IR 05000456/2019301
| ML19224C181 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 08/12/2019 |
| From: | Rhex Edwards Operations Branch III |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| Shared Package | |
| ML17214A824 | List: |
| References | |
| 50-456/19-301, 50-457/19-301 | |
| Download: ML19224C181 (33) | |
Text
ust 12, 2019
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2NRC INITIAL LICENSE EXAMINATION REPORT 05000456/2019301 AND 05000457/2019301
Dear Mr. Hanson:
On June 28, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed the initial operator licensing examination process for license applicants employed at your Braidwood Station. The enclosed report documents the results of those examinations. Preliminary observations noted during the examination process were discussed on June 19, 2019, with Ms. M. Marchionda-Palmer, Site Vice President, and other members of your staff. An exit meeting was conducted by telephone on July 3, 2019, with Ms. M. Marchionda-Palmer, other members of your staff, and Mr. J. Seymour, Operations Engineer, to review the final grading of the written examination for the license applicants. During the telephone conversation, NRC resolutions of the stations post-examination comments, received by the NRC on June 28, 2019, were discussed.
The NRC examiners administered an initial license examination operating test during the weeks of June 10 and June 17, 2019. The written examination was administered by Braidwood Station training department personnel on June 19, 2019. Eight Senior Reactor Operator applicants were administered license examinations. The results of the examinations were finalized on July 16, 2019. One applicant failed one or more sections of the administered examination and was issued a preliminary results letter. Seven applicants passed all sections of their respective examinations and were issued senior 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 June 19, 2021. 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 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Rhex A. Edwards, III, Acting Chief Operations Branch Division of Reactor Safety Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77
Enclosures:
1. OL Examination Report 05000456/2019301; 05000457/2019301 2. Post-Examination Comment, Evaluation, and Resolution 3. Simulation Facility Fidelity Report
REGION III==
Docket Nos: 05000456; 05000457 License Nos: NPF-72; NPF-77 Report No: 05000456/2019301; 05000457/2019301 Enterprise Identifier: L-2018-OLL-0004 Licensee: Exelon Generation Company, LLC Facility: Braidwood Station, Units 1 and 2 Location: Braceville, IL Dates: June 10, 2019, through June 28, 2019 Examiners: J. Seymour, Operations Engineer, Chief Examiner C. Zoia, Senior Operations Engineer, Examiner E. Cushing, Reactor Engineer, Examiner Approved By: R. Edwards, III, Acting Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY
Examination Report 05000456/2019301; 05000457/2019301; 06/10/2019-06/28/2019;
Exelon Generation Company, LLC; Braidwood Station, Units 1 and 2; Initial License Examination Report.
The announced initial operator licensing examination was conducted by regional U.S. Nuclear Regulatory Commission examiners in accordance with the guidance of NUREG-1021, Operator Licensing Examination Standards for Power Reactors,
Revision 11.
Examination Summary Seven of eight applicants passed all sections of their respective examinations.
Seven applicants were issued senior operator licenses. One applicant failed one or more sections of the administered examination and was issued a preliminary results letter. (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 were prepared by the NRC staff and were transmitted to the facility licensees staff. Members of the facility licensees staff prepared the operating test outlines and developed the written examination and operating test. The NRC examiners validated the proposed examination during the week of May 6, 2019, with the assistance of members of the facility licensees staff. During the on-site validation week, the examiners audited three 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 June 10 through June 17, 2019. The facility licensee administered the written examination on June 19, 2019.
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.
During the validation of the 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, and were documented on Form ES-401-9, Written Examination Review Worksheet. The Form ES-401-9, the written examination outlines (ES-401-2 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) on June 19, 2021, (ADAMS Accession Numbers ML17214A825, ML17214A828, ML17214A830, and ML17214A832, respectively).
On June 28, 2019, the licensee submitted documentation noting that there were six 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 documented in Enclosure 2 to this report.
The NRC examiners graded the written examination on July 2, 2019, 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, as originally proposed by the licensee, was within the range of acceptability expected for a proposed examination.
