IR 05000282/2010301
| ML101230385 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/30/2010 |
| From: | Hironori Peterson Operations Branch III |
| To: | Schimmel M A Northern States Power Co |
| References | |
| 50-282/10-301, 50-306/10-301 50-282/10-301, 50-306/10-301 | |
| Download: ML101230385 (19) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE ROAD, SUITE 210 LISLE, IL 60532-4352 April 30, 2010
Mr. Mark Site Vice President
Prairie Island Nuclear Generating Plant
Northern States Power Company, Minnesota
1717 Wakonade Drive East
Welch, MN 55089
SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC INITIAL LICENSE EXAMINATION REPORT 05000282/2010301(DRS);
Dear Mr. Schimmel:
On March 25, 2010, Nuclear Regulatory Commission (NRC) examiners completed initial
operator licensing examination process at your Prairie Island Nuclear Generating Plant.
The enclosed report documents the results of the examination. A debrief to discuss preliminary
examination observations and findings was held on March 19, 2010, with you and other
members of your staff. An exit meeting was conducted by telephone on March 25, 2010, between Mr. J. Sternisha of your staff and Mr. C. Zoia, Chief Examiner, to review the resolution
of the station
=s post examination comments and the proposed final grading of the written examination for the license applicants.
The NRC examiners administered an initial license examination operating test during the week
of March 15, 2010. The written examination was administered by Prairie Island Nuclear
Generating Plant training department personnel on March 22, 2010. Five Senior Reactor
Operator and five Reactor Operator applicants were administered license examinations. The results of the examinations were finalized on April 15, 2010. All applicants passed all sections
of their respective examinations and were issued applicable operator licenses.
In accordance with 10 CFR 2.390 of the NRC's "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 Reco rds (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). We will gladly discuss any questions you have concerning this examination.
Sincerely,/RA/
Hironori Peterson, Chief Operations Branch Division of Reactor Safety
Docket Nos. 50-282; 50-306
Enclosures:
1. Operator Licensing Examination Report 05000282/2010301 (DRS); 05000306/2010301(DRS)
w/Attachment:
Supplemental Information 2. Simulation Facility Report 3. Post Examination Comments w/ NRC Resolution 4. Written Examinations and Answer Keys (SRO)
cc w/encls: Distribution via ListServ
Enclosure 1 U. S. NUCLEAR REGULATORY COMMISSION REGION III Docket Nos. 50-282; 50-306
Report No: 05000282/2010301(DRS); 05000306/2010301(DRS)
Licensee: Northern States Power Company, Minnesota
Facility: Prairie Island Nuclear Generating Plant, Units 1 and 2
Location: Welch, MN
Dates: March 15, 2010 through March 25, 2010
Examiners: C. Zoia, Operations Engineer/Chief Examiner D. McNeil, Senior Operations Engineer
B. Palagi, Senior Operations Engineer Approved by: Hironori Peterson, Chief Operations Branch
Division of Reactor Safety
Enclosure 1 1
SUMMARY OF FINDINGS
Initial License Examination Report ER 05000282/2010301(DRS); 05000306/2010301(DRS);
03/15/2010 - 03/25/2010; Northern States Power Company, Minnesota, Prairie Island Nuclear
Generating Plant.
The announced initial operator licensing examination was conducted by regional Nuclear
Regulatory Commission examiners in accordance with the guidance of NUREG-1021, A Operator Licensing Examination Standards for Power Reactors,@ Revision 9, Supplement 1.
