ML071230743
| ML071230743 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/03/2007 |
| From: | Valos N A NRC/RGN-III/DRS/OLB |
| To: | |
| Shared Package | |
| ML070660508 | List: |
| References | |
| 50-263/07-301 | |
| Download: ML071230743 (14) | |
Text
FACILITY POST-EXAMINATION COMMENTS FOR THE MONTICELLO INITIAL EXAM - FEBRUARY 2007 Monticello Nuclear Generating Plant Operated by Nuclear Management Company, LLC Cornmined to Nuclear Excellence February 26,2007 L-MT-07-011 IO CFR 55.40 Mr. Nick Valos, Region Ill Addressee Only U.S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Monticello Nuclear Generating Plant Docket 50-263 License No. DPR-22 Operator License Post-Examination Documentation Reference 1 : NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9 Pursuant to the requirements of Reference 1, Section ES-501, "Initial Post-Examination Activities," Nuclear Management Company, LLC (NMC) is providing the NRC with the required post-examination documentation. The post-examination documentation was developed as a result of the completion of the NMC grading and reviews of the Reactor Operator (RO) and Senior Reactor Operator (SRO) License written examinations completed on February 19, 2007, at the Monticello Nuclear Generating Plant (MNGP). The following information is being provided in this letter: Applicants Comments and Facility Responses.
Written Examination Performance Analysis, Graded Written Exam Cover Sheets and Applicants Answer Sheets, Questions and Answers Given during the Written Exam, Written Exam Seating Chart, Completed NRC form ES-403-1 and Posted Exam Clarifications during the Exam. The examination grading and review was performed in accordance with Reference 1 Section ES-403, and "Grading Initial Site-Specific Written Examinations." Form ES-201-3 is not fully completed at this time.
NMC will submit the completed original form ES-201-3, "Examination Security Agreement,
in accordance with Reference 1, Section ES-501.
2807 West County Road 75 Monticello, Minnesota 55362-9637 Telephone: 763-295-51 51 Fax: 763-295-1454 USNRC Page 2 This letter makes the following new commitment:
NMC will submit the completed original form ES-201-3, "Examination Security Agreement,"
in accordance with Reference 1, Section ES-501.
Generating Plant Nuclear Management Company, LLC Enclosures (7) cc: Hironori Peterson, USNRC, Region Ill (w/o enclosures)
Enclosure 1 - Applicant's Comments and Facility Responses Enclosure 2 - Written Examination Performance Analysis Enclosure 3 - Graded Written Exam Cover Sheet and Applicants Answer Sheet Enclosure 4 - Questions and Answers Given during the Written Exam Enclosure 5 -Written Exam Seating Chart Enclosure 6 - Completed NRC form ES-403-1 Enclosure 7 - Posted Exam Clarifications during the Exam Summary of Applicant Comments on License Written Exam Administered on 02/19/07 Question ##4: Comment: Spray pumps is approximately 300 psig. The AC interlock is at 100 psig. The stem of the question states both pumps are in service. Both CS pumps must be below 100 psig to clear the alarm. Even with severe cavitation, I don't believe that both CS pumps would drop in discharge pressure at the same time to clear the alarm. Even though the C.4 basis states that this may be an indication of ECCS suction plugging, this statement assumes only one pump is running. That is NOT the case in the stem. Step 1.f states that the pump motor amperage would be erratic or decreasing for plugging strainer.
Therefore, answer A is the most correct. Answer A should also be accepted.
The discharge head of the Core Facility response:
of cavitation include fluctuations in discharge pressure and motor current. This is supported in M-8120L-114 (Fluid Statics and Dynamics) which is part of the ILT Generic Fundamentals Course. Answer "A, "Steadily lowering of Core Spray pump amps", may be an indication of suction plugging but is not an indication of pump cavitation; the question asks for indication of pump cavitation. The question grading for the exam should not change.
The effects Page 1 of 11 Question #I 7: Comment: Answer D should also be accepted as correct.
Ops Manual B.05.07-01 states that the steam and feedwater flow signals lose their accuracy below 30% power. Also, on page 7 it states that a separate single element control scheme is used when reactor power is less than 20 % of rated. The question ask per design limits of the water level control system, when can DFCS be placed in 3 element control.
