IR 05000456/2019301: Difference between revisions

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=Text=
=Text=
{{#Wiki_filter:August 12, 2019 Mr. Bryan Senior VP, Exelon Generation Company, LLC President and CNO, Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 SUBJECT: BRAIDWOOD STATION, UNITS 1 AND 2-NRC INITIAL LICENSE EXAMINATION REPORT 05000456/2019301 AND 05000457/2019301
{{#Wiki_filter:ust 12, 2019
 
==SUBJECT:==
BRAIDWOOD STATION, UNITS 1 AND 2NRC INITIAL LICENSE EXAMINATION REPORT 05000456/2019301 AND 05000457/2019301


==Dear Mr. Hanson:==
==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 examinati on for the license applicants. During the telephone conversation, NRC resolutions of the station's post-examination comments, received by the NRC on June 28, 2019, were discussed.
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 week s 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 license s. 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
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.
. 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."
 
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,
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  
/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:===
===Enclosures:===
1. OL Examination Report 05000456/2019301
1. OL Examination Report 05000456/2019301; 05000457/2019301 2. Post-Examination Comment, Evaluation, and Resolution 3. Simulation Facility Fidelity Report
; 05000457/2019301 2. Post-Examination Comment, Evaluation, and Resolution 3. Simulation Facility Fidelity Report


REGION III==
REGION III==
Docket Nos: 05000456; 05000457 License Nos
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
: 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 B y: R. Edwards , III, Acting Chief Operations Branch Division of Reactor Safety
 
2


=SUMMARY=
=SUMMARY=
Examination Report 05000456/2019301
Examination Report 05000456/2019301; 05000457/2019301; 06/10/2019-06/28/2019;


; 05000457/2019301
Exelon Generation Company, LLC; Braidwood Station, Units 1 and 2; Initial License Examination Report.
; 06/10/2019
-06/28/2019
; Exelon Generation Company, LLC
; Braidwood Station, Units 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 applicant s were issued senior operator licenses.
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.


One applicant failed one or more sections of the administered examination and was issued a preliminary results letter.  (Section 4OA5.1)
Examination Summary Seven of eight applicants passed all sections of their respective examinations.


3
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=
=REPORT DETAILS=
Line 67: Line 61:


====a. Examination Scope====
====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
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.
-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 licensee's staff.
 
Members of the facility licensee's 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 licensee's 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 licensee's 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====
====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.
: (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.


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 9, "Written Examination Review Worksheet."
Less than 20 percent of the proposed examination questions were determined to be unsatisfactory and required modification or replacement.


The Form ES-401-9, 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)on June 19, 2021 , (ADAMS Accession Numbers ML17214A825 , ML17214A828 , ML17214A830, and ML17214A832 , respectively
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, 201 9 , the licensee submitted documentation noting that there were six 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 documented in Enclosure 2 to this report.


The NRC examiners graded the written examination on July 2, 201 9 , and conducted a review of each missed question to determine the accuracy and validity of the examination questions.
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.


4
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.
: (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.
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).


All changes made to the operating test were made in accordance with NUREG
The NRC examiners completed operating test grading on July 16, 2019.
-1021, "Operator Licensing Examination Standards for Power Reactors," and were documented on Form ES 7, "Operating Test Review Worksheet."  The Form ES-301-7, the operating test outlines (ES 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
: (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.
, 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===
===.2 Examination Security===


====a. Scope====
====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.
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====
====b. Findings====
None.  
None.
{{a|4OA6}}
{{a|4OA6}}
==4OA6 Management Meetings==
==4OA6 Management Meetings==


===.1 Debrief===
===.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.
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===
===.2 Exit Meeting===


The chief examiner conducted an exit meeting on July 3, 2019, with Ms. M. Marchionda
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.
-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:
ATTACHMENT:  


