IR 05000261/2017301
| ML18026A678 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/25/2018 |
| From: | Gerald Mccoy Division of Reactor Safety II |
| To: | Kapopoulos E Carolina Power & Light Co |
| References | |
| IR 2017301 | |
| Download: ML18026A678 (18) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ary 25, 2018
SUBJECT:
H. B. ROBINSON STEAM ELECTRIC PLANT - NRC OPERATOR LICENSE.
EXAMINATION REPORT 05000261/2017301
Dear Mr. Kapopoulos:
During the period September 25 - October 5, 2017, the Nuclear Regulatory Commission (NRC)
administered operating tests to employees of your company who had applied for licenses to operate the H.B. Robinson Steam Electric 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 October 11, 2017.
Eight Reactor Operator (RO) and four Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One RO and one SRO applicant passed the operating test, but failed the written examination. One SRO applicant passed the written examination, but failed the operating test. One SRO applicant failed both the operating test and the written examination. There was one post-administration comment concerning the operating test, and five post-administration comments concerning the written examination. 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 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) 997-4551.
Sincerely,
/RA: Eugene F. Guthrie for/
Gerald J. McCoy, Chief Operations Branch 1 Division of Reactor Safety Docket No: 50-261 License No: DPR-23 Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
_________________________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRS RII:DRS SIGNATURE MKM3 EMAIL EFG FOR GJM1 NAME MMEEKS GMCCOY DATE 1/24/2018 1/25/2018 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No.: 05000261 License No.: DPR-23 Report No.: 05000261/2017301 Licensee: Carolina Power and Light Company Facility: H. B. Robinson Steam Electric Plant, Unit 2 Location: Hartsville, SC Dates: Operating Test - September 25 - October 5, 2017 Written Examination - October 11, 2017 Examiners: M. Meeks, Chief Examiner, Senior Operations Engineer D. Lanyi, Senior Operations Engineer G. Callaway, Senior Reactor Technology Instructor Approved by: Gerald J. McCoy, Chief Operations Branch 1 Division of Reactor Safety Enclosure 1
SUMMARY
ER 05000261/2017301; September 25 - October 5, 2017 & October 11, 2017; H. B. Robinson
Steam Electric Plant, Unit 2; 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 H. B. Robinson training staff developed both the operating tests and the written examination. The initial operating test, written Reactor Operator (RO) examination, and written Senior Reactor Operator (SRO) examination submittals met the quality guidelines contained in NUREG-1021.
The NRC administered the operating tests during the period September 25 - October 5, 2017.
Members of the Robinson training staff administered the written examination on October 11, 2017. Eight RO and four SRO applicants passed both the operating test and written examination. One RO applicant and one SRO applicant passed the operating test, but failed the written examination. One SRO applicant passed the written examination, but failed the operating test. One SRO applicant failed both the operating test and the written examination.
Twelve applicants were issued licenses commensurate with the level of examination administered.
There was one post-examination comment related to the operating test, and five post-examination comments related to the written examination.
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 (Reactor Operator (RO) and Senior Reactor Operator (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 administered the operating tests during the period September 25 - October 5, 2017. NRC examiners evaluated eight RO and seven SRO applicants using the guidelines contained in NUREG-1021. Members of the Robinson training staff administered the written examination on October 11, 2017. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the H. B. Robinson Steam Electric 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 H. B.
Robinson Steam Electric 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.
Revision 11 of NUREG-1021 contains newly-amplified guidance on the prevention of unnecessary challenges to the plant Reactor Protection System (RPS), as related to the assessment of critical tasks in the simulator scenario portion of the operating tests.
Specifically, NUREG-1021 Appendix D, page D-17, stated the following:
If an operator takes an action that the examiners did not expect, the examiners must further evaluate the individuals rationale for taking that action. Such preemptive actions may indicate a misunderstanding of plant conditions or a weakness in integrated plant knowledge that should be clarified with followup questions. Taking a preemptive manual action when an automatic action is imminent because of an incorrect action or inaction does not mitigate the initial incorrect action/inaction.