Following the review and validation of the operating test, minor modifications were made to several Job Performance Measures, and some 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 NRC's ADAMS on June 19, 2021 (ADAMS Accession Numbers ML17214A825, ML17214A828 , ML17214A830, and ML17214A832, respectively).
The NRC examiners completed operating test grading on July 16, 2019.
- (3) Examination Results Eight applicants at the senior reactor operator level were administered written examinations and operating tests. The results of the examinations were finalized on July 16, 2019. Seven applicants passed all portions of their examinations and were issued their respective operating licenses on July 16, 2019. One applicant failed one or more sections of the administered examination and was issued a preliminary results letter.
.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 licensees examination security activities.
b. Findings
None.
4OA6 Management Meetings
.1 Debrief
The chief examiner presented the examination team's preliminary observations and findings on June 18, 2019, to Ms. M. Marchionda-Palmer, Site Vice President, and other members of the Braidwood Station staff.
.2 Exit Meeting
The chief examiner conducted an exit meeting on July 3, 2019, with Ms. M. Marchionda-Palmer, Site Vice President, and other members of the Braidwood Station staff, by telephone. The examiners asked the licensee whether any of the material used to develop or administer the examination should be considered proprietary. Proprietary or sensitive information identified during the examination or debrief/exit meetings will be handled in accordance with the applicable requirements.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- M. Marchionda-Palmer, Site Vice President
- J. Keenan, Plant Manager
- F. Jordan, Training Director
- P. Moodie, Operations Director
- M. Spillie, Acting Regulatory Assurance Manager
- K. Lueshen, Operations Service Manager
- J. Petty, Shift Operations Superintendent
- J. Beard, Operations Training
- D. Brunswick, Operations Training
- J. Taff, Operations Training Manager
- R. Schliessmann, Regulatory Assurance
- R. Witcofski, Operations
U.S Nuclear Regulatory Commission
- R. Baker, Branch Chief (Acting)
- R. Edwards, Branch Chief (Acting)
- D. Kimble, Senior Resident Inspector
- J. Seymour, Operations Engineer, Chief Examiner
- C. Zoia, Senior Operations Engineer, Examiner
- E. Cushing, Reactor Engineer, Examiner
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
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #6
Unit 2 is in MODE 5.
2BwGP 100-1, PLANT HEATUP, is in progress.
- RCS temperature is 100°F.
- Pressurizer Level is 50 percent.
- The 2A RH train is in shutdown cooling.
- The 2B RH train is in STANDBY.
- The 2A RH pump amps are fluctuating between 50 to 60 amps.
- The 2A RH pump flow is fluctuating between 4500 to 5000 gpm.
The crew will (1) _, and (2) _ to correct the issue.
- A. (1) continue in 2BwGP 100-1, PLANT HEATUP, ONLY
(2) IMMEDIATELY trip the 2A RH pump to prevent damage
- B. (1) continue in 2BwGP 100-1, PLANT HEATUP, ONLY
(2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV,
to reduce flow
- C. (1) enter 2BwOA PRI-10, LOSS OF RH COOLING UNIT 2
(2) IMMEDIATELY trip the 2A RH pump to prevent damage
- D. (1) enter 2BwOA PRI-10, LOSS OF RH COOLING UNIT 2
(2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV,
to reduce flow
Answer D
Answer Explanation
A - Plausible: continue in 2BwGP 100-1 and immediately trip the 2A RH pump are incorrect.
The 2BwGP 100-1 directs securing all RH trains during the startup. A novice applicant may
interpret this procedural flow path as adequate to address the RH pump current issue. The 2A
RH pump amps are fluctuating near the red band. This would be correct if the 2A RH pump flow
was reduced and did not stabilize parameters.
B - Plausible: continue in 2BwGP 100-1 is incorrect and Take manual control of 2RH618
to reduce flow is correct. The 2BwGP 100-1 directs securing all RH trains during the startup.
A novice applicant may interpret this procedural flow path as adequate to address the
RH pump current issue.
