IR 05000275/2011301
| ML112640541 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 09/20/2011 |
| From: | Haire M S Operations Branch IV |
| To: | Conway J T Pacific Gas & Electric Co |
| References | |
| 50-275/11-301, 50-323/11-301 | |
| Download: ML112640541 (18) | |
Text
September 20, 2011
John Senior Vice President-Energy &
Supply and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Nuclear Plant 77 Beale Street, Mail Code B32 San Francisco, CA 94105
SUBJECT: DIABLO CANYON POWER PLANT, UNITS 1 AND 2 - NRC EXAMINATION REPORT 05000275/2011301; 05000323/2011301
Dear Mr. Conway:
On July 20, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an initial operator license examination at Diablo Canyon Power Plant, Units 1 and 2. The enclosed report documents the examination results and licensing decisions. The preliminary examination results were discussed on July 20, 2011, with Mr. James Becker, Site Vice President, and other members of your staff. A telephonic exit meeting was conducted on August 22, 2011, with Mr. Bill Hendy, Operations Training Manager, who was provided the NRC licensing decisions. The examination included the evaluation of nine applicants for reactor operator licenses, eleven applicants for instant senior reactor operator licenses and one applicant for upgrade senior reactor operator license. The license examiners determined that fifteen of the twenty-one 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," 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 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).
Sincerely,/RA/
Mark S. Haire, Chief Operations Branch Division of Reactor Safety Pacific Gas and Electric Company - 2 - Dockets: 50-275; 50-323 Licenses: DPR-80; DPR-82
Enclosures:
1. NRC Examination Report 05000275/2011301; 05000323/2011301 2. NRC Post Examination Comment Resolution cc w/enclosures: Distribution via ListServ Pacific Gas and Electric Company - 3 -Regional Administrator (Elmo.Collins@nrc.gov) Deputy Regional Administrator (Art.Howell@nrc.gov) DRP Director (Kriss.Kennedy@nrc.gov) Acting DRP Deputy Director (Jeff.Clark@nrc.gov) DRS Director (Anton.Vegel@nrc.gov) DRS Deputy Director (Tom.Blount@nrc.gov)
Senior Resident Inspector (Michael.Peck@nrc.gov) Resident Inspector (Laura.Micewski@nrc.gov) Branch Chief, DRS/OB (Mark.Haire@nrc.gov) Branch Chief, DRP/B (Geoffrey.Miller@nrc.gov) Senior Project Engineer, DRP/B (Rick.Deese@nrc.gov)
Project Engineer, DRP/B (Nestor.Makris@nrc.gov) DC Administrative Assistant (Agnes.Chan@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Public Affairs Officer (Lara.Uselding@nrc.gov) Project Manager (Alan.Wang@nrc.gov) Acting Branch Chief, DRS/TSB (Dale.Powers@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov) Congressional Affairs Officer (Jenny.Weil@nrc.gov) OEMail Resource RIV/ETA: OEDO (John.McHale@nrc.gov)
R:\ ADAMS ML ADAMS: No Yes SUNSI Review Complete Reviewer Initials: GWA Publicly Available Non-Sensitive Non-publicly Available Sensitive OE:OB SOE:OB SOE:OB OE:OB OE:OB GWApger BTLarson KDClayton TJFarina CSteely /RA/ /RA/ /RA/ /RA/ /RA/ 9/8/2011 9/8/2011 9/8/2011 9/19/2011 9/19/2011 C:PBB C:OB GMiller MSHaire /RA/ /RA/ 9/19/11 9/20/2011 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure 1 U.S. NUCLEAR REGULATORY COMMISSION REGION IV Dockets: 50-275, 50-323 Licenses: DPR-80, DPR-82 Report: 05000275/2011301; 05000323/2011301 Licensee: Pacific Gas and Electric Company Facility: Diablo Canyon Power Plant, Units 1 and 2 Location: 7 1/2 miles NW of Avila Beach Avila Beach, California Dates: July 8 - August 22, 2011 Inspectors: G. Apger, Chief Examiner B. Larson, Senior Operations Engineer K. Clayton, Senior Operations Engineer T. Farina, Operations Engineer C. Steely, Operations Engineer D. Reeser, Senior Operations Engineer, RIII Approved By: Mark Haire, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER05000275/2011301; 05000323/2011301; July 8 - August 22, 2011; Diablo Canyon Power Plant, Units 1 and 2; Initial Operator Licensing Examination Report. NRC examiners evaluated the competency of nine applicants for reactor operator licenses, eleven applicants for instant senior reactor operator licenses, and one applicant for upgrade senior reactor operator license at Diablo Canyon Power Plant, Units 1 and 2. The licensee developed the examinations using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 9, Supplement 1. The written examination was administered by the licensee on July 8, 2011. NRC examiners administered the operating tests on July 11-20, 2011. The examiners determined that fifteen of the twenty-one applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
A. NRC-Identified and Self-Revealing Findings
No findings were identified.
