IR 05000400/2011301
| ML112240145 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 08/11/2011 |
| From: | Widmann M Division of Reactor Safety II |
| To: | Jefferson W Carolina Power & Light Co |
| References | |
| 50-400/11-301 | |
| Download: ML112240145 (15) | |
Text
UNITED STATES ust 11, 2011
SUBJECT:
SHEARON HARRIS NUCLEAR POWER PLANT - NRC OPERATOR LICENSE EXAMINATION REPORT 05000400/2011301
Dear Mr. Jefferson:
During the period July 11-15, 2011, the Nuclear Regulatory Commission (NRC) administered operating tests to employees of your company who had applied for licenses to operate the Shearon Harris Nuclear Power 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 July 20, 2011.
All applicants passed both the operating test and written examination. There were three 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).
CP&L 2 If you have any questions concerning this letter, please contact me at (404) 997-4550
Sincerely,
/RA/
Malcolm T. Widmann, 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.: 05000400 License No.: NPF-63 Report No.: 05000400/2011301 Licensee: Carolina Power & Light Company (CP&L)
Facility: Shearon Harris Nuclear Power Plant Location: 5413 Shearon Harris Road New Hill, NC 27562 Dates: Operating Test - July 11 - 15, 2011 Written Examination - July 20, 2011 Examiners: G. Laska, Chief Examiner, Sr. Operations Examiner P. Presby, Senior Operations Engineer, RI K. Schaaf, Operations Engineer D. Bacon, Operations Engineer, (Certification)
Approved by: M. Widmann, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000400/2011301; July 11-15 & July 20, 2011; Shearon Harris Nuclear Power 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.
The NRC developed the written examination outline. Members of the Shearon Harris Nuclear Power Plant staff developed both the operating tests and the written examination.
The NRC administered the operating tests during the period July 11-15, 2011. Members of the Shearon Harris Nuclear Power Plant training staff administered the written examination on July 20, 2011. Seven Reactor Operator (RO) and three Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. Ten applicants were issued licenses commensurate with the level of examination administered.
There were three post-examination comments.
No findings were identified.
REPORT DETAILS
OTHER ACTIVITIES
4OA5 Operator Licensing Examinations
a. Inspection Scope
The NRC developed the written examination outline. Members of the Shearon Harris Nuclear Power Plant 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, "Operator Licensing Examination Standards for Power Reactors." 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 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 examiners evaluated seven Reactor Operator (RO) and three Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. The examiners administered the operating tests during the period July 11-15, 2011.
Members of the Shearon Harris Nuclear Power Plant training staff administered the written examination on July 20, 2011. 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 Power Plant, met the requirements specified in 10 CFR Part 55, Operators Licenses.
b. Findings
No findings were identified. The NRC determined, using NUREG-1021 that the licensees operating test and written examination submittals were both 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 three post-examination comments concerning the written examination. A copy of the final SRO and RO written examinations and answer key, with all changes incorporated, and the licensees post-examination comments may be accessed not earlier than July 22, 2013 in the ADAMS system (ADAMS Accession Number(s) ML112170233, ML112170234, and ML112170235.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On July 15, 2011 the NRC examination team discussed generic issues associated with the operating test with Mr. Ernest Kapopoulos, Shearon Harris Plant Manager, and members of the Shearon Harris Nuclear Power 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 R. Bright, Simulator Support D. Cortlett, Supervisor - Supervisor Licensing/Regulatory Programs T. Craig, Senior Operations Instructor J. Caves, Lead Engineer/Licensing W. Gunter, Manager - Shift Operations D. Griffith, Manager - Training A. Lucky, Senior Nuclear Operations Training Instructor E. Kapopoulos, Plant Manager Harris Plant M. McDade, Simulator Support R. Moore, Superintendent-Nuclear Operations Performance S. Schwindt, Supervisor Nuclear Operations Continuing Training S. Scott, Operations Initial License Training Supervisor M. Spellman, Control Room Supervisor A. Spencer, Senior Nuclear Operations Training Instructor A. Sylvester, Supervisor-Operations Initial Training M. Wallace, Senior Specialist - Licensing M. Weber, Superintendent - Operations Support J. Werner, Manager- Operations F. Womack, Manager - Operations NRC personnel P. Presby K. Schaaf D. Bacon
FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS
A complete Text of the licensees post examination comments can be found in ADAMS under
Accession Number ML112170235
Items:
(1) RO Question # 26
26. Given the following plant conditions:
- At 0530, RCS temperature was being maintained at 557°F with the plant in Mode 3
when a Small Break LOCA occurred
- At 0545, RCS temperature is 550°F
- The crew is ready to commence a cooldown to cold shutdown IAW EPP-009, Post
LOCA Cooldown and Depressurization
Which ONE of the following identifies (1) the lowest allowable temperature of the
RCS at 0630 if the crew begins the MAXIMUM permissible cooldown rate and (2) the
Basis for this temperature limit?