- Example: An applicant fails to manually control pressurizer pressure (where pressure is controllable per the validated scenario), and the pressure reaches a threshold at which the crew initiates a manual trip. This is a CT because pressure was intended to be a controllable variable in the scenario guide.
[.] Before administering the exam, developers and examiners should make an effort to identify events for which applicant inaction or common applicant error has the potential to result in an automatic RPS or ESF actuation. One method to accomplish this is to make a blanket statement in the scenario guide that Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology in Appendix D to NUREG-1021.
Using the above guidance, the NRC determined that two applicant crews created two (self-revealing) new critical tasks during the administration of the operating test. In one case, an applicant crew initiated an unnecessary manual reactor trip after failing to control pressurizer pressure during a failed-open spray valve event. In the other case, an applicant team initiated an unnecessary manual reactor trip after failing to control Volume Control Tank (VCT) level such that a manual reactor trip was required following an automatic swap over of charging pump suction alignment from the VCT to the Refueling Water Storage Tank (RWST).
Eight Reactor Operator (RO) and four Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One RO and one SRO applicant passed the operating test, but failed the written examination. One SRO applicant passed the written examination, but failed the operating test. One SRO applicant failed both the operating test and the written examination. Eight RO applicants and four 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 facility licensee submitted one post-examination comment concerning the operating test. The applicants submitted five post-examination comments concerning the written examination. A copy of the final RO and SRO written examinations and answer keys, with all changes incorporated, may be accessed not earlier than December 18, 2019, in the ADAMS system (ADAMS Accession Number(s) ML17354B267 and ML17354B270).
A copy of the Post Examination Comments (with PII redacted) may be accessed in the ADAMS system via ML17354B271.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On October 5, 2017, the NRC examination team discussed generic issues associated with the operating test with D. Hoffman, Operations Manager (acting Plant General Manager), and members of the H.B. Robinson Steam Electric Plant staff. The examiners asked the licensee if any of the examination material was proprietary. No proprietary information was identified. On December 13, 2017, the NRC discussed the results of the examination with members of the H.B. Robinson staff.
KEY POINTS OF CONTACT Licensee personnel L. Basta, Operations C. Caudell, Regulatory Affairs J. Conder, Operations G. Curtis, Assistant Operations Manager-Support R. Drehs, Supervisor Nuclear Training F. Giannone, Training Manager T. Giese, Manager Nuclear Operations Training F. Holbrook, SPD D. Hoffman, Operations Manager T. Pilo, Manager Regulatory Affairs J. Rackley, Operations Training Supervisor NRC personnel J. Rotton, Senior Resident Inspector A. Beasten, Resident Inspector
FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS
A complete text of the licensees post-examination comments can be found in ADAMS under
Accession Number ML17354B271.
1. Item
Simulator Scenario 6, Event 4; and Simulator Scenario 1, Event 3
Facility Licensee Comment:
The following comments are submitted on the RNP 2017301 Operating Examination. Both
pertain to an ITS entry into 3.3.6 Condition A for Function 4, that was not initially included in the
ES-D-2s for the scenarios identified, and it is an appropriate entry that should be added to the
exam.
N17-1-6
In simulator scenario N17-1-6, Event 4, Failure of Main Steam Line C Pressure Transmitter
(PT-495) HIGH, the exam team identified the following tech spec entry condition ONLY in the
ES-D-2: LCO 3.3.2 CONDITIONs A and
- D.
In the performance of this exam scenario on Day 2, the 3 crews that were examined, each SRO
additionally identified that they would enter ITS 3.3.6 Containment Ventilation isolation
instrumentation for Function 4, Safety Injection, shown below: (LCO 3.3.6 CONDITION A).
The exam team concluded that this ITS entry would be appropriate, and once the OWP-025,
STEAM GENERATOR PRESSURE (SGP) SGP-11 MAIN STEAM LINE C PRESSURE
TRANSMITTER PT-495 was completed this ITS would be exited. Entry into ITS 3.3.6 Condition
A for N17-1-6 Event 4 should be added to the Page 36 of the N17-1-6 ES-D-2.