B. Licensee-Identified Violations
None
1
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. The examiners also audited four 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.
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 onsite validation of the operating tests.
b. Findings
No findings were identified.
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 that 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 July 8, 2011, the licensee proctored the administration of the written examinations to all twenty-one applicants. The licensee staff graded the written examinations, analyzed the results, and presented their analysis and post examination comments to the NRC on July 25, 2011. The NRC examination team administered the various portions of the operating tests to all twenty-one applicants from July 11 through July 20, 2011.
b. Findings
No findings were identified. One reactor operator and four instant senior operators failed the written examination.
One additional instant senior reactor operator failed the simulator portion of the operating test. Therefore, fifteen of the twenty-one applicants passed the NRC license examinations. 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 the following generic weaknesses:
- Applicants showed difficulty in finding steam generator blowdown sample and isolation valves outside containment. Additionally, some of the applicants did not verify that all of the valves had closed when checking the valves closed.
- Some of the applicants displayed an inability to close a diesel generator output breaker that had failed to automatically close.
- There was a general weakness in the ability to identify the cause of a steam flow transient at low power. Specifically, none of the crews could readily identify that a terry turbine had inadvertently started.
- Most of the crews displayed weakness in using radiation monitors to diagnose problems in the plant.
.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.
.5 Examination Security
a. Scope
The NRC examiners reviewed examination security during both the onsite 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
The chief examiner presented the preliminary examination results to Messrs. James Becker, Site Vice President, James Welsch, Station Director, and other members of the staff on July 20, 2011. A telephonic exit was conducted on August 22, 2011, between Messrs. Gabriel Apger, Chief Examiner, and Bill Hendy, Operations Training Manager.
The licensee did not identify any information or materials used during the examination as proprietary.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- S. David, Site Services Director
NRC Personnel
- L. Micewski, Resident Inspector
ADAMS DOCUMENTS REFERENCED
Accession No. ML112500075 - FINAL WRITTEN EXAMS
- Accession No. ML112500077 - FINAL OPERATING TEST Accession No. ML112500068 - POST EXAM ANALYSIS AND COMMENTS
- A complete text of the licensee's post examination analysis and comments can be found in ADAMS under Accession Number ML112500068.
- RO QUESTION # 24
- LICENSEE COMMENT:
- Which of the following describes the sequence for stopping the ECCS CCPs and SI pumps in E-1.1, SI Termination?
- The answer sheet shows "C" as the correct answer.
- Nineteen percent of the candidates selected "C".
- This question is outside the target level of difficulty range of 70 to 90 percent.
- This question was on the Reactor Operator (RO) exam.
- PG&E recommends the question be deleted from the examination because the question as written has a confusing stem and distractors and was not appropriate for operator recall level of knowledge.
- Justification:
- The question required an operator to recall steps from an Emergency Operating Procedure. Since these actions were neither immediate operator actions nor frequently performed actions, recall from memory is not required.
- This question does not discriminate and has low operational validity because, in real life, an operator is not expected to have memorized this procedure.
- Although answer 'C' correctly shows the order for securing ECCS pumps as part of EOP E-1.1, this is not always the correct order for securing ECCS pumps.
- Additionally, this question should be deleted because of psychometric flaws in that it has confusing language and inconsistent punctuation.
- For example, the question asks for the sequence for stopping pumps but includes re-aligning charging and restoring letdown in the distractors.
- Restoring letdown is a recognized part of terminating SI and performed after stopping both SI pumps but is not included in answer "C".
- The distractors lead the exam taker to believe that there are four procedural steps that must be put in the correct order.
- Some students did not select "C" because it does not include restoring letdown.
