IR 05000237/2020301: Difference between revisions

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=Text=
=Text=
{{#Wiki_filter:ust 27, 2020
{{#Wiki_filter:
 
==SUBJECT:==
DRESDEN NUCLEAR POWER STATIONNRC INITIAL LICENSE EXAMINATION REPORT 05000237/2020301; 05000249/2020301
 
==Dear Mr. Hanson:==
On July 13, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed the initial operator licensing examination process for license applicants employed at your Dresden Nuclear Power Station, Units 2 and 3. The enclosed report documents the results of those examinations. Preliminary observations noted during the examination process were discussed on June 22, 2020, with you and other members of your staff. An exit meeting was conducted by telephone on July 30, 2020, between Mr. D. Thomas of your staff, and Mr. C. Zoia, Senior Operator Licensing Examiner, to review the proposed final grading of the written examination for the license applicants. We also discussed the NRCs resolution of the stations post-examination comments, received by the NRC on July 13, 2020.
 
The NRC examiners administered an initial license examination operating test during the week of June 15, 2020. The written examination was administered by Dresden Nuclear Power Station training department personnel on June 22, 2020. Six Senior Reactor Operator applicants were administered license examinations. The results of the examinations were finalized on July 31, 2020. Five applicants passed all sections of their respective examinations and were issued a senior operator license. One applicant failed one or more sections of the administered examination and was issued a preliminary results letter.
 
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 July 13, 2022. However, because an applicant received a preliminary results letter due to receiving a non-passing grade on the written examination, the applicant was provided copies of the written examination material. For examination security purposes, your staff should consider the written examination material uncontrolled and exposed to the public. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations, Part 2.390, Public Inspections, Exemptions, Requests for Withholding.
 
Sincerely, Patricia J. Digitally signed by Patricia J. Pelke Pelke Date: 2020.08.27 10:26:36 -05'00'
Patricia J. Pelke, Chief Operations Branch Division of Reactor Safety Docket Nos. 50-237; 50-249 License Nos. DPR-19; DPR-25
 
===Enclosures:===
1. OL Examination Report 05000237/2020301; 05000249/2020301 2. Post-Examination Comment, Evaluation, and Resolution 3. Simulation Facility Fidelity Report
 
REGION III==
Docket Nos: 50-237; 50-249 License Nos: DPR-19; DPR-25 Report No: 05000237/2020301; 05000249/2020301 Enterprise Identifier: L-2020-OLL-0039 Licensee: Exelon Generation Company, LLC Facility: Dresden Nuclear Power Station, Units 2 and 3 Location: Morris, IL Dates:  June 15, 2020, through July 13, 2020 Inspectors: C. Zoia, Senior Operations Engineer, Chief Examiner G. Roach, Senior Operations Engineer, Examiner R. Baker, Senior Operations Engineer, Examiner Approved By: P. Pelke, Chief Operations Branch Division of Reactor Safety Enclosure 1
 
=SUMMARY=
Examination Report 05000237/2020301; 05000249/2020301; 06/15/2020-07/13/2020;
 
Exelon Generation Company, LLC; Dresden Nuclear Power Station, Units 2 and 3;
Initial License Examination Report.
 
The announced initial operator licensing examination was conducted by 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 Five of six applicants passed all sections of their respective examinations. Five 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=
 
{{a|4OA5}}
==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 developed 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 18, 2020, with the assistance of members of the facility licensees staff. During the on-site validation week, the examiners audited two 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 week of June 15, 2020. The facility licensee administered the written examination on June 22, 2020.
 
====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 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), will be temporarily withheld from public disclosure per your request (ADAMS Accession Numbers ML19121A247, ML19121A250, ML19121A251, and ML19121A252, respectively).
 
On July 13, 2020, the licensee submitted documentation noting that there were four 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 28, 2020, and conducted a review of each missed question to determine the accuracy and validity of the examination questions.
: (2) Operating Test The NRC examiners determined that the operating test, 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, with no exam files to be temporarily withheld from public disclosure per your request (ADAMS Accession Numbers ML19121A247, ML19121A250, ML19121A251, and ML19121A252, respectively).
 
The NRC examiners completed operating test grading on July 30, 2020.
: (3) Examination Results Six applicants at the Senior Reactor Operator level were administered written examinations and operating tests. The results of the examinations were finalized on July 31, 2020. Five applicants passed all sections of their respective examinations and were issued a senior operator license. 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 Title 10 of the Code of Federal Regulations, Part 55.49, Integrity of Examinations and Tests. The examiners used the guidelines provided in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, to determine acceptability of the licensees examination security activities.
 