A. Examination Summary Ten of ten applicants passed all sections of their respective examinations. Five applicants were issued Senior Operator licenses and five applicants were issued
Operator Licenses. (Section 4OA5.1)
B. Licensee-Identified Violation A violation of very low safety significance was identified by the licensee and was reviewed by the examiners. Corrective actions planned or taken by the licensee have been entered into the licensee's corrective action program. The violation and corrective action tracking numbers are listed in Section 4OA7 of this report. (Section 4OA7)
2
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA5 Other
.1 Initial Licensing Examinations
a. Examination Scope
The Prairie Island Training Department prepared the examination outline and developed
the written examination and operating test. The NRC examiners validated the proposed
examination during the week of February 22, 2010, at Prairie Island with the assistance
of members of the licensee training staff. During the on-site validation week on
February 22, 2010, the examiners audited one license application for accuracy. The
NRC examiners conducted the operating portion of the initial license examination during
the week of March 15, 2010. Members of the Prairie Island Training Department staff
administered the written examination on March 22, 2010. The NRC examiners used the
guidance established in NUREG-1021, A Operator Licensing Examination Standards for Power Reactors,@ Revision 9, Supplement 1, to prepare, validate, revise, administer, and grade the examination.
b. Findings
Written Examination The NRC examiners determined that the written ex amination, as originally submitted by the licensee, was within the range of acceptability expected for a proposed examination.
All changes made to the submitted examination were made in accordance with
NUREG-1021, "Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1. The licensee's post examination comments on the written
examination were documented in Enclos ure 3, Post Examination Comments and Resolutions.
Operating Test The NRC examiners determined that the operating test, as originally submitted by the
licensee, was within the range of acceptability expected for a proposed examination.
All changes made to the submitted examination were made in accordance with
NUREG-1021, "Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1." The licensee had no post examination comments on the
operating test.
Examination Results Ten applicants passed all sections of their examinations resulting in the issuance of five
Senior Reactor Operator and five Reactor Operator licenses.
.2 Examination Security
a. Scope
The NRC examiners reviewed and observed the licensee's implementation of examination security requirements during t he examination validation and administration to assure compliance with 10 CFR 55.49, A Integrity of Examinations and Tests.
@ The examiners used the guidelines provided in NUREG 1021 to determine acceptability of the licensee
=s examination security activities.
b. Findings
A violation of very low significance (Severity Level IV) was identified by the licensee and
was a violation of NRC requirements which met the criteria of Section VI of the NRC
Enforcement Policy for being dispositioned as an NCV. See Section 4OA7.1 for details.
4OA6 Management Meetings
.1 Debrief The chief examiner presented the examinat
ion team's preliminary observations and
findings on March 19, 2010, to Mr. M. Schimmel and other members of the Prairie Island
Nuclear Generating Plant Operations Department and Training Department staff.
.2 Exit Meeting
The chief examiner conducted an exit meeting on March 25, 2010, with Mr. J. Sternisha, Prairie Island Nuclear Generating Plant Training Manager by telephone. The NRC
=s final disposition of the station
=s post-examination comments was discussed and the revised written examination grading key was provided to Mr. Sternisha during this
telephone discussion. The examiners asked the licensee whether any of the material
used to develop or administer the examinat ion should be considered proprietary. No proprietary or sensitive information was identified during either the examination, debrief
or exit meeting.
4OA7 Licensee-Identified Violations
The following violation of very low significance (Green) was identified by the licensee
and is a violation of NRC requirements which meet the criteria of Section VI of the
NRC Enforcement Policy, NUREG-1600, for being dispositioned as Non-Cited
Violations. Cornerstone: Mitigating Systems Title 10 CFR 55.49, stated, in part, that station personnel shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or 4 examination. This included activities related to the preparation and certification of license applications and all activities related to the preparation, administration, and grading of the tests and examinations required by this part. Contrary to the
above, during the administration of the NRC written exam, a copy of the
approved answer key with a photograph of a panel was improperly used to
identify which panel lights were lit for one question. This was done in reply to a
question asked by an applicant during the ex am. Inadvertently, the copy of the photograph of the panel with associated question distractors also included a
check mark indicating the correct answer, which immediately compromised the
question.