N17-1-1
The same ITS 3.3.6 entry also was found to be applicable in Simulator Scenario N17-1-1 in
Event 3, failure of S/G A STEAM FLOW TRANSMITTER (FT-474) LOW. Again the initially
submitted ES-D-2 ONLY identified entry into ITS 3.3.2 CONDITIONs A and D. The crews that
were examined with this scenario on Day 4, also entered ITS 3.3.6. The exam team concludes
that this ITS entry would be appropriate, and once the OWP-034, STEAM FLOW (SF) SF-1
STEAM FLOW TRANSMITTER FT-474 was completed this ITS would be exited. Entry into
Condition A for N17-1-1 Event 3 should be added to the Page 34 of the N17-1-1 ES-D-2.
NRC Resolution
The licensees recommendation was accepted.
At Robinson, there is an interlock between the automatic Safety Injection actuation function
(covered by Technical Specification (TS) LCO 3.3.2) with the Containment Ventilation Isolation
function (which has specifications applicable in TS LCO 3.3.6). During development and
validation of simulator scenarios 1 and 6, the exam developers, the NRC, and the operators
involved in the validation of the scenarios agreed that only LCO 3.3.2 was required to be
entered for scenario 1, event 3, for a steam flow transmitter failure; and scenario 6 event 4, for a
steam line pressure transmitter failure.
However, as stated above by the facility licensee, following a post-examination assessment,
entry into both LCO 3.3.2 and LCO 3.6.6 was determined to be the technically correct
application of the appropriate specifications, given the conditions present in the scenario events.
Therefore, the NRC agreed with the facility licensees recommendation and modified the ES-D-2
forms for these two scenario events to add the additional entry into LCO 3.6.6 as technically
accurate, in accordance with TS, for the given conditions.
All applicants who were given Scenarios 1 and 6 during exam administration of the operating
test were evaluated as required to enter LCO 3.6.6 along with LCO 3.3.2 for these events.
2. Item
Question 1, K/A 007EK1.04
Applicant Comment
A Robinson licensed operator applicant recommended the following concerning Question 1:
To perform the calculation in this question a GFES equation from the GFES equation sheet
(provided by the licensed operator applicant as an enclosure to the post-examination comment)
is required. These equations are not required to be memorized. During the NRC exam a GFES
equation sheet was not provided to the applicants to use but was necessary to get the correct
answer.
Due to an equation sheet not being provided to the applicants this question should be excluded
from the exam as invalid.
Facility Comment
The applicants challenge is in regards to whether or not the GFES equation sheet was required
to correctly answer the question. The facility wrote the question as a closed book, no reference
provided question, and it was approved as such. Additionally, during the exam another
candidate asked if the GFES equation sheet should be provided, and the Chief Examiner
respond that the applicant should Answer the question with the information provided. The
facility maintains that position that the equation sheet is not required.
Recommend the applicant comment be rejected based on the above.
NRC Resolution
The NRC agreed with the facility licensees recommendation, and rejected the applicants
contention.
Contrary to the applicants contention, a GFES equation sheet is not required to elicit the correct
answer for this question. The applicant is required to know/understand/recall two specific facts:
(1) first, that during the time frame of the question, the indicated Reactor Start-Up-Rate (SUR)
will remain approximately a constant -1/3 DPM; and (2) secondly, that the Robinson setpoint for
POWER ABOVE P-6 reset is approximately 8 x 10-11 Amps for Intermediate Range N-35 (the
instrument specified in the question stem). Based on these two facts, and
knowledge/understanding of what decades of nuclear power implies, an applicant can use first
principles to determine that the reset will occur approximately 3.8 decades of power away from
the hypothetical current time of the question. Simple unit conversion/division will then provide
the applicant with the correct answer (3.8 decades/0.33 DPM = ~11.4 minutes). The GFES
equation sheet is not required for this determination from basic principles and site-specific
knowledge.
Note that there may be additional ways to determine the correct answer than the above
discussion; for example, an applicant may simply remember the approximate time post-trip that
the POWER ABOVE PERMISSIVE P6 status light extinguisheswhether from experience in
the plant or from experience during simulator trainingand determine the answer via that
mental process.
In any event, as agreed by the facility licensee, a GFES equation sheet was not required for this
question. In accordance with NUREG-1021, the written examination answer key was not
changed based on this contention.
