IR 05000424/2011301
| ML111320617 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/12/2011 |
| From: | Widmann M Division of Reactor Safety II |
| To: | Tynan T Southern Nuclear Operating Co |
| References | |
| 50-424/11-301, 50-425/11-301 50-424/11-301, 50-425/11-301 | |
| Download: ML111320617 (16) | |
Text
UNITED STATES May 12, 2011
SUBJECT:
VOGTLE ELECTRIC GENERATING PLANT - REACTOR AND SENIOR REACTOR OPERATOR INITIAL EXAMINATIONS 05000424/2011301 AND 05000425/2011301
Dear Mr. Tynan:
During the period March 16 - 24, 2011, the Nuclear Regulatory Commission (NRC)
administered operating tests to employees of your company who had applied for licenses to operate the Vogtle Electric Generating 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 April 1, 2011.
Ten applicants took the examination. Two Reactor Operator (RO) and one Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. There were six post-examination comments concerning the written examination. The comments, and the NRC resolution of the comments, are summarized in Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.
The initial operating test and written SRO examination submitted by your staff failed to meet the guidelines for quality contained in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1, as described in the enclosed report.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document
SNC 2 Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
If you have any questions concerning this letter, please contact me at (404) 997-4550.
Sincerely,
/RA/
Malcolm T. Widmann, Chief Operations Branch 1 Division of Reactor Safety Docket Nos: 50-424, 50-425 License Nos: NPF-68, NPF-81
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Docket Nos.: 05000424, 05000425 License Nos.: NPF-68, NPF-81 Report No.: 05000424/2011301 and 05000425/2011301 Licensee: Southern Nuclear Operating Company, Inc.
Facility: Vogtle Electric Generating Plant Location: 7821 River Road Waynesboro, GA 30830 Dates: Operating Test - March 16 - 24, 2011 Written Examination - April 1, 2011 Examiners: P. Capehart, Chief Examiner, Senior Operations Engineer M. Meeks, Operations Engineer J. Hopkins, Reactor Technology Instructor Approved by: Malcolm T. Widmann, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS ER 05000424/2011301, 05000425/2011301; March 16 - 24, 2011, and April 1, 2011; Vogtle Electric Generating Plant Units 1 and 2; Operator License Examinations.
Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 9, Supplement 1, of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." This examination implemented the operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45 as applicable.
Members of the Vogtle Electric Generating Plant staff developed both the operating tests and the written examination. The initial operating test outline and the written Senior Reactor Operator (SRO) examination submittal did not meet the quality guidelines contained in NUREG-1021.
The NRC administered the operating tests during the period March 16 - 24, 2011. Members of the Vogtle Electric Generating Plant training staff administered the written examination on April 1, 2011. One SRO and two Reactor Operator (RO) applicants passed both the operating test and written examination, and were issued licenses commensurate with the level of examination administered.
There were six post-examination comments on the written examination.
No findings were identified.
Enclosure 1
REPORT DETAILS 4. OTHER ACTIVITIES 4OA5 Operator Licensing Examinations a. Inspection Scope Members of the Vogtle Electric Generating Plant staff developed both the operating tests and the written examination. All examination material was developed in accordance with the guidelines contained in Revision 9, Supplement 1, of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." The NRC examination team reviewed the proposed examination. Examination changes agreed upon between the NRC and the licensee were made per NUREG-1021, and incorporated into the final version of the examination materials.
The NRC reviewed the licensees examination security measures while preparing and administering the examinations in order to ensure compliance with 10 CFR §55.49, Integrity of examinations and tests.
The NRC examiners evaluated four RO applicants and six SRO applicants using the guidelines contained in NUREG-1021. The examiners administered the operating tests during the period March 16 - 24, 2011. Members of the Vogtle Electric Generating Plant training staff administered the written examination on April 1, 2011. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Vogtle Electric Generating Plant, met the requirements specified in 10 CFR Part 55, Operators Licenses.
b. Findings No findings were identified. The NRC determined that the licensees initial operating and SRO written examination submittal did not meet the guidelines for quality contained in NUREG-1021.