C - Plausible: 2BwOA PRI-10 is correct, immediately trip the 2A RH pump is incorrect.
This would be correct if the 2A RH pump flow was reduced and did not stabilize parameters.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
D - Correct: 2BwOA PRI-10 and Take manual control of 2RH618 to reduce flow are correct.
Per 2BwOA PRI-10 the first mitigative strategy performed is to reduce RH pump flow. The crew
should attempt to stabilize RH system operation prior to tripping the running RH pump and
continuing to further mitigating actions.
Technical Reference and Revision #
2BwOA PRI-10, Revision 107, Page 2.
_BwOA PRI-10 Lesson Plan (I1-OA-XL-20) Revision 13, Page 2.
Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Facility Position on Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Additional Information on Position Provided by Facility on July 2, 2019
[T]he expectation would be to attempt to lower RH system flow without exiting 2BwGP 100-1
first, since the pump is experiencing potential run out conditions. If that was successful, no
further procedure transition would take place. If not successful, the crew should enter 2BwOA
PRI-10. Entry conditions for 2BwOA PRI-10 are phrased as a may for fluctuating RH pump
amps, if the condition were corrected by adjusting RH system flow then BwOA entry would not
be needed.
The stem did not state actions taken are not successful. This creates an unstated assumption
that the action being taken could correct the issue and therefore no further procedure transitions
are required. BwOP RH-6 would have been completed prior to where the stem begins the initial
conditions for the question. Therefore, the applicant could utilize the precaution (D.4) from
memory (per BwAP 340-1) to lower RH system flow and stabilize the plant preventing damage
to the RH pump, without transitioning to another procedure. The conditions of the stem are
consistent with an excess of RH cooling (RH pump flow fluctuating between 4500-5000 gpm),
prudent operator action to correct this issue should be utilized to prevent damage or a trip of the
2A RH pump resulting in a loss of RH cooling.
Additional References
Excerpt from 2BwGP 100-1, Plant Heatup, Revision 39
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Excerpt from BwOP RH-6, Placing the RH System in Shutdown Cooling, Revision 59
Excerpts from 2BwOA PRI-10, Loss of RH Cooling Unit 2, Revision 107
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Excerpt from BwAP 340-1, Use of Procedures for Operating Department, Revision 30
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
NRC Resolution
The
- U.S Nuclear Regulatory Commission (NRC) notes that, in accordance with NUREG-1021
ES-403, Section D.1.a, and NUREG-1021 Appendix E, Section B.7, no questions were posed
by any applicants during the examination regarding Question #6.
The distractor/answer choices addressed by the applicants comment consist of the following:
Distractor B:
(1) continue in 2BwGP 100-1, PLANT HEATUP, ONLY
(2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV, to reduce flow
Answer D:
(1) enter 2BwOA PRI-10, LOSS OF RH COOLING UNIT 2
(2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV, to reduce flow
The stem of the question explicitly provided, in part, the following key information to the
applicant:
- 2BwGP 100-1 in effect
- 2A RH train aligned for shutdown cooling
- 2A RH pump amps fluctuating between 50 to 60 amps
Under the above conditions, BwOP RH-6, Placing the RH System in Shutdown Cooling, would
have been previously performed to establish shutdown cooling operations. As stated in the
information provided by the facility, BwOP RH-6 would have been completed prior to where the
stem begins the initial conditions for the question. Based upon this, the precautions of BwOP
RH-6 (including step D.4) are no longer procedurally in effect during the timeframe associated
with the stem conditions.
The first half of distractor B states continue in 2BwGP 100-1, PLANT HEATUP, ONLY.
The inclusion of the word ONLY in this distractor would limit any procedurally driven corrective
actions for addressing the oscillating RH Pump amps to solely guidance contained within
2BwGP 100-1. However, 2BwGP 100-1 does not contain any procedural guidance for
addressing conditions of oscillating RH Pump amps. Thus, the corrective action contained in
the second part of distractor B (take manual control of 2RH618, HX 1A BYP FLOW CONT
VLV, to reduce flow) would not have a procedural basis within the context of this distractor.
