IR 05000454/2019302
| ML20006E615 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 01/06/2020 |
| From: | Robert Orlikowski NRC/RGN-III |
| To: | Bryan Hanson NRC/RGN-III/DRS/OLB |
| Demarshall J | |
| Shared Package | |
| ML17214A834 | List: |
| References | |
| 50-454/19-302, 50-455/19-302 50-454/OL-19, 50-455/OL-19 | |
| Download: ML20006E615 (24) | |
Text
ary 6, 2020
SUBJECT:
BYRON STATION, UNITS 1 AND 2NRC INITIAL LICENSE EXAMINATION REPORT 05000454/2019302; 05000455/2019302
Dear Mr. Hanson:
On November 27, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed the initial operator licensing examination process for license applicants employed at your Byron Station, Units 1 and 2. The enclosed report documents the results of those examinations. Preliminary observations noted during the examination process were discussed on November 4, 2019, with Mr. Mark Kanavos, Site Vice President, and other members of your staff. An exit meeting was conducted by telephone on December 16, 2019, with Mr. Kevin Sanford, Operations Training Manager, other members of your staff, and Mr. J. DeMarshall, Chief Operator Licensing Examiner, 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 November 27, 2019, were discussed.
The NRC examiners administered an initial license examination operating test during the weeks of October 28 and November 4, 2019. The written examination was administered by Byron Station training department personnel on November 12, 2019. Four Senior Reactor Operator and three Reactor Operator applicants were administered license examinations.
The results of the examinations were finalized on December 12, 2019. Five applicants passed all sections of their respective examinations. Three applicants were issued senior operator licenses and two applicants were issued operator licenses. Two applicants failed one or more sections of the administered examination and were issued Preliminary Results Letters.
The written examination, administered 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 November 27, 2021. However, because two applicants received Preliminary Results Letters due to receiving a non-passing grade on the written examination, the applicants were 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,
/RA/
Robert J. Orlikowski, Chief Operations Branch Division of Reactor Safety Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66
Enclosures:
1. OL Examination Report 05000454/2019302; 05000455/2019302 2. Post-Examination Comments, Evaluation, and Resolutions 3. Simulator Fidelity Report
REGION III==
Docket Nos: 50-454; 50-455 License Nos: NPF-37; NPF-66 Report No: 05000454/2019302; 05000455/2019302 Enterprise Identifier: L-2019-OLL-0033 Licensee: Exelon Generation Company, LLC Facility: Byron Station, Units 1 and 2 Location: Byron, IL Dates: October 28, 2019, through November 5, 2019 Examiners: J. DeMarshall, Senior Operations Engineer, Chief Examiner B. Bergeon, Operations Engineer, Examiner D. Reeser, Operations Engineer, Examiner Approved by: R. Orlikowski, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY
Examination Report 05000454/2019302; 05000455/2019302; 10/28/2019-11/12/2019;
Exelon Generation Company, LLC, Byron Station, Unit1 and 2; Initial License Examination Report.
The announced initial operator licensing examination was conducted by regional Nuclear Regulatory Commission examiners in accordance with the guidance of NUREG-1021,
Operator Licensing Examination Standards for Power Reactors, Revision 11.
Examination Summary Five of seven applicants passed all sections of their respective examinations. Three applicants were issued senior operator licenses and two applicants were issued operator licenses.
Two applicants failed one or more sections of the administered examination and were issued Preliminary Results Letters. (Section 4OA5.1)
REPORT DETAILS
4OA5 Other Activities
.1 Initial Licensing Examinations
a. Examination Scope
The U.S. Nuclear Regulatory Commission (NRC) examiners and members of the facility licensees staff used the guidance prescribed in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 11, to develop, validate, administer, and grade the written examination and operating test. The written examination outlines and operating test outlines were prepared by the NRC staff.