====b. Findings====
None.
{{a|4OA6}}
==4OA6 Management Meetings==
 
===.1 Debrief===
 
The NRC examiners presented the examination teams preliminary observations and findings on June 22, 2020, to Mr. P. Karaba, Site Vice President, and other members of the Dresden Nuclear Power Station staff, by telephone.
 
===.2 Exit Meeting===
 
The chief examiner conducted an exit meeting on July 30, 2020, between Mr. D. Thomas, Training Director, and other members of the Dresden Nuclear Power Station staff, by telephone. The NRCs final disposition of the stations post-examination comments were disclosed and discussed. The examiners asked the licensee whether any of the material used to develop or administer the examination should be considered proprietary. No proprietary or sensitive information was identified during the examination or debrief/exit meetings.
 
ATTACHMENT:
 
=SUPPLEMENTAL INFORMATION=
 
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
: [[contact::P. Karaba]], Site Vice President
: [[contact::P. Boyle]], Plant Manager
: [[contact::D. Thomas]], Training Director
: [[contact::R. Bauman]], Operations Director
: [[contact::M. Condreay]], Manager Operations Training
: [[contact::D. Siuda]], Senior Operations Training Instructor
: [[contact::D. Heyn]], Senior Operations Training Instructor
: [[contact::F. Winter]], Senior Operations Training Instructor
: [[contact::D. Walker]], Senior Regulatory Specialist
: [[contact::J. Van Fleet]], Maintenance Director
: [[contact::W. Remiasz]], Director Organizational Performance & Regulatory
: [[contact::M. McCormick]], Shift Manager
: [[contact::H. Patel]], Shift Operations Superintendent
U.S. Nuclear Regulatory Commission
: [[contact::C. Zoia]], Chief Examiner
: [[contact::G. Roach]], Examiner
: [[contact::R. Baker]], Examiner
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, 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 23
The Operations department is working 8-hour shifts.
You are performing APPENDIX A on U2 for SHIFT 2.
To ensure required Tech Spec, TRM, and ODCM required surveillance intervals are met,
complete the required surveillance checks per this Appendix by ___(1)___. You must notify
the Unit Supervisor ___(2)___ this requirement is NOT met.
A.    (1) 1100
      (2) IF
B.    (1) 1100
      (2) BEFORE
C.    (1) 1500
      (2) IF
D.    (1) 1500
      (2) BEFORE
Answer:      B
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Applicant Comment (55-74856):
The question asks the time interval for completion of APPENDIX A in order to meet the Tech
Spec, TRM, and ODCM surveillance requirements, and the appropriate time to notify the Unit
Supervisor in the event that the requirement is NOT met.
It is recommended to consider both A and B to be correct answers because general instruction
A.2 from Attachment A of Appendix A specifically states IF any limit is exceeded OR Tech Spec
required surveillance can NOT be completed, then notify the Unit Supervisor. Additionally,
general instruction A.2 from Attachment A of Appendix A states to ensure that the Tech Spec,
TRM, and ODCM required surveillance intervals are met, the surveillance checks must be
completed within the first half of operating shift, and the Unit Supervisor be notified BEFORE
this requirement is NOT met.
Therefore, both the key answer, B, (i.e., (1) 1100, (2) BEFORE) and, A, (i.e., (1) 1100, (2) IF)
are correct.
Facility Position on Applicant Comment:
The question grading for the exam should not change. Per Operations review, the facility
has determined that A.1 and A.2 address different aspects of Surveillance Requirements,
and that the question is accurate as written. General Instruction A.1 addresses requirements
for completing Tech Spec, TRM and ODCM required surveillances within the first half of the
operating shift or notifying the Unit Supervisor BEFORE this requirement is not met (which
would by 11 am, per the stem of the question). General Instruction A.2 addresses notifying the
Unit Supervisor if any Tech Spec required surveillance can NOT be completed, or if a Tech
Spec required limit is found to be exceeded.
NRC Evaluation/Resolution:
First, the question asked the applicant to recall APPENDIX A surveillance interval requirements
for completing Tech Spec, TRM and ODCM required surveillances for given conditions. The
applicant, the facility, and the answer key all agreed that 1100, halfway through shift 2, was
the correct answer per General Instruction A.1 of APPENDIX A.
The second half of this 2-part question asked the applicant to recall the action required for
notifying the Unit Supervisor ____ this requirement is NOT met. General Instruction A.1 of
