IR 05000305/2005301
| ML060120466 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 01/10/2006 |
| From: | Hironori Peterson NRC/RGN-III/DRS/OLB |
| To: | Christian D Dominion Energy Kewaunee |
| References | |
| 50-305/05-301 50-305/05-301 | |
| Download: ML060120466 (28) | |
Text
ary 10, 2006
SUBJECT:
KEWAUNEE NUCLEAR POWER PLANT NRC INITIAL LICENSE EXAMINATION REPORT NO. 05000305/2005301(DRS)
Dear Mr. Christian:
On November 18, 2005, the NRC completed initial operator licensing examinations at your Kewaunee Nuclear Power Plant. The enclosed report documents the results of the examination which were discussed on November 18 and December 8, 2005, with Mr. M. Gaffney and Mr. S. Johnson, respectively, and with other members of your staff.
NRC examiners administered the operating test during the week of November 14, 2005.
Members of the Kewaunee Nuclear Power Plant Training Department staff administered the written examination on November 18, 2005. One Reactor Operator (RO) and six Senior Reactor Operator (SRO) applicants were administered license examinations. The results of the examinations were finalized on December 22, 2005. Five applicants passed all sections of their examinations, four of these applicants were issued respective operator or senior operator licenses. Two SRO applicants failed the written examination and will not be issued licenses.
One applicant scored less than 82 percent on the written examination; and, in accordance with the guidelines of NUREG 1021, Operator Licensing Examination Standards for Power Reactors, ES-501.D.3.c, his license will be withheld until any appeal rights of the failed applicants are exhausted.
In accordance with 10 CFR Part 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). We will gladly discuss any questions you have concerning this examination.
Sincerely,
/RA/
Hironori Peterson, Chief Operations Branch Division of Reactor Safety Docket No. 50-305 License No. DPR-43
Enclosures:
1. Operator Licensing Examination Report 050000255/2005301(DRS)
2. Simulation Facility Report 3. Post Examination Comments and Resolutions 4. Written Examinations and Answer Keys (RO & SRO)
REGION III==
Docket No.: 50-305 License No.: DPR-43 Report No.: 05000305/2005301(DRS)
Licensee: Dominion Energy Kewaunee, Inc.
Facility: Kewaunee Nuclear Power Plant Location: N490 Highway 42 Kewaunee, WI 54216 Dates: November 14 through 18, 2005 Examiners: B. Palagi, Chief Examiner C. Phillips, Examiner D. Reeser, Examiner Approved by: H. Peterson, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000305/2005301(DRS); 12/14/2005-12/18/2005; Kewaunee Nuclear Power Plant; Initial
License Examination Report.
The announced operator licensing initial examination was conducted by regional examiners in accordance with the guidance of NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9.
Examination Summary:
- Seven examinations were administered (one Reactor Operator and six Senior Reactor Operator).
- Five applicants passed all sections of their examinations, four of these applicants were issued respective operator or senior operator licenses. Two SRO applicants failed the written examination and were not issued licenses. One applicant scored less than 82 percent on the written examination; and, in accordance with the guidelines of NUREG 1021, Operator Licensing Examination Standards for Power Reactors,
ES-501.D.3.c, his license will be withheld until any appeal rights of the failed applicants are exhausted.
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA5 Other
.1 Initial Licensing Examinations
a. Examination Scope
The NRC examiners conducted an announced initial operator licensing examination during the week of November 14, 2005. The licensee used the guidance established in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, to prepare the examination outline and to develop the written examination and operating test. The NRC examiners administered the operating test November 14 through 17, 2005. Members of the Kewaunee Power Station Training Department administered the written examination on November 18, 2005. One Reactor Operator (RO) and six Senior Reactor Operator (SRO) applicants were examined.
b. Findings
Written Examination The licensee developed the written examination. During their internal review, the NRC examiners determined that the examination, as submitted, was within the range of acceptability expected for a proposed examination. Written examination comments developed during review by the NRC staff, and as a result of examination validation were incorporated into the written examination in accordance with the guidance contained in NUREG-1021.
