IR 05000321/2019301
| ML19290G553 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/17/2019 |
| From: | Eugene Guthrie Division of Reactor Safety II |
| To: | Gayheart C Southern Nuclear Operating Co |
| References | |
| 50-321/19-01, 50-366/19-01 50-321/OL-19, 50-366/OL-19 | |
| Download: ML19290G553 (19) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ber 17, 2019
SUBJECT:
EDWIN I. HATCH NUCLEAR PLANT - NRC OPERATOR LICENSE EXAMINATION REPORT 05000321/2019301 AND 05000366/2019301
Dear Mr. Dean:
During the period August 12 - 22, 2019 the Nuclear Regulatory Commission (NRC)
administered operating tests to employees of your company who had applied for licenses to operate the Edwin I. Hatch Nuclear Plant. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating tests and the written examination submittal with those members of your staff identified in the enclosed report. The written examination was administered by your staff on August 29, 2019.
Four Reactor Operator (RO) and ten Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One SRO applicant failed the written examination.
There were two post-administration comments concerning the written examination and two post-administration comments concerning the operating test. These comments, and the NRC resolution of these comments, are summarized in Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.
The initial examination submittal was within the range of acceptability expected for a proposed examination. All examination changes agreed upon between the NRC and your staff were made according to NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 11.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document
C. Gayheart 2 system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
If you have any questions concerning this letter, please contact me at (404) 997-4662.
Sincerely,
/RA/
Eugene F. Guthrie, Chief Operations Branch 2 Division of Reactor Safety Docket Nos.: 50-321, 50-366 License Nos.: DPR-57, NPF-5
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Examination Report Docket No.: 50-321, 50-366 License No.: DPR-57, NPF-5 Report No.: 05000321/2019301 and 05000366/2019301 Enterprise Identifier: L-2019-OLL-0025 Licensee: Southern Nuclear Operating Company (SNC)
Facility: Edwin I. Hatch Nuclear Plant, Units 1 and 2 Location: Baxley, GA Dates: Operating Test - August 12 - 22, 2019 Written Examination - August 29, 2019 Examiners: B. Caballero, Chief Examiner, Senior Operations Engineer D. Dumbacher, Senior Operations Engineer J. Bundy, Operations Engineer M. Emrich, Senior Reactor Technology Instructor T. Stephen, Senior Resident Inspector Approved by: Eugene F. Guthrie, Chief Operations Branch 2 Division of Reactor Safety Enclosure 1
SUMMARY
ER 05000321/2019301, 05000366/2019301; August 12 - 22, 2019 & August 29, 2019; Edwin I.
Hatch Nuclear Plant; Operator License Examinations.
Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 11, of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." This examination implemented the operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45, as applicable.
Members of the Edwin I. Hatch Nuclear Plant staff developed both the operating tests and the written examination. The initial operating test, written RO examination, and written SRO examination submittals met the quality guidelines contained in NUREG-1021.
The NRC administered the operating tests during the period August 12 - 22, 2019. Members of the Edwin I. Hatch Nuclear Plant training staff administered the written examination on August 29, 2019. Four Reactor Operator (RO) and ten Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. Fourteen applicants were issued licenses commensurate with the level of examination administered.
There were four post-examination comments.
No findings were identified.
REPORT DETAILS
OTHER ACTIVITIES
4OA5 Operator Licensing Examinations
a. Inspection Scope
The NRC evaluated the submitted operating test by combining the scenario events and JPMs in order to determine the percentage of submitted test items that required replacement or significant modification. The NRC also evaluated the submitted written examination questions (RO and SRO questions considered separately) in order to determine the percentage of submitted questions that required replacement or significant modification, or that clearly did not conform with the intent of the approved knowledge and ability (K/A) statement. Any questions that were deleted during the grading process, or for which the answer key had to be changed, were also included in the count of unacceptable questions. The percentage of submitted test items that were unacceptable was compared to the acceptance criteria of NUREG-1021, Operator Licensing Standards for Power Reactors.
