IR 05000424/2014301
| ML14197A308 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/16/2014 |
| From: | Eugene Guthrie NRC/RGN-II/DRS/EB2 |
| To: | Tynan T Southern Nuclear Operating Co |
| References | |
| 50-424/14-301, 50-425/14-301 | |
| Download: ML14197A308 (22) | |
Text
UNITED STATES uly 16, 2014
SUBJECT:
VOGTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2NRC OPERATOR LICENSE EXAMINATION REPORT 05000424/2014301 AND 05000425/2014301
Dear Mr. Tynan:
During the period May 12-15, 2014 the Nuclear Regulatory Commission (NRC) administered operating tests to employees of your company who had applied for licenses to operate the Vogtle Electric Generating Plant Units 1 and 2. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating test 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 May 29, 2014.
All applicants passed both the operating test and written examination. There were 11 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 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 9, Supplement 1.
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 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: 50-424, 50-425 License Nos: NPF-68, NPF-81
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Docket No.: 50-424, 50-425 License No.: NPF-68, NPF-81 Report No.: 05000424/2014301 and 05000425/2014301 Licensee: Southern Nuclear Operating Company, Inc.
Facility: Vogtle Electric Generating Plant Units 1 and 2 Location: Waynesboro, GA 30830 Dates: Operating TestMay 12-15, 2014 Written ExaminationMay 29, 2014 Examiners: J. Amanda Toth, Chief Examiner, Senior Operations Engineer M. Donithan, Operations Engineer A. Goldau, Operations Engineer Approved by: Eugene F. Guthrie, Chief Operations Branch 2 Division of Reactor Safety Enclosure 1
SUMMARY
ER 05000424/2014301, 05000425/2014301, Operating Test May 12-15, 2014 & Written Exam
May 29, 2014; Vogtle Electric Generating Plant Units 1 and 2; Operator License Examinations.
Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 9, Supplement 1, 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 Vogtle Electric Generating Plant Units 1 and 2 staff developed both the operating test 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 test during the period of May 12-15, 2014. Members of the Vogtle Electric Generating Plant Units 1 and 2 training staff administered the written examination on May 29, 2014. All applicants passed both the operating test and written examination. All applicants were issued licenses commensurate with the level of examination administered.
There were 11 post-examination comments.
No findings were identified.
REPORT DETAILS
OTHER ACTIVITIES
4OA5 Operator Licensing Examinations
a. Inspection Scope
The NRC administered the operating test during the period of May 12-15, 2014. The NRC examiners evaluated one Reactor Operator (RO) and six Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. One RO applicant received a waiver of the operating test. Members of the Vogtle Electric Generating Plant Units 1 and 2 training staff administered the written examination on May 29, 2014 to two RO and six SRO applicants. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate Vogtle Electric Generating Plant Units 1 and 2, met the requirements specified in 10 CFR Part 55, Operators Licenses.
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.
The NRC evaluated the performance and fidelity of the simulation facility during the preparation and conduct of the operating test.
b. Findings
No findings were identified.
The NRC developed the written examination sample plan outline. Members of the Vogtle Electric Generating Plant Units 1 and 2 training staff developed both the operating test and the written examination. All examination material was developed in accordance with the guidelines contained in Revision 9, Supplement 1, 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.
Using NUREG-1021, the NRC determined that the licensees initial examination submittal was within the range of acceptability expected for a proposed examination.
One RO applicant and six SRO applicants passed both the operating test and written examination. One RO applicant received a waiver of the operating test and passed the written examination. All applicants were issued licenses.
Copies of all individual examination reports were sent to the facility Training Director for evaluation of weaknesses and determination of appropriate remedial training.
The licensee submitted 11 post-examination comments concerning the operating test. A copy of the final written examination and answer key may be accessed not earlier than May 31, 2016 in the ADAMS system (ADAMS Accession Numbers ML14170A742, ML14170B026). A complete text of the licensees post-examination comments can be found in ADAMS under Accession Number ML14170A073.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On May 15, 2014, the NRC examination team discussed generic issues associated with the operating test with Mr. Charlie Nesbitt, Training Director, and members of the Vogtle Electric Generating Plant Units 1 and 2 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 J. Acree, Operations Training Manager M. Brown, Assistant Training Director R. Dorman, Operations Shift Manager, ILT-19 Class Mentor T. Harris, Initial License Training Supervisor F. Howard, ILT-19 Class Mentor K. Jenkins, Exam Writer K. Lewis, Operations Training Coordinator K. Morrow, Licensing Engineer C. Nesbitt, Training Director A. Sweat, Lead Exam Writer J. Todd, Operations Director
FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS
A complete text of the licensees post-examination comments can be found in ADAMS under
Accession Number ML14170A073.
Post Exam Comment #1
Item
Scenario 2, Event 1: Atmospheric Relief Valve (ARV), 1PV-3030 fails open on Loop #4
Scenario 2, Event 6: Loss of 1AA02 due to fault on bus
Licensee Comment
In Simulator Scenario #2, while in Mode 1, Loop #4 ARV, 1PV-3030, failed open and was
manually closed by the crew. Subsequently, 1AA02 de-energized due to a fault on the bus.