The NRC staff developed the written examination and members of the facility licensees staff developed the operating test. As part of the operating test development, the NRC examiners visited the facility during the week of August 12, 2019, to pre-validate the events/malfunctions constructed from the NRC prepared outlines, and to obtain feedback from your staff regarding scenario validity and capabilities of the simulator to support event performance. The NRC examiners validated the proposed examination during the week of September 30, 2019, with the assistance of members of the facility licensees staff. During the onsite validation week, the examiners audited ten 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 October 28, 2019, through November 5, 2019. The facility licensee administered the written examination on November 12, 2019.
b. Findings
- (1) Written Examination During validation of the NRC developed written examination, several questions were modified or replaced. All changes made to the written examination were made in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors. Form ES-401-9, Written Examination Review Worksheet, used primarily for the documentation of metrics on the NRC developed written examination, was updated with post-examination changes. The Form ES-401-9, the written examination outlines (ES-401-1 and ES-401-3), and both the proposed and final written examinations, will be available electronically in the NRC Public Document Room or from the Publicly Available Records component of NRCs Agencywide Documents Access and Management System (ADAMS) on November 27, 2021 (ADAMS Accession Numbers ML17214A840, ML17214A838, ML17214A839, and ML17214A837, respectively).
On November 27, 2019, the licensee submitted documentation noting that there were five post-examination comments for consideration by the NRC examiners when grading the written examination. The post-examination comments are documented in 2 of this report.
The NRC examiners completed grading of the written examination on December 9, 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, developed by the licensee from the NRC prepared outlines, was within the range of acceptability expected for a proposed examination.
Following the review and validation of the operating test, 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 NRCs ADAMS on November 27, 2021, (ADAMS Accession Numbers ML17214A840, ML17214A838, ML17214A839, and ML17214A837, respectively).
The NRC examiners completed grading of the operating test on December 12, 2019.
- (3) Examination Results Four applicants at the Senior Reactor Operator level and three applicants at the Reactor Operator level were administered written examinations and operating tests.
Five applicants passed all portions of their examinations. Five applicants were issued their respective operating licenses on December 16, 2019. One Senior Reactor Operator applicant and one Reactor Operator applicant failed the written examination portion of the administered examination and were issued Preliminary Results Letters.
.2 Examination Security
a. Scope
The NRC examiners reviewed and observed the licensee's implementation of examination security requirements during the examination validation and administration to assure compliance with Title10 of the Code of Federal Regulations, Part 55.49, Integrity of Examinations and Tests. The examiners used the guidelines provided in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, to determine acceptability of the licensees examination security activities.
b. Findings
None.
4OA6 Management Meetings
.1 Debrief
The chief examiner presented the examination team's preliminary observations and findings on November 4, 2019, to Mr. Mark Kanavos, Site Vice President, and other members of the Byron Station, Units 1 and 2 staff.
.2 Exit Meeting
The chief examiner conducted an exit meeting on December 16, 2019, with Mr. Kevin Sanford, Operations Training Manager, and other members of the Byron Station, Units 1 and 2 staff, by telephone. The chief examiner asked the licensee whether any of the material used to develop or administer the examination should be considered proprietary. No proprietary or sensitive information was identified during the examination or debrief/exit meetings.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- M. Kanavos, Site Vice President
- H. Welt, Plant Manager
- K. McGuire, Operations Director
- R. Lawlor, Senior Manager Site Training
- K. Sanford, Operations Training Manager
- B. Lewin, Shift Operations Superintendent
- M. Davis, Shift Manager, Facility Representative
- M. Kultgen, Senior Operations Training Instructor, ILT Lead Exam Author
- Z. Cox, Principle Regulatory Engineer
U.S. Nuclear Regulatory Commission
- D. Betancourt-Roldan, Senior Resident Inspector
- C. Hunt, Resident Inspector
- J. DeMarshall, Senior Operations Engineer, Chief Examiner
- B. Bergeon, Operations Engineer, Examiner
- D. Reeser, Operations Engineer, Examiner
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
None
LIST OF ACRONYMS USED
ADAMS Agencywide Documents Access and Management System
NRC U.S. Nuclear Regulatory Commission
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
RO Question 10
Byron Station Units 1 and 2 were both operating at 100% when a tornado struck the site.
- Units 1 and 2 have both tripped.
- A Loss of All AC power has occurred.
- NO Diesel Generators were able to be started.
Under these conditions, the primary load shed of DC buses must be completed within 65
minutes in order to ensure that the station batteries will last at least hours.
