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| | number = ML16273A055 | | | number = ML16273A055 |
| | issue date = 09/28/2016 | | | issue date = 09/28/2016 |
| | title = Arkansas Nuclear One, Unit 1 - NRC Examination Report 05000313/2016301 | | | title = NRC Examination Report 05000313/2016301 |
| | author name = Gaddy V G | | | author name = Gaddy V |
| | author affiliation = NRC/RGN-IV/DRS/OB | | | author affiliation = NRC/RGN-IV/DRS/OB |
| | addressee name = Anderson R | | | addressee name = Anderson R |
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| =Text= | | =Text= |
| {{#Wiki_filter: | | {{#Wiki_filter:ember 28, 2016 |
| [[Issue date::September 28, 2016]]
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| Mr. Rich Anderson, Site Vice President
| | ==SUBJECT:== |
| | ARKANSAS NUCLEAR ONE, UNIT 1 - NRC EXAMINATION REPORT 05000313/2016301 |
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| Arkansas Nuclear One Entergy Operations, Inc. | | ==Dear Mr. Anderson:== |
| | On September 1, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an initial operator license examination at Arkansas Nuclear One, Unit 1. The enclosed report documents the examination results and licensing decisions. The preliminary examination results were discussed on August 26, 2016, with Mr. D. Perkins, Senior Operations Manager, and other members of your staff. A telephonic exit meeting was conducted on September 16, 2016, with Mr. Perkins, who was provided the NRC licensing decisions. |
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| 1448 SR 333 Russellville, AR 72802-0967
| | The examination included the evaluation of nine applicants for reactor operator licenses, four applicants for instant senior reactor operator licenses, and four applicants for upgrade senior reactor operator licenses. The license examiners determined that twelve of the seventeen applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued. There were three post-examination comments submitted by your staff. |
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| SUBJECT: ARKANSAS NUCLEAR ONE, UNIT 1 - NRC EXAMINATION REPORT 05000313/2016301
| | Enclosure 1 contains details of this report and Enclosure 2 summarizes post-examination comment resolution. |
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| ==Dear Mr. Anderson:==
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| On September 1, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an initial operator license examination at Arkansas Nuclear One, Unit 1. The enclosed report documents the examination results and licensing decisions. The preliminary examination results were discussed on August 26, 2016, with Mr. D. Perkins, Senior Operations Manager , and other members of your staff. A telephonic exit meeting was conducted on September 16, 2016, with Mr. Perkins, who was provided the NRC licensing decisions.
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| The examination included the evaluation of nine applicants for reactor operator licenses, four applicants for instant senior reactor operator licenses, and four applicants for upgrade senior reactor operator licenses. The license exam iners determined that twelve of the seventeen applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued. There were three post-examinatio n comments submitted by your staff. Enclosure 1 contains details of this report and Enclosure 2 summarizes post-examination comment resolution.
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| No findings were identified during this examination. | | No findings were identified during this examination. |
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| In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice and Procedure," a copy of | | In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice and Procedure," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). |
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| this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). | |
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| Sincerely,/RA/
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| Vincent G. Gaddy, Chief Operations Branch Division of Reactor Safety
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| Docket No. 50-313 License No. DPR-51
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| ===Enclosures:===
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| 1. Examination Report 05000313/2016301
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| ===w/Attachment:===
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| Supplemental Information 2. NRC Post-Examination Comment Resolution 3. Simulator Fidelity Report
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| cc w/encl: Electronic Distribution system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
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| Sincerely,/RA/ | | Sincerely, |
| Vincent G. Gaddy, Chief Operations Branch Division of Reactor Safety | | /RA/ |
| | | Vincent G. Gaddy, Chief Operations Branch Division of Reactor Safety Docket No. 50-313 License No. DPR-51 |
| Docket No. 50-313 License No. DPR-51 | |
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| ===Enclosures:=== | | ===Enclosures:=== |
| 1. Examination Report 05000313/2016301 | | 1. Examination Report 05000313/2016301 w/Attachment: Supplemental Information 2. NRC Post-Examination Comment Resolution 3. Simulator Fidelity Report |
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| ===w/Attachment:===
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| Supplemental Information 2. NRC Post-Examination Comment Resolution 3. Simulator Fidelity Report | |
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| cc w/encl: Electronic Distribution
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| Distribution
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| : See next page ADAMS ACCESSION NUMB ER: ML16273A055 SUNSI Review By: TJFarina ADAMS Yes No Publicly Available Non-Publicly Available Non-Sensitive Sensitive Keyword: NRC-002 OFFICE SOE:OB SOE:OB OE:OB OE:OB OE:OB OE:OB C:OB NAME TFarina COsterholtz MHayes SHedger MKennard CSteely VGaddy SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ DATE 9/19/16 9/26/16 9/20/16 9/21/16 9/20/16 9/20/16 9/26/16OFFICE C:PBE C:OB NAME NOKeefe VGaddy SIGNATURE /RA/ /RA/ DATE 9/28/16 9/28/16 OFFICIAL RECORD COPY
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| Letter w/enclosures to Rich Anderson from Vincent G. Gaddy, dated September 28, 2016
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| SUBJECT: ARKANSAS NUCLEAR ONE, UNIT 1 - NRC EXAMINATION REPORT 05000313/2016301
| | REGION IV== |
| | Docket: 05000313 License: DPR-51 Report: 05000313/2016301 Licensee: Entergy Operations, Inc. |
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| Electronic distribution by RIV: Regional Administrator (Kriss.Kennedy@nrc.gov)
| | Facility: Arkansas Nuclear One, Unit 1 1448 SR 333 Location: |
| Deputy Regional Administrator (Scott.Morris@nrc.gov) DRP Director (Troy.Pruett@nrc.gov) DRS Director (Anton.Vegel@nrc.gov)
| | Russellville, AR 72802-0967 Dates: August 22 through September 16, 2016 Inspectors: T. Farina, Chief Examiner, Senior Operations Engineer C. Osterholtz, Senior Operations Engineer M. Hayes, Operations Engineer S. Hedger, Operations Engineer M. Kennard, Operations Engineer C. Steely, Operations Engineer Approved By: Vincent G. Gaddy Chief, Operations Branch Division of Reactor Safety Enclosure 1 |
| DRS Deputy Director (Jeff.Clark@nrc.gov)
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| DRP Deputy Director (Ryan.Lantz@nrc.gov)
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| Senior Resident Inspector (Brian.Tindell@nrc.gov) Resident Inspector (Margaret.Tobin@nrc.gov) Resident Inspector (Andy.Barrett@nrc.gov)
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| Branch Chief, DRP/E (Neil.OKeefe@nrc.gov)
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| Senior Project Engineer, DRP/E (John.Dixon@nrc.gov)
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| Project Engineer, DRP/E (Brian.Correll@nrc.gov) Project Engineer, DRP/E (Jackson.Choate@nrc.gov) Administrative Assistant (Mary.Bennett@nrc.gov)
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| Public Affairs Officer (Victor.Dricks@nrc.gov)
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| Project Manager (Stephen.Koenick@nrc.gov)
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| Team Leader, DRS/IPAT (Thomas.Hipschman@nrc.gov)
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| Project Engineer, DRS/IPAT (Eduardo.Uribe@nrc.gov)
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| ACES (R4Enforcement.Resource@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov)
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| Regional Counsel (Karla.Fuller@nrc.gov)
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| Technical Support Assistant (Loretta.Williams@nrc.gov)
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| Congressional Affairs Officer (Jenny.Weil@nrc.gov) RIV Congressional Affairs Officer (Angel.Moreno@nrc.gov) RIV/ETA: OEDO (Jeremy.Bowen@nrc.gov)
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| RIV RSLO (Bill.Maier@nrc.gov)
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| Enclosure 1 U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 05000313 License: DPR-51 Report: 05000313/2016301 Licensee: Entergy Operations, Inc. Facility: Arkansas Nuclear One, Unit 1 Location:
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| 1448 SR 333 Russellville, AR 72802-0967 Dates: August 22 through September 16, 2016 Inspectors: T. Farina, Chief Examiner, Senior Operations Engineer C. Osterholtz, Senior Operations Engineer M. Hayes, Operations Engineer
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| S. Hedger, Operations Engineer M. Kennard, Operations Engineer C. Steely, Operations Engineer | |
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| Approved By: Vincent G. Gaddy Chief, Operations Branch Division of Reactor Safety | |
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| =SUMMARY= | | =SUMMARY= |
| ER 05000313/2016301; 08/22/2016 - 09/16/2016; Arkansas Nuclear One, Unit 1; Initial | | ER 05000313/2016301; 08/22/2016 - 09/16/2016; Arkansas Nuclear One, Unit 1; Initial |