The Scenario Guide states the following Tech Specs are required to be entered when 1AA02
is de-energized and DG1A is emergency tripped:
Tech Spec LCO 3.8.1, AC Electrical SourcesOperating, Cond. A, B, and E Tech Spec LCO 3.8.9, Distribution SystemsOperating, Cond. A
Based on the given conditions, is Tech Spec LCD 3.7.4, Atmospheric Relief Valves (ARVs),
Cond B, required to be entered?
Licensee Discussion I Justification
Candidates are expected to address only base LCO entries and not to perform an extensive
LOSF evaluation during simulator scenarios. The ability to perform formal LOSF evaluations
is tested using Admin JPMs and the candidate would be provided 10008-C, "Recording
Limiting Conditions for Operation." Candidates are expected to recognize the existence of a
significant LOSF due to multiple failures in the simulator and ensure LCO completion times
more conservative than the base LCO are met.
Per Tech Spec LCO 3.0.6, when a supported system LCO is not met solely due to a support
system LCO not being met, the Conditions and Required Actions associated with this
supported system are not required to be entered. Only the support system LCO ACTIONS
are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In
this event, additional evaluations and limitations may be required in accordance with Tech Spec 5.5.15, "Safety Function Determination Program (SFDP)." If a loss of safety function is
determined to exist by this program, the appropriate Conditions and Required Actions of the
LCO in which the loss of safety function exists are required to be entered.
Per Tech Spec LCO 3.7.4, Atmospheric Relief Valves (ARVs), Bases, the analysis assumes
three ARVs are OPERABLE at the start of the event. One of the ARVs is on the ruptured
SG, another ARV is assumed to fail to open, and the remaining ARV is used to perform the
RCS cooldown. However, there is also a scenario where the limiting single failure is the loss
of control power for the two remaining ARVs. In this case, the ARVs cannot be controlled
from the control room to initiate cooldown. The ARVs are equipped with local hand pumps
that can be used to open them manually. Given a tube rupture on one of the steam
generators with an operable ARV, and the limiting single failure being a loss of control power
to the remaining operable ARVs, only one ARV must be capable of being manually actuated
using its hand pump. If the ARV on the ruptured generator also has one of the functional
hand pumps, then only one of the remaining ARVs needs have a functional hand pump in
order to meet the safety analysis.
Since the electrical failure is bounded by LCO 3.7.4 Bases and the hand pump operation was
unaffected, no LOSF exists as a result of this failure. Therefore, the ARV is declared
inoperable, but the actions for LCO 3.7.4 are not taken as stated in LCO 3.0.6. The support
system LCO completion times are more conservative and bound the supported system. It is
consistent with training and simulator performance on similar failures to not list the supported
LCOs as described above. Candidates are not expected to list nor take the actions of these
supported system LCOs. If a candidate lists these supported system LCOs, it is not
incorrect it is simply not necessary.
Licensee Recommendation
Based on the discussion above, the licensee does NOT recommend adding Tech Spec LCO 3.7.4, Atmospheric Relief Valves (ARVs), Cond. B, to the Scenario #2 Guide list of Tech
Specs required to be entered. Additionally, add a note to the examiner to indicate that
supported systems are inoperable, but their associated required action statements are not
required to be entered.
NRC Discussion
The first event in the scenario resulted in a failure of the Loop #4 ARV. Technical Specification
(TS) Limiting Condition for Operation (LCO) 3.7.4 was met, because the requirement to have
three operable ARVs was still met. Later in the scenario, the loss of 1AA02 resulted in entry
into action statements for TS LCO 3.8.1 Conditions A/B/C and 3.8.9 Condition A. The impact of
the loss of 1AA02 on the ARVs was that ARVs #1 and #4 no longer had 480V power to their
motor operated valves (MOVs), and therefore could not be operated from the control room.
However, while ARV #4 was still inoperable as a result of scenario 2 event 1, ARV #1 was still
operable per the Basis for TS 3.7.4, in that it still had the ability to be manually cycled via local
hand pump operation. Thus, there were still three operable ARVs, two of which had the ability
to be manually cycled from the control room.
NRC Resolution
The NRC agrees with the facility recommendation to leave TS 3.7.4 Cond B off the list of those
TS requiring entry. However, the NRC finds that entry into TS 3.7.4 Cond B would be wrong for
this scenario.
Post Exam Comment #2
Item
Scenario 2, Event 6: Loss of 1AA02 due to fault on bus
Licensee Comment
In Simulator Scenario #2, while in Mode 1, 1AA02 de-energizes due to a fault on the bus. The
Scenario Guide and 18031-C, "Loss of Class 1E Electrical Systems," Step A17 state:
A17. Verify battery charger in service for non-1E batteries:
- ND1
- ND2
- ND3A
- ND3B
The Scenario Guide states ND1 and ND3A will not have battery chargers in service and that the
crew will dispatch operators to place the required ND1 and ND3A battery chargers in service.
