IR 05000373/2020301
| ML20352A324 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/17/2020 |
| From: | Patricia Pelke NRC/RGN-III/DRS/OLB |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| Roach G | |
| Shared Package | |
| ML19127A357 | List: |
| References | |
| Download: ML20352A324 (16) | |
Text
December 17, 2020
SUBJECT:
LASALLE COUNTY STATION, UNITS 1 AND 2NRC INITIAL LICENSE EXAMINATION REPORT 05000373/2020301; 05000374/2020301
Dear Mr. Hanson:
On December 1, 2020, 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 examinations.
Preliminary observations noted during the examination process were discussed on November 12, 2020, with Mr. Philip Hansett, Plant Manager, and other members of your staff.
An exit meeting was conducted by telephone on December 10, 2020, with Mr. Philip Hansett, Plant Manager, other members of your staff, and Mr. Gregory Roach, Chief Operator Licensing Examiner, to review the final grading of the written examination for the license applicants.
During the telephone conversation, NRC resolution of the plants post-examination comments, received by the NRC on December 1, 2020, were discussed.
The NRC examiners administered an initial license examination operating test during the weeks of November 2, 2020, and November 9, 2020. The written examination was administered by training department personnel on November 13, 2020. Nine Senior Reactor Operator and three Reactor Operator applicants were administered license examinations. The results of the examinations were finalized on December 14, 2020. Twelve applicants passed all sections of their respective examinations. Nine applicants were issued senior operator licenses and three applicants were issued operator licenses.
The administered written examination and operating test, as well as documents related to the development and review (outlines, review comments and resolution, etc.) of the examination will be withheld from public disclosure until December 1, 2022. 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, Patricia J. Pelke, Chief Operations Branch Division of Reactor Safety
Docket Nos. 50-373; 50-374 License Nos. NPF-11; NPF-18
Enclosures:
1. Examination Report 05000373/2020301; 05000374/2020301 2. Post-Examination Comments, Evaluation, and Resolutions 3. Simulator Fidelity Report
REGION III==
Docket No:
50-373;05-374 License No:
05000373/2020301; 05000374/2020301 Enterprise Identifier: L-2020-OLL-0037 Licensee:
Exelon Generation Company, LLC Facility:
LaSalle County Station, Units 1 and 2 Location:
Marseilles, IL Dates:
November 2, 2020, through December 1, 2020 Examiners:
G. Roach, Senior Operations Engineer, Chief Examiner M. Kennard, Senior Operations Engineer, Examiner T. Iskierka-Boggs, Reactor Engineer, Examiner Exam Authors:
B. Bartlett, Senior Operations Engineer, RIII B. Hartle, Reactor Engineer, NRR Approved By:
P. Pelke, Chief Operations Branch Division of Reactor Safety
SUMMARY
Examination Report 05000373/2020301; 05000374/2020301; 11/02/2020-12/01/2020; 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 11.
Examination Summary Twelve of twelve applicants passed all sections of their respective examinations.
Nine applicants were issued senior operator licenses and three applicants were issued operator licenses. (Section 4OA5.1)
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 licensees staff used the guidance prescribed in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 11, to develop, validate, administer, and grade the written examination and operating test. The written examination outlines and operating test outlines were prepared by the NRC staff. The NRC staff developed the written examination and members of the facility licensees staff developed the operating test. As part of the operating test development, the NRC examiners visited the facility during the week of July 27, 2020, to pre-validate the events/malfunctions constructed from the NRC prepared outlines, and to obtain feedback from your staff regarding scenario validity and capabilities of the simulator to support event performance. The NRC examiners validated the proposed examination during the week of September 28, 2020, with the assistance of members of the facility licensees staff. During the onsite validation week, the examiners audited all twelve 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 and dynamic simulator scenarios, during the period of November 2, 2020, through November 10, 2020. The facility licensee administered the written examination on November 13, 2020.
b. Findings
- (1) Written Examination During validation of the NRC developed written examination, several questions were modified or replaced. All changes made to the written examination were made in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors. Form ES-401-9, Written Examination Review Worksheet, used primarily for the documentation of metrics on the NRC developed written examination, was updated with post-examination changes. The Form ES-401-9, the written examination outlines (ES-401-1 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 NRCs Agencywide Documents Access and Management System (ADAMS) on December 1, 2022, (ADAMS Accession Numbers ML19127A356, ML19127A359, ML19127A362, and ML19127A363, respectively).
