IR 05000259/2013301
| ML13205A410 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/23/2013 |
| From: | Mark Franke Division of Reactor Safety II |
| To: | James Shea Tennessee Valley Authority |
| References | |
| 50-259/13-301, 50-260/13-301, 50-296/13-301 | |
| Download: ML13205A410 (15) | |
Text
UNITED STATES uly 23, 2013
SUBJECT:
BROWNS FERRY NUCLEAR PLANT - NRC OPERATOR LICENSE EXAMINATION REPORT NOS 05000259/2013301, 05000260/2013301 AND 05000296/2013301
Dear Mr. Shea:
During the period June 3 - 7, 2013, the Nuclear Regulatory Commission (NRC) administered operating tests to employees of your company who had applied for licenses to operate the Browns Ferry Nuclear Plant. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating tests and the written examination submittal with those members of your staff identified in the enclosed report. The written examination was administered by your staff on June 28, 2013.
Three Reactor Operator (RO) and three Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. Two RO applicants failed the operating test. There were three post-administration comments concerning the operating test. These comments, and the NRC resolution of these comments, are summarized in Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.
The initial written SRO examination submitted by your staff failed to meet the guidelines for quality contained in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1, as described in the enclosed report.
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) 997-4436
Sincerely,
/RA/
Mark E. Franke, Chief Operations Branch 2 Division of Reactor Safety Docket Nos: 50-259, 50-260, 50-296 License Nos: DPR-33, DPR-52, DPR-68
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report (
REGION II==
Docket No.: 50-259, 50-260, AND 50-296 License No.: DPR-33, DPR-52, and DPR-68 Report No.: 05000259/2013301, 05000260/2013301, and 05000296/2013301 Licensee: Tennessee Valley Authority (TVA), LLC Facility: Browns Ferry Nuclear Plant, Units 1, 2, and 3 Location: Athens, AL 35611 Dates: Operating Test - June 3 - 7, 2013 Written Examination - June 28, 2013 Examiners: Bruno Caballero, Chief, Senior Operations Engineer, RII/DRS/OLB2 Ken Schaaf, Operations Engineer, RII/DRS/OLB1 Andreas Goldau, Operations Engineer, RII/DRS/OLB2 Matt Emrich, Examiner-in-Training, TTC Approved by: Mark E. Franke, Chief Operations Branch 2 Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000259/2013301, 05000260/2013301, and 05000296/2013301; operating test
June 3 - 7, 2013, & written exam June 28, 2013; Browns Ferry 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 Browns Ferry Nuclear Plant staff developed both the operating tests and the written examination. The initial written SRO examination submittal did not meet the quality guidelines contained in NUREG-1021.
The NRC administered the operating tests during the period June 3 - 7, 2013. Members of the Browns Ferry Nuclear Plant training staff administered the written examination on June 28, 2013. Three Reactor Operator (RO) and three Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. Four applicants were issued licenses commensurate with the level of examination administered. Issuance for two RO applicants has been delayed pending receipt of additional information.
There were three post-examination comments.
No findings were identified.
REPORT DETAILS
OTHER ACTIVITIES
4OA5 Operator Licensing Examinations
a. Inspection Scope
Members of the Browns Ferry Nuclear 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 five 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 June 3 - 7, 2013.
Members of the Browns Ferry Nuclear Plant training staff administered the written examination on June 28, 2013. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Browns Ferry Nuclear Plant, met the requirements specified in 10 CFR Part 55, Operators Licenses.
b. Findings
No findings were identified.
The NRC determined that the licensees examination submittal was outside the range of acceptable quality specified by NUREG-1021. The initial written examination submittal was outside the range of acceptable quality because more than 20% (8 of 25 sampled)of the SRO questions sampled for review contained unacceptable flaws. Individual questions were evaluated as unsatisfactory for the following reasons:
- Five questions contained two or more implausible distractors.
- One question failed to meet the K/A statement contained in the examination outline.
The NRC regional office returned the entire written examination, containing 100 questions, to the licensee for rework and correction in accordance with NUREG-1021.
Administration of the written examination was delayed, in part, because the quality of the licensees examination submittal was unacceptable. Future examination submittals need to incorporate lessons learned.
Three RO applicants and three SRO applicants passed both the operating test and written examination. Two RO applicants passed the written examination but did not pass the operating test. One RO applicant and three SRO applicants were issued licenses. Issuance of the licenses for two RO applicants has been delayed pending receipt of additional information. Details concerning the need for additional information have been sent to the individual applicants and the facility licensee.
The following generic weaknesses were discussed at the exit meeting:
- The RO applicants performance during plant evolutions with the reactor at low power was weak. For example, administrative log taking in Mode 5, response to a feed pump trip during a startup scenario, and adjustment of the cool down rate using integrated computer screens during shutdown cooling operations.