The NRC reviewed the licensees examination security measures while preparing and administering the examinations in order to ensure compliance with 10 CFR §55.49, Integrity of examinations and tests. During the on-site preparatory week (July 8 - 12, 2019) the NRC audited a sample (approximately 10%) of the license applications (i.e.,
NRC Form 398) to confirm they accurately reflected the subject applicants qualifications.
The NRC administered the operating tests during the period August 12 - 22, 2019. The NRC examiners evaluated four Reactor Operator (RO) and eleven Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. Members of the Edwin I. Hatch Nuclear Plant training staff administered the written examination on August 29, 2019. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Edwin I. Hatch Nuclear Plant, met the requirements specified in 10 CFR Part 55, Operators Licenses.
The NRC evaluated the performance or fidelity of the simulation facility during the preparation and conduct of the operating tests.
b. Findings
No findings were identified.
The NRC developed the written examination sample plan outline. Members of the Edwin I. Hatch Nuclear Plant training staff developed both the operating tests and the written examination. All examination material was developed in accordance with the guidelines contained in Revision 11, of NUREG-1021. The NRC examination team reviewed the proposed examination. Examination changes agreed upon between the NRC and the licensee were made per NUREG-1021 and incorporated into the final version of the examination materials.
The NRC determined, using NUREG-1021, that the licensees initial examination submittal was within the range of acceptability expected for a proposed examination.One issue related to examination security was identified during preparation of the operating test. On April 30, 2019 the facility licensee generated Condition Report (CR) #10606406 and notified the NRC Chief Examiner that an unauthorized camera mount was discovered in the simulator, which had been installed for approximately two weeks by the class mentor to facilitate critique feedback during applicant training scenarios. On June 11, 2019, the NRC Office of Investigations (OI) conducted interviews with the personnel involved and toured the simulator. The NRC subsequently determined that no violation of NRC requirements existed, and further investigation was not warranted.
Four RO applicants and ten SRO applicants passed both the operating test and written examination. One SRO applicant passed the operating test but did not pass the written examination. Four RO applicants and ten SRO applicants were issued licenses.
Copies of all individual examination reports were sent to the facility Training Manager for evaluation of weaknesses and determination of appropriate remedial training.
The licensee submitted two post-examination comments concerning the operating test and two comments concerning the written examination. A copy of the final written examination and answer key, with all changes incorporated, and the licensees post-examination comments, may be accessed not earlier than September 20, 2021, in the ADAMS system (ADAMS Accession Number(s) ML19275K795, ML19275K820, and ML19275K870).
4OA6 Meetings, Including Exit
Exit Meeting Summary
On August 22, 2019 the NRC examination team discussed generic issues associated with the operating test with Mr. Sonny Dean, Site Vice President, and members of the Edwin I. Hatch Nuclear Plant staff. The examiners asked the licensee if any of the examination material was proprietary. No proprietary information was identified.
KEY POINTS OF CONTACT Licensee personnel Sonny Dean, Site Vice President Daniel A. Komm, Plant Manager Tim Krienke, Operations Director Michael Torrance, Engineering Director Paul Mothena, Radiation Protection Manager Keith Long, Regulatory Affairs Manager Mike Kelly, Training Director Mark Verbeck, Operations Training Manager Gary Dudek, Operations Training Corporate Functional Area Manager Ron Wheeler, Shift Supervisor Terry Jones, Class Coordinator Hank Strahley, Initial License Training Lead John J. Payne, Lead Reactor Engineer Zachary Howell, Reactor Engineer Andrew R. Belcher, Services Manager James Love, Licensing Engineer Charlie Edmund, NRC Exam Reviewer Anthony Ball, Exam Author Arthur Genereux, Operations Training Instructor NRC personnel Clinton Jones, Senior Resident Inspector Alan Blamey, Branch Chief, Division of Reactor Projects
FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS
A complete text of the facility licensee and applicant post-examination comments can be found
in ADAMS under Accession Number ML19275K870. The facility licensees post-examination
comments were for two operating test administrative job performance measures (JPMs) and
written exam SRO test items #78 and #93; the applicants post-examination comments were
only associated with the two written exam SRO test items.