=SUPPLEMENTAL INFORMATION=
=SUPPLEMENTAL INFORMATION=


SUPPLEMENTAL INFORMATION KEY POINTS OF CONTAC
SUPPLEMENTAL INFORMATION
T  Licensee M. Marchionda
KEY POINTS OF CONTACT
-Palmer, Site Vice President
Licensee
: [[contact::J. Keenan]], Plant Manager  
: [[contact::M. Marchionda-Palmer]], Site Vice President
: [[contact::J. Keenan]], Plant Manager
: [[contact::F. Jordan]], Training Director
: [[contact::F. Jordan]], Training Director
: [[contact::P. Moodie]], Operations Director
: [[contact::P. Moodie]], Operations Director
Line 133: Line 113:
: [[contact::J. Petty]], Shift Operations Superintendent
: [[contact::J. Petty]], Shift Operations Superintendent
: [[contact::J. Beard]], Operations Training
: [[contact::J. Beard]], Operations Training
: [[contact::D. Brunswick]], Operations Training  
: [[contact::D. Brunswick]], Operations Training
: [[contact::J. Taff]], Operations Training Manager
: [[contact::J. Taff]], Operations Training Manager
: [[contact::R. Schliessmann]], Regulatory Assurance
: [[contact::R. Schliessmann]], Regulatory Assurance
Line 141: Line 121:
: [[contact::R. Edwards]], Branch Chief (Acting)
: [[contact::R. Edwards]], Branch Chief (Acting)
: [[contact::D. Kimble]], Senior Resident Inspector
: [[contact::D. Kimble]], Senior Resident Inspector
: [[contact::J. Seymour]], Operations Engineer , Chief Examiner
: [[contact::J. Seymour]], Operations Engineer, Chief Examiner
: [[contact::C. Zoia]], Senior Operations Engineer, Examiner
: [[contact::C. Zoia]], Senior Operations Engineer, Examiner
: [[contact::E. Cushing]], Reactor Engineer, Examiner
: [[contact::E. Cushing]], Reactor Engineer, Examiner
ITEMS OPENED, CLOSED, AND DISCUSSED
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
Opened, Closed, and Discussed
None   LIST OF ACRONYMS USE
None
D  ADAMS Agencywide Document Access and Management System
LIST OF ACRONYMS USED
NRC U.S. Nuclear Regulatory Commission
ADAMS           Agencywide Document Access and Management System
 
NRC             U.S. Nuclear Regulatory Commission
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #6
Question #6
Unit 2 is in MODE 5.
Unit 2 is in MODE 5.
2BwGP 100-1, PLANT HEATUP, is in progress.
2BwGP 100-1, PLANT HEATUP, is in progress.
RCS temperature is
    *  RCS temperature is 100°F.
100°F. Pressurizer Level is
    *  Pressurizer Level is 50 percent.
percent. The 2A RH train is in shutdown
    *  The 2A RH train is in shutdown cooling.
cooling. The 2B RH train is in
    *  The 2B RH train is in STANDBY.
STANDBY. The 2A RH pump amps are fluctuating between 50 to 60
    *  The 2A RH pump amps are fluctuating between 50 to 60 amps.
amps. The 2A RH pump flow is fluctuating between 4500 to 5000
    *  The 2A RH pump flow is fluctuating between 4500 to 5000 gpm.
gpm. The crew will (1) _, and (2) _ to correct the
The crew will (1) _, and (2) _ to correct the issue.
issue.          Answer Explanation
: [[contact::A.      (1) continue in 2BwGP 100-1]], PLANT HEATUP, ONLY
A - Plausible:
          (2) IMMEDIATELY trip the 2A RH pump to prevent damage
continue in 2BwGP 100
: [[contact::B.      (1) continue in 2BwGP 100-1]], PLANT HEATUP, ONLY
-1 and immediately trip the 2A RH pump are incorrect.
          (2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV,
The 2BwGP 100-1 directs securing all RH trains during the startup.
to reduce flow
A novice applicant may interpret this procedural flow path as adequate to address the RH pump current issue.
: [[contact::C.      (1) enter 2BwOA PRI-10]], LOSS OF RH COOLING UNIT 2
The 2A RH pump amps are fluctuating near the red band.
          (2) IMMEDIATELY trip the 2A RH pump to prevent damage
This would be correct if the 2A RH pump flow was reduced and did not stabilize parameters.
: [[contact::D.       (1) enter 2BwOA PRI-10]], LOSS OF RH COOLING UNIT 2
B - Plausible:
          (2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV,
continue in 2BwGP 100
to reduce flow
-1 is incorrect and Take manual control of 2RH618
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.
