IR 05000237/2020301

NRC Initial License Examination Report 05000237/2020301; 05000249/2020301

ML20239A770
Person / Time
Site:	Dresden, Sequoyah
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: IR 05000237/2020301 ML19283A116 ML20160A176
References
50-237/20-301, 50-249/20-301
Download: ML20239A770 (22)

Text

August 27, 2020

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

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.

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

P. Karaba, Site Vice President

P. Boyle, Plant Manager

D. Thomas, Training Director

R. Bauman, Operations Director

M. Condreay, Manager Operations Training

D. Siuda, Senior Operations Training Instructor

D. Heyn, Senior Operations Training Instructor

F. Winter, Senior Operations Training Instructor

D. Walker, Senior Regulatory Specialist

J. Van Fleet, Maintenance Director

W. Remiasz, Director Organizational Performance & Regulatory

M. McCormick, Shift Manager

H. Patel, Shift Operations Superintendent

U.S. Nuclear Regulatory Commission

C. Zoia, Chief Examiner

G. Roach, Examiner

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

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

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

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)___

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

T.

B. (1) DOA 5750-01, VENTILATION SYSTEM FAILURE

(2) place 2/3A SBGT to OFF AND then 2/3B SBGT to STAR

T.

C. (1) DOA 7500-01, STANDBY GAS TREATMENT SYSTEM FAN TRIP

(2) Restart Reactor Building ventilation supply and exhaust fans.

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

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

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

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

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

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