IR 05000400/2014302
| ML14364A375 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 12/30/2014 |
| From: | Gerald Mccoy Division of Reactor Safety II |
| To: | Waldrep B Duke Energy Progress |
| Shared Package | |
| ML15013A407 | List: |
| References | |
| 50-400/OL-14 | |
| Download: ML14364A375 (12) | |
Text
UNITED STATES ber 30, 2014
SUBJECT:
SHEARON HARRIS NUCLEAR POWER PLANT - NRC OPERATOR LICENSE EXAMINATION REPORT 05000400/2014302
Dear Mr. Waldrep:
During the period November 17 - 21, 2014 the Nuclear Regulatory Commission (NRC)
administered operating tests to employees of your company who had applied for licenses to operate the Shearon Harris Nuclear Plant. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating tests with those members of your staff identified in the enclosed report. The written examination was administered by your staff on November 25, 2014.
All applicants passed both the operating test and written examination. There were four 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 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) 997-4551.
Sincerely,
/RA/
Gerald J. McCoy, Chief Operations Branch 1 Division of Reactor Safety Docket No: 50-400 License No: NPF-63
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Docket No.: 50-400 License No.: NPF-63 Report No.: 05000400/2014302 Licensee: Duke Energy Progress, Inc.
Facility: Shearon Harris Nuclear Plant, Unit 1 Location: 5413 Shearon Harris Road New Hill, NC 27562 Dates: Operating Test - November 17-21, 2014 Written Examination - November 25, 2014 Examiners: David Lanyi, Chief Examiner, Senior Operations Engineer Newton Lacy, Operations Engineer Joseph Viera, Operations Engineer Approved by: Gerald McCoy Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY
ER 05000400/2014302; operating test November 17 - 21, 2014 & written exam November 21, 2014; Shearon Harris Nuclear Plant; 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 Shearon Harris 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 November 17 - 21, 2014.
Members of the Shearon Harris Nuclear Plant training staff administered the written examination on November 25, 2014. All Reactor Operator (RO) and Senior Reactor Operator (SRO)applicants passed both the operating test and written examination. All 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.
The NRC administered the operating tests during the period November 17 - 21, 2014.
The NRC examiners evaluated four Reactor Operator (RO) and six Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. Members of the Shearon Harris Nuclear Plant training staff administered the written examination on November 25, 2014. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Shearon Harris 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. Shearon Harris 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 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.
The NRC determined, using NUREG-1021, that the licensees initial examination submittal was within the range of acceptability expected for a proposed examination.
All applicants passed both the operating test and written examination and 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 four post-examination 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 November 25, 2016--two years after administration of the written exam, in the ADAMS system (ADAMS Accession Numbers ML14338A038 and ML14338A036).
4OA6 Meetings, Including Exit
Exit Meeting Summary
On November 21, 2014, the NRC examination team discussed generic issues associated with the operating test with Mr. Benjamin Waldrep, Site Vice-President, and members of the Shearon Harris 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 B. Waldrep, Site Vice-President J. Dufner, Plant General Manager D. Hayes, Operations Manager D. Griffith, Training Manager S. Schwindt, Operations Training Manager D. Corbett, Manager - Regulatory Affairs E. Betram, Operations Instructor R. Horton, Senior Nuclear Operations Instructor A. Lucky, Senior Nuclear Operations Instructor G. Pickar, Initial Training Supervisor S. Rua, NLO / Exam Supervisor S. Scott, Assistant Operations Manager - Training R. Vondenberg, Assistant Operations Manager - Shift M Wallace, Senior Technical Specialist - Regulatory Affairs NRC personnel J. Austin, Senior 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 ML14338A046.
Item
Question 5, K/A 015AA1.16
Comment
The licensee recommends that there is no correct answer for RO question #5.
The question asks whether or not a reactor trip would occur based on plant conditions described
by Bistable Status on the Bypass Permissive Light Box (BPLB) and the Trip Status Light Boxes
(TSLB). The first half of the question asked whether a reactor trip WOULD or WOULD NOT
occur. The second half of the question asked which Reactor Protection System (RPS)
Permissive caused the reactor to trip/not trip. The keyed answer is that a reactor trip would
occur because of the status of P-7 (Low Power Trips Blocked).