On December 1, 2020, the licensee submitted documentation noting that there were two post-examination comments for consideration by the NRC examiners when grading the written examination. The post-examination comments are documented in 2 of this report.
The NRC examiners completed grading of the written examination on December 9, 2020, and conducted a review of each missed question to determine the accuracy and validity of the examination questions.
- (2) Operating Test The NRC examiners determined that the operating test, developed by the licensee from the NRC prepared outlines, was within the range of acceptability expected for a proposed examination.
Following the review and validation of the operating test, extensive modifications were made to two job performance measures to reduce the time to complete these activities, minor modifications were made to several other job performance measures, and minor modifications were made to the dynamic simulator scenarios. All changes made to the operating test were made in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors, and were documented on Form ES-301-7, Operating Test Review Worksheet. The Form ES-301-7, the operating test outlines (ES-301-1, ES-301-2, and ES-D-1s), and both the proposed and final operating tests, will be available electronically in the NRC Public Document Room or from the Publicly Available Records component of NRCs ADAMS on December 1, 2022, (ADAMS Accession Numbers ML19127A356, ML19127A359, ML19127A362, and ML19127A363, respectively).
The NRC examiners completed grading of the operating test on December 9, 2020.
- (3) Examination Results Nine applicants at the Senior Reactor Operator level and three applicants at the Reactor Operator level were administered written examinations and operating tests.
Twelve applicants passed all portions of their examinations. Twelve applicants were issued their respective operating licenses on December 14, 2020.
.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, Part 55.49, Integrity of Examinations and Tests. The examiners used the guidelines provided in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, to determine acceptability of the licensees examination security activities.
b. Findings
None.
4OA6 Management Meetings
.1 Debrief
The chief examiner presented the examination teams preliminary observations and findings on November 12, 2020, to Mr. Philip Hansett, Plant Manager, and other members of the LaSalle County Station staff.
.2 Exit Meeting
The chief examiner conducted an exit meeting on December 10, 2020, with Mr. Philip Hansett, Plant Manager, and other members of the LaSalle County Station staff, by telephone. The chief examiner 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 CONTACT
Licensee
- J. Washko, Site Vice President
- P. Hansett, Plant Manager
- C. Smith, Operations Director
- J. Messina, Shift Operations Superintendent
- T. Macintyre, Manager Operations Training
- M. Miller, Shift Manager
- G. Wood, Senior Manager Site Training
- S. Boehler, Operations Instructor-Lead
- C. Betken, Senior Operations Training Instructor
- D. Mearhoff, Regulatory Assurance Manager
- J. Greenblott, Senior Regulatory Engineer
U.S. Nuclear Regulatory Commission
- W. Schaup, Senior Resident Inspector
- J. Havertape, Resident Inspector
- G. Roach, Senior Operations Engineer, Chief Examiner
- M. Kennard, Senior Operations Engineer, Examiner
- T. Iskierka-Boggs, Reactor Engineer, Examiner
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
None
LIST OF ACRONYMS USED
Agencywide Documents Access and Management System
NRC
U.S. Nuclear Regulatory Commission
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
SRO Question 83
Unit 1 was at 24% RTP:
1PM03J-B511, CNDSR VAC LO, is LIT
Subsequently,
Reactor power reduced to 20% RTP
As a result:
Condenser vacuum degraded to 10 inches backpressure and is now STABLE
(1) What is the expected automatic plant response with NO operator action?
(2) What procedure MUST be implemented to address the automatic plant response?
A. (1) Reactor power will be greater than 20% RTP
(2) LGP-3-2, Reactor SCRAM
B. (1) Reactor scram
(2) LGP-3-2, Reactor SCRAM
C. (1) Reactor power will be greater than 20% RTP
(2) LOA-TG-101, Unit 1 Turbine Generator
D. (1) Reactor scram
(2) LOA-TG-101, Unit 1 Turbine Generator
Answer:
__C__
Reference(s) provided to NRC:
LOR-1PM03J-B511, CONDENSER VACUUM LOW, Rev. 6
LOA-TG-101, Unit 1 Turbine Generator, Rev. 20
LGP-3-2, Reactor SCRAM, Rev. 74
Applicant Comment:
Answers A and C are both correct and supported per alarm response procedure LOR-1PM03J-
B511, step 4. This states that if condenser vacuum continues to degrade, decrease reactor
power as necessary to stabilize vacuum and if turbine trip is imminent then manually SCRAM
the reactor per LGP-3-2. As the turbine trip set point, 8.4 inches backpressure, had been
surpassed a reactor SCRAM is directed per the above step. Additionally, an entry condition for
LOA-CW-101 is met (degraded vacuum) which contains manual scram criteria of less than
inches Backpressure and Reactor Power less than 75%, scram the reactor.