Post-Examination Comment #1: SRO Admin JPM, Conduct of Operations 34SUV-019-2,
Surveillance & Tech Spec Required Actions
The facility licensee contended that Initial Condition #2 (2T48-N303A, Torus Temperature, is
out-of-service and inoperable) on the applicants task cue sheet was a typographical error that
was not detected prior to administration of the JPM; the facility licensee contended Initial
Condition #2 should have said 2T48-N304B (instead of -N303A). The reason the facility
licensee contended that Initial Condition #2 was a typographical error was because it did not
match the yellow colored inoperable elements on Attachment 1, Unit 2 SPDS Torus
Temperature Diagnostic, which was also provided to the applicants as Initial Condition #6. The
facility licensee contended that the typographical error caused the initial plant status presented
to the applicants to not be clear with respect to the total number of out-of-service torus
temperature elements. The facility licensee contended that the answer key should be revised to
allow two correct methods of performing the torus temperature calculation - one method for
applicants who performed the calculation with two out-of-service torus temperature elements,
since there were only two yellow elements indicated Attachment 1, and another correct method
of performing the calculation for the applicants who performed the calculation with three out-of-
service torus temperature elements because Initial Condition #2 said that 2T48-N303A was out-
of-service and inoperable.
Background
For this administrative JPM, the applicants were expected to complete Section 7.4 of 34SUV-
019-2, Surveillance Checks, to perform channel checks (critical steps) of two groups of torus
temperature elements and to calculate torus bulk average temperature (critical step). The
applicants were expected to provide the examiner with a completed copy of Section 7.4; several
line items in Section 7.4 were critical steps for satisfactory completion of the JPM. The seven
initial conditions presented to the applicants were:
According to Initial Condition #2 and #3, the only two inoperable torus temperature elements
were N303A and N308A; however, the Unit 2 SPDS Torus Water Temperature Diagnostic
(Initial Condition #6) provided to the applicants did not reflect N303A as yellow, i.e.,
inoperable. The Unit 2 SPDS Torus Water Temperature Diagnostic provided to the applicants
reflected N304B and N308A as inoperable (yellow). The discrepancy between the Initial
Conditions on the task cue sheet and the torus temperature element status on the Unit 2 SPDS
Torus Water Temperature Diagnostic (Initial Condition #6) is summarized as follows:
Task Cue Sheet Unit 2 SPDS Torus Water Temperature Diagnostic
N303A N304B (yellow)
N308A N308A (yellow)
The difference between the Task Cue Sheet and Attachment 1 was not detected during
validation of the administrative JPM.
NRC Resolution: Licensee comment accepted
NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Rev. 11,
Appendix C, Job Performance Measure Guidelines, Section B, Developing and Reviewing
JPMs, stated:
1. Specify Initial Conditions
Determine those system and plant conditions that would permit the task to be performed
realistically. They should provide sufficient information about the status of the plant and
system to facilitate task performance, without coaching the examinee.
Since there was an undetected discrepancy between the Task Cue Sheet and the Unit 2 SPDS
Torus Water Temperature Diagnostic, the initial condition information provided to the applicants
could be interpreted to mean that there were THREE inoperable torus temperature elements
(N303A, N304B, and N308A). It is operationally valid for a situation to exist where a torus
temperature element remained out-of-service (inoperable) even though the elements parameter
data field on the SPDS Torus Water Temperature Diagnostic was not yellow, i.e., appeared as
good data. For example, when a torus temperature element was previously out-of-service and
subsequently repaired, the elements parameter data field may NOT be yellow (because its
instrument signal would be valid) but the instrument would remain inoperable (out-of-service)
until post-maintenance testing was completed.
The intent of the JPM answer key was that only two elements were inoperable (N304B and
N308A) since these temperature elements parameter data fields on the Unit 2 SPDS Torus
Water Temperature Diagnostic were yellow.