Based on the given conditions, what are the primary loads supplied by ND1 and ND3A, and
what are the potential consequences if the battery chargers are not placed in service?
Additionally, would this constitute a critical task per NUREG 1021, Appendix D, which states,
"Examples of CTs involving essential safety actions include those for which a crew
demonstrates the following abilities: take one or more actions that would prevent a challenge to
plant safety?"
Licensee Discussion I Justification
An extensive review of all loads associated with 1ND1 and 1ND3A was made on one-line
drawings 1X3D-AA-H03A, H03C, H07A, H07B, F28A, G02B, and G058, and utilizing
Switchgear Notes. No supplied equipment was found that would place the plant in a significant
transient or significantly challenge power operation.
- 1ND1 provides power to the Turbine Generator backup seal oil pump.
- 1ND3A loads all have redundant power supplies from regulating transformers that
auctioneer with the exception on 1ND31 and 1ND33.
- 1ND31 loads:
o Control power to non-1E switchgear.
o Various non-1E air operated valves that fail closed.
o Some non-1E alarms functions.
- 1ND33 loads:
o Control power to non-1E switchgear.
Licensee Recommendation
Based on review of the associated load, the licensee does NOT recommend revising the
Scenario #2 Guide to indicate placing the ND1 and ND3A battery chargers in service is a critical
task.
NRC Discussion
18031-C, Loss of Class 1E Electrical Systems, step A17 requires battery chargers be placed
in service for non-1E batteries. In the scenario, the battery chargers for 1ND1 and 1ND3A were
not in service, and the batteries were discharging to maintain the loads they supplied.
NUREG 1021, Appendix D (Simulator Testing Guidelines), D.1.a (Critical Task Methodology,
Identification of Critical TasksSafety Significance) states:
In reviewing each proposed CT, assess the task to ensure that it is essential to safety. A
task is essential to safety if its improper performance or omission by an operator will
result in direct adverse consequences or significant degradation in the mitigative
capability of the plant.
The loss of the loads that are supplied by batteries 1ND1 and 1ND3A would not significantly
complicate the recovery effort if batteries 1ND1 and 1ND3A fully discharged. Additionally, the
batteries would not be expected to fully discharge during the time-frame of the scenario. The
NRC agrees with the licensee that placing the battery chargers in service for 1ND1 and 1ND3A
does not meet the criteria specified in Appendix D of NUREG 1021 for classification as a critical
task.
NRC Resolution
The NRC accepts the licensees comment, and the guide for Scenario #2 will remain unchanged
with respect to performance of step A17 of 18031-C.
Post Exam Comment #3
Item
Scenario 2, Event 6: Loss of 1AA02 due to fault on bus
Licensee Comment
In Simulator Scenario #2, while in Mode 1, 1AA02 de-energizes due to a fault on the bus.
DG1A starts, but its output breaker does not close due to the 1AA02 bus fault.
The Scenario Guide states the crew will emergency trip DG1A at Step 4 of 18031-C, "Loss of
Class 1E Electrical Systems."
Based on the given conditions, what are the potential consequences of allowing DG1A to run
with its output breaker open until ALB35-C04, DG1A HI TEMP JACKET WATER OUT, is
received? Additionally, would this constitute a critical task per NUREG 1021, Appendix D,
which states, Examples of CTs involving essential safety actions include those for which
operation or correct performance prevents the following: degraded emergency core cooling
system (ECCS) or emergency power?
Licensee Discussion I Justification
Per the System Engineer, the EDGs are designed to operate for 3 minutes, fully loaded, with
no NSCW flow through the jacket water heat exchanger without exceeding any temperature
limitations. Per ARP 17035-1, ALB35-C04 alarms at 195F, and a NOTE for the annunciator
indicates the EDG will trip at 200F if in normal mode. If an emergency start signal is present,
ALB35-C05, DG1A TRIP HI TEMP JACKET WATER, will annunciate, but the EDG will not
trip.
Since the EDG output breaker was open, the EDG was running unloaded. The minimal heat
input would greatly extend the allowable run time, although no calculations exist to determine
the exact time/temperature relationship. Since the EDG was stopped prior to ALB35-C05
alarming, the automatic trip setpoint had not been reached.
Furthermore, since 1AA02 was faulted, it would not have constituted a possible recovery
path. If the EDG were allowed to run until damage had occurred, the loss of the EDG would
not increase the degradation of the ECCS or Electrical system because that portion of the
Electrical system was already lost. Therefore, stopping the EDG does not constitute a critical
task per NUREG 1021, Appendix D.
Licensee Recommendation
Based on the discussion above, the licensee does NOT recommend revising the Scenario #2
Guide to designate tripping DG1A by the crew to be a critical task.
NRC Discussion
18031-C, Loss of Class 1E Electrical Systems, step 4RNO requires the affected emergency
diesel generator (DG1A) to be emergency tripped, since it started but did not tie to the bus due
to a fault on the bus. Additionally, since an emergency start signal was present, the EDG would
have been prevented from automatically tripping upon receipt of ALB35-C05.