APPENDIX A addressed this with the additional requirement to, Notify the Unit Supervisor
BEFORE this requirement is NOT met. Both the facility and the answer key supported the
answer BEFOR
: [[contact::E. The applicant]], however, chose the answer IF, citing General Instruction
A.2 of APPENDIX A, which stated, IF any limit is exceeded OR any Tech Spec required
surveillance can NOT be completed, THEN notify the Unit Supervisor, and recommended that
both answers be considered correct.
The A.1 requirement was intended to ensure surveillance intervals were met by requiring
surveillances to be performed within the first half of the shift that they are due or notifying the
Unit Supervisor beforehand when the early completion requirement was not expected to be met.
In contrast, the A.2 requirement addressed those notifications required if Tech Spec limits were
exceeded or if the required surveillances could not be completed (as opposed to not being
completed early). Nothing in the stem of the question implied addressing the situation where
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
the required surveillances could not be performed or had exceeded Tech Spec limits, and
thereby invoking General Instruction A.2 requirements.
It was incorrect to make these assumptions per NUREG-1021, Appendix E (Part B.7, Written
Exam Guidelines). Accordingly, the only correct choice applicable to the given stem conditions
was choice B, ((1) 1100, (2) BEFORE). Therefore, the U.S. Nuclear Regulatory Commission
(NRC) concluded that choice B, as annotated on the answer key, was the only correct answer,
and the question was considered acceptable as administered.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question 34
Unit 2 is at rated power with a normal electrical line-up.
    *  The U2 EDG is being synchronized to Bus 24-1, per DOS 6600-01, DIESEL
GENERATOR SURVEILLANCE TEST
: [[contact::S.
In order to line up the U2 EDG to Bus 24-1]], the operator would ensure that:
1) INCOMING voltage is slightly higher than RUNNING voltage to prevent_____(1)_____ .
2) To minimize the potential for motorizing the EDG, the EDG output breaker is closed when
the synchroscope is at approximately the twelve (12) o'clock position, rotating approximately
one (1) revolution every 30 seconds in the _____(2)_____ direction.
A.      (1) inductive power loading on the EDG
        (2) fast
B.      (1) inductive power loading on the EDG
        (2) slow
C.      (1) capacitive power loading on the EDG
        (2) fast
D.      (1) capacitive power loading on the EDG
        (2) slow
Answer:          C
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Applicant Comment (55-74856):
The question asks the proper method to line up the Emergency Diesel Generator (EDG) to its
respective Bus per DOA 6600-1, Diesel Generator Surveillance Test.
It is recommended to consider both, C, (i.e., (1) capacitive power loading on EDG (2) fast), and,
A, (i.e., (1) inductive power loading on the EDG (2) fast) to be correct answers.
The Note on page 36 of DOS 6600-01 states that when synchronizing the EDG to the Bus, the
Synchroscope should rotate one revolution in approximately 30 seconds in the FAST direction.
The breaker should be closed just before the pointer reaches the vertical position. Incoming
voltage should be SLIGHTLY higher than the running voltage. These conditions will help
prevent high transient current in the generator or a reverse power trip to allow the oncoming
generator to generate a small amount of reactive power, thus not weakening the generator field.
Chapter 5 of BWR Generic Fundamentals Components, Motors and Generators defines
Reactive Power as the power consumed in an AC circuit because of the expansion and
contraction of magnetic (inductive) and electrostatic (capacitive) fields, which is expressed in
volt-amperes-reactive (VAR). Lagging Power Factor is indicative of purely inductive loads such
as motors and Leading Power Factor is indicative of purely capacitive loads.
With positive VARs (Lagging Power Factor), the D/G is only SLIGHTLY inductive when closing
the breaker to the respective Bus. This is to ensure that D/G is neither overexcited (inductive)
nor under excited (capacitive), which would result in EDG overloading and subsequent breaker
trip.
Therefore, by raising the incoming voltage only SLIGHTLY higher than the running voltage,
the operator aims to prevent both capacitive and inductive power loading on the EDG so
that a breaker trip would not occur due to overloading. Thus, both the key answer, C,
(i.e., (1) capacitive power loading on EDG (2) fast), and, A, (i.e., (1) inductive power loading
on the EDG (2) fast) are correct.