The 3 element would work at > 20% feed flow and one FRV in service. Although the C.l start up procedure places the DFCS in 3 element control at approximately 40% power and after the second FRV is in service. The design of DFCS allows it to be placed in service with feedwater
> 20% per it's design. Facility response:
Per Ops Man B.05.07-01, Reactor Level Control, "Steam flow and feedwater flow signals (used in three-element control) lose their accuracy below 30% power and, therefore, become less desirable as controlling inputs". Answer "C" (exam correct answer) is the only choice describing an action above 30% power. The conditions cited by the applicant "with feedwater
> 20%" are for transition from the FW Low Flow Reg Valve to the Main Reg Valves. The question grading for the exam should not change.
Page 2 of 11 Question #20: Comment: A clarification was made to question 20 which changes the acceptable answer. Without the clarification, Answer D would be correct since 17 and 18 buses use AC control power and are not affected by loss of DC control power. However, the proctor stated to add "Control Room Control Switch" to the stem of the question.
Since there are not bus 17 or bus 18 control switches in the control room, there would be not control switches that could operate 4 kv breakers in the control room due to the loss of all DC control power. This would make answer A the only correct answer. Based upon whether a student already completed the question or did not apply the additional verbal clarification, then both answer A and D should be accepted. Facility response:
both "A and "D" responses. This is due to the information given during exam implementation, as stated above, that changed the intent of the question.
This question is correct, as written, and does not require any changes prior to incorporation into the exam bank. The question grading for the exam should be changed to accept Page 3 of 11 Question #31: Comment: to be placed in service after bypassing the 2/3 core height interlock and the Containment Spray/Cooling LPCl initiation switch. Although the DW spray procedure (Part C of 3502) states that this should be done when in the SAMGs, the logic would allow Answer D to be correct. I assumed other pumps (like HPCl and CS) were available to recover level. The basis does state DW spray and injection can be alternated to provided adequate cooling. Therefore, answer D also correctly answers the questions.
Answer D should also be accepted. The RHR logic allows drywell sprays Facility response:
procedures, other than SAMGs, do not allow bypassing of the interlock for DW spray; the question states "The SAMG's have not been entered". Although system design provides the means to bypass the interlock with the given conditions, utilization is clearly not allowed by site procedure.
The question grading for the exam should not change. Site Page 4 of 11 Question #46: Comment: Answer C should also be accepted.
Loss of the instrument air will have no effect on the valves. Therefore, the first part of the question is true. As to the alignment, instrument air is normally aligned to the system, but pressure set points have this valve closed (CV-7478).
It would automatically open if a problem exists with the Instrument N2 system. The second part of answer C states that nitrogen is used as a back up supply. This is true. Alternate N2 is used as the backup supply for these valves. Therefore, answer C is technically correct also.
Facility response:
As stated in the applicant comment above, CV-7478, in it's normal plant operating status, blocks instrument air from the supply line leading to the Vacuum Relief Dampers; therefore, the Vacuum Relief Dampers are NOT "normally aligned to air" as distractor "C" states. Answer "D" is the only correct answer. The question grading for the exam should not change. Page 5 of 11 Question #59: Comment: inches. By typical engineering convention atmospheric pressure is 30 inches. Therefore, absolute pressure in the condenser would be 6 inches, which would place you in the alert region of Figure 1 at time 0705 requiring a scram 20 minutes later. Although the C.4 procedure states that the operator should us PR-1264 when SPDS is not available scramming at 0729 would be 21 minutes later which would not be in accordance with the C.4 procedure.
Answer 6 should also be accepted.
7-6-16 and 7-B-17 alarm at 24 Facility response:
The referenced procedure, C.4-6.06.03.A (Decreasing Condenser Vacuum), under "INDICATIONS", lists the 7-B-16/17 alarm set-points of 24" Hg Vac as being equivalent to - 5 " Hg Abs on the PR-1264; this procedure would override the stated "typical engineering convention". The procedure also requires a reactor scram when in the ALERT region for longer than 20 minutes, therefore scramming at T+21 minutes is in accordance with the C.4 procedure. The question grading for the exam should not change.
Page 6 of 11 Question #62: Comment: Answer D should also be accepted as correct.
Ops Manual B.01.03-05 H.10 is used for an inadvertent control rod withdrawalhod drift out which is the condition in the stem. It states that if annunciator 5-A-I9 is not in alarm (which would be the case iaw the stem), than immediately insert the rod to 00 using normal or emergency insert. Facility response:
The question grading for the exam should not change. The question asked for the required action following a control rod mispositioning event. The correct answer ("Cy') is to "Insert the control rod to its previous position without obtaining information on fuel thermal limits"; this is per site procedures. The applicants missed this question for various reasons, not recognizing the procedural guidance for a control rod inadvertently withdrawn due to operator error and, specifically, for expediting return of the rod to its original position. Distractor "D" would not be a correct action response to the given event; with no equipment problems, there is no reason to insert the control rod further than its original position.