A. (1) 457°F
(2) to ensure that Technical Specification cooldown limits are NOT exceeded
B. (1) 457°F
(2) to ensure that a transition is NOT required to be made to FRP-P.1, Response
to Imminent Pressurized Thermal Shock
C. (1) 450°F
(2) to ensure that Technical Specification cooldown limits are NOT exceeded
D. (1) 450°F
(2) to ensure that a transition is NOT required to be made to FRP-P.1, Response
to Imminent Pressurized Thermal Shock
A was designated as the correct answer.
Facility Comment:
Facility contends that both answers A and B are correct.
In the HNP EOP users guide the basis for the100°F/hr cooldown rate limit is stated as follows:
EOPs that provide instructions for plant cooldown and depressurization to cold shutdown
conditions limit the RCS cooldown rate to 100°F/hr consistent with the Tech Spec limits for
Modes 1-3. This basis confirms answer A is correct. The stem of the question states that an
EPP-009 Post LOCA cooldown and Depressurization is in progress. The bases document for
EPP-009, (ES-1.2) states that the maximum cooldown rate of 100°F/hr will preclude violation of
the Integrity Status Tree thermal shock limits. FRP-P.1, RESPONSE TO IMMINENT
PRESSURIZED THERMAL SHOCK, would be the procedure entered if the Integrity Status Tree
limits were violated which would also make answer B Correct. Since the stem of the question
did not specify whether the basis for the limit of 457°F at 0630 or the basis for the limit in EPP-
was desired, either answer could be considered correct based on the reference used to
answer the question. EPP-09 basis document was not referenced in the development of this
question and this new technical information supports accepting both answers as correct.
Facility Recommendation:
Accept both answers A and B as correct.
NRC RESOLUTION:
The NRC does not totally agree with the licensees contention that both A and B are correct.
The basis of the cooldown step in the High Pressure version of the Westinghouse Owner Group
ES-1.2 Background document states, in part: The objective of a controlled cooldown is to
reduce the overall temperature of the RCS coolant and metal to reduce the need for supporting
plant systems and equipment required for heat removal. The maximum cooldown rate of
100°F/hr will preclude violation of the Integrity Status Tree thermal shock limits.
HNP EOP-USERS Guide, Revision 30, step 6.19 Tech Spec Cooldown and
Pressure/Temperature Limits states; EOPs that provide instructions for plant cooldown and
depressurization to cold shutdown conditions limit the RCS cooldown rate to 100°F/HR
consistent with the Tech Spec limits for Modes 1 - 3. This limit conflicts with the Tech Spec limit
in Mode 4 of 50°F/HR. Use of the less restrictive 100°F/HR limit in Mode 4 is acceptable for
implementation of the EOPs even though it does not comply with Tech Spec justification for use
of the 100°F/HR limit and non-compliance with Tech Specs is documented in Engineering
Evaluation PCR-5275 and the response to AR 00022701. The justification is based in part on
the fact that in Mode 4, Pressure-Temperature limits to protect vessel integrity have been
developed for the 100°F/HR cooldown rate as well as the 50°F/HR used as the Tech Spec limit.
The RCS pressure limit for any given RCS temperature are slightly lower for the 100°F/HR case
but well above those (actually near the PRZ SRV setpoint) expected to be present during an
operator controlled RCS cooldown. HNP has decided to use the more restrictive limit (during
normal plant cooldowns) in Mode 4 to provide more operational flexibility when LTOPS is placed
in service (References 2.2.3.4, PCR-5275, "Justification for 100°F/HR cooldown rate in EOPs"
and 2.2.3.24, AR 00022701/AR 00029783). This implies that the cooldown rate is not based
on technical specifications as stated in answer A, but is based on maintaining the cooldown rate
to prevent violation of the Integrity Status Tree thermal shock limits. This would make distractor
B the correct answer. Therefore, the NRC will grade this question with B as the only correct
answer.
(2) RO Question # 48
48.
Given the following:
- The unit is operating at 100% power
- A Fire in Aux Bus 1E1 occurs, resulting in the loss of Aux Bus E
Which ONE of the following describes the battery charger(s) that temporarily lose power and will
automatically be re-energized?
D was designated as the correct answer.