A. 2
B. 4
C. 8
D. 16
Answer: C
Reference(s) provided to NRC:
1BFSG-4, ELAP DC BUS LOAD SHED/MANAGEMENT, Unit 1 (Rev. 2)
1BCA-0.0, LOSS OF ALL AC POWER, UNIT 1 (Rev. 302)
Technical Specification 3.8.4, DC Sources-Operating (Amendment No. 198)
Applicant Comment:
Question required use of a BFSG as reference to adequately answer. The question asks
how long the battery would last as long as required loads were stripped within 65 minutes.
This 65-minute time requirement and the 8-hour capacity are only discussed in BFSG-4,
a beyond design basis document. NUREG 1122 does not require knowledge of beyond
design basis procedures for the RO level. 10 CFR 55.41 (b) does not include beyond design
knowledge in the 14 required topics to be tested for ROs. UFSAR states that the battery
contains 2320 Amp-Hours at the 8-hour rate, and that it is sized for a LOCA with a loss of
offsite power and the failure of a diesel generator for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The 3.8.4 bases only discusses
hours in relation to maintaining a 400 amp charge rate while maintaining float voltage for
hours. Question would have required a reference provided of the BFSG, or clarification
of discharge rates with BOP DC-2T1 referenced to adequately answer.
Facility Position on Applicant Comment:
The licensee agrees with the applicants comments. The 65-minute requirement time is only
found in the BFSG procedure. The BFSGs implement FLEX procedures. Flex procedures are
outside the Design Basis. The question is outside the purview of the RO section of the exam
and should be deleted.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
NRC Evaluation/Resolution:
Recommendation accepted. The question tests applicant knowledge of the operational
implications associated with battery discharge rates on capacity following a Station Blackout
Event. Specifically, the question requires the applicant to recall from memory, how long
station batteries will last (8-hours) if the primary load shed of DC Buses is completed within
a predetermined time (65 minutes) of the Loss of All AC Power.
The applicant contends that the question is at the SRO license level on the basis that:
1) the information pertaining to the 65-minute completion time and 8-hour capacity value is only
discussed in the unit specific FLEX Support Guideline procedure BFSG-4, ELAP DC Bus Load
Shed/Management; 2) NUREG-1122, Knowledge and Abilities for Nuclear Plant Operators
(Pressurized Water Reactors), does not require specific knowledge of beyond design basis
procedures at the RO level; and 3) 10 CFR 55.41, Written examination: Operators, does not
include a requirement for testing beyond design basis knowledge on the RO exam.
The NRC concludes that RO Question 10 is at the wrong license level. The question is deleted
because it requires an RO applicant to be knowledgeable of, and recall from memory, specific
beyond design basis procedural information that is only cited within the context of a Note
contained in the stations FLEX Support procedures for an Extended Loss of AC Power;
i.e., information that is not RO knowledge required by 10 CFR 55.41. The deletion is
reflected in the final written examination and answer key.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
RO Question 51
At 10:00, the 1A DG was running per 1BOSR 8.1.2-1, UNIT ONE 1A DIESEL
GENERATOR OPERABILITY SURVEILLANC
- E.
At 12:30, the EO reported a loud noise and that BOTH starting air bank pressures were rapidly
approaching 0 psig.
What is the immediate impact on the operating 1A DG?
A. The DG will ONLY trip on overspeed.
B. ONLY the DG starting capability is lost.
C. The DG will trip due to a loss of starting air.
D. ONLY the DG non-emergency trip capabilities are lost.
Answer: A
Reference(s) provided to NRC:
1BOSR 8.1.2-1, 1A Diesel Generator Operability Surveillance (Rev. 32)
N-BY-TQ-ILT-DG-9, Diesel Generator & Auxiliaries System Lesson Plan (Rev. 10)
Applicant Comment:
The answers available all say ONLY when discussing what ways are available to trip the DG
should there be a loss of the DG starting air system. Answer A states that the DG will trip
ONLY on overspeed, however, per BOP DG-11 and BOP DG-12T2 you can also trip the diesel
manually from operation of the butterfly valve and from the fuel rack trip lever. As neither the
question nor the answer reference whether we are discussing manual or automatic trips, and
there are three ways to trip the DG upon a loss of starting air, overspeed trip is not the ONLY
trip available and as such there is no correct answer for this question.
Facility Position on Applicant Comment:
The licensee agrees with the applicants comments. BOP DG-11 DIESEL GENERATOR
STARTUP states in Precautions D.2 If control Air is not available, manually operate the
Fuel Rack Trip Lever to secure fuel to the engine, or manually operate the overspeed butterfly
valve to secure air to the engine. The question has no correct answer and should be deleted.
