in part that:

NRC Final Resolution:

The NRC reviewed the aforementioned material related to this question.

The NRC agreed with the facility position concerning how this question should be dispositioned. It was determined that although the Technical Specification Bases do contain the 99.5% criteria, it is associated with the 23' level value

, and not the 22' level value. Additionally, while the 22' level

part of the correct answer is RO knowledge, all second part answers consist of SRO

-only knowledge. Based upon these considerations, the NRC concluded that no key change should be made for

SRO question #81.

B. Hanson

- 2 - Letter to Bryan C. Hanson to Robert J. Orlikowski dated December

2, 2016 SUBJECT: LASALLE COUNTY STATION, UNITS 1 AND 2

- NRC INITIAL LICENSE EXAMINATION REPORT

05000373/2016301; 05000

374/2016301 DISTRIBUTION

Jeremy Bowen

RidsNrrDorlLpl3

-2 Resource

RidsNrrPMLaSalle

RidsNrrDirsIrib Resource Cynthia Pederson

Darrell Roberts

Richard Skokowski

Allan Barker

Carole Ariano

Linda Linn

DRPIII DRSIII Nancy Salgado

Colleen Schmidt

ADAMS Accession Number: ML16357A679

OFFICE RIII RIII RIII RIII NAME JSeymour:jw

JSeymour for CZoia ROrlikowski

DATE 12/20/16 12/20/16 12/22/16 OFFICIAL RECORD COPY