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| Operator Licensing Examination Report. | | Operator Licensing Examination Report. |
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| The examiners determined that twelve of the seventeen applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued. | | The examiners determined that twelve of the seventeen applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued. |
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| ===A. NRC-Identified and Self-Revealing Findings=== | | ===NRC-Identified and Self-Revealing Findings=== |
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| None. | | None. |
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| ===B. Licensee-Identified Violations=== | | ===Licensee-Identified Violations=== |
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| None. | | None. |
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| ====a. Scope==== | | ====a. Scope==== |
| NRC examiners reviewed all license applications submitted to ensure each applicant satisfied relevant license eligibility requirements. Examiners also audited three of the license applications in detail to confirm that they accurately reflected the subject applicant's qualifications. This audit focused on the applicant's experience and on-the-job training, including control manipulations that provided significant reactivity changes. | | NRC examiners reviewed all license applications submitted to ensure each applicant satisfied relevant license eligibility requirements. Examiners also audited three of the license applications in detail to confirm that they accurately reflected the subject applicants qualifications. This audit focused on the applicants experience and on-the-job training, including control manipulations that provided significant reactivity changes. |
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| ====b. Findings==== | | ====b. Findings==== |
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| No findings were identified. | | No findings were identified. |
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| Twelve of 17 applicants passed the written examination and all parts of the operating test. Four applicants failed the written examination only, and one applicant failed both the written examination and administrative j ob performance measures (JPMs). The final written examinations and post-examination analysis and comments may be accessed in the ADAMS system under the accession numbers noted in the attachment. | | Twelve of 17 applicants passed the written examination and all parts of the operating test. Four applicants failed the written examination only, and one applicant failed both the written examination and administrative job performance measures (JPMs). The final written examinations and post-examination analysis and comments may be accessed in the ADAMS system under the accession numbers noted in the attachment. |
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| The examination team noted six generic weaknesses associated with applicant performance on the administrative JPM, simulator JPM, in-plant JPM, and dynamic scenario sections of the operating tests. Specifically:
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| | The examination team noted six generic weaknesses associated with applicant performance on the administrative JPM, simulator JPM, in-plant JPM, and dynamic scenario sections of the operating tests. Specifically: |
| 1. Three applicants demonstrated a weakness in lowering the pressurizer quench tank down to a specific level by transferring water to the clean waste receiving tank. The applicants all secured the associated transfer pump at the desired level; however, they delayed isolating the two quench tank drain valves, causing quench tank level to continue lowering to the low level alarm setpoint due to a difference in elevations between the two tanks and a 4.7 psig Nitrogen over-pressure in the quench tank. | | 1. Three applicants demonstrated a weakness in lowering the pressurizer quench tank down to a specific level by transferring water to the clean waste receiving tank. The applicants all secured the associated transfer pump at the desired level; however, they delayed isolating the two quench tank drain valves, causing quench tank level to continue lowering to the low level alarm setpoint due to a difference in elevations between the two tanks and a 4.7 psig Nitrogen over-pressure in the quench tank. |
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| 2. Seven applicants demonstrated a weakness implementing the station's work hour fitness-for-duty guidance. Given a work history of five operators, many applicants incorrectly stated that one of the operators failed to satisfy the requirement to have a minimum 34-hour break in a 9-day period and was, therefore, ineligible to stand an emergent watch. In actuality, the applicant had one full day off in between two day-shift watches, which equated to at least a | | 2. Seven applicants demonstrated a weakness implementing the stations work hour fitness-for-duty guidance. Given a work history of five operators, many applicants incorrectly stated that one of the operators failed to satisfy the requirement to have a minimum 34-hour break in a 9-day period and was, therefore, ineligible to stand an emergent watch. In actuality, the applicant had one full day off in between two day-shift watches, which equated to at least a 35-hour break, satisfying the 34-hour requirement. |
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| 35-hour break, satisfying the 34-hour requirement. | |
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| 3. When purging the main generator with CO2 following a station blackout, several operators failed to use the designated rack of 15 CO2 bottles, which is labeled "Reserved for Unit 1 EOP [Emergency Operating Procedure] Requirements," and instead used an identical rack of 15 CO2 bottles which is intended for normal daily use, not EOPs. Procedure 1106.002, Section 15.0, Generator Hydrogen System, fails to specify which bank should be used, constituting a missed opportunity to prevent operator error. | | 3. When purging the main generator with CO2 following a station blackout, several operators failed to use the designated rack of 15 CO2 bottles, which is labeled Reserved for Unit 1 EOP [Emergency Operating Procedure] |
| | Requirements, and instead used an identical rack of 15 CO2 bottles which is intended for normal daily use, not EOPs. Procedure 1106.002, Section 15.0, Generator Hydrogen System, fails to specify which bank should be used, constituting a missed opportunity to prevent operator error. |
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| 4. Three out of four crews unnecessarily tripped the reactor on a continuous rod withdrawal malfunction, which could have been mitigated by proper implementation of Alarm Response Procedure ACA 1203.003, Section 9, Control Rod Drive Malfunction Actions. Two of the crews failed to properly implement the "MAJI" sequence of operator actions, which directs the crew to verify the diamond panel in MANUAL, set the GROUP-AUXIL switch to AUXIL, set SPEED SELECTOR switch to JOG, and place manual command switch in INSERT for three seconds. These two crews failed to stop rod motion, and manually tripped the reactor. One crew properly stopped all rod motion, but misinterpreted plant indications to mean that rod motion was continuing and tripped the reactor as well. | | 4. Three out of four crews unnecessarily tripped the reactor on a continuous rod withdrawal malfunction, which could have been mitigated by proper implementation of Alarm Response Procedure ACA 1203.003, Section 9, Control Rod Drive Malfunction Actions. Two of the crews failed to properly implement the MAJI sequence of operator actions, which directs the crew to verify the diamond panel in MANUAL, set the GROUP-AUXIL switch to AUXIL, set SPEED SELECTOR switch to JOG, and place manual command switch in INSERT for three seconds. These two crews failed to stop rod motion, and manually tripped the reactor. One crew properly stopped all rod motion, but misinterpreted plant indications to mean that rod motion was continuing and tripped the reactor as well. |
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| 5. Several crews demonstrated a weak understanding of generator hydrogen temperature control valve CV-4018 controller operation. This controller is reverse acting, meaning that it generates a 0 percent signal to open the valve and a 100 percent signal to fully close the valve. Some believed that a 100 percent signal indicated open, and several applicants misunderstood this controller to indicate actual position, when it only indicates demanded position. | | 5. Several crews demonstrated a weak understanding of generator hydrogen temperature control valve CV-4018 controller operation. This controller is reverse acting, meaning that it generates a 0 percent signal to open the valve and a 100 percent signal to fully close the valve. Some believed that a 100 percent signal indicated open, and several applicants misunderstood this controller to indicate actual position, when it only indicates demanded position. |
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| Actual valve position can only be verified locally. One crew tripped the turbine instead of taking manual control of the controller, which would have mitigated the malfunction. | | Actual valve position can only be verified locally. One crew tripped the turbine instead of taking manual control of the controller, which would have mitigated the malfunction. |
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| 6. Three crews demonstrated weaknesses implementing the steam generator tube rupture procedure. Errors included tripping the reactor out of sequence, initiating | | 6. Three crews demonstrated weaknesses implementing the steam generator tube rupture procedure. Errors included tripping the reactor out of sequence, initiating emergency feedwater instead of auxiliary feedwater, unnecessary delays establishing a plant cooldown, and a failure to control steam header pressure to prevent unnecessary cycling of the bad steam generator atmospheric dump valve. |