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
A novice applicant may interpret this procedural flow path as adequate to address the
RH pump current issue. C - Plausible:
RH pump current issue.
2BwOA PRI-10 is correct, immediately trip the 2A RH pump is incorrect.
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.
This would be correct if the 2A RH pump flow was reduced and did not stabilize parameters.
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) IMMEDIATELY trip the 2A RH pump to prevent
damage  D. (1) enter 2BwOA PRI
-10, LOSS OF RH COOLING UNIT
(2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV, to reduce flow    Answer D
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
D - Correct: 2BwOA PRI-10 and Take manual control of 2RH618 to reduce flow are correct.
D - Correct: 2BwOA PRI-10 and Take manual control of 2RH618 to reduce flow are correct.
Per 2BwOA PRI
Per 2BwOA PRI-10 the first mitigative strategy performed is to reduce RH pump flow. The crew
-10 the first mitigative strategy performed is to reduce RH pump flow.
should attempt to stabilize RH system operation prior to tripping the running RH pump and
The crew should attempt to stabilize RH system operation prior to tripping the running RH pump
continuing to further mitigating actions.
and continuing to further mitigating actions.
Technical Reference and Revision #
Technical Reference and Revision #
2BwOA PRI-10, Revision 107, Page 2
2BwOA PRI-10, Revision 107, Page 2.
. _BwOA PRI-10 Lesson Plan (I1
_BwOA PRI-10 Lesson Plan (I1-OA-XL-20) Revision 13, Page 2.
-OA-XL-20) Revision 13, Page 2.
Applicant Comment
Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Facility Position on Applicant Comment
Facility Position on Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Additional Information on Position Provided by Facility on July 2, 20 19  [T]he expectation would be to attempt to lower RH system flow without exiting 2BwGP 100
Additional Information on Position Provided by Facility on July 2, 2019
-1 first, since the pump is experiencing potential run out conditions.
[T]he expectation would be to attempt to lower RH system flow without exiting 2BwGP 100-1
If that was successful, no further procedure transition would take place.
first, since the pump is experiencing potential run out conditions. If that was successful, no
If not successful, the crew should enter 2BwOA PRI-10. Entry conditions for 2BwOA PRI
further procedure transition would take place. If not successful, the crew should enter 2BwOA
-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
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.
be needed.
The stem did not state "actions taken are not successful".
The stem did not state actions taken are not successful. This creates an unstated assumption
This creates an unstated assumption that the action being taken could correct the issue and therefore no further procedure transitions are required.
that the action being taken could correct the issue and therefore no further procedure transitions
BwOP RH-6 would have been completed prior to where the stem begins the initial conditions for the question.
are required. BwOP RH-6 would have been completed prior to where the stem begins the initial
Therefore, the applicant could utilize the precaution (D.4) from memory (per BwAP 340
conditions for the question. Therefore, the applicant could utilize the precaution (D.4) from
-1) to lower RH system flow and stabilize the plant preventing damage to the RH pump, without transitioning to another procedure.
memory (per BwAP 340-1) to lower RH system flow and stabilize the plant preventing damage
The conditions of the stem are consistent with an excess of RH cooling (RH pump flow fluctuating between 4500
to the RH pump, without transitioning to another procedure. The conditions of the stem are
-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.
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
Additional References
Excerpt from 2BwGP 100
Excerpt from 2BwGP 100-1, Plant Heatup, Revision 39
-1, Plant Heatup, Revision 39
 