Facility Position on Applicant Comment:
The question grading for the exam should not change.
Per Operations review, the facility has determined the question is accurate as written.
Negative VARS would be indicative of capacitive loading, which is not desirable, since
this could potentially lead to motoring the EDG.
NRC Evaluation/Resolution:
The question asked the applicant to recall the requirements for synchronizing the Unit 2
Emergency Diesel Generator (EDG) to Bus 24-1 with a normal electrical line-up. These
requirements were summarized in the note on page 36 of DOS 6600-01, stating:
When synchronizing the D/G to the Bus, the Synchroscope should rotate one
revolution in approximately 30 seconds in the fast direction. The breaker should be
closed just before the pointer reaches the vertical position. Incoming voltage should
be slightly higher than Running voltage. These conditions will help prevent high
transient current in the generator or a reverse power trip.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
It was agreed, as noted above, that the Synchroscope should rotate one revolution in
approximately 30 seconds in the fast direction, as was specifically noted and per choices A
and C above. The difference between these two choices was identifying what INCOMING
voltage is slightly higher than RUNNING voltage prevented. The applicants position was that
choices A and C were both correct because raising the incoming voltage SLIGHTLY higher
than running voltage prevented both capacitive and inductive power loading on the EDG, and
therefore, a breaker trip would not occur due to overloading. To support that position, the
applicant noted the following information from Chapter 5 of the BWR Generic Fundamentals
Components text:
        *    Lagging Power Factor is indicative of purely inductive loads such as motors and
Leading Power Factor is indicative of purely capacitive loads.
        *    With positive VARs (Lagging Power Factor), the D/G is only SLIGHTLY inductive
when closing the breaker to the respective Bus.
The applicant further stated that This is to ensure that D/G is neither overexcited (inductive) nor
underexcited (capacitive), which would result in EDG overloading and subsequent breaker trip.
The facility position, in contrast, stated:
Negative VARS would be indicative of capacitive loading, which is not desirable,
since this could potentially lead to motoring the ED
: [[contact::G.
Therefore]], the applicant and the facility differed on the reasons for having incoming voltage
slightly higher than running voltage, which was the first part of what this question asked. To
resolve this, note Chapter 5 of the BWR Generic Fundamentals Components text, page 62:
If the incoming generator voltage lags the grid voltage, current will flow from
the grid to the generator and accelerate it to synchronous speed.
In summary, to avoid having the grid attempt to accelerate the EDG to synchronous speed
(i.e. motoring it), negative VARS (capacitive loading) would be undesirable and some positive
VARS (inductive loading) would be required. This supports having the incoming voltage
SLIGHTLY higher than running voltage. Therefore, the U.S. Nuclear Regulatory Commission
(NRC) concluded that choice C, as annotated on the answer key, was the only correct answer,
and the question was considered acceptable as administered.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question 81
Unit 2 was operating at near rated power, with the SBGT system control switches in the
following positions:
    *  2/3A - PRI
    *  2/3B - STBY
A transient occurs causing RPV water level trend down to -15 inches.
Two minutes later:
    *  Aux NSO reports that the 2/3A and 2/3 B SBGT are de-energized.
    *  RPV water level has recovered to 10 inches.
The Unit Supervisor will direct entering __(1)___ and take action to ___(2)___
: [[contact::A.      (1) DOA 7500-01]], STANDBY GAS TREATMENT SYSTEM FAN TRIP
        (2) place 2/3A SBGT to OFF AND then 2/3B SBGT to STAR
: [[contact::T.
B.      (1) DOA 5750-01]], VENTILATION SYSTEM FAILURE
        (2) place 2/3A SBGT to OFF AND then 2/3B SBGT to STAR
: [[contact::T.
C.      (1) DOA 7500-01]], STANDBY GAS TREATMENT SYSTEM FAN TRIP
        (2) Restart Reactor Building ventilation supply and exhaust fans.
: [[contact::D.      (1) DOA 5750-01]], VENTILATION SYSTEM FAILURE
        (2) Restart Reactor Building ventilation supply and exhaust fans.
Answer:        A
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Applicant Comment (55-74856):
The question asks what abnormal operating procedure is entered and what action is taken if
both Primary and Standby trains of SBGT are De-Energized after receipt of a valid Group 2