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| emergency feedwater instead of auxilia ry feedwater, unnecessary delays establishing a plant cooldown, and a failure to control steam header pressure to prevent unnecessary cycling of the bad steam generator atmospheric dump valve. | |
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| Copies of all individual examination reports were sent to the facility training manager for evaluation and determination of appropriate remedial training. | | Copies of all individual examination reports were sent to the facility training manager for evaluation and determination of appropriate remedial training. |
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| ====a. Scope==== | | ====a. Scope==== |
| The NRC examiners observed simulator performanc e with regard to plant fidelity during examination validation and administration. | | The NRC examiners observed simulator performance with regard to plant fidelity during examination validation and administration. |
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| ====b. Findings==== | | ====b. Findings==== |
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| During nine scenarios, this log was flooded due to noisy analog inputs on control boards C03/C04, causing it to exceed its capacity. The trainee action monitor log can accommodate a maximum of 1,000 entries in a scenario, but the noisy analog inputs caused multiple log entries per second which filled the log to capacity prior to scenario completion. This program had no impact on simulator performance or fidelity, but it resulted in an incomplete record of applicant actions for the scenarios which were affected. The licensee initiated simulator deficiency report DR16-0120 to address the issue. | | During nine scenarios, this log was flooded due to noisy analog inputs on control boards C03/C04, causing it to exceed its capacity. The trainee action monitor log can accommodate a maximum of 1,000 entries in a scenario, but the noisy analog inputs caused multiple log entries per second which filled the log to capacity prior to scenario completion. This program had no impact on simulator performance or fidelity, but it resulted in an incomplete record of applicant actions for the scenarios which were affected. The licensee initiated simulator deficiency report DR16-0120 to address the issue. |
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| ===2. During one administration of JPM S-6, "Transfer Buses from the Unit Aux Transformer to a Startup Transformer," an applicant was required to open breaker H-14 on 6900V bus H1 to correct a condition where the bus was aligned to two power sources simultaneously. The applicant correctly identified that === | | ===2. During one administration of JPM S-6, Transfer Buses from the Unit Aux=== |
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| breaker H-14 failed to open automatically after the H-15 Synchronize switch was taken to off, and attempted to manually open breaker H-14 twice. Both attempts appeared to be a full turn of the handswitch, but the breaker remained closed on the control board. Therefore the applicant correctly implemented step 8.4.6.A.1 of Procedure OP-1107.001, and manually reopened breaker H-15 from the startup transformer #1 to ensure that bus H1 remained energized from only one source. | | Transformer to a Startup Transformer, an applicant was required to open breaker H-14 on 6900V bus H1 to correct a condition where the bus was aligned to two power sources simultaneously. The applicant correctly identified that breaker H-14 failed to open automatically after the H-15 Synchronize switch was taken to off, and attempted to manually open breaker H-14 twice. Both attempts appeared to be a full turn of the handswitch, but the breaker remained closed on the control board. Therefore the applicant correctly implemented step 8.4.6.A.1 of Procedure OP-1107.001, and manually reopened breaker H-15 from the startup transformer #1 to ensure that bus H1 remained energized from only one source. |
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| Post-scenario, the examiner reviewed available simulator data with the licensee to determine what actions were captured by the machine. The trainee action monitor log records all manipulations which are performed on the simulator. The trainee action monitor log clearly showed that component C10 152-14/CS (breaker H-14 control switch) was taken to Trip multiple times and returned to normal. The simulator did not pick up this action, however, and continued to display the breaker as closed. The licensee believes that the issue may be related to a difference in scan operating speeds between the simulator operating system (10 Hz) and the simulator subroutine that controls large breaker logic (2 Hz). The licensee initiated simulator deficiency reports DR16-0122 and DR16-0123 to address the issue. | | Post-scenario, the examiner reviewed available simulator data with the licensee to determine what actions were captured by the machine. The trainee action monitor log records all manipulations which are performed on the simulator. The trainee action monitor log clearly showed that component C10 152-14/CS (breaker H-14 control switch) was taken to Trip multiple times and returned to normal. The simulator did not pick up this action, however, and continued to display the breaker as closed. The licensee believes that the issue may be related to a difference in scan operating speeds between the simulator operating system (10 Hz) and the simulator subroutine that controls large breaker logic (2 Hz). The licensee initiated simulator deficiency reports DR16-0122 and DR16-0123 to address the issue. |
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| ====b. Findings==== | | ====b. Findings==== |
| No findings were identified. | | No findings were identified. |
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| {{a|4OA6}} | |
| ==4OA6 Meetings, Including Exit== | | ==4OA6 Meetings, Including Exit== |
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| The licensee did not identify any information or materials used during the examination as proprietary. | | The licensee did not identify any information or materials used during the examination as proprietary. |
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| ATTACHMENT: | | ATTACHMENT: |
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| =SUPPLEMENTAL INFORMATION= | | =SUPPLEMENTAL INFORMATION= |
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| ===Licensee Personnel=== | | ===Licensee Personnel=== |
| : [[contact::R. Martin]], Operations Training Superintendent | | : [[contact::R. Martin]], Operations Training Superintendent |
| : [[contact::B. Possage]], Examination Writer | | : [[contact::B. Possage]], Examination Writer |
| : [[contact::J. Cork]], Examination Writer | | : [[contact::J. Cork]], Examination Writer |
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| ===NRC Personnel=== | | ===NRC Personnel=== |
| : [[contact::B. Tindell]], Senior Resident Inspector | | : [[contact::B. Tindell]], Senior Resident Inspector |
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| ==ADAMS DOCUMENTS REFERENCED== | | ==ADAMS DOCUMENTS REFERENCED== |
| Accession No. ML16259A363 - FINAL WRITTEN EXAMS Accession No. ML16259A379 - FINAL OPERATING TEST
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| : Accession No. ML16259A348 -
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| : POST-EXAMINATION ANALYSIS-COMMENTS
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| : Note:
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| : The final written examination and operating test are withheld from public release until September 1, 2018.
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| : NRC Resolution to the Arkansas Nuclear One Unit 1 Post-Examination Comments
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| : A complete text of the licensee's post-examination analysis and comments can be found in
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| : ADAMS under Accession Number ML16259A348.
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| : RO QUESTION # 1
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| - ESAS actuated on low RCS pressure. - RCS Tave 560 °F and stable
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| - Pressurizer level 320" and stable
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| - RCS pressure 1350 psig and rising rapidly
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| - RB sump level 55% and rising - Fuel failure of 1 % is indicated Considering the above conditions, which of the following methods, and reason behind the method, will be used to mitigate the RCS pressure transient in accordance with RT-
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| : 14?
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| : A. Cycle ERV as required, this prevents challenges to the PZR safeties.
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| : B. Raise PZR spray flow, this condenses steam in PZR vapor space.
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| : C. Throttle HPI flow, this reduces input of mass into RCS to match RCS leakage.
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| : D. Raise letdown flow, this lowers RCS mass and thus reduces pressure.
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| : Answer:
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| : A.
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| : LICENSEE COMMENT:
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| : 7/17 total candidates incorrect (41.1%).
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| : All seven chose "C". The correct answer is "A".
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| : This question involves a steam space leak and asks which of the choices given, and reason for that choice, will be used to control RCS pressure in accordance with
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| : RT-14.
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| : The correct answer "A" was to cycle the ERV to prevent challenges to the safeties. Candidate feedback during the exam debrief on Friday, September 2, revealed that the reason the seven candidates chose "C" was based on the wording of the question.
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| : The question asks, "- which of the
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| following methods...