NUREG 1021, Appendix D (Simulator Testing Guidelines), D.1.a (Critical Task
Methodology, Identification of Critical Tasks-Safety Significance) states:
In reviewing each proposed CT, assess the task to ensure that it is essential to safety. A
task is essential to safety if its improper performance or omission by an operator will
result in direct adverse consequences or significant degradation in the mitigative
capability of the plant.
Additionally, NUREG 1021, Appendix D, D.1.a states:
Examples of CTs involving essential safety actions include those for which operation or
correct performance prevents the following:
- degraded emergency core cooling system (ECCS) or emergency power capacity
The loss of an EDG, in general, would be safety significant. Reaching the 200F trip setpoint
without operator action to emergency trip the EDG (because with an emergency start signal
present, the EDG will not automatically secure) could result in damage to the EDG, and would
normally meet the criteria in Appendix D for a critical task. However, in this scenario, the fault
was on 1AA02, and DG1A could not have supplied power to the bus. Thus, even if DG1A had
reached 200F, the loss of DG1A would not have complicated the recovery effort in this specific
scenario.
NRC Resolution
The NRC accepts the licensees comment, and the guide for Scenario #2 will remain unchanged
with respect to emergency tripping DG1A.
Post Exam Comment #4
Item
Scenario 4, Event 5: SG Pressure Instrument PI-525A fails low
Licensee Comment
In Simulator Scenario #4, while in Mode 1, 1PT-525, Steam Generator Pressure, fails low.
This is the only steam generator pressure instrument failed.
The Scenario Guide states the following Tech Specs are required to be entered for this
instrument failure:
3.3.2, ESFAS Instrumentation, Fu 1.e, Cond.
- D.
3.3.3, Post Accident Monitoring Instrumentation, Fu 8, Cond. 8, G, H, I. (INFO only)
3.3.4, Remote Shutdown System, Fu 13, Cond A
Based on the given conditions, is Tech Spec LCD 3.3.2, ESFAS Instrumentation, Fu 4.d.(1),
Cond. D, and Fu 4.d.(2), Cond. D, required to be entered?
Licensee Discussion I Justification
A review of the scenario indicates that all the correct Tech Specs were listed for the specific
event on Form ES-D-2, but not all were included in the Scenario Outline, Form ES-0-1.
On Page 2 of 2 of Form ES-D-2 for Event #5 (Step # F8):
- Reference Tech Specs for ESFAS Instrumentation, PAMS Instrumentation, and
Remote Shutdown Instrumentation.
- LCO 3.3.2, Fu 1e, Condition D
- LCO 3.3.2, Fu 4d(1), Condition D
- LCO 3.3.2, Fu 4d(2), Condition D
- LCO 3.3.3, Fu 8, Condition B
- LCO 3.3.4, Fu 13, Condition A
Licensee Recommendation
The licensee recommends adding Tech Spec LCO 3.3.2, ESFAS Instrumentation, Fu 4.d.(1),
Cond. D, and Fu 4.d.(2), Cond. D, to the Scenario #4 Outline, Form ES-D-1, list of Tech
Specs required to be entered for Event 5.
NRC Discussion
NUREG 1021 Appendix D (Simulator Testing Guidelines), B.3 (Integrated Scenario
DevelopmentSelect and Document Events) states:
For each event listed on Form ES-D-1, prepare a Form ES-D-2, Required Operator
Actions (or equivalent), by entering the scenario, event, and page numbers and a brief
description of the event at the top of the form.
Every required operator action should be included on Form ES-D-2; this is particularly
important for the critical tasks (refer to Section D, Critical Task Methodology) and other
verifiable actions and behaviors that will provide a useful basis for evaluating the
operators competence.
The inclusion of an events TS on the D-1 form, while helpful, is not required. The NRC
recognizes that the licensee included TS 3.3.2, ESFAS Instrumentation, Fu 4.d.(1) Condition D,
and Fu 4.d.(2) Condition D on the D-2 form for Event #5, but did not include them on the D-1
form. This oversight is administrative in nature, and administratively adding them to the D-1
form does not constitute a change in the scenario guide. However, while reviewing the modes of
applicability for Fu 4.d.(2), it was determined that Fu 4.d.(2) did not apply due to the event being
outside the mode of applicability.
NRC Resolution
The NRC accepts a portion of the licensees comment: LCO 3.3.2, Fu 4d(1), Condition D will be
administratively added to the D-1 form (pen-and-ink change) for consistency, and LCO 3.3.2, Fu
4d(2), Condition D will be removed from the D-2 form (pen-and-ink change).
Post Exam Comment #5
Item
Scenario 4, Event 7: Loop #4 Bypass Feed Regulating Valve (BFRV) fails closed
Licensee Comment
In Simulator Scenario #4, while in Mode 1, the Loop #4 BFRV fails to control in automatic.
The BFRV can be controlled successfully when taken to manual.
The Scenario Guide states no Tech Specs apply for this condition.
Based on the given conditions, is Tech Spec LCO 3.7.3, MFIVs and MFRVs and Associated
Bypass Valves, Cond. C, required to be entered?