NRC Evaluation/Resolution:
Recommendation not accepted. The question tests applicant knowledge of the immediate
impact (emphasis added) on the operating Diesel Generator (DG) resulting from the loss of
air pressure in both starting air banks/receivers.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
The applicant contends that choice A, The DG will ONLY trip on overspeed, is incorrect,
and therefore no correct answer exists, because: 1) the DG, in addition to having overspeed
trip capability, can also be stopped by: (a) manually closing the overspeed butterfly damper
(air intake butterfly) to secure air to the engine; or (b) operation of the fuel rack trip lever to
secure fuel to the engine; and 2) neither the question nor the answer specify manual or
automatic trip capability.
The applicant did not answer the question as written. The focus of the question was not to
test applicant knowledge of local operator actions that can be taken to manually stop the DG
in response to the loss of starting air (i.e., physical repositioning of components at the DG), but
rather to test applicant knowledge of what the immediate impact is on the running DG. The
immediate impact is a loss of Pneumatic Protection Subsystem trip functionality and resultant
inability to trip the DG on any automatic or manual signal (emphasis added). Lesson Plan
N-BY-TQ-ILT-DG-9, Diesel Generator & Auxiliaries System,Section III.C, Pneumatic
Protection Circuit, (Pages 14-15) states:
The Fuel-Control Cylinder is spring-loaded and air-operated with air lines on both sides
of an internal piston.
If air pressure is lost (i.e., is removed from both sides), the spring returns the
Fuel-Control Cylinder to its retracted position. This means the engine can operate
without control air but will not trip (except overspeed) on any automatic or manual
signal. An overspeed will shut the air intake butterfly, which will stop the engine.
Note that the overspeed trip referenced in the above cited text is a mechanical overspeed trip.
The immediate impact to the operating DG resulting from the loss of air pressure is exactly as
described in the N-BY-TQ-ILT-DG-9 Lesson Plan; i.e., the loss of all automatic and manual trip
signal functionality and preservation of mechanical overspeed trip capability. Manual closure
of the fuel racks and the overspeed butterfly damper are ways to manually shutdown the DG
and are NOT trips. Seventeen DG trips (identified below) are listed in the N-BY-TQ-ILT-DG-9
Lesson Plan (Note: manual operation of the fuel racks and overspeed butterfly damper are
excluded from this list).
Lube Oil Press Low
Turbocharger Lube Oil Press Low
Main Bearing Temperature High
Generator Bearing Temperature High
Connecting Rod Bearing Temperature High
Thrust Bearing Failure
High Jacket Water Temperature
Reverse Power
Emergency Stop Pushbutton
Underfrequency
Generator Differential
Loss of Field
Overload
High Crankcase Pressure
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
Ground Fault
Incomplete Sequence
NUREG-1021, Operator Licensing Examination Standards for Power Reactor (Rev. 11),
Appendix E, Policies and Guidelines for Taking NRC Examination, Subsection B, Written
Examination Guidelines, Paragraph B.7, states, in part, If you have any questions concerning
the intent or the initial conditions of a question, do not hesitate to ask them before answering
the question. All applicants were briefed on the contents of APPENDIX E prior to exam
administration, and all paragraph items contained in Subsection B were read verbatim.
No questions concerning any aspect of Question 51 were raised by any of the applicants during
administration of the exam.
Based on the above, the NRC concludes that choice A, as annotated on the answer key,
is correct, and that the question is considered acceptable as administered.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
RO Question 68
Unit 2 is operating at 100% power.
- A plant transient occurs.
systems behaving unexpectedly.
Which of the following practices by the ATC RO would be consistent with OP-AA-101-111-1001
OPERATIONS STANDARDS AND EXPECTATIONS for conservative decision-making and
action?
A. NOT becoming totally involved in any single operation.
B. Implementing Immediate Actions only AFTER verbalizing them.
C. Reducing power OR initiating a reactor scram without delay.
D. Referencing hard cards BEFORE taking immediate operator actions.
Answer: C
Reference(s) provided to NRC:
OP-AA-101-111, Roles and Responsibilities of On-Shift Personnel (Rev. 12)
OP-AA-101-111-1001, Operations Standards and Expectations (Rev. 21)
Applicant Comment:
Question asks for what the RO would do per OP-AA-101-111-1001 in the event of unexpected
controls during a transient. Per the reference, the RO will take conservative action,
INCLUDING tripping the reactor or reducing power. This step references SOER 94-1,
recommendation 2.