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| will be used ..." (italics and bold emphasis added for this report).
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| : The candidates missing this question reasoned this wording implied what will be used to control
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| : RCS pressure considering a given stable RCS temperature of 560 °F and a rapidly rising RCS
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| pressure of 1350 psig.
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| : The candidates reasoned that subcooling margin (SCM) would thus be restored quickly since the RCS pressure and temperature given was just below the SCM line on Figure 1 of 1202.013 in their handout (please refer to the attached figure).
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| : We maintain the correct answer "A" is still correct considering the parameters as a snapshot in time. We are therefore requesting a change to the key to allow both "A" and "C" to be correct. This is in accordance with
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| : NUREG-1021,
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| : ES-403, D.1.c.
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| : We will, of course, revise this question to ensure it is suitable for future use. We will also be performing training needs analysis on this
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| question.
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| - 2 -
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| : NRC RESOLUTION:
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| : The NRC has determined to delete this question from the exam.
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| : This question has missing information in the stem that is needed to answer the question.
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| : The conditions given in the stem are representative of a steam space leak (TMI event).
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| : EOP Repetitive Task 14 (RT-14), "Control RCS Pressure", is used to combat these conditions.
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| : The stem indicates reactor coolant system pressure rapidly increasing as the pressurizer goes solid, with some fuel failure, implying a bubble in the vessel - also consistent with TMI events.
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| : Break flow is assumed to be the method of cooling the core under the given conditions.
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| : Step 3.B of procedure
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| : RT-14 directs that High Pressure Injection (HPI) may be throttled once subcooling margin (SCM) is restored (this is the licensee's requested additional correct answer,
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| "C").
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| : The EOP basis document for this step, GEOG Bases Section V.B, RCS Pressure Control, Step 3.3 specifies that use of this guidance also requires adherence to GEOG Bases Section VI
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| : Rule 2.0, "HPI Throttling/Termination Rule".
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| : This Rule states that "HPI flow may not be throttled unless SCM exists," and further notes that "HPI may not be throttled, even with SCM, if HPI cooling is in progress until core exit thermocouple temperatures are decreasing, except to prevent violating the RV P-T limit."
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| : Therefore, it is necessary to know both the status of SCM
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| and CET temperature trend in order to make the determination to throttle HPI flow.
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| : However, only Tave is given in the stem; without CET temperature or its trend given, the applicant cannot make this determination.
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| : Tave would not accurately track CET temperature with a bubble in the vessel, RCPs secured, and fuel damage.
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| : Further, per the licensee, the determination of SCM
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| using
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| : EOP 1202.013 Figure 1, "Saturation and Adequate SCM", requires the use of CET
| |
| temperature on the RCS Pressure - Temperature plot, not Tave.
| |
| : For these reasons, inadequate information is given in the stem to determine any correct answer, and the question is therefore deleted from the exam.
| |
| : RO QUESTION # 19
| |
| : A dropped rod event has occurred (one CRA in Group 7) and the following conditions
| |
|
| |
| exist: - Reactor power = 30% and decreasing. - Turbine output = 320 MWe and decreasing.
| |
| - Annunciator (K07-C3) HIGH LOAD LIMIT is in fast flash.
| |
| - Turbine runback is in progress.
| |
| : What operator action is procedurally required?
| |
| : A. Allow the runback to terminate normally.
| |
| : B. Take manual control of the turbine and raise load.
| |
| : C. Take manual control of SG/RX master.
| |
| : D. Trip the reactor.
| |
| : Answer: C.
| |
| : LICENSEE COMMENT:
| |
| : 6/17 total candidates incorrect (35.2%).
| |
| : One chose "A" and five chose "D".
| |
| : The correct answer is "C".
| |
| : This question involves a dropped rod event which causes a plant runback to 40% of 902 MWe (the runback final power was not given).
| |
| : The parameters given show reactor power and turbine
| |
| - 3 - load at 30%, the student was asked which action was procedurally required.
| |
| : The correct answer "C" is to take manual control of the SG/RX master, an ICS station which will stop the runback when taken to manual.
| |
| : Candidate feedback during the exam de-brief on Friday, September 2, revealed that the reason the five candidates chose "D" (trip the reactor) was due to recent OE at Grand Gulf where the operating crew continued to operate the plant during a severe transient instead of tripping the unit.
| |
| : This OE was used by ANO's acting Site Vice President during recent meetings with plant staff where he reinforced the conservative position to trip the unit when plant control is not present.
| |
| : This action is procedurally supported by the entry conditions of 1202.001, Reactor Trip, where it is stated that a manual trip is required due to "... a system degradation that requires a manual reactor trip based on operator judgement." Please refer to the attached page of 1202.001.
| |
| : We are therefore requesting a change to the key to allow both "C" and "D" to be correct.
| |
| : This is in accordance with
| |
| : NUREG-1021,
| |
| : ES-403
| |
| : D.1.c.
| |
| : Although
| |
| : ES-403 D.1.c gives an example where the question would be deleted we ascertain question 19 is different from the example since answer "C" does not overtly state "a manual trip is not required."
| |
| : We will, of course, revise this question to ensure it is suitable for future use.
| |
| : We will also be performing training needs analysis on this question.
| |
| : NRC RESOLUTION:
| |
| : The licensee requests to accept two correct answers, one where the reactor is tripped (Distracter 'D') and one where it is not tripped by taking manual control of the condition and stabilizing the plant (correct answer 'C').
| |
| : These are opposing choices, and if both were true then the question would be required to be deleted based on
| |
| : NUREG-1021 guidance for directly opposing selections (Section D.1.c).
| |
| : This is not necessary for question 19 however, because the stem clearly states that the annunciator K07-C3 "HIGH LOAD LIMIT" is alarming.
| |
| : In the case of a dropped rod with a failure of turbine runback to terminate at 40% reactor power, explicit procedural direction exists in annunciator response procedure 1203.012F, K07-C3, "HI LOAD LIMIT IN EFFECT," to mitigate the malfunction without a reactor trip.
| |
| : Specifically, if the high load limit is caused by an ICS failure or ICS input signal failure, the operator is required to take manual control of the affected ICS station (in this case, SG/RX Master):
| |
| : 1. Verify Integrated Control System (ICS) is in track and running back to the maximum load limit setpoint.
| |
| : 2. If high load limit is clearly caused by an ICS failure - then take manual control of the affected ICS station and return the plant to steady state condition.
| |
| : The licensee states that per the entry conditions of 1202.001, Reactor Trip, a manual reactor trip is required for "a system degradation that requires manual reactor trip based on operator judgment."
| |
| : Recognizing that unplanned reactor trips have a direct impact on core damage frequency (CDF,
| |
| : SECY-99-007), the NRC maintains the "Unplanned Scrams per 7000 critical hours" performance indicator to monitor plant performance in this area.
| |
| : A subjective operator decision to trip the reactor without first attempting to take manual control of the SG/RX Master
| |
| : ICS station would therefore not be a conservative decision, would unnecessarily introduce additional transient risk, and would be contrary to the specific required procedural guidance of
| |
| : 203.012F, K07-3 for a condition well within the capability of the operator to recognize and correct.
| |
| : Therefore, the NRC has determined to retain question 19 with no changes.
| |
|
| |
| - 4 -
| |
| : RO QUESTION # 59
| |
| : Given: - An overheating event has been in progress.
| |
| - 1202.005, Inadequate Core Cooling, is in use.
| |
| - Core Exit Thermocouples= 1460 degrees F (average) and rising.
| |
| - RCS pressure = 2350 psig
| |
| - All actions have been performed for the current Region.
| |
| : Critical parameters have been updated by the ATC:
| |
| - Core Exit Thermocouples= 1520 degrees F (average) and rising.
| |
| - RCS pressure = 2400 psig
| |
| : CBOT reports multiple CETs are alarming on the plant computer with the status "INVALID" or "FAIL_LO".
| |
| : What is occurring and what procedural action is required for the above conditions?