During administration of the JPM, some of the applicants interpreted the initial condition
information to mean there were THREE inoperable torus temperature elements whereas other
applicants performed the channel check and bulk average temperature calculation assuming
that the TWO yellow temperature elements were the only inoperable elements. Therefore, the
initial condition information provided to the applicants was not clear.
The facility licensees recommendation to revise the answer key to allow two correct methods of
performing the torus temperature calculation - one method for applicants who performed the
calculation with two out-of-service torus temperature elements, since there were only two yellow
elements indicated on the Unit 2 SPDS Torus Water Temperature Diagnostic, and another
correct method of performing the calculation for the applicants who performed the calculation
with three out-of-service torus temperature elements, was accepted.
Post-Examination Comment #2: SRO Admin JPM, Emergency Classification - Complete
NMP-EP-142-F01, Emergency Notification Form
The facility licensee contended that the Unit 2 Shutdown Time (Line Item #12) on Emergency
Notification Form (NMP-EP-142-F01) was undefinable based on the Initial Conditions provided
to the applicants. For this reason, the facility licensee contended that the corresponding JPM
Step #34 should NOT be a critical step.
Background
For this administrative JPM, the applicants were first expected to identify emergency declaration
identifier SS1 for Unit 2, including the basis for the classification, within 15 minutes (time
critical step). After the applicants identified an emergency declaration identifier, the applicants
were provided with an Emergency Notification Form (ENF) and were expected to complete Line
Items 2 through 7, and Line Items 9 through 12, except unaffected unit status, within 15 minutes
(time critical steps). The facility licensees post exam comment was related to the Shutdown at
Time portion of Line #12 (see below).
The corresponding answer key for Line #12 was JPM Step 34, which was identified as a critical
step.
NRC Resolution: Licensee comment accepted
NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Rev. 11,
Appendix C, Job Performance Measure Guidelines, Section B, Developing and Reviewing
JPMs, states:
2. Specify Initial Conditions
Determine those system and plant conditions that would permit the task to be performed
realistically. They should provide sufficient information about the status of the plant and
system to facilitate task performance, without coaching the examinee.
The Unit 2 shutdown time was not provided to the applicants in the Initial Conditions; the ten
Initial Conditions provided to the applicants were:
Initial Condition #3 stated that the reactor scrammed and all rods inserted; however, the time
of the reactor shutdown was not included. The absence of a clear time of shutdown for Unit 2
was not detected during validation of the administrative JP
several applicants asked the proctor for the Unit 2 Shutdown Time. Therefore, the facility
licensees recommendation that JPM Step #34 (i.e., Unit 2 Shutdown Time portion of ENF Line
- 12) was not critical was accepted.
Post-Examination Comment #3: SRO Question #78
Six applicants contended that Question #78 should be deleted because the question stem did
not include all the necessary information to answer the first part of the question. Specifically,
the applicants contended that the question stem did not include information that 1) a second
licensed operator was assigned to verify rod withdrawal sequence and 2) the Rod Worth
Minimizer (RWM) was operable. The applicants contended that these two items were
necessary for the minimum allowable number of operable Intermediate Range Monitors (IRMs)
to be FOUR (one in each core quadrant) when the plant was in Mode 2. The applicants
contended that since the two items were missing from the stem of the question, an assumption
was required to be made that these two conditions existed. The applicants contended that
making the assumption that a second licensed operator was available and that the RWM was
operable was contrary to the guidance in NUREG-1021, Appendix E, Policies and Guidelines
For Taking NRC Examinations, Section B.7, which stated, in part:
When answering a question, do not make assumptions regarding conditions that are not
specified in the question unless they occur as a consequence of other conditions that
are stated in the question.