[Recommendation 2 quoted by the applicant was removed since INPO SOER 94-1
is an OUO-Not Publicly Available document.]
The commitment to SOER 94-1 requires more than just reducing power and requires
specific criteria to tripping the reactor. OP-AA-101-111-1001 recognizes this by stating to take
conservative action, including reducing power and tripping the reactor, instead of just directing
only the action of reducing power or tripping the reactor. As this procedure is a corporate wide
procedure, further clarification is required to be made at the site level, which would transfer
you to OP-BY-101-0004. This procedure defines the specifics of how to implement
OP AA-101-111-1001s commitments to SOER 94-1. This includes taking actions per posted
and approved hard cards in the event of unexpected conditions (as stated in the stem) and
tripping the reactor, and thus performing immediate actions, once a critical parameter is
reached, and not as the initial step unless such conditions dictate. Conditions were not
adequately defined to reach the decision to immediately trip the reactor. Reducing power
would be covered within our hard card procedures.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
Facility Position on Applicant Comment:
The licensee agrees with the applicants comments.
OP-BY-101-0004, Strategies for Successful Transient Mitigation provides site specific guidance
for implementation of hard cards. Specifically, section 4.4.6 Prompt Operator Actions, states
that Prompt Operator Actions are taken in response to an upset condition in order to stabilize
the plant when the delay involved in normal procedural response would result in a major plant
transient or potentially damage plant equipment. The expected action from a licensed
individual is to perform the prompt actions via the hard cards to preclude putting the plant in a
state where immediate actions are required (e.g. taking prompt action to restore Steam
Generator Water Level prior to reaching a reactor trip setpoint).
The stem does not provide sufficient discrimination to identify the event. Since the event cannot
be identified, the correct actions to adhere to OP-AA-101-111-1001 for the ATC operator cannot
be properly discerned between the answer C and the distractor D. Since the two answers do
not conflict, both C and D should be accepted as answers.
NRC Evaluation/Resolution:
Recommendation not accepted. The question specifically tests applicant knowledge
of conservative decision-making guidance contained in Exelon Corporate procedure
OP-AA-101-111-1001, Operations Standards and Expectations. The question requires
the applicant to determine which of the listed practices would be consistent with
OP-AA-101-111-1001, given that a plant transient occurs at 100 percent power
(emphasis added), and unexpected behavior of the reactor control systems
(emphasis added) is observed while responding to the transient.
The applicant contends that the question stem contains insufficient information to conclude that
an immediate action response (i.e., without delay) to the unexpected behavior of the reactor
control systems (and in progress plant transient), was required. The applicant further contends
that because no information was provided to indicate that any thresholds/trip criteria were
challenged or exceeded, that immediate actions (i.e., actions committed to memory and
allowed to be performed without reference to procedure) would not be performed as the initial
response, and that a station Prompt Action Hard Card procedure would instead be utilized to
perform any mitigating action(s). Therefore, the applicant selected choice D as the answer to
the question. The applicants comments also include a discussion of several references (see
above) which he contends supports the selection of choice
- D.
OP-AA-101-111-1001, Section 4.11, Conservative Decision-Making, states in part:
4.11.3 When faced with unexpected or anomalous behavior of the reactor, or its
control systems, take conservative action, including reducing power or
initiating a reactor scram without hesitation. (CM-2)
The applicant points out that Step 4.11.3 of OP-AA-101-111-1001 [the basis for the identified
correct answer] is a facility commitment (CM-2) to INPO SOER 94-1, Recommendation 2, and
further states that OP-BY-101-0004, Strategies for Successful Transient Mitigation, provides
clarification on the guidance contained in OP-AA-101-111-1001, adding that OP-BY-101-0004
defines the specifics of how to implement OP-AA-101-111-1001s commitments to
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
SOER 94-1. OP-BY-101-0004 actually provides station specific guidance to complement the
corporate procedure of the same name; i.e., OP-AA-103-102-1001, Strategies for Successful
Transient Mitigation.