The violation was of very low safety significance because the error was
discovered shortly after the copies were distributed to the applicants, the NRC
was immediately informed, and the compromised question was deleted from the
examination. Additionally, after deleting the compromised question, the NRC
determined that because the examination's question distribution still supported a wide and adequate variety of plant knowledge items, the examination was still considered to be a valid examination. Immediate actions taken by the licensee's
training department included entering this condition into the corrective action
program as AR 1223729. The licensee's training personnel were again briefed
concerning examination security requirements and the need to comply with
examination security procedures was stressed.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- M. Schimmel, Site Vice President
- J. Sternisha, Training Manager
- T. Ouret, General Supervisor Operations Training
- M. Peterson, Fleet General Supervisor-Simulator / NRC Examinations
- J. Sorenson, General Manager Nuclear Training
- J. Lash, Operations Manager
- M. Davis, Regulatory Affairs
NRC
- C. Zoia, Chief Examiner
- P. Zurawski, Resident Inspector
- D. Betancourt, Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
None
LIST OF ACRONYMS USED
ADAMS Agency-Wide Document
Access and Management System DRS Division of Reactor Safety
NRC Nuclear Regulatory Commission
ALARA As Low As Reasonably Achievable IR Inspection Report
SIMULATION FACILITY REPORT
Facility Licensee: Prairie Island Nuclear Generating Plant
Facility Docket No: 50-282; 50-306
Operating Tests Administered: March 15 through 19, 2010
The following documents observations made by t
he NRC examination team during the initial
operator license examination. These observations do not constitute audit or inspection
findings and are not, without further verification and review, indicative of non-compliance with
CFR 55.45(b). These observations do not affect NRC certification or approval of the
simulation facility other than to provide information which may be used in future evaluations.
No licensee action is required in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were
observed:
ITEM DESCRIPTION
Unexpected
Condenser Hotwell
Level Alarms
Unexpected condenser hotwell level alarms occurred during
Scenario 3, which could neither be explained nor eliminated by the
simulator staff. The alarms caused a significant delay for the crew
being evaluated. Due to the anticipated alarms and expected delays
when this scenario was repeated, the normal evolution for starting the
Condensate Pump was eliminated in subsequent scenarios.
Simulator Work Order (SWO) B0D-019 was written to address these
unexpected alarms.
Post Examination Comments and Resolutions
Question 2
- Given the following conditions:
- Unit 2 is at 30% power and stable.
- Control rod K7 is 15 steps lower than the other rods in control bank D.
- The decision has been made to realign control rod K7 to control bank D per 2C5
AOP5, Misaligned Rod, Stuck Rod, and/or RPI Failure or Drift.
To realign rod K-7, the crew will disconnect the lift coil(s) for:
a. the affected GROUP (except K7) and adjust the affected GROUP step counter to
the misaligned rod position.
b. the affected BANK (except K7) and determine the average RPI position for all
rods in the affected bank.
c. control rod K7 and determine the average RPI position for all rods in the affected
bank. d. control rod K7 and adjust both control Bank D step counters to the misaligned
rod position.
Answer - b.
Reference: 1C5 AOP5, Section 2.5.4
Applicant Comment
- The word "both" in distractor "d" is a misprint or typographical error.
Facility Comment
The Station recommended deleting "both" and the "s" off "counters" in distractor "d."
NRC Resolution
The NRC agreed with the station response to delete "both" and the "s" off the word
"counters" in distractor "d." It was also noted that "d," was an incorrect choice for
answering this question with or without the recommended changes. No change was
made to the answer key as a result of this examination change.
Question 28
- Given the following conditions:
- Unit 1 is at 100% power
- C47017, 11 STM GEN LO-LO LVL Reactor Trip, First out annunciator is LIT.
The required crew response is to . . .
a. initiate a manual Safety Injection and enter 1E-0.
b. manually insert control rods if power is grater than 5%.
c. manually open the FRV to feed 11 S/G back into normal band.
d. verify S/G levels are below the reactor trip setpoints, THEN manually trip the
reactor.
Answer - d.