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Excerpt from BwOP RH
Excerpt from BwOP RH-6, Placing the RH System in Shutdown Cooling, Revision 59
-6, Placing the RH System in Shutdown Cooling, Revision 59
Excerpts from 2BwOA PRI-10, Loss of RH Cooling Unit 2, Revision 107
Excerpts from 2BwOA PRI
-10, Loss of RH Cooling Unit 2, Revision 107
 
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Excerpt from BwAP 340
Excerpt from BwAP 340-1, Use of Procedures for Operating Department, Revision 30
-1, Use of Procedures for Operating Department, Revision 30
 
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
NRC Resolution
NRC Resolution
The U.S Nuclear Regulatory
The  
Commission (NRC)
: [[contact::U.S Nuclear Regulatory Commission (NRC) notes that]], in accordance with NUREG-1021
notes that, in accordance with NUREG
ES-403, Section D.1.a, and NUREG-1021 Appendix E, Section B.7, no questions were posed
-1021 E S-403, Section D.1.a, and NUREG
by any applicants during the examination regarding Question #6.
-1021 Appendix E, Section B.7, no questions were
The distractor/answer choices addressed by the applicants comment consist of the following:
posed by any applicants during
Distractor B:
the examination
  (1) continue in 2BwGP 100-1, PLANT HEATUP, ONLY
regarding Question #6.
  (2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV, to reduce flow
The distractor/answer choices addressed by the applicant's comment consist of the following:
Answer D:
Distractor B
  (1) enter 2BwOA PRI-10, LOSS OF RH COOLING UNIT 2
:  (1) continue in 2BwGP 100
  (2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV, to reduce flow
-1, PLANT HEATUP, ONLY (2) take manual control of 2RH618, HX 1A BYP FLOW CONT VLV, to
The stem of the question explicitly provided, in part, the following key information to the
reduce flow
applicant:
Answer D: (1) enter 2BwOA PRI
  * 2BwGP 100-1 in effect
-10, LOSS OF RH COOLING UNIT
  * 2A RH train aligned for shutdown cooling
(2) take manual control of 2RH618, HX
  * 2A RH pump amps fluctuating between 50 to 60 amps
1A BYP FLOW CONT VLV, to
Under the above conditions, BwOP RH-6, Placing the RH System in Shutdown Cooling, would
reduce flow
have been previously performed to establish shutdown cooling operations. As stated in the
The stem of the question explicitly provided, in part, the following key information to the applicant:
information provided by the facility, BwOP RH-6 would have been completed prior to where the
2BwGP 100-1 in effect 2A RH train aligned for shutdown
stem begins the initial conditions for the question. Based upon this, the precautions of BwOP
cooling  2A RH pump amps fluctuating between 50 to 60
RH-6 (including step D.4) are no longer procedurally in effect during the timeframe associated
amps  Under the above conditions, BwOP RH
with the stem conditions.
-6, "Placing the RH System in Shutdown Cooling," would have been previously performed
The first half of distractor B states continue in 2BwGP 100-1, PLANT HEATUP, ONLY.
to establish shutdown cooling operations. As stated in the information provided by the facility
The inclusion of the word ONLY in this distractor would limit any procedurally driven corrective
, "BwOP RH-6 would have been completed
actions for addressing the oscillating RH Pump amps to solely guidance contained within
prior to where the stem begins the initial conditions for the question."  Based upon this, the precautions of BwOP
2BwGP 100-1. However, 2BwGP 100-1 does not contain any procedural guidance for
RH-6 (including step D.4) are no longer
addressing conditions of oscillating RH Pump amps. Thus, the corrective action contained in
procedurally in effect during the timeframe associated with the stem conditions.
the second part of distractor B (take manual control of 2RH618, HX 1A BYP FLOW CONT
The first half of distractor "B" states "continue in 2BwGP 100
VLV, to reduce flow) would not have a procedural basis within the context of this distractor.
-1, PLANT HEATUP, ONLY."  The inclusion of the word "ONLY" in this distractor would limit
In contrast, 2BwOA PRI-10, Loss of RH Cooling Unit 2, lists the symptom of oscillating
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
RH Pump amps occurring as an entry condition. Based upon this, it would be appropriate
to enter this
to enter this Procedure 2BwOA PRI-10, RNO step 1.b, subsequently directs the required
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.
reduction in RH pump flow. Thus, answer D provides a procedurally directed means of
Therefore, the NRC concludes that no change should be made to the key regarding this exam question.
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
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #10
Question #10
Unit 1 is at 100
Unit 1 is at 100 percent power.
percent power. The following annunciators have just alarmed:
The following annunciators have just alarmed:
1-1-A2, CNMT DRAIN LEAK DETECT FLOW
  * 1-1-A2, CNMT DRAIN LEAK DETECT FLOW HIGH
HIGH  1-10-E4, OVATION SYSTEM
  * 1-10-E4, OVATION SYSTEM TROUBLE
TROUBLE  1-10-E5, OVATION ALTERNATE
  * 1-10-E5, OVATION ALTERNATE ACTION
ACTION  The RO reviews OWS graphic 6040, FW OVERVIEW, and notes the following:
The RO reviews OWS graphic 6040, FW OVERVIEW, and notes the following:
The crew will...
The crew will...
1. Reduce Unit 1 Turbine
1. Reduce Unit 1 Turbine Loading
Loading 2. Trip Unit 1 Reactor
2. Trip Unit 1 Reactor
3. Initiate Safety
3. Initiate Safety Injection
Injection 4. Actuate Main Steamline
4. Actuate Main Steamline Isolation
Isolation         Answer Explanation
A.       1 ONLY.
A - Plausible:
B.       2 ONLY.
Reduce turbine loading only is incorrect.
C.       2 and 3 ONL
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. A. 1 ONLY. B. 2 ONLY. C. 2 and 3 ONL
: [[contact::Y.
: [[contact::Y.
D. 2]], 3, and 4.
D.       2]], 3, and 4.
Answer D
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
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
B - Plausible:
B - Plausible: Trip unit 1 reactor only is incorrect. The indications provided shows the
Trip unit 1 reactor only is incorrect.
containment leak detection flow high alarm in, coupled with the 1A SG level at 54.5 percent
The indications provided shows the containment leak detection flow high alarm in, coupled with the 1A SG level at 54.5
and 106.3 percent feed flow in the 1A SG. These conditions are indicative of a feedline break
percent and 106.3 percent feed flow in the 1A SG.
in containment, requiring a reactor trip. Incorrect because tripping the reactor, initiating SI and
These conditions are indicative of a feedline break
MSI are all high-level actions to mitigate the event in progress.
in containment, requiring a reactor trip.
C - Plausible: Trip the reactor and initiate SI only is incorrect. The indications provided shows
Incorrect because tripping the reactor, initiating SI and MSI are all high
the containment leak detection flow high alarm in, coupled with the 1A SG level at 54.5 percent
-level actions to mitigate the event in progress.
and 106.3 percent feed flow in the 1A SG. These conditions are indicative of a feedline break in
C - Plausible:
containment, requiring a reactor trip. The examinee may plausibly conclude that only a reactor
Trip the reactor and initiate SI only is incorrect.
trip and SI is required to address this casualty since MSI does not close FWIVs. Incorrect
The indications provided shows the containment leak detection flow high alarm in, coupled with the 1A SG level at 54.5
because tripping the reactor, initiating SI and MSI are all high-level actions to mitigate the event
percent and 106.3 percent feed flow in the 1A SG
in progress.
. These conditions are indicative of a feedline break in containment, requiring a reactor trip.
D - Correct: Trip the reactor, SI and MSI is correct. The indications provided shows the
The examinee may plausibly conclude that only a reactor trip and SI is required to address this casualty since MSI does not close FWIVs.
containment leak detection flow high alarm in, coupled with the 1A SG level at 54.5 percent and
Incorrect because tripping the reactor, initiating SI and MSI are all high
106.3 percent feed flow in the 1A SG. These conditions are indicative of a feedline break in
-level actions to mitigate the event in progress.
containment. The crew will trip the reactor, initiate SI and main steam isolation (MSI) as
D - Correct: Trip the reactor, SI and MSI is correct.
high-level actions to mitigate the event in progress.
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 #
Technical Reference and Revision #
1BwEP-2, Revision 300, Page 13
1BwEP-2, Revision 300, Page 13.
. 1BwEP-0, Revision 303, Page 5
1BwEP-0, Revision 303, Page 5.
. _BwEP-2 Faulted Steam Generator Isolation Lesson Plan (I1-EP-XL-03), Revision 14, Page 6.
_BwEP-2 Faulted Steam Generator Isolation Lesson Plan (I1-EP-XL-03), Revision 14, Page 6.
Applicant Comment
Applicant Comment
Facility Position on Applicant Comment
Facility Position on Applicant Comment
NRC Resolution
NRC Resolution
The NRC notes that, in accordance with NUREG
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
-1021 ES-403, Section D.1.a, and
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
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
the examination regarding Question #10. The NRC also notes that, in accordance with
-1021 ES-501, Section D.2.d, the facility's performance analysis indicates that only 25
NUREG-1021 ES-403, Section D.3.a and NUREG-1021 ES-501, Section D.2.d, the facilitys
percent of applicants answered Question #10 incorrectly.
performance analysis indicates that only 25 percent of applicants answered Question #10
 