Licensee Discussion I Justification
The solenoid valve that vents air and allows the BFRV valves to close automatically on a
FWI signal is upstream of the control air functions. Therefore, the isolation function will still
occur. However, the specific knowledge of the failure mechanism is not known to the
candidate at the time of failure. As such, the Operability Determination process is entered
per NMP-AD-012, "Operability Determinations and Functionality Assessments." A component
remains operable as long as a reasonable expectation of operability remains.
NMP-AD-012 states a subsequent determination of operability should be based on the
licensee's "reasonable expectation," from the evidence collected, that the Systems,
Structures, or Components (SSCs) are operable and that the Operability Determination will
support that expectation. Reasonable expectation does not mean absolute assurance that
the SSCs are operable. The SSCs may be considered operable when there is evidence that
the possibility of failure of an SSC has increased, but not to the point of eroding confidence in
the reasonable expectation that the SSC remains operable. The supporting basis for the
reasonable expectation of SSC operability should provide a high degree of confidence that
the SSCs remain operable. It should be noted that the standard of "reasonable expectation"
is a high standard, and that there is no such thing as an indeterminate state of operability; an
SSC is either operable or inoperable.
Without the specific information needed for the operability call and the BFRV control failure, it
is understandable that a candidate would conclude that reasonable expectation of operability
no longer existed, and concurrent with NMP-AD-012 guidance, immediately declare the
BFRV inoperable.
Licensee Recommendation
The licensee recommends adding Tech Spec LCO 3.7.3, MFIVs and MFRVs and Associated
Bypass Valves, Cond. C, to the Scenario #4 Guide list of Tech Specs with a note to the
examiner indicating the candidate may enter LCO 3.7.3 if the operability of the BFRV is in
question (awaiting an operability determination).
NRC Discussion
Event 7 of the scenario resulted in Loop #4 BFRV failing to control in automatic with the valve
travelling in the shut direction (the malfunction was that the valve failed closed while the
controller was selected to Automatic). The failure mechanism was such that the BFRV could
be controlled successfully in manual. According to the basis for TS 3.7.3, the Main Feed
Isolation Valves (MFIVs) and Main Feed Regulating Valves (MFRVs), and their associated
bypass valves, are considered OPERABLE when isolation times are within limits and capable of
closing on an isolation actuation signal. The Loop #4 BFRV still operated in manual, and the
solenoid valve that vents air and allows the BFRV to isolate on a Feedwater Isolation (FWI)
signal is separate from the control air function for the BFRV. Since NMP-AD-012 states that
determination of operability should be based on the licensees reasonable expectation from
the evidence collected, it would not be reasonable to conclude that both the control air valve
and the solenoid valve associated with FWI failed, especially given indication that the Loop #4
BFRV operated in manual.
NRC Resolution
The NRC rejects the licensees comment, and the guide for Scenario #4 will remain unchanged
with respect to entry into TS 3.7.3 Condition C.
Post Exam Comment #6
Item
Simulator JPM 'a': Perform a Manual Makeup to the VCT
Licensee Comment
In Simulator JPM 'a', "Perform a Manual Makeup to the VCT," the candidate is directed to
raise VCT level from 32% to 50%. JPM 'a', Step 2, is a critical step that reads:
NOTE
Volumetric change in VCT is equal to 19.2 gallons per percent change in level.
4.6.1.1 Set TOTAL MAKEUP Integrator 1-FQI-0111 to the desired amount of Total
Makeup Water.
Standard: Candidate reviews CAUTIONS and NOTE and calculates the total makeup
required to raise VCT level from 32% to 50% (18%) using:
(19.2 gallons / % x 18% = 345.6 gallons) and sets 1-FQI-0111 to 345 - 346
gallons by depressing the gray pushbutton under the digit to be
changed (red pushbutton will reset reading to all zeros).
Based on the given conditions, does rounding the calculated value to 350 gallons meet the
expectation for completion of this step?
Licensee Discussion I Justification
Rounding the total required makeup up to 350 gallons is an acceptable practice. The
additional 4.4 gallons would have resulted in a VCT level change from 50% to 50.2%, which
is neither significant nor detectable using QMCB indication.
Licensee Recommendation
To more closely represent the initiating cue of raising VCT level from 32% to 50%, the
licensee recommends revising Simulator JPM 'a', Step 2, Standard to a new range of 326.4-
364.8 gallons of makeup to be added, which is equivalent to 50+/-1% VCT level change.
NRC Discussion
NUREG 1021 Appendix C (Job Performance Measure Guidelines), B.3 (Developing and
Reviewing JPMsDevelop Performance Criteria) states:
The JPM must clearly identify the task standard (i.e., the predetermined qualitative
and/or quantitative outcome) against which task performance will be measured. Every
procedural step that the examinee must perform correctly (i.e., accurately, in the proper
sequence, and at the proper time) in order to accomplish the task standard shall be
identified as a critical step and shall have an associated performance standard.
The task standard for Simulator JPM a, Perform a Manual Makeup to the VCT, JPM step 2
directs the applicant to raise VCT level from 32% to 50%, and the standard includes a
calculated value of 345.6 gallons as being required to accomplish the desired change in volume.