containment isolation signal due to RPV Level dropping below the Low setpoint. It is also stated
that level is found to have recovered to above the Low Level Setpoint after two minutes.
It is recommended to consider both A and D to be correct answers because both DOA 5750-01,
Ventilation System Failure, and DOA 7500-01, Standby Gas Treatment System Fan Trip,
entry conditions exist in the stem of the question. DOA 5750-01 lists Group 2 isolation as a
symptom that constitutes an entry condition. DOA 7500-01 is entered due to both SBGT trains
being De-Energized when an initiation signal is present.
Additionally, per step D.1.b of DOA 7500-01, the Primary train of SBGT is placed in OFF and
the Standby train is placed in STAR
: [[contact::T. However]], due to the fact that both trains are known to be
De-Energized per the stem of the question, neither train will initiate, which translates to a
safety function not fulfilled with Reactor Building DP rendered less negative. With knowledge of
SBGT being Inoperable/Unavailable, the Unit Supervisor may direct the operator to restart
Reactor Building Ventilation.
Since Reactor Water Level has recovered to above the Group 2 containment isolation setpoint,
a Group 2 isolation signal is NO longer present, and the 2/3A and 2/3B trains of SBGT are
De-Energized per the stem. General guidance D.3.b of DOA 5750-01 can be used to restart
ventilation fans in accordance with the appropriate DO
: [[contact::P.
Step B.2 of DAN 902(3)-5 E-5 identifies DOP 0500-13]], Plant Restoration from PCIS Group 2
Isolation, as the appropriate DOP to restore the plant if the Group 2 isolation signal is NO
longer present. Step G.1 of DOP 0500-13 provides the guidance necessary to reset the PCIS
Group 2 isolation logic, and step G.3 outlines the steps necessary to restore Reactor Building
ventilation and securing SBG
: [[contact::T.
Therefore]], both the key answer, A, (i.e., (1) DOA 7500-01 STANDBY GAS TREATMENT
SYSTEM FAN TRIP (2) Place 2/3A SBGT to OFF AND then 2/3B to START), and D
(i.e., (1) DOA 5750-01, VENTILATION SYSTEM FAILURE (2) Restart Reactor Building
ventilation supply and exhaust fans) are correct.
Facility Position on Applicant Comment:
The station agrees with the challenge that A and D are both correct answer choices.
A is a correct choice due to the direction provided in DOA 7500-01 subsequent actions to start
the standby train if it did not start.
D is also correct due to the following. DOA 5750-01, VENTILATION SYSTEM FAILURE,
would be entered due to a loss of Reactor Building Ventilation caused by a group two isolation
on reactor water level. Per the question stem, the condition driving a group 2 isolation signal
has cleared, which would allow the group 2 isolation to be reset and Reactor Building Ventilation
to be re-started. DOA 5750-01 directs restoring ventilation per the appropriate DOP. DOP
0500-13, PLANT RESTORATION FROM PCIS GROUP 2 ISOLATION, provides the guidance
to reset the group 2 isolation and restart Reactor Building Ventilation.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
NRC Evaluation/Resolution:
The question asked the applicant what abnormal operating procedure should be entered and
what action should be taken for the given stem conditions. All agree that choice A, as per the
original answer key, was correct. The focus of the remaining evaluation will be whether choice
D was also correct, as both the facility and the applicant affirm.
The stem conditions were meant to lead the applicant into selecting DOA 7500-01, STANDBY
GAS TREATMENT SYSTEM FAN TRIP, and recognize the need to attempt a restart of Standby
Gas Treatment (SBGT) without providing any additional references. Distractor D, was
intended to be a plausible distractor because DOA 5750-01, VENTILATION SYSTEM FAILURE
met the stems entry conditions, but then incorrect because it provided no guidance to restart
Reactor Building Ventilation. This was only partially true upon further review.
The SUBSEQUENT OPERATOR ACTION General guidance of DOA 5750-01 could be used to
restart ventilation fans, though indirectly, per step D.3.b, Re-start affected fans once power is
restored, in accordance with the appropriate DO
: [[contact::P. The appropriate DOP]], specified in DAN
2(3)-5 E-5, GROUP 2 ISOLATION INITIATED, was DOP 0500-13, PLANT RESTORATION
FROM PCIS GROUP 2 ISOLATION, where the guidance to reset the group 2 isolation and
restart Reactor Building Ventilation was ultimately provided. While this procedure path was
technically correct, it was not an intended outcome for a question without additional references