In contrast, 2BwOA PRI-10, Loss of RH Cooling Unit 2, lists the symptom of oscillating
RH Pump amps occurring as an entry condition. Based upon this, it would be appropriate
to enter this Procedure 2BwOA PRI-10, RNO step 1.b, subsequently directs the required
reduction in RH pump flow. Thus, answer D provides a procedurally directed means of
correcting the issue presented in the stem.
Therefore, the NRC concludes that no change should be made to the key regarding this exam
question.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #10
Unit 1 is at 100 percent power.
The following annunciators have just alarmed:
- 1-1-A2, CNMT DRAIN LEAK DETECT FLOW HIGH
- 1-10-E4, OVATION SYSTEM TROUBLE
- 1-10-E5, OVATION ALTERNATE ACTION
The RO reviews OWS graphic 6040, FW OVERVIEW, and notes the following:
The crew will...
1. Reduce Unit 1 Turbine Loading
2. Trip Unit 1 Reactor
3. Initiate Safety Injection
4. Actuate Main Steamline Isolation
A. 1 ONLY.
B. 2 ONLY.
C. 2 and 3 ONL
- Y.
D. 2, 3, and 4.
Answer D
Answer Explanation
A - Plausible: Reduce turbine loading only is incorrect. BwOA INST-2, OPERATION WITH A
FAILED INSTRUMENT CHANNEL UNIT 1, Attachment E, NARROW RANGE SG LEVEL
CHANNEL FAILURE, Step 2 RNO has actions to reduce turbine load. The examinee may
plausibly conclude the narrow range SG level shown has been caused by a failed instrument
and actions are needed to reduce power to less than 100 percent.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
B - Plausible: Trip unit 1 reactor only is incorrect. The indications provided shows the
containment leak detection flow high alarm in, coupled with the 1A SG level at 54.5 percent
and 106.3 percent feed flow in the 1A SG. These conditions are indicative of a feedline break
in containment, requiring a reactor trip. Incorrect because tripping the reactor, initiating SI and
MSI are all high-level actions to mitigate the event in progress.
C - Plausible: Trip the reactor and initiate SI only is incorrect. The indications provided shows
the containment leak detection flow high alarm in, coupled with the 1A SG level at 54.5 percent
and 106.3 percent feed flow in the 1A SG. These conditions are indicative of a feedline break in
containment, requiring a reactor trip. The examinee may plausibly conclude that only a reactor
trip and SI is required to address this casualty since MSI does not close FWIVs. Incorrect
because tripping the reactor, initiating SI and MSI are all high-level actions to mitigate the event
in progress.
D - Correct: Trip the reactor, SI and MSI is correct. The indications provided shows the
containment leak detection flow high alarm in, coupled with the 1A SG level at 54.5 percent and
106.3 percent feed flow in the 1A SG. These conditions are indicative of a feedline break in
containment. The crew will trip the reactor, initiate SI and main steam isolation (MSI) as
high-level actions to mitigate the event in progress.
Technical Reference and Revision #
1BwEP-2, Revision 300, Page 13.
1BwEP-0, Revision 303, Page 5.
_BwEP-2 Faulted Steam Generator Isolation Lesson Plan (I1-EP-XL-03), Revision 14, Page 6.
Applicant Comment
Facility Position on Applicant Comment
NRC Resolution
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
the examination regarding Question #10. The NRC also notes that, in accordance with
NUREG-1021 ES-403, Section D.3.a and NUREG-1021 ES-501, Section D.2.d, the facilitys
performance analysis indicates that only 25 percent of applicants answered Question #10
incorrectly.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Although the applicant contends that the question stem does not specify a timeframe, this is
irrelevant since the use of the word only in the answer/distractor choices logically results in D
being the only correct answer to the question. Therefore, the NRC concludes that no change
should be made to the key regarding this exam question. This is also consistent with the
recommendation of the facility regarding this specific applicant comment.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #24
Both Units are at 100 percent power.