A note in the same procedure step referenced in the applicant comment states "In this case the timer malfunction circuit is deficient"; this reinforces that this is the incorrect procedure for this event. CAP 01078752 has been initiated to reinforce practical training on this type of event and to clarify procedural guidance. It is recommended that the wording for the correct answer be revised to state: Insert the control rod to its previous position prior to obtaining information on fuel thermal limits.
The current wording could be implied to mean that fuel thermal limits would not be checked even after rod re-positioning.
Page 7 of 11 Question #70: Comment: There is no correct answer. Answer B is a Reactor Core safety limit (or conditions) at which MCPR limits do not apply.
It is NOT the MCPR limit. The MCPR limits are 1 .IO for two loop ops or 1 .I2 for single loops ops. The first two distractors only identify when MCPR may or may not apply. The question specifically asked which of the following describes the MCPR safety limits. Not when the MCPR limits do not apply. I thought the values were reversed in my chosen answer, I went with it, because the first two did not answer the question. Facility response:
The question grading for the exam should not change. The Bases for the Reactor Core SLs, BACKGROUND, states that the limits of 2.1.1.1, Fuel Cladding Integrity, provide "a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1 .OO)". In the APPLICABLE SAFETY ANALYSES of Tech Spec Section B2.1.1.1 Reactor Core SLs, the Bases for Safety Limits, the following is stated: The fuel cladding must not sustain damage as a result of normal operation and AOOs. The reactor core SLs are established to preclude violation of the fuel design criterion that a MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling. The argument in the applicant comment contradicts these statements in the Bases.
Page 8 of 11 Question #79: Comment: Per Fleet Tagging (FP-OP-TAG-OI), the "note" in section 3.7 states that "Ops Shift Supervision" may be carried out by an SRO or persons designated by Ops Management in the WCC acting for Ops shift supervision, providing the operating crew is kept informed and involved in decisions, when necessary. Ops Shift Supervision and the WCC should be considered to be the same.
WCC personnel hours are different than the shift schedule and the WCC person would be there at some time during the shift. Since, in answer "D", I ask for the WCC permission, this tells me that WCC person was available to perform the task. For normal day to day operation at the plant, as a CRS, I would direct the WCC person to perform the necessary actions and have the tags removed; at a minimum, since most work starts and ends with the WCC, I would seek their concurrence and/or permission. Facility response:
The question grading for the exam should not change.
The question specifically asks for actions required for the CRS to direct removal of the tags and place the unit in service. Although the WCC may also perform these actions, and the WCC may be communicated with prior to the CRS performing these actions, distractor "D" is incorrect as the permission of the WCC is not required.
Page 9 of 11 Question #87: Comment: The "A' answer is the most correct, since there is procedural direction to refill the SBLC tank with boron.
The "B" answer is not correct, because Cold S/D Boron weight is required to depressurize. The "C" answer is not correct, because securing SBLC pump when reactor level indicates "0" inches is wrong. Securing SBLC pump has nothing to do with Reactor Level. The "D" answer is not correct, because Hot S/D Boron weight is required to raise level and Securing SLBC pump has nothing to do with Reactor Level at '0" inches. Facility response: The question should be removed from the exam due to there being no correct answer as worded. The correct answer, as stated on the exam, states I'. . . secure SBLC pump when level indicates 0 inches". This is not specific as to which "level" indicates 0 inches. The SBLC level instruments are in units of gallons, not inches. These inconsistencies could lead the applicants to assume that the level indication is referencing reactor water level. In this case, there is no correct answer.
The question should be revised, such that the correct answer states "...when SBLC tank level indicates 0 gallons", prior to inclusion to INPO Exam Bank. PageIOof11 Summary: Results of the applicant exam feedback response identified one question
(#20) for consideration to accept two answers. Results also identified one question (#87) for consideration to be removed from the exam and the same question having a wording change prior to incorporation to the INPO Exam Bank. One question (#62) is suggested for a wording change prior to incorporation to the INPO Exam Bank. No changes are needed due to the remaining feedback items.
Prepared by: .&ZJd 2/23/0 7 Olaf 61son Supervisor bperatons Training - Initial Reviewed by: Training Shawn Halbert Training Manager Page 11 of 11