The stem of the question states that the BPLB and TSLB both indicate that both the P-7 and P-
(Power Range > 10%) bistables were lit. This is not possible unless there is a fault, because
one illuminated light indicates greater than 10% power and the other indicates less than 10%
power.
Clarification for this question was given during the exam in that the Bypass Permissive Light
Box window names for each RPS Permissive were written down. The clarification provided
implied that the P-7 BPLB light was being referenced in the question.
The stem of the question and the additional guidance provided in the clarification led candidates
to answer the question based on BPLB status. Since the conditions in the stem of the question
could not exist in any plant condition using BPLB indications, this is an invalid question without a
correct answer.
NRC Resolution
The licensees recommendation was accepted.
The question was written upon the mistaken supposition that if the P-7 permissive light were lit,
the interlock would be met. The stem also stated that the P-10 permissive light was lit.
Further review of the wiring diagrams reveal that P-7 would be extinguished when power is
greater than 10% and that P-10 would be lit when power was greater than 10%. The
configuration of having both lights lit at the same time is not possible without a circuitry failure.
The question was written to test the applicants knowledge of low-power reactor trip block status
lights during a loss of Reactor Coolant flow. Specifically, the intent was to examine their
knowledge that the reactor trip was due to loss of two out of three loops when power was
between the P-7 power and P-8 (Single Loop Low Flow Trip Blocked) power (approximately
49% power). The concept was that with the P-7 light on and the P-8 light off, the applicant was
to infer that power was between these two permissives. They would then be able to choose the
correct answer that a trip would occur because two of the three Reactor Coolant Pumps would
have lost power (P-7 permissive). However with the P-7 light illuminated, the stem of the
question led the applicants to conclude that power was actually less than 10% and the P-7
induced loss of flow trips were not valid. No trip would have occurred in this case. No answer
was provided that would have allowed no trip occurring due to the status of P-7.
Since the P-7 permissive appeared to not be met, there were no correct answers provided and
the question was deleted.
Item
Question 27, K/A WE15EA1.3
Comment
The licensee recommends that there is no correct answer for RO question #27.
The first part of the question requires the candidate to evaluate containment parameters and
choose the appropriate Function Restoration (FR) Procedure to implement. The current
construction of the Containment Critical Safety Function Status Tree (CSFST) as adopted by
the licensee evaluates containment pressure before containment sump level or radiation levels.
CSFST rules of usage as described in EOP USERS GUIDE section 5.2 states At any given
time, a Critical Safety Function status is represented by a single path through its tree. Since
each path is unique, it is uniquely labeled at its end point, or terminus. This labeling consists of
color coding and/or line-pattern-coding of the terminus and last branch line, plus a transition to
an appropriate FR if required by that safety status. Since containment pressure is evaluated
before evaluation of sump levels, the Containment Status Tree would result in a YELLOW
terminus requiring transition to FR-Z.1. This YELLOW path effectively blocks evaluation of the
ORANGE path terminus for Containment flooding. This results in an entry into EOP-FR-Z.1
Response to High Containment Pressure, eliminating answers C and D from being correct.
The second part of this question requires the candidate to know what needs to be sampled as
required by the implementing procedure. The stem of the question stated that bus 1A2-SA was
de-energized due to a fault. This fault results in a loss of one train of Containment Spray Pumps
and Emergency Service Water (ESW) Booster Pumps. Since one train of ESW booster pumps
remain in service (B train), service water for the de-energized train is simply isolated per step
9.a RNO, not sampled. Since nothing is sampled in EOP-FR-Z.1 with the current plant
conditions, A and B are also incorrect leaving no correct answer for this question.
NRC Resolution
The licensees recommendation was accepted.
The Westinghouse Emergency Response Guidelines (ERG) Background Document for F-0.5,
Section 2 states When the status tree rules of usage are applied to F-0.5, CONTAINMENT,
with a spray pump running and containment pressure between the spray actuation pressure
(T.02) and the design pressure (T.03), then a YELLOW priority will result. The operator should
be aware that this YELLOW priority can be reached without evaluating the ORANGE priority for
entry into FR-Z.2, RESPONSE TO CONTAINMENT FLOODING, based on high containment
sump level. This priority scheme should not present conflicts for plants with a large, dry
containment (like the reference plant) due to the containment pressure behavior following an
event that releases sufficient mass and energy into the containment atmosphere to actuate
containment spray, and the value of footnote (T.06) for entry into FR-Z.2, RESPONSE TO
CONTAINMENT FLOODING. The background document goes on to provide options for how
to change the Containment Status Tree if it were determined that containment flooding should
be evaluated before pressure was reduced less than the Containment Spray actuation setpoint.