incorrectly.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
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.
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
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #24
Question #24
Both Units are at 100
Both Units are at 100 percent power.
percent power. A fire occurs and the main control room (MCR)
A fire occurs and the main control room (MCR) requires IMMEDIATE evacuation per 1BwOA
requires IMMEDIATE evacuation per 1BwOA PRI-5, CONTROL ROOM INACCESSIBILITY UNIT 1.
PRI-5, CONTROL ROOM INACCESSIBILITY UNIT 1.
  (1) Prior to leaving the MCR the reactor trip ______ be verified.
(1) Prior to leaving the MCR the reactor trip ______ be verified.
(2) Pressurizer LEVEL indication at the remote shutdown panel (shown below)  
(2) Pressurizer LEVEL indication at the remote shutdown panel (shown below)
-    A. (1) will (2) will NOT require temperature
A.     (1) will
correction.
                    (2) will NOT require temperature correction.
B. (1) will (2) WILL require temperature correction utilizing BwCB
B.     (1) will
-1 FIGURE 31, PRESSURIZER
                    (2) WILL require temperature correction utilizing BwCB-1 FIGURE 31,
LEVEL. C. (1) will NOT (2) will NOT require temperature
PRESSURIZER LEVEL.
correction.
C.     (1) will NOT
D. (1) will NOT (2) WILL require temperature correction utilizing BwCB
                    (2) will NOT require temperature correction.
-1 FIGURE 31, PRESSURIZER
D.     (1) will NOT
LEVEL.       Answer A
                    (2) WILL require temperature correction utilizing BwCB-1 FIGURE 31,
PRESSURIZER LEVEL.
Answer         A
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Answer Explanation
Answer Explanation
A - Correct: will and will NOT require temperature correction, are correct.
A - Correct: will and will NOT require temperature correction, are correct. Per the mitigating
Per the mitigating strategy of 1BwOA PRI
strategy of 1BwOA PRI-5 ONLY the reactor trip will be verified when an immediate evacuation
-5 ONLY the reactor trip will be verified when an immediate evacuation of the MCR is required.
of the MCR is required. Per 1BwGP 100-5 only the cold cal pressurizer level (1LT-462) is
Per 1BwGP 100
required to be corrected for pressurizer vessel liquid temperature.
-5 only the cold cal pressurizer level (1LT
B - Plausible: will is correct, WILL require temperature correction is incorrect.
-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
Temperature correction would be correct if the stem asked for monitoring the cold calibrated
pressurizer level channel (1LT
pressurizer level channel (1LT-462) during a normal shutdown per 1BwGP 100-5.
-462) during a normal shutdown per 1BwGP 100
C - Plausible: will NOT is incorrect, will NOT require temperature correction is correct.
-5. Technical Reference and Revision #
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
1BwOA PRI-5, Revision 109, Page 3
Control Room Inaccessibility (_BwOA PRI
Control Room Inaccessibility (_BwOA PRI-5) Lesson Plan, Revision 8, Page 10
-5) Lesson Plan, Revision 8, Page 10
1BwGP 100-5, Revision 58, Page 49
1BwGP 100-5, Revision 58, Page 49
Applicant Comment
Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Facility Position on Applicant Comment
Facility Position on Applicant Comment
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Additional References
Additional References
BwCB-1 Figure 31, Pressurizer Level 462 Cold Calibration, Revision 1
BwCB-1 Figure 31, Pressurizer Level 462 Cold Calibration, Revision 1
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Excerpt from 1BwGP 100
Excerpt from 1BwGP 100-5, Plant Shutdown and Cooldown, Revision 58
-5, Plant Shutdown and Cooldown, Revision 58
NRC Resolution
NRC Resolution
The NRC notes that, in accordance with NUREG
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
-1021 E S-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
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
-1021 ES-501, Section D.2.d, the facility's performance analysis indicates that only 37.5
the examination regarding Question #24. The NRC also notes that, in accordance with
percent of applicants answered Question #24
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.
incorrectly.
Regardless of any reference made to "BwCB
Regardless of any reference made to BwCB-1 FIGURE 31 in the answer/distractor
-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
combinations, the key wording in the available choices consists of WILL require temperature
-1 FIGURE 31 was ultimately unnecessary, since only the understanding that 1LI-462 requires temperature correction during a cooldown (as specified in 1BwGP 100
correction or will NOT require temperature correction. 1BwGP 100-5 discusses correcting
-5) was necessary to select between the options presented by the second half of the two
only the 1LI-462 (the cold calibrated instrument) level indication based upon pressurizer
-part question. Therefore, the NRC concludes that no change should be made to the key regarding this exam question.
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
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #37
Question #37
Unit 2 is at 100
Unit 2 is at 100 percent power.
percent power. 2BwOSR 3.3.1.4
2BwOSR 3.3.1.4-1, UNIT TWO SSPS, REACTOR TRIP BREAKER, AND REACTOR TRIP
-1, UNIT TWO SSPS, REACTOR TRIP BREAKER, AND REACTOR TRIP BYPASS BREAKER SURVEILLANCE (TRAIN A) is in progress.
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?
*    The EO has racked the Train A Reactor Trip Bypass Breaker to the TEST position and has
1)
just completed step F.2.2.c, AT 2RD05E CLOSE THE TRAIN A REACTOR TRIP BYPASS
2)                   3)                   4)
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
Answer Explanation
A - Plausible:
A - Plausible: 1 is incorrect. With the BYA racked to the test position the indicating lights will
is incorrect.
have power. However, this configuration shows the RTA as red and BYA as green. This
With the BYA racked to the test position the indicating lights will have power.
configuration is opposite of the correct configuration of RTA being green and BYA being red.
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
The examinee may confuse the indications given and select this answer, since they are
opposite. B - Plausible:
opposite.
is incorrect.
B - Plausible: 2 is incorrect. With the BYA racked to the test position the closed indicating lights
With the BYA racked to the test position the closed indicating lights will have power.
will have power. However, this configuration shows RTA as red and BYA as red. The indication
However, this configuration shows RTA as red and BYA as red.
for BYA is correct for the condition in the stem. The indication for RTA is incorrect. This answer