3. Item
Question 12, K/A 055EG2.1.19
Applicant Comment
A Robinson licensed operator applicant recommended the following concerning Question 12:
The answer key states C is the correct answer. Contrary to this part one of the question
describes the function of ERFIS to use adverse containment set-points when evaluating
CSFSTs via SPDS is written as a general statement independent of plant conditions.
Regardless of plant conditions, ERFIS DOES use adverse containment set-points when
evaluating the CSFSTs via SPDS (reference material provided by the licensed operator
applicant as an enclosure to the post-examination comment). Containment pressure reduction
below 4 Psig does not remove the ability of SPDS to perform this function. If the intent of the
questions was to describe the current state of ERFIS and the SPDS program under current
plant conditions then the answers should have stated: EFRIS IS or IS NOT using adverse
containment set-points when evaluating CSFSTs via SPDS. These above conditions support
A as the correct answer.
Correct answer is A.
Facility Comment
Which ONE of the following correctly completes the statements below based on the current
plant conditions?
It is important to note that the stem of the question specifically asks the candidate to answer
based on current plant conditions.
The current plant condition provided in stem is,
- CURRENT CV pressure is 3.5 psig and slowly lowering
The ERFIS Safety Parameter Display System (SPDS) Critical Safety Function Status Trees
(CSFSTs) use adverse set-point values when adverse containment conditions are present.
Adverse containment conditions are defined as containment pressure greater than or equal to
four psig. The SPDS CSFSTs returns to normal set-point values when the adverse containment
conditions are no longer present. This is in accordance with the SPDS Software Requirement
Specification (RNP2-6004-SPDS-0001).
The stem of the question specifically asks about current plant conditions, therefore, for the
current plant containment pressure of 3.5 psig ERFIS does NOT use adverse containment set-
points when evaluating CSFSTs via SPDS.
Recommend the applicant comment be rejected based on above.
NRC Resolution
The NRC agreed with the facility licensees recommendation, and rejected the applicants
contention.
Contrary to the applicants contention, as the facility licensee pointed out, the question stem
specifically directed the applicant to answer the question based on the current plant
conditions [emphasis added]. The question stem stated that peak containment pressure
during the event was 4.5 psig, which requires the use of adverse containment setpoints in
determining critical safety function status (CSFST), and then containment pressure lowered to
3.5 psig, which is below the adverse setpoint requirement. Under the current plant conditions of
3.5 psig in containment, the automatic system (ERFIS) will NOT be using adverse setpoints in
its displays for CSFS
- T. However, procedure OMM-022 requires the use of adverse setpoints
for the current conditions because once adverse setpoints are required to be used they are
required to be used throughout the eventeven if containment pressure was to lower.
Therefore, the question highlights an important point of knowledge to the operators that they
must be able to interpret and understand computer indications for these given conditions.
Therefore, the NRC determined that C remains the only technically correct answer to the
question. In accordance with NUREG-1021, the written examination answer key was not
changed based on this contention.
4. Item
Question 23, K/A 060AA2.03
Comment
A Robinson licensed operator applicant recommended the following concerning Question 23:
The key states B is correct. Contrary to this part one of the question asks if a proposed
clearance COMPLETELY isolated the A Waste Gas Compressor from the Gaseous Waste
Disposal System. The clearance uses the following valves: WD-1611, 1670, 1665, 1669, and
3335. This clearance does not COMPLETELY isolate the A Waste Gas Compressor from the
Gaseious Waste Disposal system, as shown in the reference provided. To COMPLETELY
Isolate the A Waste Gas Compressor from the Gaseous Waste Disposal System the following
valves must be included on the clearance as well: WD-1643, WD-3336 and LCV-1030 A/B.
Piping runs 3/4-WD-739 and 1-1/2-WD-152R-175 are part of the Gaseous Waste Disposal
System. Additionally, WD-1643 and LCV-1030B and nomenclature were not mentioned in the
stem of the question so there was no way to tell what part of the system they are. These valves
go to a leg of piping which ends in an arrow pointing down. This arrow is unlabeled on the
reference drawing so there is no way to tell where the rest of this piping ends nor what system it
goes to. These above conditions support C as the correct answer.
