IR 05000335/2009301
| ML100430839 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/12/2010 |
| From: | Widmann M NRC/RGN-II/DRS/OLB |
| To: | Nazar M Florida Power & Light Co |
| References | |
| ER-09-301 | |
| Download: ML100430839 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION uary 12, 2010
SUBJECT:
ST. LUCIE NUCLEAR PLANT - NRC OPERATOR LICENSE EXAMINATION REPORT 05000335/2009301 AND 05000389/2009301
Dear Mr. Nazar:
During the period of October 19, - 22, 2009, the Nuclear Regulatory Commission (NRC)
administered operating tests to employees of your company who had applied for licenses to operate the St. Lucie 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 December 15, 2009.
Four Senior Reactor Operator (SRO) applicants and two Reactor Operator (RO) applicants passed both the operating and the written examinations. There were three post-examination comments concerning the written examination. These comments, and the NRC resolution of these comments, are summarized in Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.
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).
FP&L 2 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) 562-4550.
Sincerely,
/RA By Richard S. Baldwin For/
Malcolm T. Widmann, Chief Operations Branch Division of Reactor Safety Docket Nos: 50-335, 50-389 License Nos: DPR-67, NPF-16
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Docket No.: 50-335, 50-389 License No.: DPR-67, NPF-16 Report No.: 05000335/2009301, 05000389/2009301 Licensee: Florida Power and Light Company (FP&L)
Facility: St. Lucie Nuclear Plant, Units 1 & 2 Location: 6351 S. Ocean Drive Jensen Beach, FL 34957 Dates: Operating Test - October 19 - 22, 2009 Written Examination - December 15, 2009 Examiners: R. Baldwin, Chief Examiner, Senior Operations Engineer M. Bates, Senior Operations Engineer P. Capehart, Operations Engineer Approved by: Malcolm T. Widmann, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS ER 05000335/2009301, 05000389/2009301, 10/19-22/2009 and 12/15/2009; St. Lucie Nuclear Plant, Licensed Operator Examinations.
The NRC examiners conducted operator licensing initial examinations in accordance with the guidance in NUREG-1021, Revision 9, Supp.1, Operator Licensing Examination Standards for Power Reactors. This examination implemented the operator licensing requirements of 10 CFR §55.41, §55.43, and §55.45.
The NRC administered the operating tests during the period of October 19 - 22, 2009.
Members of the St. Lucie Power Plant training staff administered the written examination on December 15, 2009. The written examination outline, written examination, operating test outlines and operating test details were developed by the St. Lucie Nuclear Plant training staff.
Four Senior Reactor Operator (SRO) and two Reactor Operator (RO) applicants passed both the written and operating examinations.
There were three post examination comments.
No findings of significance were identified.
Enclosure 1
REPORT DETAILS 4. OTHER ACTIVITIES 4OA5 Operator Licensing Initial Examinations a. Inspection Scope The St. Lucie Nuclear Plant training staff developed the written examination and operating test. NRC regional examiners reviewed the proposed examination material to determine whether it was developed in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1.
Examination changes agreed upon between the NRC and the licensee were made according to NUREG-1021 and incorporated into the final version of the examination materials.
The examiners reviewed the licensees examination security measures while preparing and administering the examinations to ensure examination security and integrity complied with 10 CFR 55.49, Integrity of Examinations and Tests.
The examiners evaluated four SRO and two RO applicants who were being assessed under the guidelines specified in NUREG-1021. The examiners administered the operating tests during the period of October 19-22, 2009. Members of the St. Lucie Power Plant training staff administered the written examination on December 15, 2009.
The evaluations of the applicants and review of documentation were performed to determine if the applicants, who applied for licenses to operate the St. Lucie Nuclear Plant, met requirements specified in 10 CFR Part 55, Operators Licenses.
b. Findings The NRC determined that the details provided by the licensee for the written exam, walkthrough, and simulator tests were within the range of acceptability expected for a proposed examination.
Four Senior Reactor Operator (SRO) and two RO applicants passed both the written and operating examinations.
A copy of the final as-given RO and SRO written examinations and answer keys, with all changes incorporated, and the licensees post-examination comments, may be accessed in the ADAMS system (ADAMS Accession Numbers, ML100410029, ML100410030 and ML1004128).