- The RO and SRO applicants implementation of the requirement to stop rod withdrawal prior to reaching the rod block monitor (RBM) set point was weak.
That is, the applicants failed to stop rod withdrawals prior to receiving the RBM High/Inop alarm.
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 operating test. 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 July 9, 2015, in the ADAMS system (ADAMS Accession Number(s)
ML13191A869, ML13191A879, and ML13191A882.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On June 7, 2013 the NRC examination team discussed generic issues associated with the operating test with Lang Hughes, Operations Manager, and members of the Browns Ferry 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 Lang Hughes, Operations Senior Manager James Emens, Site Licensing Manager Steve Austin, Licensing Manager Russell Joplin, Corporate Training Director Daniel Laing, Site Training Director Hal Higgins, Nuclear Operations Training Supervisor Doug Hakenewerth, Operations Shift Manager NRC personnel Dave Dumbacker, NRC 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 ML13191A882.
Item #1: Walk-Through - Job Performance Measure (JPM) Administrative Topic b, SR-2
Operator Logs in Mode 5
Comment
The licensee recommended that Steps 5 and 8 of this JPM were NOT critical steps.
- The licensees basis for why JPM Step 5 was not a critical step was that the applicability
listed in SR-2, Instrument Checks and Observations, Table 4.5, Mode Switch Position,
was:
o Mode 5 with the Reactor Mode Switch in the REFUEL position and any control rod
withdrawn OR
o Mode 4 when in Special Operation LCO 3.10.4
Because the actual plant condition presented to the applicants (on the simulator) was
that the Mode Switch was locked in the REFUEL (Mode 5) position, with all rods fully
inserted, the licensee contended that the applicant could record either SAT or NOT
APPLICABLE for JPM Step 5.
- The licensees basis for why JPM Step 8 was not a critical step was that the actual plant
condition presented to the applicants (on the simulator) was the vessel head removed
and the cavity flooded to greater than 22 feet above the RPV flange. The licensee
contended that the potential for thermal stratification could not, and did not, exist;
therefore, performing the RPV differential temperature calculation in JPM Step 8 was not
critical.
NRC Resolution
The licensees recommendation was accepted.
For this administrative JPM, the applicant was expected to perform operator logs in accordance
with SR-2, Instrument Checks and Observations, for Tables 4.1 through 4.7 while the unit was
in Mode 5, Refueling, and use the table notes to determine whether acceptance criteria was
satisfied. The following items were required to be logged and identified by the applicant:
- Table 4.1, IRM Instrumentation
- Table 4.2, SRM Instrumentation (identify A SRM inoperable; critical step)
- Table 4.3, Level Instrumentation
- Table 4.4.a, Control Rod Position
o write All Rods In for Column A (critical step)
o write not applicable for Column B (critical step because local observation of
hydraulic control unit (HCU) pressure indicator was not required when all rods
were inserted)
- JPM Step 5: Table 4.5, Mode Switch Position
- Table 4.6, Reactor Coolant Conductivity (record between 4 - 6 µmhos; critical step)
- Table 4.7, Part 1, RHR Shutdown Cooling (SDC) (identify flow requirements not met;
critical step)
- JPM Step 8: Table 4.7, Part 2, Vessel Differential Temperature (Record the bottom and
top RPV temperatures, then subtract to obtain the overall RPV temperature difference)
NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Rev.9,
Supplement 1, Appendix C, JPM Guidelines, Section B.3 requires that 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. The
task standard was to perform operator logs in accordance with SR-2, Instrument Checks and
Observations, for log tables 4.1 through 4.7 and to verify acceptance criteria were satisfied in
accordance with notes.
For JPM Step 5, because no control rods were withdrawn, Table 4.5, Mode Switch Position was
not required to be performed. Therefore, completion of JPM Step 5 was not required to
accomplish the task standard because, with all rods fully inserted, Table 4.5 was not applicable.
For JPM Step 8, the actual plant condition presented to the applicants (on the simulator) was
the vessel head as removed and the cavity flooded to greater than 22 feet above the RPV
flange. The actual temperature difference across the RPV (bottom to top) was 10.9 °F. Based
on Note 6, a temperature differential 50°F was indicative of inadequate mixing and
stratification of the water in the RPV; however, this value was impossible to achieve since the
vessel head was removed and cavity flooded. Because the plant condition presented to the
applicants (on the simulator) was not affiliated with a situation where thermal stratification could
ever occur, performance of JPM Step 8 was determined to be not critical.