A fire occurs and the main control room (MCR) requires IMMEDIATE evacuation per 1BwOA
PRI-5, CONTROL ROOM INACCESSIBILITY UNIT 1.
(1) Prior to leaving the MCR the reactor trip ______ be verified.
(2) Pressurizer LEVEL indication at the remote shutdown panel (shown below)
A. (1) will
(2) will NOT require temperature correction.
B. (1) will
(2) WILL require temperature correction utilizing BwCB-1 FIGURE 31,
PRESSURIZER LEVEL.
C. (1) will NOT
(2) will NOT require temperature correction.
D. (1) will NOT
(2) WILL require temperature correction utilizing BwCB-1 FIGURE 31,
PRESSURIZER LEVEL.
Answer A
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Answer Explanation
A - Correct: will and will NOT require temperature correction, are correct. Per the mitigating
strategy of 1BwOA PRI-5 ONLY the reactor trip will be verified when an immediate evacuation
of the MCR is required. Per 1BwGP 100-5 only the cold cal pressurizer level (1LT-462) is
required to be corrected for pressurizer vessel liquid temperature.
B - Plausible: will is correct, WILL require temperature correction is incorrect.
Temperature correction would be correct if the stem asked for monitoring the cold calibrated
pressurizer level channel (1LT-462) during a normal shutdown per 1BwGP 100-5.
C - Plausible: will NOT is incorrect, will NOT require temperature correction is correct.
This would be correct if the stem asked if the turbine trip will be verified during an
immediate evacuation.
D - Plausible: will NOT and WILL require temperature correction are incorrect. This would be
correct if the stem asked if the turbine trip will be verified during an immediate evacuation.
Temperature correction would be correct if the stem asked for monitoring the cold calibrated
pressurizer level channel (1LT-462) during a normal shutdown per 1BwGP 100-5.
Technical Reference and Revision #
1BwOA PRI-5, Revision 109, Page 3
Control Room Inaccessibility (_BwOA PRI-5) Lesson Plan, Revision 8, Page 10
1BwGP 100-5, Revision 58, Page 49
Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Facility Position on Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Additional References
BwCB-1 Figure 31, Pressurizer Level 462 Cold Calibration, Revision 1
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Excerpt from 1BwGP 100-5, Plant Shutdown and Cooldown, Revision 58
NRC Resolution
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
the examination regarding Question #24. The NRC also notes that, in accordance with
NUREG-1021 ES-403, Section D.3.a and NUREG-1021 ES-501, Section D.2.d, the facilitys
performance analysis indicates that only 37.5 percent of applicants answered Question #24
incorrectly.
Regardless of any reference made to BwCB-1 FIGURE 31 in the answer/distractor
combinations, the key wording in the available choices consists of WILL require temperature
correction or will NOT require temperature correction. 1BwGP 100-5 discusses correcting
only the 1LI-462 (the cold calibrated instrument) level indication based upon pressurizer
temperature conditions. Thus, including any reference to BwCB-1 FIGURE 31 was ultimately
unnecessary, since only the understanding that 1LI-462 requires temperature correction during
a cooldown (as specified in 1BwGP 100-5) was necessary to select between the options
presented by the second half of the two-part question. Therefore, the NRC concludes that no
change should be made to the key regarding this exam question.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #37
Unit 2 is at 100 percent power.
2BwOSR 3.3.1.4-1, UNIT TWO SSPS, REACTOR TRIP BREAKER, AND REACTOR TRIP
BYPASS BREAKER SURVEILLANCE (TRAIN A) is in progress.
- The EO has racked the Train A Reactor Trip Bypass Breaker to the TEST position and has
just completed step F.2.2.c, AT 2RD05E CLOSE THE TRAIN A REACTOR TRIP BYPASS
BREAKER (BYA).
Which of the following indications at 2PM05J reflect the current status?