| |
| : A. CETs are experiencing thermionic emission, trip all running RCPs. B. CETS are failing due to short circuits, trip all running RCPs.
| |
| : C. CETs are experiencing thermionic emission, use ADVs to reduce SG T-sat to ~100°F
| |
|
| |
| below current value.
| |
| : D. CETS are failing due to short circuits, use ADVs to reduce SG T-sat to ~100°F below
| |
|
| |
| current value.
| |
| : Answer: B.
| |
| : LICENSEE COMMENT:
| |
| : 17/17 total candidates incorrect (100.0%).
| |
| : Twelve chose "A", and five chose "C". The correct answer is "B".
| |
| : This question is about an Inadequate Core Cooling (ICC) scenario testing the applicants' knowledge of Core Exit Thermocouples (CETs) during a core damaging event as well as actions to take when CET indications exceed specific thresholds on Figure 4 of 1202.013. The correct answer "B" states, "CETs are failing due to short circuits, trip all running RCPs."
| |
| : Choice "A" contains the correct action (trip all RCPs due to entering Region IV in an ICC event) and the reason "CETs are experiencing thermionic emission."
| |
| : Thermionic emission was chosen as a distracter since the
| |
| : ANO-1 system training manual on incore Self Powered Neutron Detectors (SPNDs) stated thermionic emission affected SPNDs but did not state this affected CETs (both are contained within the same insulating material).
| |
| : Following exam administration, a discussion with an instructor who is also an Electrical Engineer revealed that thermionic emission will have an effect on thermocouples similar to that of a short.
| |
| : Thermocouples operate on the Seebeck effect.
| |
| : Two dissimilar metals (in this case chromel and alumel), one with a greater affinity for electrons result in a small voltage being generated where the wires are joined (junction).
| |
| : As temperature increases the average energy of the electrons increase resulting in a greater migration to the wire with the greater affinity.
| |
| : This results in a larger voltage at the junction.
| |
| : As the temperature further increases, the electron energies approach the work function energy of the conductor.
| |
| : Thermionic emission
| |
| - 5 - occurs when the electrons have sufficient energy to breach the work function and leave the conduction surface. This free movement of electrons neutralizes the potential previously generated at the thermocouple junction resulting in no or little voltage.
| |
| : This lowering of voltage at the hot junction will be indicated as a rapid lowering of the measured temperatures.
| |
| : We maintain the correct answer "B" is still correct since melting of the CETs will still cause a short and a short will cause a CET to fail low.
| |
| : We are therefore requesting a change to the key to allow both "A" and "B" to be correct. This is in accordance with
| |
| : NUREG-1021,
| |
| : ES-403, D.1.c.
| |
| : We will, of course, revise this question to ensure it is suitable for future use.
| |
| : We will also be performing training needs analysis on this question.
| |
|
| |
| [Follow-on Comment]:
| |
| : Another aspect of the question #59 - is the level of knowledge on the thermocouple failure and to what level of detail the operators should know from memory.
| |
| : When the operators were asked about this during the post exam debrief, they recalled the impact of thermionic emission within the incore instrument string which contains the CET thermocouple.
| |
| : Based on that information and the discussion with [the exam writer] - I believe introduced a level of difficulty into #59 that is possibly beyond what is necessary for the licensed operator to recall from memory. The proposal to accept the two answers (A and B) that contain the correct response to part B of the
| |
| : KA is what we are presenting as the more essential piece of the question and that the question does test the level of knowledge for the operator in the control room. [Part B is the ability to use procedures to correct, control or mitigate the consequences of core damage on the in-core temperature monitoring system]
| |
| : NRC RESOLUTION: The licensee's proposed basis for accepting "A" as an additional correct answer relies on the opinion of an ANO training department instructor, but does not include any formal technical background documentation to support this basis.
| |
| : The licensee could not provide any design documentation or formal technical analysis to demonstrate that the
| |
| : ANO-1 CETs would experience thermionic emission under accident conditions, and ANO training materials do not discuss this as a failure mode of the CETs.
| |
| : The NRC independently reviewed ANO training materials, as well as NUREG/CR-3386, "DETECTION OF INADEQUATE CORE
| |
| : COOLING WITH CORE EXIT THERMOCOUPLES: LOFT PWR EXPERIENCE," published in November of 1983.
| |
| : This study doesn't mention thermionic emission as a failure mechanism of
| |
| : CETs during accident conditions.
| |
| : Lacking supporting evidence, there is no basis to consider thermionic emission a known major contributor to the failure mechanism of the CETs, and therefore insufficient justification for accepting "A" as an additional correct answer on technical
| |
|
| |
| grounds.
| |
| : The licensee secondarily proposed that only part 2 of the question (procedural action required for entering Region 4 of Inadequate Core Cooling: the cause of CET damage) is expected operator knowledge, and proposed accepting both "A" and "B" as correct answers based on the second part of the question alone.
| |
| : This justification, if accepted, would more appropriately result in a deletion of the question from the examination because it would constitute only a
| |
| : 2-choice question.
| |
| : However, both parts of the question as-written test to the K/A directly, specifically the effect of core damage conditions on in-core temperature monitoring systems, differentiating between the failure modes of two different
| |
| : ANO-1 in-core instruments: CETs vs. SPNDs.
| |
| : In addition to the selected K/A, "Failure modes of thermocouples" is required operator
| |
| - 6 - knowledge listed in the Component section of the K/A catalog,
| |
| : NUREG-1122, Revision 2, page 5-3, K/A K1.14.
| |
| : This K/A has an RO importance rating of 2.8, implying its acceptability for examination inclusion when combined with ANO-specific operational material, which the second part satisfies.
| |
| : While every applicant incorrectly chose the SPND failure mode instead of the CET failure mode, this is an indication of a highly plausible distractor rather than a non-applicable knowledge check.
| |
| : For the above reasons, the NRC has determined to retain question 59 with no changes.
| |
| : Simulator Fidelity Report Facility Licensee: Arkansas Nuclear One, Unit 1
| |
| : Facility Docket No.: 50-313
| |
| : Operating Tests Administered on: August 22 to 27, 2016
| |
| : 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
| |
| : IP 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 tests, examiners observed the following
| |
|
| |
| items:
| |
| : Item
| |
| : Description Trainee Action
| |
| : Monitor Log Over-Capacity due to Noisy Inputs. The simulator control software includes an administrative trainee action monitor (TAM) program which logs simulator switch manipulations during the course of a scenario.
| |
| : During nine scenarios, this log was flooded due to noisy analog inputs on control boards C03/C04, causing it to exceed its capacity.
| |
| : The TAM log can accommodate a maximum of 1000 entries in a scenario, but the noisy analog inputs caused multiple log entries per second which filled the log to capacity prior to scenario completion.
| |
| : This program had no impact on simulator performance or fidelity, but it resulted in an incomplete record of applicant actions for the scenarios which were affected.
| |
| : The licensee initiated simulator deficiency report DR16-0120 to address the issue.
| |
| : Simulator Failed
| |
|
| |
| to Recognize
| |
| : Breaker Manipulation. During one administration of JPM S-6, "Transfer Buses from the Unit Aux Transformer to a Startup Transformer," an applicant was required to open breaker H-14 on 6900V bus H1 to correct a condition where the bus was aligned to two power sources simultaneously.
| |
| : The applicant correctly identified that breaker H-14 failed to open automatically after the H-15
| |
| : Synchronize switch was taken to off, and attempted to manually open breaker H-14 twice.
| |
| : Both attempts appeared to be a full turn of the handswitch, but the breaker remained closed on the control board.
| |
| : Therefore the applicant correctly implemented step 8.4.6.A.1 of procedure
| |
| : OP-1107.001, and manually reopened breaker H-15 from the Startup Transformer #1 to ensure that bus H1 remained energized from only one
| |
|
| |
|
| source.
| |
| : Post-scenario, the examiner reviewed available simulator data with the licensee to determine what actions were captured by the machine.
| |
| : The Trainee Action Monitor (TAM) log records all manipulations which are performed on the simulator.