Reference: FP-OP-COO-01, 1E-0
Applicant Comment
Distractor "d." should be asking to verify 11 SG level is below reactor trip setpoint. The
way the distractor is worded it makes it sound like you need both SG levels to be low in
order to trip the reactor.
Facility Comment
The station recommends that distractor "d." wording be changed to "11 S/G level is" vice
"S/G levels are" to make the distractor technically accurate.
NRC Response
The NRC agreed with the facility's proposed change to distractor "d." The distractor, as
written, appeared to require the operator verify both S/G levels were below reactor trip
setpoint, before manually tripping the reactor. The correct action was to trip the reactor
with either S/G below the reactor trip setpoints. Since the question referred to 11 S/G in
the stem, editing distractor "d." to read "verify 11 S/G levels-" from "verify S/G levels-"
was correct. No change was made to the answer key as a result of this examination
change.
Question 35
- Given the following conditions: (A photograph of panel controls for Unit 1 Air Ejectors was
provided to the applicants).
- The conditions in the above photograph are seen on the control board.
- A Unit 2 startup is in progress.
- Condenser vacuum is being established.
- Condenser vacuum is 21 in. Hg.
What operator action (if any) is required and why?
a. No action is required until vacuum reaches 24.5 in. Hg.
b. No action is required, vacuum is established with the given conditions.
c. Place Normal Service First Stage Jets in service with the given conditions.
d. Place Normal Service First Stage Jets in service to finish drawing vacuum.
Answer - d.
Reference: C26, 2C1.2
Facility Comment
During administration of the examination, an applicant asked for clarification on which
lights are lit. The photograph provided did not have sufficient clarity to determine which
lights were illuminated and which lights were extinguished. The facility proctor provided
the applicants with a revised photograph which included circles around the lights that
were lit. It was then discovered that the revised photograph given to students included a
check mark next to the correct answer. The facility recommend deleting question
because the correct answer was inadvertently disclosed to some of the applicants.
NRC Response
The NRC agreed with the facility personnel that additional clarification was required to
distinguish which controller lights were illuminated and which controller lights were
extinguished. The NRC also agreed that the question should be deleted from the
examination because the question answer was compromised. Because the number of
applicants that saw the question answer could not be determined, the question cannot
be deleted for only the applicants that saw the answer; it had to be deleted for all of the
applicants. Because a compromise of examination material occurred, the NRC issued a
Non-Cited Violation (NCV) in accordance with 10 CFR 55.49, "Integrity of examinations
and tests." The answer key was modified to remove question 35 from the answer key.
Question 49
- Given the following conditions:
- Unit 1 is at 100% power.
- 11 Steam Generator tube rupture occurs.
- The Instrument Air header is depressurized.
- 1E-3, Steam Generator Tube Rupture, is in progress.
The RCS cooldown initiate due to the opening of the . . .
a. Condenser Steam Dump.
b. Atmospheric Steam Dumps.
c. Steam Generator PORVs.
d. Steam Generator Safety Valves.
Answer - c.
Reference: 1E-3
Facility Comment
The facility recommends that the typographical error "iniate" in the stem of the question
be changes to "initiates."
One applicant contended that the question stem was unclear as to whether the
cooldown would be from a manual or automatic action. The applicant contended that
the S/G PORVs are fail closed valves per P8174L-001. The applicant further contended
that MSIVs also fail closed so distractors "a." "b." and "c." are isolated and will not auto
open to begin a plant cooldown. While the S/G PORV has an accumulator, the MSIVs
also do and plant OE shows that on a loss of air the MSIVs still fail closed. The only
entirely correct answer is "d." since the loss of air will not affect the safety from opening.
At a minimum the stem should clarify how the cooldown will be initiated, automatically or
by operator control. Also refer to logic NF-40322-3 which shows S/G PORV fails closed
and NF-40322-1 which shows MSIVs fail closed and NF-40322-2 which show steam
dumps fail closed.