The facility licensee concurred with the applicants contention and recommended deleting the
question because the stem did not include information that a second licensed operator was
available and that the RWM was operable. The facility licensee based their recommendation on
NUREG-1021, ES-403, Grading Initial Site-Specific Written Examinations, Section D.1.b, which
stated, in part:
Despite the extensive reviews performed by both the NRC and the facility licensee
before examination administrationit is possible that a few isolated errors may be
discovered only after an examination has been administered. The following types of
errors areare most likely to result in post-examination changes agreeable to the NRC:
- A question with an unclear stem that confused the applicant or did not provide all
the necessary information.
Background
Question #78 was:
The answer key indicated that Choice C was the correct answer.
NRC Resolution: Licensee comment NOT accepted
Question #78 tested the applicants knowledge of the MINIMUM allowable number of IRMs; the
question was NOT testing the requirements needed to support operation with the minimum
number of IRMs. In accordance with 34GO-OPS-001-2, Plant Startup, the MINIMUM number of
IRMs required to be operable in Mode 2 was one channel in each quadrant of the core, i.e.,
FOUR IRMs. No assumption was required to answer the first part of Question #78 because 1)
control room staffing during a plant startup does not preclude a second licensed operator from
being assigned to verify the control rod withdrawal sequence, and 2) the stem of the question
did not contain any information that directly or indirectly implied the RWM was inoperable.
NUREG-1021, Appendix E, Policies and Guidelines For Taking NRC Examinations, Section B.7,
which stated, in part:
When answering a question, do not make assumptions regarding conditions that are not
specified in the question unless they occur as a consequence of other conditions that
are stated in the question.
The stem of the question did not contain any conditions/information that would, as a
consequence, lead to a lack of control room staffing or the RWM being inoperable.
The first part of the question tested the applicants knowledge of 34GO-OPS-001-2, Plant
Startup, Precaution & Limitation 5.1.3, which stated:
A minimum of four IRM channels must be operable for the control rod withdrawal block
instrumentation to be considered operable.
One channel in each quadrant of the core must be OPERABLE whenever the IRMs are
required to be OPERABLE. Both the RWM and a second Licensed Operator must verify
compliance with the withdrawal sequence when less than three channels in any trip
system are OPERABL
- E.
Additionally, 34GO-OPS-001-2, Step 7.1.14.2 and 7.1.14.3 stated:
(N/A IF < 3 IRM channels per trip system operable).
(N/A IF 3 IRM channels per trip system are operable).
- 7.1.14.3.1 Assign a second Licensed Operator to verify compliance with the
withdrawal sequence.
- 7.1.14.3.3 Check one IRM channel in each quadrant of the core is
Tech Spec 3.3.1.1 Reactor Protection System (RPS) Instrumentation required TWO
Channels per RPS trip system in Mode 2. This Tech Spec requirement was modified by
NOTE d, which stated:
Tech Spec 3.3.2.1, Control Rod Block Instrumentation, did NOT include a requirement
for IRM operability; Tech Spec 3.3.2.1 only included requirements for the Rod Block
Monitor (RBM), RWM, and Reactor Mode Switch Shutdown Position functions.
The first part of Question #78 asked for the MINIMUM number of IRMs required to be
operable in Mode 2. 34GO-OPS-001-2, Step 7.1.14.3 allowed for a minimum of TWO
IRMs to be operable, provided that at least one IRM channel was operable in each
quadrant of the core. Choice C (correct answer) stated: 1 IRM channel in each core
quadrant. Since there are four core quadrants, a minimum of four IRMs were required
to be operable. Choice C met the Tech Spec 3.3.1.1 RPS requirement that two
channels per RPS trip system were required in Mode 2 because Choice C specified a
minimum of four IRM channels. Therefore, the applicants and facility licensees
recommendation that Question #78 be deleted was NOT accepted.
Post-Examination Comment #4: SRO Question #93
Five applicants contended that the answer to Question #93 should be changed from
Choice A to Choice B because technical information, which was not known by the
exam authors during the written exam development, was subsequently discovered.
Specifically, the applicants contended that Reactor Building Component Cooling Water
(RBCCW) heat exchanger outlet valves (2P41-F440A and -F440B) are normally full-open
(instead of throttled) and the common downstream valve (2P41-F491) was used to adjust
RBCCW temperature (instead of the heat exchanger outlet valve).