Facility Comment:
Facility contends that the question should be deleted from the exam. The question was unclear
as to what power supply was reenergized. An assumption could be made that it was the output
of the charger which is correct, or the input to the chargers from the motor control centers that
are reenergized by the Emergency Diesel Generators following a loss of off-site power. In
addition, the stem of the question is worded such that if the correct answer, a single charger, is
selected the question is not grammatically correct. The use of lose vice lose(s) in the stem
would require the correct answer to be plural to be grammatically correct. The lack of clarity
and the grammatical incorrectness of the stem lead to confusion among the applicants.
Facility Recommendation:
Delete question 48.
NRC RESOLUTION:
The NRC does not agree with the licensees contention. Appendix E (Policies and Guidelines
for Taking RCS Examinations) of NUREG-1021 Operator Licensing Examination Standards for
Power Reactors states in Part B paragraph 7: When answering a question, do not make
assumptions regarding conditions that are not specified in the question unless they occur as a
consequence of other conditions that are stated in the question. The question asked which
battery charger(s) would temporarily lose power and will automatically be reenergized?
(implying that it would be charging the batteries). The off service charger must be manually
aligned to allow it to charge the batteries. In accordance with OP-156.01 only one charger (1A-
SB or 1B-SB) may have the AC input and DC output breakers closed. Therefore only one
battery charger would be energized. The use of the word lose versus lose(s) should not have
had an impact on how the applicant answered the question. The NRC determined that the
question will remain as written with answer D as the only correct answer.
(3) RO Question # 49
Given the following plant conditions:
- A loss of AC power has occurred
- 125 VDC battery 1A-SA is currently loaded at rated load
DC load shedding has been performed to reduce the battery load to half of rated load.
How long will the battery be available to supply the remaining loads?
A. More than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
B. More than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> but less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
C. More than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> but less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
D. Up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
A was designated as the correct answer.
Facility Comment:
Facility contends that both answers A and B are correct because insufficient information is
provided in the stem of the question to determine when the DC load shed occurs. The
assumption was made that the stem of the question was asking the operator to determine the
effect on safety battery life as our station procedures are implemented. In accordance with site
procedures and the Station Blackout Coping Analysis, our 125VDC Class 1E batteries are rated
to last for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at rated load (Calc 8S44-P-101 STATION BLACKOUT COPING ANALYSIS
REPORT section 7.2.2.1) and the DC load shed occurs within 60 minutes following the loss of
AC power (EOP-EPP-001 LOSS OF AC POWER TO 1A-SA AND 1B-SB BUSES). If the DC
load shed to half rated capacity occurs at time +60 minutes (as assumed in Calculation E-4
SAFETY BATTERIES 1A-SA & 1B-SB LOAD PROFILE DETERMINATION (LOCA/SBO)
section 4.2.2.4), then 25% of battery capacity would be used prior to loads being shed, this
would extend remaining battery life to greater than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Battery life cannot be greater than
hours because of the 60 minutes at rated capacity already used. Therefore, answer B would
be correct.
Facility Recommendation:
Accept both answers A and B as correct.
NRC RESOLUTION:
The NRC does not agree with the licensees contention. In a review of documentation
submitted after the exam was administered the NRC determined based on when the load shed
was conducted (question was not specific as to when the load shed occurred), any of the
answers presented could be correct. ES-403 D.1.c states: If three or more answers could be
considered correct, or there is not a correct answer, the question shall be deleted. Therefore,
the NRC determined the question should be deleted from the examination.
SIMULATOR FIDELITY REPORT
Facility Licensee: Shearon Harris Nuclear Power Plant
Facility Docket No.: 05000400
Operating Test Administered: July 11-15, 2011
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
Steam Generator Pressure During three separate scenarios differences in simulator response
response during a 50% were observed between the crews, this difference caused multiple
feed line break. paths through the EOP network. The licensee review of feed
breaks since exam delivery has revealed that the 50% size break
does not provide liquid flow from the steam generator to the break
as would be expected after the feed system is isolated. NCR
00479941 and SSR 11-0270 were written to address the issue.
AFW Idle suction pressure A difference of approximately 2 psig does exist. SSR 11-0276
differs between the was written to address the issue
pumps.
OSI Pi display of turbine The simulator does not have an explicit DEH computer so OSI Pi
reference and demand did points of importance to the operator are transmitted to OSI Pi
not indicate the same as through an interface routine. In the interface routine the points
reference and demand on were mapped incorrectly. SSR 11-0266 was written to address
the MCB DEH display. the concern and was fixed prior to the next day scenarios.
AFW suction pressure AFW suction pressure when supplied by ESW indicated lower for
when supplied by ESW the static (non running) SDAFW pump than for the running
was lower for the static MDAFW pump. SSR 11-0272 was written to address this issue.
SDAFW than the operating
MDAFW pump.
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