The NRC understands that, if given the direction in the Control Room to raise VCT level to 50%
(as given in the initiating cue), it would not be against the expectation of the Operations
Department for the final volume of the VCT to be between 49% and 51%. The NRC agrees with
the facility that it is acceptable for the total allowable calculated volume for this step, in gallons,
to reflect a VCT final volume in the range of 49-51%.
NRC Resolution
The NRC accepts the licensees comment, and the performance standard for JPM step 2 will be
revised to allow a range of 326.4-364.8 gallons of makeup water.
Post Exam Comment #7
Item
Simulator JPM 'a: Perform a Manual Makeup to the VCT
Licensee Comment
In Simulator JPM 'a', "Perform a Manual Makeup to the VCT," the candidate is directed to raise
VCT level from 32% to 50%. JPM 'a', Step 3, is a critical step that reads:
Based on the given conditions and using 350 gallons from the previous step in the equation,
is a value of 43.3 gallons of boric acid acceptable if the calculated value should be 45.35
gallons?
Licensee Discussion I Justification
Based on the discussion and recommendation of Post-Exam Comment #6, a value of 326.4 -
364.8 gallons of makeup is acceptable for satisfactory completion of Step 4.6.1.1. Therefore, the
acceptable range for the completion of the calculation in Step 4.6.1.2 is +/- 0.1 gallon based on the
value determined in Step 4.6.1.1. As described in Post-Exam Comment #6, the acceptable
range for Step 4.6.1.2 would be 45.35 +/- 0.1 gallons using 350 gallons of boric acid in the
calculation.
Licensee Recommendation
The licensee recommends revising Simulator JPM 'a', Step 3 Standard to indicate a range of +/-0.1
gallons based on the total makeup value determined by the candidate in Step 4.6.1.1.
NRC Discussion
NUREG 1021 Appendix C (Job Performance Measure Guidelines), B.3 (Developing and
Reviewing JPMsDevelop Performance Criteria) states:
The JPM must clearly identify the task standard (i.e., the predetermined qualitative
and/or quantitative outcome) against which task performance will be measured. Every
procedural step that the examinee must perform correctly (i.e., accurately, in the proper
sequence, and at the proper time) in order to accomplish the task standard shall be
identified as a critical step and shall have an associated performance standard.
The task standard for Simulator JPM a, Perform a Manual Makeup to the VCT, JPM step 3
directs the applicant to calculate the required volume of boric acid for the total volume of
makeup determined from the previous step. The existing standard includes an allowable
calculated range of 44.7-44.9 gallons of boric acid, based on a calculation of 345.6 gallons of
boric acid in the previous step. Based on the resolution to post exam comment #6, the NRC
agrees that the calculated volume of boric acid may change, and the NRC agrees that +/-0.1
gallon is an acceptable range for the calculated value. Moreover, in accordance with NMP-OS-
006, Operations Performance Indicators, addition of boric acid outside of this range in the
actual plant would require screening in accordance with Attachment 1, Reactivity
ManagementPWR, to determine if a Vogtle Performance Indicator threshold was exceeded
due to a calculation error; therefore expansion of the boric acid band beyond this range would
be unacceptable.
NRC Resolution
The NRC accepts the licensees comment that a calculated value for volume of boric acid based
on the total makeup volume determined in the previous JPM step, including a band of +/-0.1
gallons, is acceptable. The task standard in simulator JPM a will be annotated to reflect the
allowance of a wider band from the previous JPM step (post exam comment #6). The intent of
the JPM step remains unchanged.
Post Exam Comment #8
Item
Simulator JPM f: Synchronize Main Generator to the Grid
Licensee Comment
In Simulator JPM f, "Synchronize Main Generator to the Grid," the candidate is directed in
Step 10 to:
Simulator JPM f, Step 14, is a critical step that reads:
Based on the given conditions, what rotation speed in the FAST (clockwise) direction is
acceptable to meet the intent of this step?
Licensee Discussion I Justification
Per NMP-AP-002, "SNC Fleet Procedures Writers' Guide," critical steps are procedure steps,
series of steps, or actions that, if performed incorrectly, will cause irreversible, intolerable
harm to plant equipment, people, or significantly impact plant operation. The critical aspect of
the task is the direction of rotation and the general speed of rotation. The generator auto-sync
circuit is equipped with a sync check relay. When the sync scope rotation is correct and the
rotation speed is within an allowable range, the AUTO SYNC PERMISSIVE LIGHT
immediately below the sync scope illuminates when the sync scope reaches the 12 o'clock
position. If any of the required parameters is not correct, this light will not illuminate and
consequentially, the generator output breakers will not automatically close when the AUTO
SYNC PERMISSIVE pushbutton is depressed. Since incorrect performance cannot close the
generator output breakers, this cannot be a critical step.
Licensee Recommendation
A procedure revision suggestion for 13830-1/2, Step 4.1.3.10.c., has been submitted to
remove the "critical" step designator (CR #822337).