The scenario submittal did not meet the criteria for each applicant to receive the minimum number of transients and events or Technical Specifications as specified on Form ES-301-5. The three scenarios submitted contained events that did not meet the ES-301 criteria for verifiable actions. These events would not have provided sufficient insight of an applicants competence. Examples of other deficiencies with the initial submittal of the simulator scenarios and JPMs included the following items:
- Incorrect simulator setup.
- Incorrect JPM cue sheets.
- All of the needed references to perform the administrative JPMs were not initially provided.
- Incorrect valve and/or component numbers.
- Inadequate pass/fail criteria for critical steps.
- Initial conditions missing important details.
Enclosure 1
- In-plant JPM delays were created due to the Shift Manager and Shift Supervisor not being previously briefed and signed for on the security agreement for JPMs conducted inside the main control room area.
- JPM Steps either inappropriately designated as "critical" or should have been designated as "critical."
The NRC determined that the licensees initial SRO written examination submittal was outside the range of acceptable quality specified by NUREG-1021. More than 20% (11 of the 25 or 44%) questions contained unacceptable flaws. Individual questions were evaluated as unsatisfactory for the following reasons:
- (3) Questions failed to meet the K/A statement contained in the examination outline.
- (3) Questions contained two or more implausible distracters.
- (2) Questions contained other unacceptable psychometric flaws.
- (2) Questions contained multiple unacceptable flaws.
Of the six exam post exam comments, three affected the SRO written exam. Of these three, one required a change to the SRO exam key (question deletion or answer key change), and is therefore also considered in the total count of unsatisfactory SRO questions. Future examination submittals need to incorporate lessons learned.
Two RO applicants and one SRO applicant passed both the operating test and written examination, and were issued licenses.
Copies of all individual examination reports were sent to the facility Training Manager for evaluation of weaknesses and determination of appropriate remedial training.
The licensee submitted six post-examination comments. A copy of the final written examination and answer key, with all changes incorporated, and the licensees post-examination comments may be accessed not earlier than April 2, 2013, in the ADAMS system (ADAMS Accession Number(s) ML111260716 and ML111260719.)
4OA6 Meetings, Including Exit Exit Meeting Summary On March 4, 2011, the NRC examination team discussed generic issues associated with the operating test with Mr. Richard D. Brigdon, Training Manager, and members of the Vogtle Electric Generating Plant staff. The examiners asked the licensee if any of the examination material was proprietary. No proprietary information was identified.
Enclosure 1
KEY POINTS OF CONTACT Licensee personnel R. Brigdon, Training Manager R. Dorman, Operations Shift Manager G. Gunn, Nuclear Operations Training Supervisor T. Parton, Operations Support Superintendent P. Tucker, Nuclear Operations Plant Instructor, Lead (Initial)
G. Wainwright, Nuclear Operations Plant Instructor, Lead (Exams)
Enclosure 1
FACILITY COMMENTS AND NRC RESOLUTION NRC RESOLUTION TO THE FACILITY COMMENT:
A complete text of the licensee's post examination comments can be found in ADAMS under Accession Number ML111260716 and ML111260719.
RO QUESTION # 10 LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers.
The question states that the MCR has been evacuated and all switches at the Post Shutdown Panel (PSD) train B have been taken to local, and all switches on PSD train A remain in the REMOTE position. An automatic SI occurs after the evacuation. This will cause Train A SI to automatically actuate due to its handswitches position still being in the REMOTE position. Train B will not automatically actuate due to its handswitches being in the LOCAL position.
The licensee contends that ECCS flow would be necessary to control PRZR pressure and level based on the conditions given in the stem of the question. Step 30 RNO step g. substep 2 of procedure 18038-C states to start ECCS pumps as necessary to maintain Pressurizer (PRZR)
pressure, and step 39 to start ECCS pumps to maintain PRZR level between 50% -70%. The licensee contends that this could have already occurred and since the question does not give a timeline, choice D (Both Trains of SI would be running at this point in time) is also correct.