NRC Response
The NRC agreed to add an "s" to "initiate" in order to make the question stem read
correctly. The NRC also agreed with the applicant's contention that the stem did not
clearly state whether the cooldown was from manual or automatic action. However, the
NRC determined that it did not matter whether the cooldown was conducted manually or
allowed to occur automatically. Either manual or automatic action would result in the
Steam Generator PORV being the initial source of the cooldown. The Steam Generator
PORV would initially automatically open due
to its accumulator. The cooldown would
then be manually controlled per E-3 Step 7, local operation of the PORV.
The answer key was not modified in response to this typographical error correction, nor
in response to the applicant's contention that the stem was unclear.
Question 81
- Given the following conditions:
- Unit 1 is at 50% power following a refueling outage.
- 47012-0601, RCP OIL RESERVOIR HI/LO LVL, is in alarm.
- 11 RCP Upper Thrust Bearing temperature on recorder 1TR-2001 is LIT.
- 11 RCP Upper Thrust Bearing temperat
ure is currently reading 180°F and slowly
rising. - 11 RCP seal injection flow is 6 gpm.
- 11 RCP No. 1 seal leakoff is 1.2 gpm.
What action is required?
a. Perform an emergency containment entry to add oil to 11 RCP per F2, Radiation
Safety.
b. Initiate a controlled shutdown per 1C1.4, Unit 1 power Operation. When the
reactor is shutdown, stop 11 RCP and close the associated spray valve.
c. Lower Component Cooling system temperature to minimum per 1C14, Component Cooling System - Unit 1.
d. Trip Unit 1 Reactor and enter 1E-0, Reactor Trip or Safety Injection. When the reactor trip is verified, stop 11 RCP and close associated spray valve.
Answer - b.
Reference: C47012-0601 Annunciator Response
Applicant Comment
One applicant contended that per ARP 47012 for alarm 47012-0601, the correct
response should be to monitor RCP 11 bearing temperatures and vibrations, to contact
I&C to determine which reservoir is alarming, and then check conditions locally when
conditions permit, and repair if possible. The applicant stated that PINGP has a history
of having to add oil to the RCPs at power, to the extent that a modification was installed
to allow oil to be added to the upper and lower RCP reservoirs from outside the RCP
vaults. The applicant referred to CAP 395684. The applicant stated that an emergency
containment entry is defined as "-as an entry which is not controlled by the Radiation
Protection Group," and is a "-non-routine entry for inspection or operation such as a fire
alarm or limit switch position check. He further asserted that if ARP C47012-0601 was
followed, an emergency containment entry would be made to validate the condition while
monitoring RCP bearing temperatures and vibrations. The ARP assumes that bearing
temperatures remain below 200°F during the entry. Once it is determined that an oil
reservoir level is low, oil would be added under a work order, still as an emergency
containment entry. The applicant contends that by following this line of reasoning,
answer "a." would be correct.
Another applicant contended that distractor "c." was the correct answer. The applicant
stated that although there was not a step in ARP 47012-0601 to lower CC temperatures, the first action was to monitor bearing temperatures. Temperatures that were higher
than normal would require operators to look at the cooling medium (CC) and evaluate if
adjustments were needed. Per procedure 1C14, CC was maintained between 80°F and
105°
- F. From the above, the applicant believed it would be expected that operators
would consider lowering CC temperature per distractor "c.," to control bearing
temperature while preparing for the remaining actions of the AR
therefore, contended the remaining actions would consist of the actions found in
distractor "a.," to check the oil reservoir status and correction. The applicant maintained
that answers "b.," and/or "d." would be correct if bearing conditions continued to
degrade.
Facility Follow-up Comment
The station agreed with the with the first applicant's comment and recommend accepting
distractors "a." and "b." as correct answers. The facility disagreed with the second
applicant's comment as there is no reference within ARP 47012-0601 to adjust
Component Cooling (CC) temperatures. Per procedure 1C14, normal operation of the
Component Cooling system maintains sy
stem temperature between 80°F-105°F.