The licensee recommends revising Simulator JPM 'f, Step 14 to remove the Critical Step
designation. The licensee also recommends revising Simulator JPM 'f, Step 14, Standard to
require rotation in the fast (clockwise) direction. Subsequent steps of the JPM will require
checking the auto-sync relay scheme as indicated by the red indicating light lit and the output
breaker closure.
NRC Discussion
NUREG 1021 Appendix C (Job Performance Measure Guidelines), B.3 (Developing and
Reviewing JPMsDevelop Performance Criteria) states:
The JPM must clearly identify the task standard (i.e., the predetermined qualitative
and/or quantitative outcome) against which task performance will be measured. Every
procedural step that the examinee must perform correctly (i.e., accurately, in the proper
sequence, and at the proper time) in order to accomplish the task standard shall be
identified as a critical step and shall have an associated performance standard.
The NRC understands that the generator auto-sync circuit prevents closure of the generator
output breakers if any of the required parameters are outside the range of acceptability.
Additionally, since the synchroscope was already rotating in the correct direction, regardless of
an applicants manipulation of the speed of the synchroscope, the auto-sync circuit would
prevent them from closing the generator output breaker inappropriately (i.e., if they tried to close
the breaker when the required conditions were not met, the breaker would not close). As such,
the NRC agrees that JPM step 14 should not be designated as a critical step. However, the
procedure specifically gives a band of acceptability of 10-15 seconds per rotation, and
regardless of an outstanding procedure revision request to remove the Critical Step notation in
the procedure, the procedure used by the applicants in the performance of this JPM had the 10-
second requirement that was annotated as a Critical Step in the procedure itself.
NRC Resolution
The NRC accepts the licensees comment that JPM step 14 should be revised to remove the
critical step designation, because there is no way to incorrectly perform the step (i.e., it is not
failable). However, the NRC rejects the licensees comment to revise the JPM step 14 task
standard to simply require rotation in the fast direction, irrespective of rotation speed. The
procedure the applicants were given specifies a rotation speed, and while the speed may not be
critical, it is still a procedurally directed action.
Post Exam Comment #9
Item
Simulator JPM 'g': Manually Align Control Room Isolation on High Radiation
Licensee Comment
In Simulator JPM 'g', "Manually Align Control Room Isolation on High Radiation," high
radioactivity levels are detected in the incoming duct for the Unit 1 Control Room (1RE-12116
and 1RE-12117 are in high alarm). Normal Control Room HVAC is in service, and an
automatic Control Room Isolation does not occur.
1HV-12130, Control Room Return Fan Inlet Air Damper (CREFS 'A'), and 1HV-12131,
Control Room Return Fan Inlet Air Damper (CREFS 'B'), are simultaneously opened
(manually). The dampers then re-close when the handswitch is released.
Based on the given conditions and with no previous action taken, what are the potential
consequences of partially opening 1HV-12130, Control Room Return Fan Inlet Air Damper
(CREFS 'A), and 1HV-12131, Control Room Return Fan Inlet Air Damper (CREFS 'B),
simultaneously? Would this action increase the radioactivity levels in the Control Room?
Licensee Discussion I Justification
A review of P&IDs AX4D8206-1, 2, and 3 determined that a backdraft damper exists
immediately downstream of both 1HV-12130 and 1HV-12131. These backdraft dampers
would prevent any reverse flow through the associated ductwork. Furthermore, both suction
paths originate from a common supply header. Having both dampers open simultaneously
would not result in a change in the airflow into or out of the control room envelope.
Additionally, the flow path created by having both paths open does not create a "short circuit"
flow path. Worst case, a parallel flow path could be established that would decrease the
Control Room envelope positive pressure slightly due to increasing the amount of
recirculation air flow and reducing the amount of makeup air. However, Control Room dose
rate calculations take into account the air intakes being completely isolated and therefore this
condition is bounded.
Licensee Recommendation
The licensee has determined there were NO negative consequences due to partially opening
1HV-12130, Control Room Return Fan Inlet Air Damper (CREFS 'A'), and 1HV-12131,
Control Room Return Fan Inlet Air Damper (CREFS 'B'), simultaneously.
NRC Discussion
The NRC understands that review of the applicable P&ID documents determined that backdraft
dampers exist that would prevent reverse flow through the associated ductwork.
NRC Resolution
This comment does not require resolution. The NRC accepts the technical analysis that there
are no negative consequences to simultaneously opening 1HV-12130 and 1HV-12131.
Post Exam Comment #10
Item
Admin RO JPM 'e': Perform ERO Recall and ENN Notification
Licensee Comment
In Admin RO JPM 'e', "Perform ERO Recall and ENN Notification," the candidate is directed to
notify and confirm receipt of the Emergency Notification Form using the manual method. Admin
RO JPM 'e', Step 17, is a critical step that reads:
Based on the given conditions, is it acceptable for the candidate to wait for an agency to
request an authentication code (i.e. ENN communicator does not query the agencies
concerning message authentication)?