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
Facility Position on Applicant Comment:
With the Reactor at 24% RTP and condenser vacuum degrading, when back pressure reaches
23.7 of vacuum (6.3 Backpressure), 1PM03J-B511 will annunciate. In order to reach 10
backpressure, vacuum must continue to degrade and reach the Main Turbine Trip setpoint of
8.4 backpressure. As stated in the annunciator response procedure (LOR-1PM03J-B511)
step 4: If vacuum cannot be stabilized and Turbine Trip is imminent, MANUALLY Scram
reactor per LGP-3-2, Reactor SCRA
- M. Therefore, when condenser Backpressure
approaches 8.4 of Backpressure, LGP-3-2 should be entered, and the reactor scrammed
per step 4 of LOR-1PM03J-B511.
STATION RECOMMENDATION: ACCEPT (A) and (C) as correct answers.
NRC Evaluation/Resolution:
The applicant selected Distractor A as the correct answer when responding to Question 83.
In order to arrive at this answer, the applicant would have had to correctly assess part (1) of the
answer. Specifically, based on plant conditions described in the stem with no operator actions,
reactor power at 20% RTP, and condenser backpressure degrading to 10 and stabilizing that a
turbine trip would have automatically occurred. Part (1) of Distractor A states, Reactor power
will be greater than 20% RTP which means the applicant would have correctly recognized that
an automatic main turbine trip with reactor power < 25% and therefore within the capacity of the
Turbine Bypass Valves would not result in a Reactor Protection System (RPS) initiation to
SCRAM the reactor. In addition, the applicant would have had to correctly recognize that the
automatic turbine trip would result in a loss of Extraction Steam to the Feedwater Heaters
resulting in a lowering of feedwater temperature entering the reactor vessel and therefore a
positive reactivity insertion, hence part (1) of the distractor stating that reactor power would be
greater than 20% RT
- P. If an automatic reactor SCRAM had occurred, power would be less
than 20% RTP as no indications of an ATWS were provided in the question stem.
Part (2) of Distractor A incorrectly states that procedure LGP 3-2, Reactor SCRAM would be
the procedure that MUST be implemented to address the automatic actuation from part (1) of
the question. As indicated above, the only automatic actuation which would have occurred
based on the conditions in the question stem was the trip of the main turbine. The applicant
notes in their contention that step 4 of annunciator response procedure LOR-1PM03J-B511,
Condenser Low Vacuum states, in part, If vacuum cannot be stabilized and Turbine Trip is
imminent, MANUALLY Scram reactor per LGP-3-2, Reactor SCRAM. This step in the
procedure addresses the Concept of Operations philosophy that if an automatic RPS actuation
is about to occur, then operators should take manual action to insert a SCRAM and not rely on
the automatic action to initiate. The piece that is missing from the statement in the annunciator
response procedure, is when reactor power is < 25% RTP or within the capacity of the Turbine
Bypass Valves. In the situation provided in the question stem, a reactor SCRAM will not
automatically occur when the main turbine trips, so it would not be expected for the operators
to manually insert one. Therefore, it is recommended that the Facility enhance LOR-1PM03J-
B511, step 4 to include this additional caveat for conditions below 25% RT
- P. In addition, LOR-
1PM03J-B511, step 5 directs operators to refer to LOA-TG-101, Unit Turbine Generator if the
condenser backpressure degraded to the turbine trip setpoint, which it did when 10 inches back
pressure was achieved. LOA-TG-101 entry condition A.1 states, Loss of Turbine/Generator.
In response to a loss of the Turbine/Generator step B.1 states, CHECK Reactor Power - below
25% and stable. At the time of the turbine trip reactor power was indicating 20% RTP and
stable. Since this step was met, then the Response Not Obtained column would not be entered.