Note: The question does not ask if the clearance is adequate.
Correct Answer is C.
Facility Comment
Which ONE of the following correctly completes the statements below?
The proposed clearance ________ completely isolates the A Waste Gas Compressor from the
Gaseous Waste Disposal System.
The proposed clearance does completely isolate the A Waste Gas Compressor from the
Gaseous Waste Disposal System. Neither, WD-1643 (Compressor A Moisture Separator
Drain), WD-3336 (Test Connection at PC-1028), nor LCV-1030B (WGC A Moisture Separator
Level Control) are required to isolate the A Waste Gas Compressor for the Gaseous Waste
disposal system because they dont isolate any sources of energy associated with the
compressor, nor do they establish a boundary between the compressor and the Gaseous Waste
Disposal System.
Recommend the applicant comment be rejected based on above.
NRC Resolution
The NRC agreed with the facility licensees recommendation, and rejected the applicants
contention.
Contrary to the applicants contention, as the facility licensee noted, the proposed isolations in
the question stem do isolate the A Waste Gas Compressor (WGC) from the Gaseous Waste
Disposal (GWD) system. During the exam administration, there were no questions from any
applicant concerning the nomenclature of the additional components mentioned by the applicant
above. Furthermore, the additional components identified by the applicant above as required to
isolate the A WGC do not, in fact, interface with the GWD system.
Therefore, the NRC determined that B remains the only technically correct answer to the
question. In accordance with NUREG-1021, the written examination answer key was not
changed based on this contention.
5. Item
Question 62, K/A 029A1.02
Comment
A Robinson licensed operator applicant recommended the following concerning Question 62:
The key states C is correct. Contrary to this part two of the question states the equipment
hatch is removed per OMM-033. No other components were referenced in the stem of the
question so it cannot be assumed that any other components were manipulated. Additionally,
there are times when the equipment hatch is removed and the leads for R-11 and R-12 are NOT
lifted (i.e., performance of OST-163 [Enclosure provided by the licensed operator applicant].
With the information provided in the stem of the question an R-11 alarm would cause HVE-1A to
automatically stop. These above conditions support A as the correct answer
Correct Answer is A
Facility Comment
Which ONE of the following correctly completes the statements below?
To commence the Purge operation ________ MUST be in operation.
If R-11, CV AIR & PLANT VENT PARTICULATE alarms, HVE-1A ________ automatically stop.
Per OMM-033 (Implementation of CV Closure), Section 5.6.1
If the Equipment Hatch has been removed, then the automatic closure signal
from R-11 and 12 has been defeated per Attachment 6 Section 1.0 Step 2.b,
in support of CM-603, Disassembly and Assembly of the Containment Equipment
Hatch and Missile Barrier. The automatic signal from SI and the Manual signal
are still available.
Section 1.0 Step 2.b states,
2. Ensure CV Purge has been established as follows: . ______
a. CV Purge is in progress per OP-921, Containment Air Handling ______
b. R-11 and R-12 leads are lifted as follows: . ______
(1) At Safeguards Cabinet, Rack 52 (rear), Train A, lift and tape
Cable C2279N at Terminal Board 6L, Terminals .. ________
(2) At Safeguards Cabinet, Rack 64 (rear), Train B, lift and tape
Cable C2279T at Terminal Board 6L, Terminals 1 and 2 .. ______
There is nothing in the stem of the question about any other procedures being in progress. With
the leads lifted an alarm on R-11 would not result in HVE-1A automatically stopping. As stated
in OMM-033 the automatic signal from SI and manual signals are available.
Recommend the applicant comment be rejected based on above.
NRC Resolution
The NRC agreed with the facility licensees recommendation, and rejected the applicants
contention.
The applicants contention is centered on unsupported assumptions that other procedures are in
progress that are not stated in the question stem. The insistence of an unsupported assumption
of unstated procedures in progress is contrary to the guidance contained in Appendix E to
The question stem clearly specifies that The equipment hatch is removed IAW OMM-033,
IMPLEMENTATION OF CV CLOSURE. Procedure OMM-033 clearly specifies that the
automatic closure signals from R-11 and R-12 are defeated for the given plant conditions due to
the procedural requirement to lift the leads for these signals. Furthermore, OMM-033 requires
R-14C to be in operation for the special test OST-163 mentioned in the applicants contention to
be performed. However, the question stem states that R-14C is out-of-service (OOS).