NRC DISCUSSION:
The question asked for the applicant to identify 1) SI Pump status (one train or both trains running), and 2) the location where it is possible to shutdown the SI pumps that are running based on the given conditions. Only the first part of this two part question is being challenged.
The intent of the governing plant procedure, (18038-1, U) per VP-LO-PP-60328 is to:
Bottle up the RCS; Protect RCS inventory; Protect RCS subcooling; Protect against uncontrolled cooldown; Insure critical components in AUTO; and Protect against excessive RCS cooldown.
Since no additional information is given in the stem of the question that the MCR was evacuated due to a fire, the procedure would direct the applicant per step 5 of 18038, to verify the trip of the Main Turbine to prevent excessive cooldown and verify PRZR pressure control in AUTO.
Normal Charging and Letdown are still in service at this time. Also, with no other information given in the stem of the question, the operator should only consider the conditions for those created in accordance with the procedure prior to exiting the MCR up to the point for placing the local handswitches in LOCAL exist. At this point in time, there would be no response required from the outside operators to control PRZR Pressure or Level. A manual SI would not be a consequence of the conditions stated in the question. Since no information is given in the stem Enclosure 2
of the question to identify the cause of the automatic SI signal, the source of the automatic SI signal could therefore be a condition that leads to an SI signal but has not jeopardized PRZR pressure or level. Since no additional information is given in the stem of the question that infers a potential loss of PRZR level or pressure, the applicant should address this question based on the Caution statement given on page 12 of 18038-1 that informs the operator that, If any SI actuation occurs, any components previously transferred to the Shutdown Panels will only actuate if manually actuated by the Operator. Components not transferred to the Shutdown Panel should realign to the safety configuration.
NRC RESOLUTION Comment not accepted. In accordance with NUREG 1021, Rev 9, Supplement 1, ES-403, Section D.1.d, Question # 10 is correct as administered with only one correct answer.
RO QUESTION # 52 LICENSEE COMMENT:
In summary, the licensee requested that this question be deleted from the exam as there is no correct answer provided.
The question states that the MCR has been evacuated due to a fire that threatens personnel safety and the Shift Supervisor has made the decision at step 42 of 18038-1, to locally trip the MFPTs to limit the plant cooldown from steam usage. The next step (step 43) requires that all but one condensate pump be shutdown. Prior to these steps, step 25 requires the isolation of MSIVs and BSIVs. This isolates the main steam and auxiliary steam paths to the affected Unit.
Since the MCRs are shared, a fire would impact both Units MCRs and therefore requires both to be abandoned. Actions per 18038-1/2 would be taken on Unit 1 and Unit 2. Since both units are affected, neither would have main or auxiliary steam available, and the possibility of supplying auxiliary steam from the other unaffected unit is not plausible. Auxiliary steam to maintain condenser vacuum would not be available to either unit and therefore the reason for keeping one condensate pump in operation is no longer valid for this condition.
The licensee contends that even though the procedure directs that the MFPT may be secured using the guidance of procedure 18038-1/2 to minimize the cooldown, the subsequent reason for the second part of the answer given to keep one condensate pump running is not applicable for the conditions given and is therefore incorrect.
NRC DISCUSSION:
The question asked for the applicant to identify 1) the final condensate pump alignment (how many running), and 2) the reason for the final alignment. Only the second part of this two part question is being challenged.
NRC RESOLUTION Comment accepted. In accordance with NUREG 1021, Rev 9, Supplement1, ES-403, Section D.1.c, Question # 52 does not have a correct two part response for the given conditions in the stem and therefore will be deleted from the exam.
Enclosure 2
RO QUESTION # 66 LICENSEE COMMENT:
In summary, the licensee requested that this question be deleted from the exam because the question being asked is at the SRO level and is therefore not appropriate for the RO exam.