OP-AA-103-102-1001 defines Immediate Actions as:
Actions that are taken from memory. They must be actions that are otherwise
contained in a procedure, but are urgently needed to place a system or the
reactor in a safe and stable condition.
The procedure further states that Any procedurally directed actions that are not specified as
Immediate Actions should be performed via use of a controlling document.
OP-BY-101-0004 provides guidance on performance of Prompt Operator Actions, which fall
within the category of procedurally directed actions that are not specified as Immediate
Actions. OP-BY-101-0004, Section 4.4.6, states in part:
4.4.6.1 Prompt Operator Actions are actions taken in response to an upset condition
in order to stabilize the plant when the delay involved in normal procedural
response would result in a major plant transient or potentially damage plant
equipment.
4.4.6.4 In deciding if taking prompt action is appropriate, the following elements
should be considered:
B. Prompt Actions DO NOT take priority over Immediate Actions but may
be performed as time permits and if the Prompt Actions DO NOT delay
completion of the Immediate Actions.
The question is not asking for an event specific response. Therefore, the applicant is not
required to determine whether the response requires the implementation of immediate actions.
The question statement specifically asks the applicant to identify, based on the information
provided, which of the listed practices, performed by a Reactor Operator, would be consistent
with OP-AA-101-111-1001, Operations Standards and Expectations, for conservative decision-
making and action.
The facility, citing OP-BY-101-0004, Section 4.4.6 (above), states that they agree with the
applicants comment. The facility contends that the information provided in the stem was
insufficient to determine whether the response should consist of hard card implementation to
stabilize the plant (choice D) OR immediate actions to trip the reactor (choice C). The facility
maintains that since prompt operator actions (i.e., hard card actions) can be taken to stabilize
the plant and mitigate the effects of the transient enough to preclude putting the plant in a state
where immediate actions are required, that choice D, Referencing hard cards BEFORE taking
immediate operator actions, is a second correct answer.
Both the applicant and the facility responses require an assumption that the initial conditions
specified in the question stem, need to describe a specific event to differentiate between
choices C and D in the selection of the correct answer. Without any specifics regarding
the nature of the anomalous reactor control systems behavior (and in progress plant transient),
no sequence of events can be implied as suggested by the applicants or facilitys response.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
Answer choice C: 1) encompasses the full range of conservative operator responses to a
transient (defined in OP-103-102-1001 as an unexpected or undesired change in plant
operational parameters caused by equipment failures, external or internal events, or human
performance errors that present an immediate challenge to the normal operating limits); and
2) does not require the applicant to determine whether immediate actions are required.
Choice C is clearly the only correct answer to the question. The statement in choice D is
incorrect because it is not a conservative decision-making action for placing the reactor in a
safe and stable condition in accordance with Exelon corporate and site-specific procedures for
successful transient mitigation and conduct of operations standards/expectations. If immediate
actions are deemed necessary, operators are not required to reference procedures prior to
implementation of mitigating action(s). The selection of answer choice D requires the applicant
to assume the occurrence of plant-specific events that would preclude an immediate action
response.
NUREG-1021, Operator Licensing Examination Standards for Power Reactor (Rev. 11),
Appendix E, Policies and Guidelines for Taking NRC Examination, Subsection B, Written
Examination Guidelines, Paragraph B.7, states, in part, If you have any questions concerning
the intent or the initial conditions of a question, do not hesitate to ask them before answering
the question. All applicants were briefed on the contents of APPENDIX E prior to exam
administration, and all paragraph items contained in Subsection B were read verbatim.
No questions associated with the adequacy of conditions, or any other aspect of Question 68,
were raised by any of the applicants during administration of the exam.
In summary, choice D, Referencing hard cards BEFORE taking immediate operator actions,
is incorrect because the use of hard cards when immediate actions are warranted is not in
accordance with the conservative decision-making guidance delineated in Exelon corporate
and site-specific procedures. Based on the above, the NRC concludes that choice C, as
annotated on the answer key, is the only correct answer, and that the question is considered
acceptable as administered.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
SRO Question 92
The Unit 1 crew is preparing to synchronize the main turbine in accordance with 1BGP 100-3,
POWER ASCENSION.
- Permissive P10 (3.4) Bypass Permissive Light is ILLUMINATED.
- Reactor power is being slowly raised.