A total of nine post-examination comments (6 RO; 3 SRO comments) were submitted by the applicants and station training department personnel on November 28, 2005. The results of the NRCs review of the comments are documented in Attachment 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 for a proposed examination. The examiners validated the operating test during the validation week and replaced or modified several items in the proposed operating test. Test changes, agreed upon between the NRC and the licensee, were made in accordance with NUREG-1021 guidelines.
Examination Results Five applicants passed all sections of their examinations, four of these applicants were issued respective operator or senior operator licenses. Two SRO applicants failed the written examination and were not be issued a licenses. One applicant scored less than 82 percent on the written examination; and, in accordance with the guidelines of NUREG 1021, Operator Licensing Examination Standards for Power Reactors, ES-501.D.3.c, his licenses will be withheld until any appeal rights of the failed applicants are exhausted.
.2 Examination Security
a. Inspection Scope
The NRC examiners briefed the facility contact on the NRCs requirements and guidelines related to examination physical security (e.g., access restrictions and simulator considerations). The examiners observed the implementation of examination security and integrity measures (e.g., security agreements) throughout the examination process.
b. Findings
No findings were noted in this area. The licensee staff was observed to be enforcing correct examination security procedures.
4OA6 Meetings
.1 Exit Meeting
The chief examiner presented the examination team's preliminary observations and findings on November 18, 2005, to Mr. M. Gaffney and other members of the Operations and Training Department staff. A subsequent exit via teleconference was held on December 8, 2005, with Mr. S. Johnson following review of the site post examination comments. The licensee acknowledged the observations and findings presented. No proprietary information was identified by the stations staff during the exit meeting.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
Enclosure 1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- M. Gaffney, Site Vice President
- K. Hoops, Site Director
F Winks, Manager Nuclear Training
- J. Ruttar, Manager Nuclear Operations
- D. Fitzwater, Supervisor Nuclear Operations Training
- S. Johnson, Operations Training
NRC
- S. Burton, Senior 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
PARS Publicly Available Records
RO Reactor Operator
SRO Senior Reactor Operator
Attachment
SIMULATION FACILITY REPORT
Facility Licensee: Kewaunee Power Station
Facility Docket No.: 50-305
Operating Tests Administered: November 14 - 17, 2005
The following documents observations made by the 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 item was
observed:
ITEM DESCRIPTION
A problem was noted with the use of the Emergency Diesel Generator
Emergency Speed Control switch. When attempting to adjust frequency from 62 to
Diesel 65 Hz, frequency increased uncontrollably resulting in a Diesel
Generator Generator trip. This problem occurred on both the A and B Diesel
Generators.
Question Number 32.
Given the following:
- The plant startup is in progress.
- Reactor power is 1.4%.
- FW-07A & B, S/G Main Feed valves, AND FW-10A & B, S/G Bypass Feed
valves, indicate closed.
- Feedwater Pump A has just been started.
- SG B level begins to slowly rise.
What action is required?
- A. Direct the NAO to locally close FW-9B, Feedwater Main Control Valve 1B Bypass
Valve Inlet.
- B. Cycle FW-10B, S/G B Bypass, fully open and closed.
- C. Close FW-12B, S/G B Feedwater Isolation Valve.
D. Stop Feedwater Pump A.
Answer: C
Applicant Comment:
An applicant commented that answer D should also be accepted as correct. He
argued that if level was near the high level trip of 67% stopping the Feedwater Pump
would be a viable answer to preclude damage to the secondary system and the turbine.
Facility Proposed Resolution:
The facility management argued that there is only one correct answer for the given
conditions. Based on the given conditions, the guiding procedures are N-O-02, Plant
Startup From Hot Shutdown to 35% power, and N-FW-05A, Feedwater System Normal
Operation. Per N-O-02, the initial conditions are Steam Generator (SG) level is
maintained between 33% and 50%. Therefore, based on the question stem a normal
plant startup level band should be assumed. Furthermore, the stem states that SG B
level begins to slowly rise. Thus, there is no basis for assuming SG level is near or
approaching the SG Feedwater Isolation setpoint of 67%.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to accept only the original correct answer. Based on the plant
conditions given in the question stem (normal startup in progress) it is unreasonable to
postulate that steam generator level is near the high level trip. A slowly rising steam
generator level is anticipated by plant procedure under the conditions given this
question, and procedural guidance is provided to close the feedwater isolation valve
under these conditions (answer C). Therefore, C is the only correct answer.