- Since the question is only asking for the correct order for securing ECCS and SI pumps, then "A" could also be a correct answer.
- Additionally, the use of periods, commas and semi-colons is inconsistent between answers.
- The question has low discrimination validity.
- When comparing candidates who scored above or below 85 percent, seven candidates scoring 85 percent or above chose "D", while nine candidates scoring less than 85 percent chose "D".
- This question was on the Reactor Operator (RO) exam. This is not a Reactor Operator knowledge level question.
- One of nine (11 percent) reactor operator candidates picked "C".
- NRC RESOLUTION TO QUESTION #24:
- Recommendation not accepted.
- The stem of the question clearly asks the applicant to determine the correct sequence for stopping ECCS centrifugal charging pumps (CCPs) and safety injection pumps (SI) while operating in Emergency Operating Procedure E-1.1, SI Termination.
- The stated correct answer was the sequence listed in distractor C.
- Distractors A and C look to be identical sequences of operation with the exception of restoring letdown in distractor A.
- In E-1.1, SI pumps are stopped before restoring letdown; therefore, distractor A is incorrect.
- Because letdown is restored seven steps after SI pumps are stopped, letdown is not germane to distractor C.
- Distractor D, which was chosen by a majority of the applicants, is the sequence of operation while operating in E-3, Tube Rupture.
- This is an incorrect answer because the stem clearly asks about E-1.1, SI Termination.
- Distractor B is simply an incorrect distractor.
- The answer choices have inconsistent use of punctuation; however, none of the 21 applicants asked for clarification of this question while taking the examination and the question did not receive comments from the facility validators.
- Notwithstanding the punctuation, by analyzing the list of equipment operations in each distractor (whether separated by a comma or semicolon), only one correct answer can be identified.
- 10CFR55.41 (b) (10) requires operators be examined on emergency operating procedures, and this question tests the applicants' knowledge of the major sequence of operations in E-1.1.
- It does not ask the applicants to assess plant conditions, to have knowledge of the EOP basis, or to make procedure transition decisions; therefore, it is testing at the RO license level.
- The contention was made that the high failure rate should be used as justification to remove the question from the exam for being too difficult.
- This is not a valid argument unless there were a technical reason for removal (the stem clearly makes an additional distractor correct that has a conflicting answer, etc.).
- After extensive validation by the licensee and review by the NRC, both NRC Region IV and the licensee agreed that this was a fair and valid question before the exam was administered.
- No additional technical justification was presented that would refute the correct answer or make another choice valid.
- The stem does not appear to be confusing when it asks for the sequence in E-1.1.
- The question is technically valid and the only correct answer is 'C': Stop one ECCS CCP; realign charging; then stop both SI pumps.
- RO QUESTION # 32
- LICENSEE COMMENT:
- A Pressurizer PORV has failed open and cannot be isolated, causing a reactor trip and safety injection.
- As the crew transitions from E-O, Reactor Trip or Safety Injection, the following conditions exist:
- PRT pressure after an initial rapid rise is now 5 psig and lowering
- Tailpipe temperature after an initial rise is now 240 F and lowering
- RCS pressure is 1800 psig and lowering Which of the following describes the expected current Containment conditions?
- A. Normal Containment parameters
- B. Rising radiation levels and temperatures
- C. Rising radiation levels; lowering temperatures after an initial rapid rise
- D. After an initial rapid rise, lowering radiation levels and temperatures
- The answer sheet shows "B" as the correct answer. 71 percent of the candidates selected "B". This question was on the Reactor Operator (RO) exam. PG&E recommends accepting answers "B" and "C" as correct.
- Justification:
- The time line as described in the stem of the question is not clear. The point in time in which applicants are being asked to predict Containment conditions is when the crew transitions from E-O. Since the transition from E-O has consistently been shown to be 5-10 minutes after a Reactor Trip/Safety Injection, the plant parameters as described would not occur in the time frame suggested.
- With Reactor Coolant discharging into containment it is reasonable to conclude that radiation levels would rise. Because of the phase "A", the sensitive (low range) Radiation Monitors are isolated and will not immediately detect the rise in radiation levels.
- Because the timeline is not well defined and the fact that containment temperature will initially rise and then quickly lower, due to the start of CFCU fans on the SI signal, and then begin to rise about 15 minutes after the rupture disk blows, PG&E recommends accepting answers "B" or "C" as the correct answer.