1) 2) 3) 4)
A. 1
B. 2
C. 3
D. 4
Answer C
Answer Explanation
A - Plausible: 1 is incorrect. With the BYA racked to the test position the indicating lights will
have power. However, this configuration shows the RTA as red and BYA as green. This
configuration is opposite of the correct configuration of RTA being green and BYA being red.
The examinee may confuse the indications given and select this answer, since they are
opposite.
B - Plausible: 2 is incorrect. With the BYA racked to the test position the closed indicating lights
will have power. However, this configuration shows RTA as red and BYA as red. The indication
for BYA is correct for the condition in the stem. The indication for RTA is incorrect. This answer
would be correct if performance of 2BwOSR 3.3.1.4-1, Section 3.8.e had occurred.
C - Correct: 3 is correct. With the BYA racked to the test position the indicating lights will have
power and the red indicating light will be lit (BYA closed). This answer shows this configuration.
D - Plausible: 4 is incorrect. With the BYA racked to the test position the indicating lights will
have power. However, the green indicating light will not be lit (indicates BYA is open). The dark
board / green board concept is frequently misunderstood by novice applicants and could cause
them to select this answer. This answer would be correct if the EO had not closed the breaker.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Technical Reference and Revision #
2BwOSR 3.3.1.4-1, Revision 043
Applicant Comment
Facility Position on Applicant Comment
NRC Resolution
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
the examination regarding Question #37. The NRC also notes that, in accordance with
NUREG-1021 ES-403, Section D.3.a and NUREG-1021 ES-501, Section D.2.d, the facilitys
performance analysis indicates that only 12.5 percent of applicants answered Question #37
incorrectly.
The applicant contends that their ability to answer this question was negatively affected by the
print quality of the graphics on the administered examination. To validate this, the NRC was
provided the actual page from the applicants examination containing the graphics in question.
A review of this page indicated the print quality was satisfactory and that the colors used in the
graphics provided sufficient fidelity to those that would appear in the actual plant to allow for
accurate interpretation by applicants. As previously noted, no clarification was requested by
any applicant during the exam for this question and, furthermore, the facilitys performance
analysis indicates that only a single applicant answered this question incorrectly. Therefore, the
NRC concludes that no change should be made to the key regarding this exam question. This
is also consistent with the recommendation of the facility regarding this specific applicant
comment.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #45
Per 1BwGP 100-3, POWER ASCENSION 5 PERCENT TO 100 PERCENT, what is the
approximate steam flow from a SG, when the FW Bypass Reg Valves (1FW510A/520A/530A
and 540A) automatically close?
A. 5%
B. 10%
C. 20%
D. 30%
Answer D
Answer Explanation
A - Plausible: 5 percent is plausible since the unit 2 MFW system allows tempering flow only to
the SG less than 5 percent.
B - Plausible: 10 percent is plausible since this is the max power level the startup feedwater
pump can go to.
C - Plausible: 20 percent is plausible since this is the approximate steam dump demand when
the main generator is synchronized.
D - Correct: This is the power limit 1BwGP 100-3 states as the approximate power level where
feed flow is transferred from the FW Bypass valves to the Main FW Reg Valves.
Technical Reference and Revision #
1BwGP 100-3, Revision 077, Page 16
Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Facility Position on Applicant Comment
Additional References
Simulator Response Data Provided by Facility
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Excerpt from 1BwGP 100-3, Power Ascension 5 percent to 100 percent, Revision 77
NRC Resolution
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
the examination regarding Question #45.
The stem of the question states [p]er 1BwGP 100-3; this focuses the context of the solicited
response to information contained in that procedure 1BwGP 100-3, Section E.4.h, in turn states
that [w]hen the individual steam flow to a steam generator reaches approx. 30 percent (approx.
1.2 MLBM) the Feedwater Reg Bypass valve will get a closed signal transferring control solely
to the Main Feedwater valve. It should be noted that the stem of the question does not ask at
what power level that the FW Bypass Reg Valves begin to throttle; the stem instead asks what
is the approximate steam flow from a SG, when the FW Bypass Reg Valves automatically
close? Thus, the question is clear in asking for the value provided by 1BwGP 100-3 for
automatic Feedwater Reg Bypass valve closure. Therefore, the NRC concludes that no change
should be made to the key regarding this exam question.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #53
Unit 1 is at 100 percent power.