Therefore, with R-14C OOS, special test OST-163 cannot be in progress under the given
conditions, and the leads would be in a lifted status. It is technically correct to assert that
given the conditions present in the questions stem, if R-11 were to alarm, then HVE-1A will
NOT automatically stop, as specified in the question.
Therefore, the NRC determined that C remains the only technically correct answer to the
question. In accordance with NUREG-1021, the written examination answer key was not
changed based on this contention.
6. Item
Question 90, K/A 103A2.02
Comment
A Robinson licensed operator applicant recommended the following concerning Question 90:
The key states A is correct. Contrary to this part two of this question asks if a containment
purge may be started under the current permit. Under the plant conditions described in the
stem of this question a containment ventilation isolation signal from R-12 has isolated the purge.
In the current plant conditions, IAW OP-921 CONTAINMENT AIR HANDLING, the Containment
purge MAY NOT be restarted (step 6.4.1.24 e [Enclosure provided by licensed operator
applicant] under the current permit until steps 6.4.1.24 a., b., c., and d of OP-921 have been
completed (i.e. the containment isolation signal has been reset). These above conditions
support B as the correct answer.
Correct answer is B
Facility Comment
Which ONE of the following correctly completes the statements below?
The CV Purge ________ be restarted under the CURRENT permit.
The question is asking if the CV Purge may be restarted. In this case the fact that the verb
may was used is relevant. May means; have permission to. The question is asking of [sic] the
procedure gives the operator permission to restart the CV purge under the CURRENT permit.
OP-921 (Containment Air Handling) contains the following steps which permit the CV purge to
be restarted under the CURRENT permit:
10. IF AT ANY TIME the following occurs:
-R-11 (CV and Plant Vent Air Particular Monitor) alarms during Purge, AND
Setpoint are lower than setpoint on release permit,
-R-12 (CV and Plant Vent Radioactive Gas Monitor) alarms during Purge, AND
Setpoint are lower than setpoint on release permit,
THEN perform the following: ..________
a. Adjust RMS setpoints as required IAW values on release permit .. ________
b. RESET Containment Isolation Vent Iso ________
c. Check the following Safeguards Relay lights EXTINGUISHED: .. ________
-CV Ventilation Isolation Signal V-1 . ________
-CV Ventilation Isolation Signal V-2 . ________
d. Place CV Press Relieve V12-10 and V12-11 Control Switch to
OPEN position _________
e. Check, by position indicating lights, Containment Pressure Relief
valves are OPEN _________
-V12-10 (Containment Pressure Relief) OPEN _________
-V12-11 (Containment Pressure Relief) OPEN . _________
Recommend the applicant comment be rejected based on above.
NRC Resolution
The NRC agreed with the facility licensees recommendation, and rejected the applicants
contention.
The applicant is correct that additional steps in OP-921, CONTAINMENT AIR HANDLING, are
required to be performed in order to restart the containment purge under the conditions
specified in the question stem. However, the applicant is incorrect in the statement that the
additional steps that are required to be performed make distractor B a correct answer. The
second part question statement clearly states: The CV purge [may or may NOT] be restarted
under the CURRENT permit. It is technically correct that the CV purge may be restarted under
the CURRENT permit (once the additional steps in OP-021 are performed as required). It is
technically incorrect to state that the CV purge may NOT be restarted, which is the alternative
choice. Therefore, as stated above by the facility licensee, the applicants contention is
technically inaccurate. The question is asking whether or not the procedure allows the
operators to restart the CV purge under the current permit; OP-921 contains steps (listed above)
that allow the CV purge to be restarted under the current permit.
Therefore, the NRC determined that A remains the only technically correct answer to the
question. In accordance with NUREG-1021, the written examination answer key was not
changed based on this contention.
SIMULATOR FIDELITY REPORT
Facility Licensee:
Facility Docket No.: 05000261
Operating Test Administered: September 25 - October 5, 2017
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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