At this time, steam dump valve 1MS004A sticks in the fully open position.
The Steam Dumps will initially control steam header pressure (1) the setpoint.
Manual isolation of the stuck open steam dump valve can be done in accordance with
the instructions found in (2) .
A. (1) BELOW
(2) OP-BY-101-0004, STRATEGIES FOR SUCCESSFUL TRANSIENT
MITIGATION
B. (1) BELOW
(2) 1BGP 100-3, POWER ASCENSION
C. (1) AT
(2) OP-BY-101-0004, STRATEGIES FOR SUCCESSFUL TRANSIENT
MITIGATION
D. (1) AT
(2) 1BGP 100-3, POWER ASCENSION
Answer: C
Reference(s) provided to NRC:
1BGP 100-3, Power Ascension (Rev. 98)
Horse Notes MS-4, Main Steam Dumps (Rev. 26)
OP-AA-103-102-1001, Strategies for Successful Transient Mitigation (Rev. 2)
BOP MS-100, Main Steam Dump Operation (Rev. 0)
Applicant Comment:
Ambiguity relating to the use of initially. This refers to what happens first, and with MS004A fully
opening at less than P-10, more steam flow will occur and header pressure will drop. Initially, the
proportional controller portion will control the working steam dumps and operate below the current
setpoint, with a proportional band of 100psi. As the integral builds in, then the steam dumps will
throttle to control at the setpoint, but this would not be the initial controlling band as it takes time
for the integral portion of the controller to take over. The integral for this controller is 180 seconds,
and an integral controller takes approximately 5 time constants to control at its setpoint and
therefore would take between 3 and 15 minutes to control at the setpoint.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
Facility Position on Applicant Comment:
The licensee agrees with the applicants comments. A plant reference simulator was used to fail
open 1MS004A at approximately 9% power. The steam dumps controlled pressure less than
setpoint for approximately 3 minutes, overshot and returned to setpoint at about the 5 minute
mark. In addition, 1FW- 0507 CALIBRATION OF STEAM GENERATOR HEADER PRESSURE
STEAM DUMP TO CONDENSER (FW) TEST REPORT PACKAGE BIP 2500-038 Rev. 11
supports the proportional band and reset values identified. This would change the acceptable
answer from AT to BELOW. The correct answer for the question should be changed from C to A.
NRC Evaluation/Resolution:
Recommendation accepted. The question tests the ability of the applicant to: 1) predict the
impacts of a stuck open steam dump valve at low power (approximately 9 percent); and 2) select
the appropriate procedure under which to accomplish manual isolation of the valve to mitigate the
consequences of the malfunction.
The applicant contends that the correct answer for the question should be changed from C to A,
on that basis that use of the word initially in Part 1 of the question statement introduces an
element of ambiguity with respect to how the Steam Dump Valves will be controlling steam header
pressure relative to the setpoint, based on the operating characteristics of the Steam Pressure
Proportional-Integral (PI) Controller.
Part 1 of the question statement requires the applicant to determine whether the Steam Dumps will
initially control steam header pressure AT" or BELOW the setpoint after Steam Dump Valve
1MS004A fails in the open" position. Choice C (controls AT setpoint), was designated as the
correct answer on the administered version of the exam. The NRC concludes that use of the word
initially changes the correct answer from C to A, because the proportional action of the PI
controller is the feature that provides the initial response to the difference (error) between the
desired value (setpoint) and the measured value (process variable). The proportional action
however, does not eliminate the residual error, resulting in a steam header pressure that will
initially be controlled BELOW the setpoint. The integral action of the controller, which takes time
to build in, is the feature that provides the needed correction to eliminate the residual error by
changing the output of the controller as necessary until the process variable (i.e., steam header
pressure) is returned to, and being controlled AT, the setpoint value.
In summary, use of the word initially in Part 1 of the question statement, resulted in an unclear
stem that confused the applicant. The final written examination and answer key is revised to
reflect the change from choice C to A.
POST-EXAM COMMENTS, EVALUATIONS, AND RESOLUTIONS
SRO Question 94
Unit 1 was loading fuel when the audible source range indication inside the containment building
became silent.
Both source range instruments were otherwise functioning normally.