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
If step B.1 had not been met, the Response Not Obtained column would have directed the
operator to SCRAM the reactor. The applicant also noted that in abnormal operating procedure,
LOA-CW-101, Unit 1 Circulating Water System Abnormal there is manual SCRAM criteria for,
Power below 75% AND backpressure above 5. Again, as described above, this is assuming
power is above the RPS actuation value for a main turbine trip and the degraded vacuum in the
condenser is driving operators to trip the main turbine.
In summary, the only automatic actuation which would have occurred based on the conditions
listed in the question stem was a main turbine trip. As a result, the procedure which MUST be
entered is LOA-TG-101 as entry condition A.1 loss of Turbine/Generator has occurred. It is
recommended that LOR-1PM03J-B511 and LOA-CW-101 be enhanced as appropriate to
address a degrading condenser vacuum with reactor power < 25% RTP. Answer C stands
as the only correct answer.
Therefore, the NRC concluded that no change to the key for this exam question was required.
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
SRO Question 85
Unit 1 was at 100% RTP when a LOCA occurs:
HCVS is NOT available
LGA-VQ-102, Unit 1 Emergency Containment Vent is in progress using Standby Gas
Treatment (VG)
Drywell pressure is 45 psig and slowly LOWERING
1H13-P601-F401, DIV 1 MSL PIPE TUNNEL AMB TEMP HI, is LIT
1H13-P601-B111, TB/AUX RAD HI, is LIT
Rx Bldg Standby Gas ARM is in alarm
Turb Bldg Off Gas Equip and Sample Station ARM is in alarm
The radiation alarms are due to (1) and ODCM samples are (2) PRIOR to
venting.
A.
(1) leakage on a MSL outside of containment
(2) required
B.
(1) leakage on a MSL outside of containment
(2) NOT required
C.
(1) venting the containment at elevated pressure
(2) required
D.
(1) venting the containment at elevated pressure
(2) NOT required
Answer:
__D__
Reference(s) provided to NRC:
LGA-VQ-102, Unit 1 Emergency Containment Vent, Rev 3
LGA-003, Primary Containment Control, Rev 18
Drawing M-89, Sheet 1, Schematic Diagram, Standby Gas Treatment System
Drawing M-92, Sheet 1, Schematic Diagram, Primary Containment Vent & Purge
System
Applicant Comment:
For the stem issue, VG is only to be used to vent in LGA-VQ-101 if both VQ trains are
unavailable. Further, LGA-VQ-101 must be exited once DW pressure is above 1.97 (or any
group IV isolation) and the 1/2VG023 reclosed (See LGA-VQ-101 C.1 and discussion section).
Once you're out of LGA-VQ-101 and into LGA-VQ-102 to stay below PCPL (or H2), HCVS and
VQ are the only options, with HCVS preferred. Therefore, the framing of the question puts the
plant in an invalid line-up. This also makes is difficult to determine the position of 1VQ041,
which would be closed in VG venting per LGA-VQ-101 but would be open in LGA-VQ-102.
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
This damper position is consequential to determining communication of containment steam
through VG with the RB exhaust riser and OG/TB sample station.
If we look past the issue in the plant-line-up in the stem, the issue of MSL break indications
must also be considered. Given the parameters to address are VG rads, OG/TB Sample Panel
Rads, and MSL temperature alarms it would need to be demonstrated that a MSL break is not
plausible. VG rads can be associated with any VG operation with containment steam
post-accident, given the stem is operating VG, this is plausible regardless of the selected
distractor. Steam tunnel temperatures (hi or differential) can be associated with leakage of
high energy steam into the tunnel, or affects to ventilation (group IV, VR tripped). Therefore,
the MST temperature alarms cannot be discerned between distractors. The TB/OG alarms
are likely induced by changing radiological conditions in the turbine building. This may be
associated with MSL rupture leakage into the Heater Bay (expected on a MSL break), including
areas of 687 and 754 near sample panels. This mirrors the condition of a potential containment
steam in the RB return air riser due to VG rupture. This potential is described in the UFSAR
when discussing the VR design basis (Page 9.1-19, section 9.4.2.3.i):
(i)
Airflow check dampers are provided in the main steam pipe tunnel to check the
steam flow in the reactor building following the pipe break in the main steam pipe
tunnel. The steam is released through the blowout panels to the turbine building
and outdoors, and through the pressure relief damper to the auxiliary building
HVAC equipment area.