NRC DISCUSSION:
The question required applicants to correctly recognize: (1) a situation where internal notification to the OMOC was required; and (2) minimum qualifications required to serve as the OMOC. Both knowledge items were as stated in Southern Nuclear Operating Company procedure NMP-OS-007-001, Conduct of Operations Standards and Expectations, section 6.2, Operations Manager On-Call. Specifically, section 6.2.2.1 states, in part: An individual [i.e.,
the OMOC], assigned by the Operations Manager is on-call to assist the duty Shift Manager in the resolution of problems. Therefore, it is apparent that it is the responsibility of the site Operations Manager to appoint (and verify the qualifications of) the OMOC. In accordance with NMP-OS-007-001, it is specifically the Shift Manager (i.e., a SRO) who is responsible for recognizing situations that would require notification to the OMOC, and for conducting discussions concerning operational decisions and insights with the OMOC. Therefore, it is unreasonable to expect ROs to be conducting calls with the OMOC during the course of their on-shift duties; and it is unreasonable to expect an RO would be responsible for verifying minimum qualifications of the individual OMOC. Therefore, required knowledge to correctly answer this question is beyond the scope of what is expected for a written examination question at the RO level.
NRC RESOLUTION Comment accepted. Because the question was written at the SRO level and is not linked to RO job requirements, this question will be deleted from the written examination in accordance with NUREG-1021, ES-403, paragraph D.1.b.
SRO QUESTION # 78 LICENSEE COMMENT:
In summary, the licensee requested that this questions correct response should be C instead of D.
The question asks if the PORV is OPERABLE based on the PORV being isolated with the block valve due to excessive seat leakage. The licensee contends that the T.S. Bases for 3.4.11 states, An OPERABLE PORV is required to be capable of manually opening and closing and is not experiencing excessive seat leakage. Excessive seat leakage is given as an initial condition given in the stem of the question. Therefore, based on the T.S. Bases, C is the only correct answer.
NRC DISCUSSION:
The question asked for the applicant to identify whether it is required to close the 1) PORV or PORV block valve to stop excessive leakage from the PORV, and if the PORV is now considered to be 2) OPERABLE or INOPERABLE.
Enclosure 2
NRC RESOLUTION Comment accepted. In accordance with NUREG 1021, Rev 9, Supplement1, ES-403, Section D.1.b, based on additional information provided by the licensee, Question # 78 had an incorrect answer annotated and therefore the correct answer should be changed from D to C.
SRO QUESTION # 90 LICENSEE COMMENT:
In summary, the licensee requested that this question be deleted from the exam because of the unclear question stem.
The licensee contends that the stem of the question should have directly asked if the SG Pressure at the Remote Shutdown Panel is Fire Event Qualified. The question therefore has an unclear stem that confused the applicants and did not ask the question in a more direct manner.
NRC DISCUSSION:
The question asked for the applicant to identify Which ONE of the following is correct concerning both (1) the Control Room indicators and (2) the Remote Shutdown Panel indicators for SG Pressure? Control Room indicators are Post Accident Monitor qualified and the B Remote Shutdown Panel (RSP) indicators are NOT Fire Event Qualified. In the question stem is a statement that there are Red Bezels at the bottom of some Control Room SG Pressure indicators, and that SG Pressure is also indicated on the Remote Shutdown Panels. This statement was placed in the stem of the question to imply that there are also some Red Bezels at the bottom of some indicators (i.e. SG 2/3 WR Level) at the RSP. The Red Bezels at the RSP indicate that these indicators are Fire Event Qualified. To answer the question, the applicant had to use acquired knowledge that the SG Pressure indicators at the RSP do not have Red Bezels on their indicators.
ES-403.D.1.b states that a question that contains psychometric errors that would be unsatisfactory during the exam pre-review would not justify deleting a question after exam administration. The responses are a collection of true/false statements however; they do not in of themselves increase the difficulty level of the question. The reason for disallowing a series of true/false distractors is that this makes the question stem irrelevant. The applicant can arrive at the answer for the question by eliminating all of the false statements in the distractors.