The indication for BYA is correct for the condition in the stem.
would be correct if performance of 2BwOSR 3.3.1.4-1, Section 3.8.e had occurred.
The indication for RTA is incorrect.
C - Correct: 3 is correct. With the BYA racked to the test position the indicating lights will have
This answer would be correct if performance of 2BwOSR 3.3.1.4
power and the red indicating light will be lit (BYA closed). This answer shows this configuration.
-1, Section 3.8.e had occurred.
D - Plausible: 4 is incorrect. With the BYA racked to the test position the indicating lights will
C - Correct: 3 is correct.
have power. However, the green indicating light will not be lit (indicates BYA is open). The dark
With the BYA racked to the test position the indicating lights will have power and the red indicating light will be lit (BYA closed).
board / green board concept is frequently misunderstood by novice applicants and could cause
This answer shows this configuration.
them to select this answer. This answer would be correct if the EO had not closed the breaker.
D - Plausible:
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.
A. 1  B. 2  C. 3  D. 4  Answer C
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Technical Reference and Revision #
Technical Reference and Revision #
2BwOSR 3.3.1.4
2BwOSR 3.3.1.4-1, Revision 043
-1, Revision 043 Applicant Comment
Applicant Comment
Facility Position on Applicant Comment
Facility Position on Applicant Comment
NRC Resolution
NRC Resolution
The NRC notes that, in accordance with NUREG
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
-1021 E S-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
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
-1021 ES-501, Section D.2.d, the facility's performance analysis indicates that only 12.5
the examination regarding Question #37. The NRC also notes that, in accordance with
percent of applicants
NUREG-1021 ES-403, Section D.3.a and NUREG-1021 ES-501, Section D.2.d, the facilitys
answered Question #37 incorrectly.
performance analysis indicates that only 12.5 percent of applicants answered Question #37
The applicant contends that their ability to answer this question was negatively affected by the print quality of the graphics on the administered examination.
incorrectly.
To validate this, the NRC was provided the actual page from the applicant's examination containing the graphics in question. A review of this page indicated the print quality was satisfactory and
The applicant contends that their ability to answer this question was negatively affected by the
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 facility's performance analysis indicates that only a single applicant answered this question incorrectly.
print quality of the graphics on the administered examination. To validate this, the NRC was
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.
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
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #45
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?
Per 1BwGP 100-3, POWER ASCENSION 5 PERCENT TO 100 PERCENT, what is the
D. 30%   Answer Explanation
approximate steam flow from a SG, when the FW Bypass Reg Valves (1FW510A/520A/530A
A - Plausible:
and 540A) automatically close?
percent is plausible since the unit 2 MFW system allows tempering flow only to the SG less than 5
A.      5%
percent. B - Plausible:
B.      10%
percent is plausible since this is the max power level the startup feedwater pump can go to.
C.      20%
C - Plausible:
D.     30%
percent is plausible since this is the approximate steam dump demand when the main generator is synchronized.
Answer          D
D - Correct: This is the power limit 1BwGP 100
Answer Explanation
-3 states as the approximate power level where
A - Plausible: 5 percent is plausible since the unit 2 MFW system allows tempering flow only to
feed flow is transferred from the FW
the SG less than 5 percent.
Bypass valves to the Main FW Reg Valves.
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 #
Technical Reference and Revision #
1BwGP 100-3, Revision 077, Page 16
1BwGP 100-3, Revision 077, Page 16
Applicant Comment
Applicant Comment
A. 5%  B. 10%  C. 20%  Answer D
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Facility Position on Applicant Comment
Facility Position on Applicant Comment
Additional References
Additional References
Simulator Response Data Provided by Facility
Simulator Response Data Provided by Facility
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Excerpt from 1BwGP 100
Excerpt from 1BwGP 100-3, Power Ascension 5 percent to 100 percent, Revision 77
-3, Power A
scension 5 percent to 100 percent, Revision 77
NRC Resolution
NRC Resolution
The NRC notes that, in accordance with NUREG
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
-1021 E S-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.
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
The stem of the question states "[p]er 1BwGP 100
the examination regarding Question #45.
-3-"; this focuses the context of the solicited response to information contained in that procedure
The stem of the question states [p]er 1BwGP 100-3; this focuses the context of the solicited
1BwGP 100-3, Section E.4.h, in turn states that "[w]hen the individual steam flow to a steam generator reaches approx. 30
response to information contained in that procedure 1BwGP 100-3, Section E.4.h, in turn states
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
that [w]hen the individual steam flow to a steam generator reaches approx. 30 percent (approx.
-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.
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
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question #53
Question #53
Unit 1 is at 100
Unit 1 is at 100 percent power.
percent power. Unit 2 is DEFUELED during a refueling outage.
Unit 2 is DEFUELED during a refueling outage.
The 1A and 2B SX pumps are
    *  The 1A and 2B SX pumps are RUNNING.
RUNNING. The 1B SX pump is in
    *  The 1B SX pump is in STANDBY.
STANDBY. The 2A SX pump is OOS for
    *  The 2A SX pump is OOS for maintenance.
maintenance.
    *  The 2B SX pump TRIPS on overcurrent.
The 2B SX pump TRIPS on
    *  The 1B SX pump is STARTE
overcurrent.
: [[contact::D.
The 1B SX pump is
Conditions of LCO 3.7.8]], ESSENTIAL SERVICE WATER SYSTEM, are...
STARTE
A.     NOT met on Unit 1 ONLY.
: [[contact::D. Conditions of LCO 3.7.8]], ESSENTIAL SERVICE WATER SYSTEM, are...
B.     NOT met on Unit 2 ONLY.
A. NOT met on Unit 1 ONLY.
C.     NOT met on BOTH Unit 1 and Unit 2.
B. NOT met on Unit 2 ONLY.
D.     MET on BOTH Unit 1 and Unit 2.
C. NOT met on BOTH Unit 1 and Unit 2.
Answer        A
D. MET on BOTH Unit 1 and Unit 2.
Answer Explanation
Answer Explanation