Question Number 34.
Given the following:
- The plant is at 100% power.
- AFW Pump B is running for a surveillance test in progress.
- Annunciator 47061-M, AFW PUMP B LOW OIL PRESS, alarms.
What is the expected operator response for this condition?
- A. Trip AFW Pump B, and go to N-FW-05B, Auxiliary Feedwater System.
- B. Trip AFW Pump B, and go to A-FW-05B, Abnormal Auxiliary Feedwater System
Operation.
- C. Verify the Auxiliary Lube Oil Pump is running, and go to N-FW-05B, Auxiliary
Feedwater System.
- D. Verify the Auxiliary Lube Oil Pump is running, and go to, A-FW-05B, Abnormal
Auxiliary Feedwater System Operation.
Answer: D
Applicant Comment:
An applicant commented that answer B should also be accepted as correct. He
argued that per the stem of the question, a surveillance test was in progress with
Annunciator 47061-M, AFW PUMP B LOW OIL PRESS, O
- N. Since, the stem provides
no information as to whether the aux lube oil pump starts and the alarm clears, and
knowing the Auxiliary Feedwater (AFW) Pump is being run only for testing, it would be
correct to trip the AFW pump. The applicant sights UG-0 (Rev. F), Users Guide For
Emergency And Abnormal Procedures, section 6.1.2, which states that during abnormal
conditions, an operator has the authority to take action that is clearly needed based on
observed condition. With the annunciator 47061-M not clearing, there is a lube oil
problem. Conservative Decision Making, would require the AFW pump be tripped.
Facility Proposed Resolution:
The facility management agreed with the applicant that, without information about the
current oil pressure value and trend, an operator could not make a valid decision as to
whether the Auxiliary Feedwater (AFW) Pump should be tripped or not. The question
would need to be revised to make only one of the answers correct. It was
recommended that two answers be accepted, B and D.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to accept two answers as correct. Because, the stem of the
question states that the auxiliary feedwater pump low oil pressure alarm annunciates it
was expected that the applicants would select answer D which contained the actions
called out in the annunciator procedure (verify the auxiliary oil pump is running).
However, neither the question stem nor answer D provided information as to whether
oil pressure recovered. Therefore, it is equally reasonable to assume oil pressure has
not recovered. Knowing that there has been a low oil pressure alarm, believing oil
pressure to still be low, and from the stem knowing there is no emergency condition
because it is stated that the auxiliary feedwater pump is being run for a surveillance
test, it would also be correct to trip the auxiliary feedwater pump (answer B).
Therefore both answers B and D were accepted as correct.
Question Number 44.
Which of the following situations requires entry into a Technical Specification LCO
Action?
A. Pressurizer Pressure Transmitter P-429 fails low while at 38% power.
B. Pressurizer level decreases to less than 17% with a reactor startup in progress.
C. Pressurizer Backup Heaters energize after a 8% load reduction.
D. Pressurizer Pressure Transmitter P-449 fails high while on RHR Cooling.
Answer: A
Applicant Comment:
An applicant commented that answer D should also be accepted as correct. He
argued that answer D requires entry into a Technical Specification LCO Action to
MAINTAIN HOT SHUTDOWN.
Facility Proposed Resolution:
The facility management disagreed with the applicant. They argued that with the plant
on RHR Cooling, as stated in answer D, the reactor would be shutdown with Reactor
Coolant System (RCS) temperature is less than 400ºF and RCS pressure less than
25 psig. In these conditions the plant is in INTERMEDIATE SHUTDOWN, and
therefore, the LCO for PT-449 does not apply.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to accept only the original correct answer. The applicant
contends that answer D should be considered correct because, although no action is
required, the failure would place a Mode change restriction on the plant. A review of
Technical Specification 3.5, Table 3.5-2, and Table 3.5-3 confirmed the facility position
that entry into a Technical Specification LCO Action was not required for the conditions
listed in distractor D. The fact that Transmitter P-449 has a Technical Specification
function in another plant Mode of operation is what makes it a good distractor. Only the
conditions given in answer A would require Technical Specification directed actions
and therefore A was retained as the only correct answer.