| |
| : The TAM log clearly showed that component
| |
| : C10 152-14/CS (breaker H-14 control switch) was taken to Trip multiple times, and returned to normal.
| |
| : The simulator did not pick up this action
| |
| - 2 - however, and continued to display the breaker as closed.
| |
| : The licensee believes that the issue may be related to a difference in scan operating speeds between the simulator operating system (10 Hz) and the simulator subroutine that controls large breaker logic (2 Hz).
| |
| : The licensee initiated simulator deficiency reports DR16-0122 and DR16-0123 to address the issue.
| |
| }} | | }} |
|
---|
Category:Inspection Report
MONTHYEARIR 05000313/20234202024-01-10010 January 2024 Security Baseline Inspection Report 05000313/2023420 and 05000368/2023420 IR 05000313/20234022024-01-0202 January 2024 Security Baseline Inspection Report 05000313/2023402 and 05000368/2023402 IR 05000313/20234052023-12-12012 December 2023 Security Baseline Inspection Report 05000313/2023405 and 05000368/2023405 IR 05000313/20230032023-11-21021 November 2023 Revised - ANO Revised Integrated Inspection Report 05000313/2023003 and 05000368/2023003 and Independent Spent Fuel Storage Installation Inspection Report 07200013/ 2023001 ML23313A0962023-11-13013 November 2023 Integrated Inspection Report 05000313/2023003 and 05000368/2023003 and Independent Spent Fuel Storage Installation Inspection Report 07200013/2023001 IR 05000313/20230112023-10-10010 October 2023 Commercial Grade Dedication Inspection Report 05000313/2023011 and 05000368/2023011 IR 05000313/20230102023-09-20020 September 2023 Biennial Problem Identification and Resolution Inspection Report 05000313/2023010 and 05000368/2023010 IR 05000313/20230052023-08-21021 August 2023 Updated Inspection Plan for Arkansas Nuclear One, Units 1 and 2 (Report 05000313/2023005 and 05000368/2023005) - Mid Cycle IR 05000313/20243012023-08-14014 August 2023 Notification of NRC Initial Operator Licensing Examination 05000313/2024301 IR 05000313/20230022023-08-11011 August 2023 Integrated Inspection Report 05000313/2023002 and 05000368/2023002 and Notice of Violation IR 05000416/20230022023-08-0808 August 2023 Integrated Inspection Report 05000416/2023002 and Independent Spent Fuel Storage Installation Inspection Report 07200050/2023001 IR 05000313/20234032023-06-28028 June 2023 Security Baseline Inspection Report 05000313/2023403 and 05000368/2023403 IR 05000313/20234012023-06-13013 June 2023 Cyber Security Inspection Report 05000313/2023401 and 05000368/2023401 IR 05000313/20230012023-05-0909 May 2023 Integrated Inspection Report 05000313/2023001 and 05000368/2023001 IR 05000368/20233012023-05-0101 May 2023 NRC Examination Report 05000368/2023301 IR 05000313/20220062023-03-0101 March 2023 Annual Assessment Letter for Arkansas Nuclear One Units 1 and 2 (Report 05000313/2022006 and 05000368/2022006) IR 05000313/20220042023-01-26026 January 2023 Integrated Inspection Report 05000313/2022004 and 05000368/2022004 - 4th Quarter IR 05000313/20220032022-11-0909 November 2022 Integrated Inspection Report 05000313/2022003 and 05000368/2022003 and Exercise of Enforcement Discretion IR 05000313/20220112022-11-0909 November 2022 Notice of Violation NRC Inspection Report 05000313/2022011 and 05000368/2022011 ML22301A1532022-11-0909 November 2022 Notice of Violation NRC Inspection Report 05000313/2022011 and 05000368/2022011 ML22308A1912022-11-0606 November 2022 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000313/2023401 and 05000368/2023401 IR 05000313/20224012022-10-20020 October 2022 Security Baseline Inspection Report 05000313/2022401 and 05000368/2022401 IR 05000313/20220502022-09-26026 September 2022 Special Inspection Reactive Report 05000313 2022050 and 05000368 2022050 Cover Letter Only ML22208A1282022-08-0404 August 2022 (2nd Quarter) Inspection Report 2022 002 IR 05000313/20224032022-06-23023 June 2022 Security Baseline Inspection Report 05000313/2022403 and 05000368/2022403 IR 05000313/20223012022-06-15015 June 2022 NRC Examination Report 05000313/2022301 IR 05000313/20210132022-05-17017 May 2022 License Renewal Phase 4 Inspection Report 05000313/2021013 IR 05000313/20220012022-05-10010 May 2022 Integrated Inspection Report 05000313/2022001 and 05000368/2022001 IR 05000313/20220102022-03-14014 March 2022 Triennial Fire Protection Inspection Report 05000313/2022010 and 05000368/2022010 IR 05000313/20210062022-03-0202 March 2022 Annual Assessment Letter for Arkansas Nuclear One, Units 1 & 2 (Report 05000313/2021006 and 05000368/2021006) ML22047A2702022-02-23023 February 2022 Security Baseline Inspection Report 05000313/2022403 and 05000368/2022403 IR 05000313/20210042022-02-0404 February 2022 Integrated Inspection Report 05000313/2021004 and 05000368/2021004 ML22011A0992022-01-14014 January 2022 IR 2021012 EOI CO3 IR 05000313/20210102022-01-13013 January 2022 Biennial Problem Identification and Resolution Inspection Report 05000313/2021010 and 05000368/2021010 IR 05000313/20210032021-11-0505 November 2021 Integrated Inspection Report 05000313/2021003 and 05000368/2021003 IR 05000313/20210112021-09-23023 September 2021 Design Basis Assurance Inspection (Teams) Inspection Report 05000313/2021011 and 05000368/2021011 IR 05000313/20214012021-08-10010 August 2021 Security Baseline Inspection Report 05000313/2021401 and 05000368/2021401 IR 05000313/20210022021-08-0909 August 2021 Integrated Inspection Report 05000313/2021002 and 05000368/2021002 IR 05000368/20213012021-05-17017 May 2021 NRC Examination Report 05000368/2021301 IR 05000313/20210012021-05-0606 May 2021 Integrated Inspection Report 05000313/2021001 and 05000368/2021001 ML21095A3162021-04-13013 April 2021 ANO 2021402-Non SGI Rcl IR 05000313/20200062021-03-0303 March 2021 Annual Assessment Letter for Arkansas Nuclear One, Units 1 and 2 (Report 05000313/2020006 and 05000368/2020006) IR 05000313/20200042021-02-0303 February 2021 Integrated Inspection Report 05000313/2020004 and 05000368/2020004 IR 05000313/20200032020-11-0909 November 2020 Integrated Inspection Report 05000313/2020003 and 05000368/2020003 and Independent Spent Fuel Storage Installation Aging Management Inspection Report 07200013/2020001 IR 05000313/20204112020-10-22022 October 2020 Security Baseline Inspection Report 05000313/2020411 and 05000368/2020411 IR 05000313/20204042020-09-23023 September 2020 Notice of Violation; NRC Inspection Report 05000313/2020404 and 05000368/2020404 and NRC Investigation Report 4-2019-009 IR 05000313/20204102020-09-14014 September 2020 Security Baseline Inspection Report 05000313/2020410 and 05000368/2020410 IR 05000313/20200052020-09-0101 September 2020 Updated Inspection Plan for Arkansas Nuclear One, Units 1 and 2 (Reports 05000313/2020005 and 05000368/2020005) IR 05000313/20200022020-08-0505 August 2020 Integrated Inspection Report 05000313/2020002 and 05000368/2020002 IR 05000313/20204032020-07-30030 July 2020 Material Control and Accounting Inspection Report 05000313/2020403 and 05000368/2020403 2024-01-02
[Table view] Category:Letter