provided. Nevertheless, the U.S. Nuclear Regulatory Commission (NRC) concluded that both
choices A and D were correct answers, and the answer key was changed to reflect this.
In accordance with ES-501 E.3.a:
Any questions that were deleted during the grading process, or for which the
answer key had to be changed, will also be included in the count of unacceptable
questions.
Therefore, question 81 was rated as an UNSAT question on Form ES-401-9.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question 88
Unit 2 was operating at near rated power when a transient occurred, resulting in the following
conditions:
    *  Torus water level is 30 feet.
    *  Drywell pressure is 1.31 psig and increasing 0.1 psig/10 minutes.
    *  Current Iodine-131 sample is 2.0 X 10-8 uCi/cc.
    *  Current Beta/Gamma (total particulate) is 5.0 X 10-7 uCi/cc.
    *  Radiation protection is unavailable to perform an off-site dose calculation.
The Unit Supervisor is required to direct the Operating team to vent the Drywell to the
___(1)___ system in accordance with ___(2)___ , to reduce Drywell pressure.
A.      (1) SBGT;
        (2) DEOP 500-4, CONTAINMENT VENTING
B.      (1) SBGT;
        (2) DOP 1600-01, NORMAL PRESSURE CONTROL OF THE DRYWELL
C.      (1) Rx Building Vent;
        (2) DEOP 500-4, CONTAINMENT VENTING
D.      (1) Rx Building Vent;
        (2) DOP 1600-01, NORMAL PRESSURE CONTROL OF THE DRYWELL
Answer:        B
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Applicant Comment (55-74856):
The question asks the appropriate Drywell vent path and procedure to use for the conditions
described in the stem.
It is recommended to change the correct answer from B (i.e., (1) SBGT (2) DOP 1600-1,
NORMAL PRESSURE CONTROL OF THE DRYWELL) to A (i.e., (1) SBGT (2) DEOP 0500-04,
CONTAINMENT VENTING). First part of both A and B are correct because with the given
radiation levels, the Drywell is required to be vented to the SBGT per the limitations and actions
of DOP 1600-05.
However, per section F.8 of the limitation and actions of DEOP 0500-04, Containment Venting,
DEOP 0500-04 is the preferred procedure for venting containment in DEOPs and SAMGs.
With Torus Water Level at 30 ft, a DEOP 200-1, Primary Containment Control, entry condition
exists. DEOP 200-1 requires a reactor Scram before Torus Water Level rises to 18.5 ft.
Following the Scram, setpoint setdown actuated by the Feedwater Level Control (FWLC)
system will lower RPV Water Level below the Group 2 containment isolation setpoint, which
causes a Group 2 containment isolation. Subsequently, DEOP 100-1, RPV Control, is entered
and if Torus Water Level cannot be restored and held below 18.5 ft, DEOP 200-1 directs a
blowdown in accordance with DEOP 400-2, Emergency Depressurization. Therefore,
DEOP 0500-04 is the preferred procedure for venting the Drywell for the conditions described in
the stem of the question.
In contrast, prerequisite D.2 of DOP 1600-1, Normal Pressure Control of the Drywell or
Torus, states that RPV level must be greater than the Group 2 containment isolation setpoint.
DOP 1600-1 does NOT provide guidance to reset the Group 2 containment isolation. As
previously stated, RPV Water Level is lowered below the Group 2 containment isolation setpoint
via the setpoint setdown function of the FWLC system. Therefore, DOP 1600-1 alone is NOT
the preferred procedure to vent the Drywell for the conditions described in the stem of the
question.
In conclusion, A (i.e., (1) SBGT (2) DEOP 0500-04, CONTAINMENT VENTING) is the correct
answer as the Drywell is vented to SBGT using DEOP 0500-4 given the condition in the stem
of the question.
Facility Position on Applicant Comment:
The question grading for the exam should not change.
Per Operations review, the facility has determined that A is not correct. Although
DEOP 0500-04 can be used to vent primary containment via the SBGT, operators are not
allowed to vent primary containment using DEOP 0500-04 unless the Primary Containment
Pressure Limit (PCPL) is being challenged; or a decision has been made to perform early
venting of containment per the override in the pressure leg of DEOP 0200-01. Given the
conditions in the stem, PCPL is not being challenged, and conditions for early venting are
not met.
NRC Evaluation/Resolution:
In the first part of the 2-part question, the applicant was asked to identify where to vent the
Drywell for the given plant conditions following a transient. The applicant, the facility, and the
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
answer key all agreed that SBGT was the correct response due to the given radiation levels.
Rx Building Vent was incorrect because the current Iodine-131 sample results and