Copies of all individual examination reports were sent to the facility Training Manager for evaluation and determination of appropriate remedial training.
Enclosure 1
4OA6 Meetings Exit Meeting Summary On October 22, 2009, the examination team discussed generic issues associated with the operating test with Mr. B. Hughes, Plant General Manager, and members of the St.
Lucie Nuclear Plant staff. The examiners asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
PARTIAL LIST OF PERSONS CONTACTED Licensee personnel T. Benton - Operations Training Supervisor S. Duston - Training Manager F. Forrest - Asst. Operations Manager M. Hicks - Operations Manager B. Hughes - Plant General Manager J. Klauck - Asst. Operations Manager M. Verbeck - ILT Training Supervisor NRC personnel T. Hoag, SRI M. Sanchez,RI Enclosure 1
A complete text of the licensees post examination comments can be found in ADAMS under Accession Number ML100410028.
RO QUESTION # 23 LICENSEE COMMENT:
The question concerns itself with a Steam Generator (S/G) Tube Rupture in the 1 B S/G (pressure 840 psia) where a Reactor Coolant System (RCS) depressurization is to be performed in accordance with 1-EOP-04, Steam Generator Tube Rupture. The applicant, (first part) was expected to identify the appropriate pressure band (depressurization strategy), as well as, the reason (second) for maintaining the S/G in this pressure band during RCS depressurization.
The licensee contends, in their post examination comment, that the question has two correct answers, the original answer B and an additional answer D.
The licensee identifies in procedure, 1-EOP-04, Step 11 criteria/parameters that are to be evaluated during the RCS depressurization. These criteria/parameters (Figure 1A pressure limits, Maximum pressure 930 psia, RCS minimum pressure for RCP operation, ) require the crew to maintain the RCS pressure with ALL the criteria. Specifically S/G pressure must be maintained within 50 psia of the ruptured S/G pressure. The crew is to use these for stopping the RCS depressurization. The licensee identifies that the goal, during depressurization, is to provide subcooling during heat removal while avoiding over pressurization. This is done because it could lead to Pressurized Thermal Shock (PTS) and Reference Transition Nil-Ductility Temperature (RT NDT) issues while minimizing the differential pressure between the RCS and the S/Gs. Furthermore, it is done to cause a secondary to primary leak (backflow) to control S/G level or reduce S/G pressure.
For the second part of the answers licensee identifies, that the original answer Bs, second part as well as, answer Ds second part of the question is derived from the same sentence in CEN-152, "Combustion Engineering Emergency Procedures Guidelines, thus making the second part of answer D also correct.
Therefore, the licensee contends that since both parts of the question is elicited in two answers, the original answer, B, there is an additional correct answer, D.
NRC RESOLUTION:
The NRC agrees with the licensees recommendation. After reviewing the licensees contention and supporting documentation, it appears to support that either answer B or D would be allowed in accordance with procedures, 1-EOP-04 and CEN-152. Since the applicants could answer the question and both answers were listed in two distractors, the comment appears valid.
Based on the above discussion, the licensees recommendation was accepted; answer choices B and D will be considered as correct answers. The answer key was changed to reflect this change.
Enclosure 2
RO QUESTION # 42 LICENSEE COMMENT:
The question concerns itself with the plant at 100% power with a rupture in the 1 A CCW (Component Cooling Water) Header. The question identifies that the crew was using procedure, 1-310030, Component Cooling Water (CCW) OFF Normal Operation and asks how the CCW system, based on a rupture of the 1A CCW header, was compromised and what actions should be carried out within 45 minutes if the Containment temperature can not be lowered to 120 degrees Fahrenheit (F) or lower. Part one of two answers (A and B) indicates that a partial loss of CCW flow and for the other two answers (C and D) indicates a total loss of CCW flow to the A and B Containment Coolers. Part 2 of each distractor asks what procedure the crew should either initiate; EOP-01, Standard Post Trip Actions, (A and C) or 2-ONP-22.01, Rapid Down Power (B and D)
The licensee contends, in their post examination comment, that the applicant has to identify the impact on containment cooling and the procedural directions based on the containment exceeding 120 degrees F. The licensee noted that for the first part of the question for distractors C and D are the same (concerning CCW), in that a total loss of CCW flow to the A and B Containment coolers, making answers C and D correct. They additionally identify that distractors A and B are the same (concerning CCW) in that a partial loss of CCW flow to the A and B Containment Coolers) making answers.A and B incorrect.