MONTHYEARML23326A0392024-01-24024 January 2024 Issuance of Amendment No. 281 Revision to Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML24017A1582024-01-17017 January 2024 Submittal of Emergency Plan Revision 50 IR 05000313/20234202024-01-10010 January 2024 Security Baseline Inspection Report 05000313/2023420 and 05000368/2023420 IR 05000313/20234022024-01-0202 January 2024 Security Baseline Inspection Report 05000313/2023402 and 05000368/2023402 ML23354A0022023-12-21021 December 2023 Request for Withholding Information from Public Disclosure ML23349A1672023-12-21021 December 2023 Request for Withholding Information from Public Disclosure ML23348A3572023-12-14014 December 2023 Application to Revise Technical Specifications to Use Online Monitoring Methodology Slides and Affidavit for Pre-Submittal Meeting ML23352A0292023-12-13013 December 2023 Entergy - 2024 Nuclear Energy Liability Evidence of Financial Protection ML23340A1592023-12-13013 December 2023 Entergy Operations, Inc. - Entergy Fleet Project Manager Assignment IR 05000313/20234052023-12-12012 December 2023 Security Baseline Inspection Report 05000313/2023405 and 05000368/2023405 ML23341A0832023-12-11011 December 2023 Material Control and Accounting Program Inspection Report 05000313/368/2023404- Cover Letter ML23305A0922023-12-0707 December 2023 Summary of Regulatory Audit Regarding License Amendment Request to Revise Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times RITSTF Initiative 4b ML23333A1362023-11-29029 November 2023 Supplement to Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23275A2072023-11-28028 November 2023 Issuance of Amendment No. 280 Removal of Technical Specification Condition Allowing Two Reactor Coolant Pump Operation ML23325A1412023-11-21021 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000313/20230032023-11-21021 November 2023 Revised - ANO Revised Integrated Inspection Report 05000313/2023003 and 05000368/2023003 and Independent Spent Fuel Storage Installation Inspection Report 07200013/ 2023001 ML23243B0452023-11-13013 November 2023 Request for Withholding Information from Public Disclosure ML23313A0962023-11-13013 November 2023 Integrated Inspection Report 05000313/2023003 and 05000368/2023003 and Independent Spent Fuel Storage Installation Inspection Report 07200013/2023001 ML23311A2082023-11-0909 November 2023 Reassignment of U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch IV IR 05000313/20230112023-10-10010 October 2023 Commercial Grade Dedication Inspection Report 05000313/2023011 and 05000368/2023011 IR 05000313/20230052023-08-21021 August 2023 Updated Inspection Plan for Arkansas Nuclear One, Units 1 and 2 (Report 05000313/2023005 and 05000368/2023005) - Mid Cycle IR 05000313/20243012023-08-14014 August 2023 Notification of NRC Initial Operator Licensing Examination 05000313/2024301 IR 05000313/20230022023-08-11011 August 2023 Integrated Inspection Report 05000313/2023002 and 05000368/2023002 and Notice of Violation ML23209A6022023-08-0909 August 2023 Regulatory Audit Plan in Support of License Amendment Request to Revise Technical Specifications to Adopt Risk- Informed Completion Times IR 05000416/20230022023-08-0808 August 2023 Integrated Inspection Report 05000416/2023002 and Independent Spent Fuel Storage Installation Inspection Report 07200050/2023001 ML23208A2132023-08-0303 August 2023 Regulatory Audit Summary Concerning Review of License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation ML23208A2112023-07-27027 July 2023 Entergy Operations Inc., Application to Revise Technical Specifications to Adopt TSTF-205, Revision of Channel Calibration, Channel Functional Test, and Related Definitions ML23188A1732023-07-12012 July 2023 Request for Withholding Information from Public Disclosure ML23142A2022023-06-29029 June 2023 Issuance of Amendment Nos. 279 and 332 Emergency Plan Staffing Requirements IR 05000313/20234032023-06-28028 June 2023 Security Baseline Inspection Report 05000313/2023403 and 05000368/2023403 ML23178A0892023-06-26026 June 2023 August 2023 Emergency Preparedness Program Inspection-Request for Information ML23166B0902023-06-21021 June 2023 Request for Relief ANO1 ISI 036 Regarding Volumetric Examination Requirements ML23180A1082023-06-20020 June 2023 ANO Unit 1 SAR Amendment 31, TRM, TS Bases, 10 CFR 50.59 Report, and Commitment Change Summary Report IR 05000313/20234012023-06-13013 June 2023 Cyber Security Inspection Report 05000313/2023401 and 05000368/2023401 ML23130A2972023-05-17017 May 2023 Notification of Inspection (NRC Inspection Report 05000313/2023003) and Request for Information ML23121A3012023-05-10010 May 2023 Regulatory Audit Plan in Support of License Amendment Request to Revise Technical Specifications to Adopt Risk - Informed Completion Times ML23114A3572023-05-10010 May 2023 Regulatory Audit Summary Concerning Review of Request for Alternative ANO-CISI-002 Regarding Containment Inservice Inspection Frequency ML23130A2732023-05-10010 May 2023 and Waterford 3 Steam Electric Station - Response to Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations IR 05000313/20230012023-05-0909 May 2023 Integrated Inspection Report 05000313/2023001 and 05000368/2023001 ML23118A3202023-05-0808 May 2023 Notification of Biennial Problem Identification and Resolution Inspection and Request for Information (05000313;05000368/2023010) IR 05000368/20233012023-05-0101 May 2023 NRC Examination Report 05000368/2023301 ML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment ML23108A1362023-04-24024 April 2023 Approval of Request for Half-Nozzle Repair of Reactor Vessel Closure Head Penetration No. 46 ML23110A1122023-04-20020 April 2023 Annual Report for Entergy Quality Assurance Program Manual Changes Under 10 CFR 50.54(a)(3), 10 CFR 71.106, and 10 CFR 72.140(d). Notification of Application of Approved Appendix B to 10 CFR 72 Subpart G 2024-01-24
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Text
ember 28, 2016
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT 1 - NRC EXAMINATION REPORT 05000313/2016301
Dear Mr. Anderson:
On September 1, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an initial operator license examination at Arkansas Nuclear One, Unit 1. The enclosed report documents the examination results and licensing decisions. The preliminary examination results were discussed on August 26, 2016, with Mr. D. Perkins, Senior Operations Manager, and other members of your staff. A telephonic exit meeting was conducted on September 16, 2016, with Mr. Perkins, who was provided the NRC licensing decisions.
The examination included the evaluation of nine applicants for reactor operator licenses, four applicants for instant senior reactor operator licenses, and four applicants for upgrade senior reactor operator licenses. The license examiners determined that twelve of the seventeen applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued. There were three post-examination comments submitted by your staff.
Enclosure 1 contains details of this report and Enclosure 2 summarizes post-examination comment resolution.
No findings were identified during this examination.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice and Procedure," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Vincent G. Gaddy, Chief Operations Branch Division of Reactor Safety Docket No. 50-313 License No. DPR-51
Enclosures:
1. Examination Report 05000313/2016301 w/Attachment: Supplemental Information 2. NRC Post-Examination Comment Resolution 3. Simulator Fidelity Report
REGION IV==
Docket: 05000313 License: DPR-51 Report: 05000313/2016301 Licensee: Entergy Operations, Inc.
Facility: Arkansas Nuclear One, Unit 1 1448 SR 333 Location:
Russellville, AR 72802-0967 Dates: August 22 through September 16, 2016 Inspectors: T. Farina, Chief Examiner, Senior Operations Engineer C. Osterholtz, Senior Operations Engineer M. Hayes, Operations Engineer S. Hedger, Operations Engineer M. Kennard, Operations Engineer C. Steely, Operations Engineer Approved By: Vincent G. Gaddy Chief, Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY
ER 05000313/2016301; 08/22/2016 - 09/16/2016; Arkansas Nuclear One, Unit 1; Initial
Operator Licensing Examination Report.
NRC examiners evaluated the competency of nine applicants for reactor operator licenses, four applicants for instant senior reactor operator licenses, and four applicants for upgrade senior reactor operator licenses at Arkansas Nuclear One, Unit 1.