Beta/Gamma radiation levels were above the limits for venting via that flow path.
The second part of the question, where there was disagreement, asked the applicant which
procedure to use for these conditions. The answer key and the facility supported
DOP 1600-01, NORMAL PRESSURE CONTROL OF THE DRYWELL because of the need
to vent to stay below 2.0 psig per the first box of the PRIMARY CONTAIMENT PRESSURE leg
of DEOP 200-1 with Primary Containment Pressure rising - Hold drywell and torus pressures
below 2.0 psig using SBGT and drywell purge (DOP 1600-1). The applicant, on the other
hand, chose DEOP 0500-04, CONTAINMENT VENTING for the following reasons:
    *  DEOP 0500-04 is the preferred procedure for venting containment in DEOPs and
SAMGs
    *  DOP 1600-1 alone is NOT the preferred procedure to vent the Drywell for the given
plant conditions
The facility determined that DEOP 0500-04, CONTAINMENT VENTING was not correct.
Although DEOP 0500-04 can be used to vent primary containment via the SBGT, operators
would not be allowed to vent primary containment using DEOP 0500-04 unless the Primary
Containment Pressure Limit (PCPL) was being challenged, or for other specific (extreme)
exceptions; and these thresholds were not met for the given conditions. This position was
supported by DEOP 0500-04, H.3:
provides direction for controlling Primary Containment pressure
below the Primary Containment Pressure Limit. Normally, venting is only
performed, as needed, to keep the pressure below the design pressure.
However, in extreme cases, such as an extended station blackout (beyond the
plant design basis), earlier or more extensive primary containment pressure
reductions (per Attachment 1) may be appropriate to restore and maintain
adequate core cooling, limit the total radioactivity release due to Primary
Containment degradation, or if significant fuel damage is anticipated.
In summary, DOP 1600-01, NORMAL PRESSURE CONTROL OF THE DRYWELL was
referenced by DEOP 200-1 and was appropriate for current plant conditions, whereas
DEOP 0500-04, CONTAINMENT VENTING, while a preferred procedure for specific
circumstances, was not relevant in this case. Another point of contention by the applicant was
that DOP 1600-1 alone was not the preferred procedure under given plant conditions, however,
nothing in the question specified that only the selected procedure was necessary. The analysis
for Question 81, for example, required other referenced procedures to fully address all issues in
the correct responses. It was incorrect to make that assumption per NUREG-1021, Appendix E
(Part B.7, Written Exam Guidelines). Accordingly, the only correct choice applicable to the
given stem conditions was choice B, ((1) SBGT; (2) DOP 1600-01, NORMAL PRESSURE
CONTROL OF THE DRYWELL). Therefore, the U.S. Nuclear Regulatory Commission (NRC)
concluded that choice B, as annotated on the answer key, was the only correct answer, and
the question was considered acceptable as administered.
SIMULATION FACILITY FIDELITY REPORT
Facility Licensee:                      Dresden Nuclear Power Station, Units 2 and 3
Facility Docket Nos:                    50-237; 50-249
Operating Tests Administered:          June 15, 2020 through June 18, 2020
The following documents observations made by the U.S. Nuclear Regulatory Commission
examination team during the initial operator license examination. These observations do not
constitute audit or inspection findings and are not, without further verification and review,
indicative of non-compliance with Title 10 of the Code of Federal Regulations, Part 55.45(b).
These observations do not affect U.S. Nuclear Regulatory Commission certification or approval
of the simulation facility other than to provide information, which may be used in future
evaluations. No licensee action is required in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were
observed:
ITEM                                        DESCRIPTION
SWR 0136211              SBLC computer switch command does not work properly
SWR 0136210              Valve 2-5418 showing dual indication
SWR 0136209              SER shows incorrect alarm
SWR 0136184              SRM Indication Discrepancy
3
}}
}}

Revision as of 22:37, 24 September 2020

NRC Initial License Examination Report 05000237/2020301; 05000249/2020301
ML20239A770
Person / Time
Site: Dresden, Sequoyah  Constellation icon.png
Issue date: 08/27/2020
From: Patricia Pelke
Operations Branch III
To: Bryan Hanson
Exelon Generation Co, Exelon Nuclear
Zoia C
Shared Package
ML19121A248 List:
References
50-237/20-301, 50-249/20-301
Download: ML20239A770 (22)
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