However, a CC system temperature rise is not occurring in the question and no
adjustment is necessary to CC system tem
perature. The facility recommends accepting
answers "a." and "b." based on the above comments.
NRC Response
The NRC disagreed with the station response recommending both distractors "a." and
"b." be considered correct. The argument for considering "a." to be correct assumed that
it was necessary to perform an emergency containment entry to add oil to investigate
and repair the RCP. The applicant pointed out that adding oil to the RCPs occurred with
such regularity that a plant modification was installed to allow oil addition with the plant
at power. The NRC determined that such containment entries to add oil were not
conducted as emergency containment entries. Because distractor "a." denoted the need
to invoke an emergency containment entry, it was an incorrect distractor. Therefore,
distractor "a." was considered to be incorrect. The NRC disagreed with the applicant
that contended distractor "c." was correct. The NRC agreed with the station response to
disallow distractor "c." as a correct answer because ARP 47012-0601 did not reference
adjusting CC temperatures and a CC temperature rise was not specified in the stem of
the question. The applicant would have needed to assume that CC temperatures were
high out of their normal band to see a need to lower CC temperature. Since the
question did not reference CC temperatures, the applicant cannot assume the CC
temperatures were outside their normal temperature band. NUREG 1021, Appendix E,
Part B.7, which was read to the applicants prior to administering the exam states: "When
answering a question, do not make assumptions that are not specified in the question-"
For the reasons specified above, distractors "a." and "c." are considered incorrect. The
answer key was not modified; distractor "b." was retained as the only correct answer.
Question 86
- Given the following conditions:
- Unit 1 is at 100% power.
- Voltage on 4.16KV Safeguards Bus 16 is 3955 volts.
After _____ seconds, D2 Diesel Generator will auto start and load shedding will be initiated on
4.16KV Safeguards Bus 16.
AFTER grid voltage recovers, the Shift Supervisor will direct performance of _________ to
respond to this event.
a. 8 1C20.5, Unit 1 - 4.16KV System
b. 60
1C20.5, Unit 1 - 4.16KV System
c. 8
1C20.5 AOP2, Reenergizing 4.16KV Bus 16
d. 60
1C20.5 AOP2, Reenergizing 4.16KV Bus 16
Answer - b.
Reference: B20.5; 1C20.5, C47024-0304
Facility Comment
The facility determined that there was no correct answer provided to this question. After
post-examination review, it was determined that no section of procedure 1C20.5 results
in a transfer of Bus 16 back to CT11 from D2 - the procedure for this transfer is found in
1C20.7. Additionally, 1C20.5 AOP2 is only used if the bus is de-energized. This makes
distractors "a." "b." "c." and "d." incorrect answers. The facility recommended deleting
this question from the examination because no correct answer was provided in the
distractors.
NRC Response
The NRC reviewed 1C20.5 and found no section of the procedure that the SRO would
direct to return Bus 16 to CT11 from D2. This eliminated distractors "a." and "b." as
correct answers. Bus 16 was not de-energized as part of the question stem and
question conditions. Because 1C20.5 AOP2 was only performed if Bus 16 was
de-energized, distractors "c." and "d." were also incorrect. Because none of the
distractors matched the correct answer (Use of procedure 1C20.7), there was no correct
answer provided for this question. The answer key was modified to delete this question
from the examination.
WRITTEN EXAMINATIONS AND ANSWER KEYS (SRO)
SRO Initial Examination ADAMS Accession # ML101130329
M. Schimmel
-2-
We will gladly discuss any questions you
have concerning this examination.
Sincerely, /RA/
Hironori Peterson, Chief
Operations Branch
Division of Reactor Safety
Docket Nos. 50-282; 50-306
Enclosures: 1. Operator Licensing Examination
Report 05000282/2010301 (DRS); 05000306/2010301(DRS)
w/Attachment: Supplemental Information 2. Simulation Facility Report
3. Post Examination Comments w/ NRC Resolution
4. Written Examinations and Answer Keys (SRO)
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