The licensee developed the examinations using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 10. The written examination was administered by the licensee on September 1, 2016. NRC examiners administered the operating tests on August 22-27, 2016.
The examiners determined that twelve of the seventeen applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
NRC-Identified and Self-Revealing Findings
None.
Licensee-Identified Violations
None.
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA5 Other Activities (Initial Operator License Examination)
.1 License Applications
a. Scope
NRC examiners reviewed all license applications submitted to ensure each applicant satisfied relevant license eligibility requirements. Examiners also audited three of the license applications in detail to confirm that they accurately reflected the subject applicants qualifications. This audit focused on the applicants experience and on-the-job training, including control manipulations that provided significant reactivity changes.
b. Findings
No findings were identified.
.2 Examination Development
a. Scope
NRC examiners reviewed integrated examination outlines and draft examinations submitted by the licensee against the requirements of NUREG-1021. The NRC examination team conducted an on-site validation of the operating tests.
b. Findings
NRC examiners provided outline, draft examination, and post-validation comments to the licensee. The licensee satisfactorily completed comment resolution prior to examination administration.
NRC examiners determined the written examinations and operating tests initially submitted by the licensee were within the range of acceptability expected for a proposed examination.
.3 Operator Knowledge and Performance
a. Scope
On September 1, 2016, the licensee proctored the administration of the written examinations to all 17 applicants. The licensee staff graded the written examinations, analyzed the results, and presented their analysis and post-examination comments to the NRC on September 6, 2016.
The NRC examination team administered the various portions of the operating tests to all applicants on August 22-27, 2016.
b. Findings
No findings were identified.
Twelve of 17 applicants passed the written examination and all parts of the operating test. Four applicants failed the written examination only, and one applicant failed both the written examination and administrative job performance measures (JPMs). The final written examinations and post-examination analysis and comments may be accessed in the ADAMS system under the accession numbers noted in the attachment.
The examination team noted six generic weaknesses associated with applicant performance on the administrative JPM, simulator JPM, in-plant JPM, and dynamic scenario sections of the operating tests. Specifically:
1. Three applicants demonstrated a weakness in lowering the pressurizer quench tank down to a specific level by transferring water to the clean waste receiving tank. The applicants all secured the associated transfer pump at the desired level; however, they delayed isolating the two quench tank drain valves, causing quench tank level to continue lowering to the low level alarm setpoint due to a difference in elevations between the two tanks and a 4.7 psig Nitrogen over-pressure in the quench tank.
2. Seven applicants demonstrated a weakness implementing the stations work hour fitness-for-duty guidance. Given a work history of five operators, many applicants incorrectly stated that one of the operators failed to satisfy the requirement to have a minimum 34-hour break in a 9-day period and was, therefore, ineligible to stand an emergent watch. In actuality, the applicant had one full day off in between two day-shift watches, which equated to at least a 35-hour break, satisfying the 34-hour requirement.
3. When purging the main generator with CO2 following a station blackout, several operators failed to use the designated rack of 15 CO2 bottles, which is labeled Reserved for Unit 1 EOP [Emergency Operating Procedure]
Requirements, and instead used an identical rack of 15 CO2 bottles which is intended for normal daily use, not EOPs. Procedure 1106.002, Section 15.0, Generator Hydrogen System, fails to specify which bank should be used, constituting a missed opportunity to prevent operator error.
4. Three out of four crews unnecessarily tripped the reactor on a continuous rod withdrawal malfunction, which could have been mitigated by proper implementation of Alarm Response Procedure ACA 1203.003, Section 9, Control Rod Drive Malfunction Actions. Two of the crews failed to properly implement the MAJI sequence of operator actions, which directs the crew to verify the diamond panel in MANUAL, set the GROUP-AUXIL switch to AUXIL, set SPEED SELECTOR switch to JOG, and place manual command switch in INSERT for three seconds. These two crews failed to stop rod motion, and manually tripped the reactor. One crew properly stopped all rod motion, but misinterpreted plant indications to mean that rod motion was continuing and tripped the reactor as well.
5. Several crews demonstrated a weak understanding of generator hydrogen temperature control valve CV-4018 controller operation. This controller is reverse acting, meaning that it generates a 0 percent signal to open the valve and a 100 percent signal to fully close the valve. Some believed that a 100 percent signal indicated open, and several applicants misunderstood this controller to indicate actual position, when it only indicates demanded position.
Actual valve position can only be verified locally. One crew tripped the turbine instead of taking manual control of the controller, which would have mitigated the malfunction.
6. Three crews demonstrated weaknesses implementing the steam generator tube rupture procedure. Errors included tripping the reactor out of sequence, initiating emergency feedwater instead of auxiliary feedwater, unnecessary delays establishing a plant cooldown, and a failure to control steam header pressure to prevent unnecessary cycling of the bad steam generator atmospheric dump valve.
Copies of all individual examination reports were sent to the facility training manager for evaluation and determination of appropriate remedial training.
.4 Simulation Facility Performance
a. Scope
The NRC examiners observed simulator performance with regard to plant fidelity during examination validation and administration.
b. Findings
No findings were identified. However, two simulator-related issues occurred during examination administration that warrant discussion.
1. The simulator control software includes an administrative trainee action monitor program which logs simulator switch manipulations during the course of a scenario.
During nine scenarios, this log was flooded due to noisy analog inputs on control boards C03/C04, causing it to exceed its capacity. The trainee action monitor log can accommodate a maximum of 1,000 entries in a scenario, but the noisy analog inputs caused multiple log entries per second which filled the log to capacity prior to scenario completion. This program had no impact on simulator performance or fidelity, but it resulted in an incomplete record of applicant actions for the scenarios which were affected. The licensee initiated simulator deficiency report DR16-0120 to address the issue.
2. During one administration of JPM S-6, Transfer Buses from the Unit Aux
Transformer to a Startup Transformer, an applicant was required to open breaker H-14 on 6900V bus H1 to correct a condition where the bus was aligned to two power sources simultaneously. The applicant correctly identified that breaker H-14 failed to open automatically after the H-15 Synchronize switch was taken to off, and attempted to manually open breaker H-14 twice. Both attempts appeared to be a full turn of the handswitch, but the breaker remained closed on the control board. Therefore the applicant correctly implemented step 8.4.6.A.1 of Procedure OP-1107.001, and manually reopened breaker H-15 from the startup transformer #1 to ensure that bus H1 remained energized from only one source.
Post-scenario, the examiner reviewed available simulator data with the licensee to determine what actions were captured by the machine. The trainee action monitor log records all manipulations which are performed on the simulator. The trainee action monitor log clearly showed that component C10 152-14/CS (breaker H-14 control switch) was taken to Trip multiple times and returned to normal. The simulator did not pick up this action, however, and continued to display the breaker as closed. The licensee believes that the issue may be related to a difference in scan operating speeds between the simulator operating system (10 Hz) and the simulator subroutine that controls large breaker logic (2 Hz). The licensee initiated simulator deficiency reports DR16-0122 and DR16-0123 to address the issue.
.5 Examination Security
a. Scope
The NRC examiners reviewed examination security for examination development during both the on-site preparation week and examination administration week for compliance with 10 CFR 55.49 and NUREG-1021. Plans for simulator security and applicant control were reviewed and discussed with licensee personnel.
b. Findings
No findings were identified.
4OA6 Meetings, Including Exit
Exit Meeting Summary
The chief examiner presented the preliminary examination results to Mr. D. Perkins, Senior Operations Manager, and other members of the staff on August 26, 2016. A telephonic exit was conducted on September 16, 2016, between Mr. T. Farina, Chief Examiner, and Mr. Perkins.
The licensee did not identify any information or materials used during the examination as proprietary.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- R. Martin, Operations Training Superintendent
- B. Possage, Examination Writer
- J. Cork, Examination Writer
NRC Personnel
- B. Tindell, Senior Resident Inspector
ADAMS DOCUMENTS REFERENCED