- NRC RESOLUTION TO QUESTION 32:
- Recommendation not accepted.
- The stem of the question asks the applicants to make a determination about containment parameters after the Pressurizer Relief Tank (PRT) rupture disc ruptures.
- The cause of the in-leakage to the PRT is a failed open Pressurizer Power Operated Relief Valve (PORV).
- The stated correct answer is B.
- After the rupture disc breaks, there would be a direct path from the primary system to the containment atmosphere.
- Because of this, radiation levels would continually rise in containment.
- Therefore, choices B and C are the only plausible choices.
- As the primary system water releases energy into containment, the long-term effect would be rising containment temperature.
- This would make B correct.
- However, the short-term effect that the Containment Fan Cooler Units (CFCU) would have on containment temperature was not considered when developing the question.
- The CFCUs would start on the Safety Injection (SI) signal which, in the stem of this question, actuates before the rupture disc breaks.
- The conditions listed in the stem indicate that the rupture disc had just ruptured, so an initial rise in temperature may have occurred.
- However, this is not conclusive based on the initial conditions and possible assumptions that can be made by the applicants concerning the timeline and size of the leak (i.e., how much was the PORV open).
- The stem of the question does not provide adequate information about what time after the PRT rupture the applicants should focus.
- If the applicants assume immediately after, then temperatures would still be lowering.
- However, shortly after, there may be a short leveling off or even a small rise in temperature before temperature would continue lowering until equilibrium is reached.
- Because the simulator is not an engineering model and the exact conditions given in the stem are not reproducible, the NRC is not analyzing its data for determining exactly what actually will happen.
- It is conclusive, however, that the long-term effect would be a rise in containment temperature.
- Additionally, the term 'rapidly' in distractor C is subjective and could be argued correct for a small rise in temperature during a relatively short period of time.
- Based on this, there is the possibility of three conflicting outcomes:
- 1) the temperature goes up; 2) the temperature goes down; and 3) the temperature rises and then goes down.
- Additionally, the conditions in the stem may have been unrealistic to determine any answer; therefore, the question will be deleted from the exam.
- RO QUESTION # 72
- LICENSEE COMMENT:
- Unit 1 is at full power.
- Power is lost to CCW radiation monitor,
- RM-17A and PK11-22, Rad Mon Sys Failure/CVI Bypass, alarms.
- What action, if any, must be taken by the crew to maintain adequate sample flow through
- RM-17B?
- A. Verify CCW heat exchanger 1-1 is in service.
- B. Verify CCW heat exchanger 1-2 is in service.
- C. Place a second CCW heat exchanger in service.
- D. No action required, there is adequate flow through
- RM-17B with either heat exchanger in service.
- The answer sheet shows "B" as the correct answer. 29 percent of the candidates selected "B".
- This question is outside the target level of difficulty range of 70 to 90 percent.
- This question was on the Reactor Operator (RO) exam.
- PG&E recommends accepting answers "B" and "D" as correct.
- Justification:
- OVID 1 06714 sheet 2 shows that radiation monitors
- RE 17A and
- RE 17B on the inlet side of the heat exchangers.
- There is a crosstie header at the discharger of each pump that is normally lined up so that either pump can provide flow through either heat exchanger.
- With this (normal)
lineup there is flow through both
- RE-17A and
- RE-17B regardless of the pump and/or heat exchanger in service.
- The sample flow rate through either RE is not dependent on which heat exchanger is in service. Adequate sample flow implies that flow rate changes.
- The flow rate does not change when the other heat exchanger is placed into service.
- There is a delayed response in the opposite RM.
- A better choice of words for the question would have been "valid sample flow."
- Many of the operators are former non-licensed operators.
- Part of their rounds is to check flow through
- RM-17A and
- RM-17B.
- This could easily lead them to believe no action was necessary because there is always flow through both radiation monitors.
- The question required applicants to recall a step from an alarm response procedure.
- Since this action is neither an immediate operator action nor frequently performed action, recall from memory is not required.
- This question does not discriminate and has operational validity because, in real life, an operator would not be expected to have memorized the alarm response procedure.
- The question has low discrimination validity.