Because of the failure of the containment building audible source range monitor,
fuel loading may
A. continue only if an Inverse Count Rate Ratio (ICRR) is performed.
B. continue only if the Source Range High Flux at Shutdown alarm is OPERABLE.
C. continue only if the Boron Dilution Protection System (BDPS) is OPERABLE.
D. NOT continue until the audible source range indication is restored to the
containment building.
Answer: A
Reference(s) provided to NRC:
1BGP 100-6, Refueling Outage (Rev. 53)
Applicant Comment:
These two precautions list requirements to suspend core alterations. Loss of audible count rate
in containment is not contained in either of these lists of requirements.
BGP 100-6 Step E.8.d:
While E.8.d contains the requirement to perform an ICRR when audible counts are
lost in containment, no requirement is given to stop fuel movement until the ICRR
is performed.
BGP 100-6 Step F.4:
Audible indication from an OPERABLE Channel N-31 and N-32 in the Containment
and the Control Room should be maintained, though not required, to aid the operators
in monitoring core reactivity.
Additionally, OU-AP-200, OU-AP-205, OU-AP-209, and OU-AP-4001 were reviewed;
no requirement to suspend fuel movement based on audible count were found.
Based on the wording of the answers provided, there is no correct answer to this question since
they all stipulated that fuel movement had to be suspended.
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
Facility Position on Applicant Comment:
The licensee agrees with the applicants comments. There are no requirements for fuel moves
to be stopped as implied by the stem and the correct answer. This then leaves no correct
answer. This question should be deleted from the exam.
NRC Evaluation/Resolution:
Recommendation accepted. The question tests applicant knowledge of SRO fuel-handling
responsibilities. Specifically, the question requires the applicant to apply knowledge of the
procedural guidance in 1BGP 100-6, Refueling Outage, to determine whether fuel loading
activities may continue following the loss of continuous audible source range (SR) indication
inside the Containment Building.
The applicant contends that there is no correct answer and that the question should be deleted
on the basis that: 1) information contained in 1BGP 100-6 confirms that no requirements exist
to suspend fuel loading after the loss of audible SR indication; and 2) the wording of the
answers stipulates that fuel movement had to be suspended.
The question is flawed as written, because: 1) unclear wording in the stem and answer choices
incorrectly imply that fuel loading was suspended; and 2) correct Answer A incorrectly
assumes, based on the following excerpts from Section E of 1BGP 100-6, that an Inverse
Count Rate Ratio (ICCR) had to be performed before fuel loading activities were allowed to
continue, given the loss of audible SR indication.
Section E, Subsection 8, Step d, states that an ICCR shall be performed if the following
condition is not satisfied:
Audible Source Range indication exists continuously in the main control
room and inside the containment building.
Subsection 5, Step d, states that Core Alterations during Core Re-Load shall be
IMMEDIATELY suspended:
If an Inverse Count Rate Ratio is being performed and the data indicates
that Criticality may occur with the addition of the next few Fuel
Assemblies.
1BGP 100-6, Section E, Limitations and Actions, Subsections 4 and 5, provide specific
direction for the suspension of Core Alterations. The loss of audible SR indication is not
included in any of the conditions listed. In addition, 1BGP 100-6, Section F, Main Body,
Step 4, states:
Audible indication from an OPERABLE Channel N-31 and N-32 in the
Containment and the Control Room should be maintained, though not
required, to aid the operators in monitoring core reactivity.
Audible indication is not a requirement for channel operability per Refueling Tech Spec
LCO 3.9.3, Nuclear Instrumentation. LCO 3.9.3 specifically requires two SR Neutron Flux
Monitors to be operable, each with continuous visual indication in the control room. The source
range neutron flux monitors have no safety function in MODE 6 and are not assumed to function
POST-EXAMINATION COMMENTS, EVALUATION, AND RESOLUTIONS
during any UFSAR design basis accident or transient. The source range neutron flux monitors
provide the only on-scale monitoring of the neutron flux level during refueling, and for that
reason alone are retained in the Tech Specs.
Based on the above, the NRC concludes that choice A is incorrect and that there are no correct
answers. Therefore, SRO Question 94 is deleted. The deletion is reflected in the final written
examination and answer key.
SIMULATOR FIDELITY REPORT
Facility Licensee: Byron Station, Units 1 and 2
Facility Docket Nos: 050-454; 050-455
Operating Tests Administered: October 28, 2019, through November 5, 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, 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
None N/A
3