Item #2: Walk-Through - Job Performance Measure (JPM) Administrative Topic a, Work Hour
Limitations - SRO Version
Comment
The licensee recommended that a typographical error existed in the standard for JPM Step 1.
The basis for the licensees recommendation was that the operator first exceeded the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
in a 7 day period work limitation on April 20 at 15:00. The licensee contended that standard for
this JPM step incorrectly listed that the operator first exceeded this work hour limitation on April
at 11:00.
NRC Resolution
The licensees recommendation was accepted.
For this administrative JPM, the applicant was expected to analyze two operators work
schedules and identify the date and time that one reactor operator exceeded 72 work hours in a
day period (critical step). Additionally, the applicant was expected to identify the date and
time that the same operator also failed to meet the requirement for 3 days off in a 15 day period
(critical step).
After identifying the date and times of the reactor operators non-compliance with the Fatigue
Rule, the applicant was expected to:
- Notify the Nuclear Fatigue Rule (NFR) Administrator, Operations Manager, and Site NFR
Subject Matter Expert (critical step).
- Generate a problem evaluation report (PER) (critical step)
- Determine that Tech Spec 5.2.2, Unit Staff, required another operator to replace the
operator within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, because control room staffing was below minimum (critical step).
The examiners verified, based on the work schedules presented to the applicants, the operator
first exceeded the 72 work hour in a 7 day period work limitation on April 20th at 15:00 and the
same operator also failed to meet the requirement for 3 days off in a 15 day period on April 20th
at 07:00. Therefore, the licensees recommendation that the standard for JPM Step 1 contained
a typographical error was accepted.
Item #3: Walk-Through - Job Performance Measure (JPM) Systems - Control Room Topic e,
Verify Traversing Incore Probe (TIP) Isolation
Comment
The licensee recommended that Steps 6 and 12 of this JPM were NOT critical steps.
For JPM Step 6, the licensee contended that placing the Manual TIP Drive Control Switch to the
OFF position, after the TIP had been manually retracted, was not a critical step because the in-
shield limit switch turned off the detector drive motor. Because the detector drive motor was
stopped by the in-shield limit switch, the licensee contended that JPM Step 6 was not a critical
step.
For JPM Step 12, the licensee contended that placing the TIP C & E Manual Valve Control
Switches to the CLOSED position was not critical because the ball valve had already
automatically closed for TIP C and because the shear valve was activated for TIP E.
NRC Resolution
The licensees recommendation was accepted.
For this JPM, the applicant was expected to recognize that TIP detectors A, B, D, and E failed to
automatically retract (TIP C did auto-retract) and then manually retract and isolate TIPs in
accordance with 2-AOI-64-2E, Traversing Incore Probe Isolation. The applicant was also
expected to identify that TIP E failed to manually retract and then activate its associated
explosive shear valve. The following expected actions were designated as critical steps in the
JPM:
- Place Mode Switch to the MANUAL position for TIP drives A, B, D, and E
- Place the Manual Switch to the REV position for TIP drives A, B, D, and E (identifying
TIP E fails to retract)
- Place Man Valve Control Switch to the CLOSED position for TIP drives A, B, and D
- Obtain key PA-235
- Insert key into the key lock switch for the TIP E shear valve and turn the key to the FIRE
position
NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Rev.9,
Supplement 1, Appendix C, JPM Guidelines, Section B.3 requires that 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. The
task standard was 1) TIPs A, B, and D are manually driven inward and their associated ball
isolation valves closed after the TIP was moved to the In-Shield position and 2) the TIP E shear
valve was activated.
For JPM Step 6, an in-shield position limit switch de-energized the detector drive motor.
Therefore, placing the Manual Switch to the OFF position was not required to complete the task.
JPM Step 6 was not a critical step.
For JPM Step 12, placing the MAN VALVE CONTROL switch to the CLOSED position for TIP C
was not critical because TIP C had already automatically retracted and its ball isolation valve
was already closed, based on the initial plant conditions (on the simulator) presented to the
applicants. Placing the MAN VALVE CONTROL switch to the CLOSED position for TIP E was
not critical because TIP E was manually isolated via the explosive shear valve, which effectively
isolates the TIP penetration. TIPs A, B, and D MAN VALVE CONTROL switches had already
been placed to the CLOSED position in a previous procedure step. Therefore, JPM Step 12 was
not a critical step.
SIMULATOR FIDELITY REPORT
Facility Licensee: Browns Ferry Nuclear Plant
Facility Docket No.: 50-259, 50-260, AND 50-296
Operating Test Administered: June 3 - 7, 2013
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.
During the onsite preparatory visit during the period of May 6 - 10, 2013, the examiners
observed the following:
Item Description
Problem Report # 5348 U2 simulator FW flow oscillations at low
power during scenario validation
3