The applicants contended that the RBCCW HX OUTLET TEMP HIGH (34AR-650-249-2)
annunciator response procedure did NOT contain a specific step that provided direct
guidance for adjusting the correct Plant Service Water Valve since it provided guidance
for adjusting 2P41-F440A or -F440B, instead of adjusting the common downstream valve
2P41-F491. Therefore, the applicants contended that the answer key should be changed
from Choice A to Choice
- B.
The facility licensee concurred with the applicants contention and cited NUREG-1021, ES-403,
Grading Initial Site-Specific Written Examinations, Section D.1.b, which stated, in part:
Despite the extensive reviews performed by both the NRC and the facility licensee
before examination administrationit is possible that a few isolated errors may be
discovered only after an examination has been administered. The following types of
errors areare most likely to result in post-examination changes agreeable to the NRC:
- Newly discovered technical information that supports a change in the
answer key.
Background
Question #93 was:
The answer key indicated that Choice A was the correct answer. Only the second part
of Question #93 was contested.
NRC Resolution: Licensee comment NOT accepted; question deleted
The guidance in the annunciator response procedure [RBCCW HX OUTLET TEMP
HIGH (34AR-650-249-2)] was inadequate for a high RBCCW heat exchanger outlet
temperature condition because it directed the operator to throttle a valve that was already
fully open; therefore, the question was technically inaccurate. The facility licensee
generated condition report (CR) #10647129 to correct the annunciator procedure.
The second part of Question #93 tested the applicants knowledge of whether the
RBCCW HX OUTLET TEMP HIGH (34AR-650-249-2) annunciator response procedure
contained a specific step that provides the direct guidance for adjusting the Plant
Service Water valve.
The answer key indicated that Choice A was correct based on Step 5.2 of the RBCCW
HX OUTLET TEMP HIGH (34AR-650-249-2) annunciator response procedure which
stated:
However, 34SO-P41-001-2, Plant Service Water System, Attachment 2, Plant Service
Water System Valve Lineup, identified that the following normal valve positions:
The PSW Return from RBCCW Heat Exchanger 2P42-B001A and -B001B Isolation
Valves (2P41-F440A and -F440B) were required to be full open even though the
RBCCW HX OUTLET TEMP HIGH (34AR-650-249-2) annunciator response procedure
provided guidance to throttle open the valves during a high temperature condition.
Therefore, the annunciator response procedure guidance was inadequate for a high
RBCCW heat exchanger outlet temperature condition because it directed the operator to
throttle a valve that was already fully open.
34SO-P41-001-2, Section 7.1.2 (Initial Division 1 Pump Startup) and Section 7.1.3 (Initial
Division 2 Pump Startup) included the following Step for correcting a high RBCCW heat
exchanger outlet temperature condition:
The applicants and facility licensees recommendation to change the answer key from Choice
A to Choice B was not incorporated because the test item was based on an inadequate
procedure. The facility licensee initiated condition report (CR) 10647129 to correct the
annunciator procedure. Although Step 5.2 of RBCCW HX OUTLET TEMP HIGH (34AR-650-
249-2) annunciator response procedure did technically include a specific step that provides the
direct guidance for adjusting the Plant Service Water valve, the guidance was incorrect
because it directed adjusting Isolation Valves 2P41-F440A and -F440B even though these
valves are already fully open. Therefore, the question was deleted from the exam.
SIMULATOR FIDELITY REPORT
Facility Licensee: Hatch
Facility Docket No.: 05000321/2019301 & 05000366/2019301
Operating Test Administered: August 12 - 22, 2019
This form is to be used only to report observations. These observations do not constitute audit
or inspection findings and, without further verification and review in accordance with Inspection
Procedure 71111.11 are not indicative of noncompliance with 10 CFR 55.46. No licensee
action is required in response to these observations.
No simulator fidelity or configuration issues were identified.
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