Additionally, the licensee identifies that an operator would have a different view point which would allow use of the Off Normal Procedure(s) (ONP(s)). Because of this allowance, the operator could take action in accordance with 1-ONP-25.01, Loss of RCB Cooling Fans, which would allow the crew to perform a down power (Contingency Action Step 6.4.2.A) in accordance with procedure 1-ONP-22.01, Rapid Down Power. Because of this allowance of using the ONP, the licensee identified that the applicant could make an assumption since there is no time frame identified in the question, between one and forty-four minutes into the event (any time prior to the 45 minute action requirement). By assuming a finite time before the 45 minutes limit, this would have allowed the crew time to perform a rapid down power in order to attempt the reduction of containment temperature to less than 120 degrees F. This would then allow the second part of answer D becomes also a correct answer.
Therefore, the licensee contends that since both parts of the question are elicited in two answers, the original answer, C, there is an additional correct answer, D.
NRC RESOLUTION:
The NRC agrees with the licensees recommendation. After reviewing the licensees contention and supporting documentation appears to support that either the original answer C, or answer D should be considered correct. It appears that procedure 1-NOP-25-01 as well as 1-310030 provides guidance for answering this question. While procedure 1-310030 identifies the response for the operators to perform a reactor trip and Standard Post Trip Actions (SPTAs)
Procedure 1-NOP-25.01 also provides the down power requirement (in contingency Step 6.4.2.A) as well as requiring the reactor trip and performance of the SPTAs (in a note prior to Enclosure 2
Step 6.4.1). Based on the discussion, it appears that either procedural direction could be used to answer this question.
Based on the above discussion, the licensees recommendation was accepted; answer choices C and D will be considered as correct answers. The answer key was changed to reflect this change.
RO QUESTION # 51 LICENSEE COMMENT:
The question concerns itself with a Unit 1 process monitor if it fails to perform its control function and because of this failure which one failed monitor would result in an UNMONITORED release of radioactive effluent. The examinee was required to select from the four different systems, the correct answer.
The licensee contends that the original answer B and answer C (S/G Blowdown Monitor)
should be considered correct. The licensee provided information that identified that procedure 1-NOP-26.01, Process Radiation Monitors, identifies valves that receive a close signal when a valid high radiation signal is indicated. These valves are associated with Channel 44 (Component Cooling Water (CCW) Monitor) and Channel 45 (S/G Blowdown Monitor). Both protective function failures would cause an UNMONITORED release in the CCW and S/G Blowdown systems. The licensee provided plant Piping and Instrument diagrams (P&IDs) that identified the pathway for the S/G blowdown system and is sent to the Monitor Storage Tank (MST) which is vented to atmosphere and they indicate that the system is not monitored.
The licensee concludes that answers B and C both have control functions associated with their respective radiation monitors (as identified in 1-NOP-26.01) and a failure of these functions would have the same end result of an unmonitored release.
Therefore, the licensee contends there are two correct answers for this question, choice B and choice C.
NRC RESOLUTION:
The NRC agrees with the licensees recommendation. After reviewing the licensees post-examination comment and supporting documentation, it appears to support that either the original answer choice B, choice C are correct. The inspectors reviewed the P&IDs provided and with further discussions with the licensee training department personnel, it was determined that S/G blowdown does in fact goes to the MST and vents to atmosphere and is not monitored.
Based on the above discussion, the licensees recommendation was accepted; answer choices B and C will be considered as correct answers. The answer key was changed to reflect this change.
Enclosure 2
SIMULATION FACILITY REPORT Facility Licensee: St. Lucie Nuclear Plant Facility Docket Nos.: 05000335/05000389 Operating Tests Administered on: October 19 -22, 2009 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.
No simulator fidelity or configuration items were identified.
Item Description None Enclosure 3