Upon review of the prints, procedures, and abnormal procedures, it seems that this question
intends to measure knowledge of a precaution of LGA-VQ-101 and the VG UFSAR description.
In the way it is delivered and the selected distractors, the plant is in a line-up not supported by
procedure in such a way that a MSL break outside containment may present the given
symptoms. My determination is that there is no distinguishing information given in the stem to
justify distractor B or D from the other.
Facility Position on Applicant Comment:
The stem of the question states that the plant is venting per LGA-VQ-102 using the Standby
Gas System (VG), however this procedure does not allow VG to be used for venting. Further,
the allowed lineup with HCVS not available (using VQ) requires 1VQ041, Reactor Building
Exhaust Discharge Valve, to be OPE
- N. With 1VQ041 OPEN, there would be a path from the
Main Steam Tunnel (MSL Leak in distractor A) to the running VG system (with 1VG023, SBGT
VQ crosstie, OPEN), and thus produce alarms in the same fashion as when venting using VG.
The lineup given in the stem would support that the given alarms could be from either venting
the containment at elevated pressure or a Main Steam Line Leak. Since the lineup in the stem
is not allowed with drywell pressure at 45 psig and the system lineup is needed to determine the
cause of the rad alarms, the question is not valid as written.
DCPs 9700402/403 added 1(2)VG023, SBGT VQ XTIE Valve, to isolate the primary
containment during VQ inserting and purging. This will ensure that the VG system is not
over-pressurized during a LOCA with the 26 VQ valves open for inserting or purging.
Additionally, the piping downstream of 1VG023, SBGT VQ XTIE Valve, has a design
pressure of 10 psig (well below the 45 psig stated in the question).
STATION RECOMMENDATION: REMOVE question from exam.
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
NRC Evaluation/Resolution:
Based on information provided in the question stem, the primary containment was being
emergency ventilated to prevent exceeding the Primary Containment Pressure Limit using the
standby gas treatment system (VG) per LGA-VQ-102, Unit 1 Emergency Containment Vent.
Emergency procedure LGA-VQ-102 does not provide specific steps to cross tie the VG system
to the containment purge system to emergency ventilate the primary containment under high
containment pressure conditions as are described in the question stem. Contrary to this, there
are multiple references to potentially using the VG system under these conditions. Specifically,
section E.2, note 2 states in part, [a] pressure band of 10 psig immediately below the limit, such
as 50 to 60 psig while in LGA-003, is recommended to reduce the amount of damage in/to the
Reactor Building when using VQ or VG [emphasis added] for emergency venting. Additional
technical information provided by the facility indicates that the 1VG023, SBGT VQ XTIE Valve,
would have to be manually opened under accident radiological conditions against a differential
pressure of up to 60 psig which is questionable as to whether this would be possible in order to
align the VG system for emergency ventilation. In addition, the VG system downstream of
1VG023 has a design pressure of 10 psig which is well below the 45 psig conditions described
in the question stem. Based on this additional information, the use of the VG system under the
plant conditions described is not operationally valid and created confusion for the applicant as
they attempted to answer the question. Based on the additional information provided by the
facility, it is recommended that the facility review the note described above and other references
to the VG system in LGA-VQ-102 to ensure they remain accurate.
Therefore, the NRC concluded that with an operationally invalid plant line-up in the stem,
the applicant could be confused with the plant conditions described in the question stem
and therefore, the question was DELETED from the administered examination.
SIMULATOR FIDELITY REPORT
Facility Licensee:
LaSalle County Station, Units 1 and 2
Facility Docket No:
50-373; 50-374
Operating Tests Administered:
November 2, 2020, through November 10, 2020
The following documents observations made by the U.S. Nuclear Regulatory Commission
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, Part 55.45(b).
These observations do not affect U.S. Nuclear Regulatory Commission 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
During dynamic Scenario 2 an applicant crew noted, prior to taking the
shift, that the 1D Main Steam Line Radiation Monitor display at the
1H13-P636 Panel was not working. This deficiency did not impact the
applicant crews ability to perform the dynamic scenario and was
repaired with the 1D Main Steam Line Radiation Monitor functional for
the remainder of the exam.