A - Correct: NOT met on Unit 1 ONLY is correct.
A - Correct: NOT met on Unit 1 ONLY is correct. LCO 3.7.8 requires One opposite-unit SX
LCO 3.7.8 requires  
train for unit-specific support, the conditions in the stem have both Unit 2 SX pumps inoperable.
"One opposite
B - Plausible: NOT met on Unit 2 ONLY is incorrect. LCO 3.7.8 is only applicable in modes
-unit SX train for unit
1-4. This would be the correct answer if the outage and online unit were reversed. The
-specific support
applicability of LCO 3.7.8, in various plant conditions, is frequently misunderstood by novice
", the conditions in the stem have both Unit 2 SX pumps inoperable.
applicants.
B - Plausible:
C - Plausible: NOT met on BOTH Unit 1 and Unit 2 is incorrect. This would be the correct
NOT met on Unit 2 ONLY is incorrect.
answer if Unit 2 was in mode 1-4. The applicability of LCO 3.7.8, in various plant conditions,
LCO 3.7.8 is only applicable in modes
is frequently misunderstood by novice applicants.
1-4. This would be the correct answer if the outage and online unit were reversed.
D - Plausible: MET on BOTH Unit 1 and Unit 2 is incorrect. This would be the correct answer
The applicability of LCO 3.7.8, in various plant conditions, is frequently misunderstood by novice applicants.
if only one Unit 2 SX pump were inoperable or if both units were in mode 5. The applicability
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.
of LCO 3.7.8, in various plant conditions, is frequently misunderstood by novice applicants.
Technical Reference and Revision #
Technical Reference and Revision #
TS 3.7.8, Amendment 193 TS 3.7.9, Amendment 189
TS 3.7.8, Amendment 193
Answer A
TS 3.7.9, Amendment 189
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Applicant Comment
Applicant Comment
Facility Position on Applicant Comment
Facility Position on Applicant Comment
Additional References
Additional References
Excerpt from Braidwood Technical Specifications
Excerpt from Braidwood Technical Specifications:
:    NRC Resolution
NRC Resolution
The NRC notes that, in accordance with NUREG
The NRC notes that, in accordance with NUREG-1021 ES-403, Section D.1.a, and
-1021 E S-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
NUREG-1021 Appendix E, Section B.7, no questions were posed by any applicants during
-1021 ES-501, Section D.2.d, the facility's performance analysis indicates that only 37.5
the examination regarding Question #53. The NRC also notes that, in accordance with
percent of applicants answered Question #53 incorrectly.
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
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
The applicant contends that distractor "C" should
The applicant contends that distractor C should be a second correct answer. However, the
be a second correct answer. However, the applicant's 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.
applicants comment appears to not recognize that Unit 2 (listed as being defueled in the stem
, 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.
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
SIMULATION FACILITY FIDELITY REPORT
Facility Licensee:
Facility Licensee:                     Braidwood Station
Braidwood Station Facility Docket No
Facility Docket Nos:                   50-456 and 50-457
s:   50-456 and 50-457 Operating Tests Administered:
Operating Tests Administered:           June 10, 2019, through June 17, 2019
June 10, 2019, through June
The following documents observations made by the U.S Nuclear Regulatory Commission
17, 201 9 The following documents observations made by the U.S Nuclear Regulatory
examination team during the initial operator license examination. These observations do
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,
not constitute audit or inspection findings and are not, without further verification and review, indicative of non
indicative of non-compliance with Title 10 of the Code of Federal Regulations 55.45(b). These
-compliance with Title
observations do not affect U.S. Nuclear Regulatory Commission certification or approval of the
of the Code of Federal Regulations 55.45(b). These observations do not affect U.S. Nuclear Regulatory
simulation facility other than to provide information which may be used in future evaluations.
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.
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 the conduct of the simulator portion of the operating tests, the following items were
Simulator Work Request #135325
observed:
Action Request
ITEM                                         DESCRIPTION
#04259276 On June 11, 2019, an issue with the software code for the simulator resulted in an unplanned reactor trip in the middle of an applicant operating test scenario, necessitating cancellation of the remaining sessions of that scenario for the remainder of the day, as well as the
Simulator Work         On June 11, 2019, an issue with the software code for the simulator
administration of a "spare" scenario on a subsequent day.
Request #135325        resulted in an unplanned reactor trip in the middle of an applicant
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
operating test scenario, necessitating cancellation of the remaining
experienced a shaft shear event.
Action Request        sessions of that scenario for the remainder of the day, as well as the
The overall effect of this issue was that an erroneous boron calculation caused the modeling of an
      #04259276          administration of a spare scenario on a subsequent day. The issue
extremely large boron concentration
involved a simulator error created by the way in which the simulator
in the discharge flow path of the standby
modeled Chemical Volume Control System flow and boron
CV pump. Subsequently, when this
concentrations when a Centrifugal Charging Pump experienced a
standby CV
shaft shear event. The overall effect of this issue was that an
pump was started by an applicant crew, a rapid Reactor Coolant System temperature and pressure transient
erroneous boron calculation caused the modeling of an extremely
resulted in a reactor trip
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
}}
}}

Latest revision as of 10:29, 18 December 2019

NRC Initial License Examination Report 05000456/2019301 and 05000457/2019301
ML19224C181
Person / Time
Site: Braidwood  Constellation icon.png
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)
v • d • e


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 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 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 2A SX pump is OOS for maintenance.
  • 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

  1. 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

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