NRC RESOLUTION Comment not accepted. In accordance with NUREG 1021, Rev 9, Supplement 1, ES-403, Section D.1.b, based on additional information provided by the licensee, Question # 90 is noted to contain psychometric errors. However, these errors do not disqualify the question from the exam. Therefore, the question is considered satisfactory and will not be deleted from the exam.
SRO QUESTION # 97 LICENSEE COMMENT:
In summary, the licensee requests that two answers for this (SRO) question be accepted.
Enclosure 2
There are two correct answers for this question. The second part of one of the distractors is a numerical subset of the correct answer choice and therefore this response is also correct.
Specifically, the licensee stated that if Unit 2 was held in MODE 3 for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in MODE 3 for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, all Technical Specification (TS) required actions and completion times would be met.
NRC DISCUSSION:
The question asked the applicant to identify, given a certain level for each of the U2 Condensate Storage Tanks (CST #1 and CST #2), whether the Unit 2 CSTs are 1) OPERABLE or INOPERABLE and 2) The Tech Spec Bases for the minimum level requirements.
Only the second part of this two part question is being challenged. The licensee agrees that the Unit 2 CSTs are INOPERABLE.
At Vogtle, there is a specific unit difference in the TS minimum level requirements for Unit 2 which is stated in Unit 2 TS Bases p. B 3.7.6-3 as follows:
The basis for requiring an additional 38,000 gallons of safety-related usable CST inventory is to support the elimination of the bypass line and associated valve bonnet depressurization line for the 2HV-8701B RHR suction isolation valve.
The elimination of the bypass line and valve bonnet depressurization line requires an additional 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for a total of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to placing RHR Train A in service. The additional time ensures that the 2HV-8701B valve bonnet and the space between the 2HV-8701B and 2HV-8701A RHR suction isolation valves have depressurized sufficiently to allow the suction isolation valves to be opened.
The licensee is correct in that holding the unit in MODE 3 for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> would comply with the TS requirements for the required actions and completion times. Furthermore, the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> requirement would be the correct basis for the Unit 1 CST. However, the second part of this question specifically asks for the TS basis (not the TS completion times) for the Unit 2 minimum level requirements, which is listed above. Therefore, the only correct answer for the second part is that the basis for the TS limits is holding the unit in MODE 3 for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> (i.e., the additional 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> plus the 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> cooldown to achieve the total minimum time of [4 hrs in hot standy + 3 hrs additional time + 5 hrs cooldown time to RHR entry] = 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />). Due to the question being unit specific, there is no subset issue between responses C and D. The minimum total time required prior to RHR entry is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The minimum total time for response C is only 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Therefore, this response does not meet the minimum total time (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) and is therefore not correct and is not a subset of D, which is the correct response.
NRC RESOLUTION Comment not accepted. In accordance with NUREG 1021, Rev 9, Supplement 1, ES-403, Section D.1.d, Question # 97 is correct as administered with only one correct answer.
Enclosure 2
SIMULATOR FIDELITY REPORT Facility Licensee: Vogtle Electric Generating Plant Facility Docket No.: 05000424 and 05000425 Operating Test Administered: March 16 - 24, 2011.
This form is to be used only to report observations. These observations do not constitute audit or inspection findings, and without further verification and review in accordance with Inspection Procedure 71111.11, are not indicative of noncompliance with 10 CFR 55.46.
During the NRC administration of one job performance measure (JPM), the examiner noted that a label was missing from what appeared to be the Remote H2 Monitor Reset pushbutton.
Based on discussion with the licensee, this alarm was originally approved for installation on 10/30/2000. The label in the simulator is missing; the one in the plant control room is labeled as HS-7 (a vendor alias). The control room label is not in accordance with the plant annunciator response (Window F05, 17062-1(2)), which states the Hand Switch number as 1(2)-HS-87024A(B) (Simulator Condition Report #2011106609).
Enclosure 3