The licensee has not adopted any of these changes and currently uses the ERG version of the
Containment Status Tree as is. Therefore, the question as written would recommend entry into
FR-Z.1. Based upon that procedure, no guidance would be given on sampling any water in
containment. Therefore, none of the answers provided were correct.
Since no correct answers were provided, the question has been deleted.
Item
Question 51, K/A 064A2.04
Comment
The licensee recommends that there is no correct answer based on the information provided in
the stem of the question.
The first part of the question requires the candidate to evaluate the time required to shut down
the Emergency Diesel Generator (EDG) from 35% load. The second part of the question asks
for the impact of taking the action per the first part of the question. There is no comment on the
second part of the question. The basis for the timely shutdown of the EDG is to minimize
carbon buildup which is both B and D answers.
The comment on the first part of the question is there was not enough information provided in
the stem of the question to provide an operationally accurate response. Section 7.1.2, step 9 of
OP-155, Diesel Generator Emergency Power System states At the MCB (Main Control
Board), WHEN stack exhaust temperatures are less than 500°F,THEN POSITION DIESEL
GENERATOR A-SA (B-SB) control switch to STOP. Since these temperatures were not
provided there was not enough information provided for the candidate to determine if the note
that states The EDG should be shutdown from 35% load in less than 5 minutes to minimize
carbon buildup was applicable.
Without the stack temperatures it is not possible to determine the time the EDG should be
shutdown.
NRC Resolution
The licensees recommendation was rejected.
The lack of stack exhaust temperature data was irrelevant for the question asked. Although
there have been occasions in the past where the EDGs had to be operated unloaded for periods
greater than five minutes. The note to shutdown the EDG within 5 minutes is always applicable
unless other conditions prevent performing it in that time frame. In this case, no information was
provided on the stack exhaust temperature, therefore the logical assumption would be that the
temperature did not hinder a timely shutdown.
The lack of information in the stem would not cause confusion about what was being asked. If
any confusion was experienced by the applicants by the stem of the question, they were
repeatedly informed that they should ask for clarification. The applicants asked no clarifying
questions about lack of stack exhaust temperature data. Therefore there was adequate
information in the stem to answer the question.
Item
Question 73, K/A G2.4.17
Comment
The licensee recommends that the correct answer to this question is A and not B as
keyed.
The question required the applicant to evaluate the condition of Reactor Coolant System (RCS)
RCS pressure during an operator controlled cooldown following a small break Loss of Coolant
Accident (LOCA) and the basis for that decision. The EOP USERS GUIDE section 6.5 states
The operator is frequently asked to check RCS and SG (Steam Generator) pressures and
temperature as STABLE (or RISING). STABLE does not necessarily imply constant. RCS
and/or SG pressure may be dropping slowly due to an operator-controlled cooldown and still be
considered stable. If the operator can control the rate and magnitude of the pressure change,
then pressure should be considered stable. Therefore RCS pressure should be considered
STABLE due to the pressure drop seen being a direct result of an operator controlled cooldown,
which makes answer A correct and answers C and D incorrect.
Answer B provides a basis for calling RCS pressure stable as RCS subcooling is rising. While
a rising subcooling is a diverse indication of a stable/rising pressure, it is not an indication that is
referenced in the EOP users guide revision that was used to write this exam question or by the
applicants as they prepared for the exam.
Based on the information provided in the EOP Users Guide answer A is the more correct
answer to this question.
NRC Resolution
The licensees recommendation was partially accepted
Based upon the assumption made by the applicants during the test that a controlled cooldown
was in progress, it is reasonable to understand that RCS pressure would be slowly lowering due
to the operators actions. However, rising subcooling with a slowly lowering RCS pressure
continues to be a legitimate indication of stable pressure.
Both answer A and B were accepted as correct answers.
SIMULATOR FIDELITY REPORT
Facility Licensee: Shearon Harris Nuclear Plant
Facility Docket No.: 50-400
Operating Test Administered: November 17 - 21, 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.
No simulator fidelity or configuration issues were identified.
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