Unit 2 is DEFUELED during a refueling outage.
- The 1A and 2B SX pumps are RUNNING.
- The 1B SX pump is in STANDBY.
- The 2B SX pump TRIPS on overcurrent.
- The 1B SX pump is STARTE
- D.
Conditions of LCO 3.7.8, ESSENTIAL SERVICE WATER SYSTEM, are...
A. NOT met on Unit 1 ONLY.
B. NOT met on Unit 2 ONLY.
C. NOT met on BOTH Unit 1 and Unit 2.
D. MET on BOTH Unit 1 and Unit 2.
Answer A
Answer Explanation
A - Correct: NOT met on Unit 1 ONLY is correct. LCO 3.7.8 requires One opposite-unit SX
train for unit-specific support, the conditions in the stem have both Unit 2 SX pumps inoperable.
B - Plausible: NOT met on Unit 2 ONLY is incorrect. LCO 3.7.8 is only applicable in modes
1-4. This would be the correct answer if the outage and online unit were reversed. The
applicability of LCO 3.7.8, in various plant conditions, is frequently misunderstood by novice
applicants.
C - Plausible: NOT met on BOTH Unit 1 and Unit 2 is incorrect. This would be the correct
answer if Unit 2 was in mode 1-4. The applicability of LCO 3.7.8, in various plant conditions,
is frequently misunderstood by novice applicants.
D - Plausible: MET on BOTH Unit 1 and Unit 2 is incorrect. This would be the correct answer
if only one Unit 2 SX pump were inoperable or if both units were in mode 5. The applicability
of LCO 3.7.8, in various plant conditions, is frequently misunderstood by novice applicants.
Technical Reference and Revision #
TS 3.7.8, Amendment 193
TS 3.7.9, Amendment 189
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Applicant Comment
Facility Position on Applicant Comment
Additional References
Excerpt from Braidwood Technical Specifications:
NRC Resolution
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
the examination regarding Question #53. The NRC also notes that, in accordance with
NUREG-1021 ES-403, Section D.3.a and NUREG-1021 ES-501, Section D.2.d, the facilitys
performance analysis indicates that only 37.5 percent of applicants answered Question #53
incorrectly.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
The applicant contends that distractor C should be a second correct answer. However, the
applicants comment appears to not recognize that Unit 2 (listed as being defueled in the stem
and thus being in no mode) is no longer within the Modes of Applicability of Technical Specification 3.7.8 (e.g., Modes 1, 2, 3, and 4). For distractor C to be correct, LCO 3.7.8
would also need to be applicable to Unit 2; based upon the conditions provided in the stem, it is
not. Therefore, the NRC concludes that no change should be made to the key regarding this
exam question. This is also consistent with the recommendation of the facility regarding this
specific applicant comment.
SIMULATION FACILITY FIDELITY REPORT
Facility Licensee: Braidwood Station
Facility Docket Nos: 50-456 and 50-457
Operating Tests Administered: June 10, 2019, through June 17, 2019
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 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
Simulator Work On June 11, 2019, an issue with the software code for the simulator
Request #135325 resulted in an unplanned reactor trip in the middle of an applicant
operating test scenario, necessitating cancellation of the remaining
Action Request sessions of that scenario for the remainder of the day, as well as the
- 04259276 administration of a spare scenario on a subsequent day. The issue
involved a simulator error created by the way in which the simulator
modeled Chemical Volume Control System flow and boron
concentrations when a Centrifugal Charging Pump experienced a
shaft shear event. The overall effect of this issue was that an
erroneous boron calculation caused the modeling of an extremely
large boron concentration in the discharge flow path of the standby
CV pump. Subsequently, when this standby CV pump was started by
an applicant crew, a rapid Reactor Coolant System temperature and
pressure transient resulted in a reactor trip.
3