Licensee Discussion I Justification
During Emergency Notification Form (ENF) transmission, the State of South Carolina typically
requires an authentication code early in the communication. ENN Communicators are trained to
respond to the authentication code request. The ENN Communicator is permitted to wait for the
code to be requested by the state agency to avoid jeopardizing the 15-minute notification time.
The authentication code is not a Federal requirement for ENN transmission; it is a requirement of
the State of South Carolina EM
- A. Therefore, it is acceptable for the ENN Communicator to wait
for an authentication code to be requested from the state agency. As such, a cue from the
Simulator Operator would be necessary.
Licensee Recommendation
The licensee recommends revising Admin RO JPM 'e', Step 17, Standard to provide a note to the
examiner indicating that it is acceptable for the candidate to wait for an agency to request
authentication. When this occurs, a cue from the Simulator Operator will request an
authentication code as the State of South Carolina.
NRC Discussion
The NRC understands that request of an authentication code is a requirement of the State of
South Carolina EMA, rather than a Federal requirement for ENN transmission. Additionally,
NMP-EP-111-F06, Manual Emergency Notification Transmission and Confirmation
Instructions, does not specifically require the ENN communicator to request authentication.
5.0.c states, Provide the authentication code supplied by the system to the agency requiring
authentication verbally over the ENN. This sentence indicates that a code word will be
provided to the agency that requests authentication, and NMP-EP-111-F06 is silent on the ENN
communicator requesting authentication.
NRC Resolution
The NRC accepts the licensees comment that it is not critical for the applicant, as ENN
communicator, to request authentication, given the wording in step 5.0.c of NMP-EP-111-F06.
Based on the lack of an expectation from the Operations Department for the ENN
Communicator to specifically request authentication if a separate agency does not request
authentication, and based on the wording in step 5.0.c of NMP-EP-111-F06, the JPM guide for
Admin RO JPM e does not require revision, because other elements of step 5.0 of NMP-EP-
111-F06 are critical.
Post Exam Comment #11
Item
Admin SRO JPM 'e': Classify an Emergency Event and Complete the Emergency Notification
Form
Licensee Comment
In Admin SRO JPM 'e', "Classify an Emergency Event and Complete the Emergency
Notification Form," the candidate is directed to classify an emergency event and complete the
Emergency Notification Form. Admin SRO JPM 'e', Step 10, is a critical step and its Standard
reads:
Standard: Candidate marks Block 'A' (DRILL) and enters MESSAGE #1.
Based on the given conditions, is it acceptable for the candidate to mark Block 'B' (ACTUAL
EVENT) on Line 1 of the Emergency Notification Form?
Licensee Discussion I Justification
Per NMP-TR-416, "Licensed Operator Continuing Training Program Administration," Step
4.11.11, satisfactory completion of the Emergency Notification Form (ENF) requires correctly
declaring the event as either drill or actual event. Unless an actual event is in progress, drill
should be selected. This is consistent with the training and evaluation of emergency
declarations in both the Initial and Requal training programs.
Licensee Recommendation
The licensee does NOT recommend revising Admin SRO JPM 'e', Step 10, Standard.
NRC Discussion
The NRC understands that the expectation at Vogtle is that, in the absence of an actual event in
progress, satisfactory completion of Line 1 on the ENF requires Block A (DRILL) to be
selected. However, the intent of this JPM on the NRC exam is to determine whether an
applicant would correctly complete the ENF form during an actual event. As such, the critical
aspect of Line 1 is that the block is not left blankthat is, the applicant understands that Line 1
is required to be filled in completely, and the applicant uses the information they are given in the
JPM to decide which block to mark (Block A or Block B).
NRC Resolution
The NRC rejects the licensees comment that the guide for SRO Admin JPM e should remain
unchanged with respect to the applicant marking Drill vice Actual. The guide for SRO Admin
JPM e step 10 standard will be revised to state, Candidate marks Block A (DRILL) OR Block
B (ACTUAL) and enters MESSAGE #1.
SIMULATOR FIDELITY REPORT
Facility Licensee: Vogtle Nuclear Generating Plant, Unit Nos. 1 & 2
Facility Docket No.: 50-424, 50-425
Operating Test Administered: May 12-15, 2014
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.
While conducting the simulator portion of the operating test, examiners observed the following:
Item Description
MDAFW-B did not respond During an exam scenario, MDAFW-B unexpectedly failed to start
properly to an automatic in AUT
- O. Upon simulator reset, the switch indicated that
start signal. repositioning was required and then cleared its flashing indication
without repositioning the switch. A simulator change request (mod
number 2014-05-016) was initiated to replace the switch.
Three MSIVs travelled During an exam scenario, three MSIVs were observed to travel in
shut during a simulator the shut direction without simulator booth operator manipulation.
scenario without simulator This resulted in the operating crew manually tripping the reactor
booth operator and taking actions in accordance with E-0. The simulator anomaly
manipulation. did not impact the major transient, and simulator performance after
the manual reactor trip was consistent with other scenarios run
that day. A hard-reset of the simulator was performed following
the scenario, and the subsequent scenario that day had no issues.
A simulator change request (mod number 2014-05-013) was
initiated to investigate the anomaly.
3