- Of the candidates that scored 90 percent or above, only 33 percent (2 of 6) chose "B". When comparing candidates who scored above or below 85 percent, eight candidates scoring 85 percent or above chose "D", while seven candidates scoring less than 85 percent chose "D".
- An equal amount (3) of candidates (scoring above and below 85 percent) chose "B".
- Additional information:
- Answer "B" is a step from an alarm response procedure that can be selected as the correct answer. Alarm procedure AR
- PK 11-22 step 2.3.2 states, "IF
- RM-17A
failure is indicated, THEN verify CCW Hx 1-2 in service. This action is taken because of the CCW mass transient time delay affecting RE response time.
- The concept of CCW mass transient time delay affecting RE alarm and reset is not in the initial license program lesson guides.
- The information was taken from the System Training Guides
(STG) which is not part of the initial license training program.
- NRC RESOLUTION TO QUESTION 72:
- Recommendation not accepted.
- The stem of the question asks the applicant to determine what action must be taken to maintain adequate sample flow through
- RM-17B (RE-17B) after Component Cooling Water (CCW) radiation monitor
- RM-17A (RE-17A) loses power and
- PK11-22 alarms.
- The stated correct answer is B: verify CCW heat exchanger 1-2 is in service.
- There are three CCW pumps.
- The discharge of each pump splits into two headers, A and B.
- The pumps can be isolated from each header by normally-sealed-open cross-connect valves (1-15 through 1-20).
- The discharge of each cross-connect valve is connected to a check valve which prevents back-flow between headers A and B (1-601 through 1-603 and 1-607 through
- 1-609).
- A small portion of the flow from headers A and B is diverted back to the suction lines, and this is where
- RE-17A (from header A) and
- RE-17B (from header B) obtain their samples.
- When the heat exchanger associated with a header is isolated, flow in that header is significantly reduced.
- Therefore, the flow through that header's
- RE-17 radiation monitor does not represent the bulk flow through the CCW system until a certain time delay (approximately seven minutes).
- The flow rate through the RE, however, did not change.
- Based on the stem of the question and the given choices, distractor 'D' could be considered correct.
- However, this would only be true if station operating procedures were not considered when choosing 'D'.
- While it is true that adequate flow rate exists through the RE, station procedures take into account the time delay associated with the header having an isolated heat exchanger.
- Therefore, station procedures caution against having the wrong heat exchanger in service.
- They also direct (in the alarm response procedure) verifying the correct heat exchanger in service under the condition given in the stem.
- The argument was made by the facility licensee that the question was too hard, requiring knowledge deep within an alarm response procedure (making it minutia knowledge).
- The statement was also made that the licensee does not train on the details contained in the stem of the question (time delay associated with the CCW header and isolated heat exchangers).
- In accordance with 10
- CFR 55.41 (b) (10), applicants are required to have knowledge of normal operating procedures at the facility.
- This specifically includes the precautions and limitations associated with operating their systems.
- The following is precaution 5.4 from OP E-5:IV, "Auxiliary Saltwater System - Changing Over Pump and Heat Exchanger Trains:"
- If
- RE-17A or
- RE-17B is out of service when swapping HXs, note that its response time to CCW System in-leakage will change drastically.
- Due to CCW pump discharge piping configuration, only the
- RE-17 monitor associated with the in-service CCW heat exchanger monitors flow representative of the bulk system.
- When placing a CCW heat exchanger in service, the
- RE-17 monitor for that flow path should also be in service.
- Based on this, it is reasonable to assume applicants should have known about this specific feature of operating the CCW system, regardless of what step in the alarm response procedure the question referenced.
- Therefore, based on station procedural compliance, performing
- NO ACTION is not correct, making 'D' incorrect.
- The stem of the question was specifically asked this way to make 'D' seem plausible; thereby, testing whether or not the applicants truly understood procedural requirements.
- Additionally, based on the above reference, 'A' is also incorrect.
- OP E-5:II, "Auxiliary Saltwater System -Two CCW Heat Exchanger Operation," step 2.1 lists three reasons to operate with two heat exchangers in service.
- None of the three reasons were mentioned in the stem; therefore, distractor 'C' (placing a second heat exchanger in service) would be an incorrect answer.
- The question and answer are technically valid and the only correct answer is 'B':
- Verify heat exchanger 1-2 is in service.