Question Number 62.
Given the following:
- Plant is at 90% power.
- Power Range Instrument N-42 fails and is removed from service per A-MI-87,
Bistable Tripping for Failed Reactor Protection or Safeguards Inst.
- The surveillance for SP-47-011A, Reactor Coolant Temperature and
Pressurizer Pressure Instrument Channel I (Red) Calibration, comes due.
- The decision is to bypass the failed NIS channel inputs to allow the calibration
(Technical Specification 3.5.d).
What is the coincidence for the OTT reactor trip while the surveillance is in progress?
A. 2 out of 2
B. 2 out of 3
C. 1 out of 4
D. 1 out of 3
Answer: B
Applicant Comment:
An applicant commented that answer D should also be accepted as correct. He
argued that, in a normal lineup, the OTDT trip coincidence is 2 out of 4 (4 channels
provide input, 2 required to actuate). With N-42 input to Channel II - OTDT trip
bypassed, only 3 channels provide input, 2 are still required to actuate to trip the
reactor, therefore coincidence is 2 out of 3. During SP-47-011A, after initial switch
lineups and verifications, the bistables associated with the RED Channel Pressure /
Temperature trips are placed in TES
- T. After that step, 3 channels still provide input if a
BLUE or a YELLOW channel actuates the reactor will trip, therefore he argued that the
coincidence becomes 1/3. Therefore he believes the correct answer depends on which
actions of SP-47-011A have been completed. Because, it was not stated in the
question stem which actions of SP-47-011A have been completed, he contends answer
D should also be accepted as correct.
Facility Proposed Resolution:
The facility management disagreed with the applicant. They argued that the
coincidence for the OTT reactor trip will always remain 2 out of 3 for the conditions
given in the question stem. While it is true that the number of bistables that must
actuate to cause a reactor trip varies between one and two at various times during the
performance of SP-47-011A, the overall coincidence for the trip is still 2 out of 3. There
will always be 3 channels, at times during the performance of SP-47-011A one of the
three bistables is actuated and only one additional bistable needs to actuate to
complete the 2 out of 3 logic.
It was also stated that an applicant requested clarification of the status of performing
the surveillance. It was answered that the surveillance had not yet been entered. This
information was provided to all applicants in the room at that time. The individual
providing this feedback was not in the room, as he had completed his examination.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to accept only the original correct answer. The applicant
commented that answer D should be considered correct because, at times in the
conditions provided in the question stem, the trip of one more instrument would cause a
reactor trip. While it is true that at times during the completion of SP-47-011A the trip of
one more instrument would cause a reactor trip, that was not the question asked. The
question asked what was the reactor trip coincidence, which in this case was 2 out of
at all times, answer B. The concept of what trip coincidence the plant is in is an
important concept for Technical Specification compliance. Because only answer B
provided the correct reactor trip coincidence for the conditions given in the stem it was
retained as the only correct answer.
The Facility noted that a clarification to this question was provided after the applicant
providing this comment had left the room having completed the examination. However,
the clarification did not provide information necessary to select the correct answer.
Therefore, again it was determined B was the only correct answer.
Question Number 63
Given the following:
- Power level is 99.9% (1771.5 MWt).
- The current UFMD and RTO OPERATING LIMITs are 1772 MWt.
- SP-87-125, Shift Instrument Channel Checks - Operating, calorimetric was
completed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ago.
- The signal is lost to PPCS from Feedwater Flow channel FT-476.
- The RTO 1-minute average, PPCS point R5110G, drops to 1200 MWt.
- The Computer Group reports the signal will be restored within 10 minutes.
What is the affect on the UFMD and RTO Operating Limits, and what action is taken
when the flow channel input is restored and R5110G reads normal?
A. Both the UFMD and RTO Operating limits will read 1749 Mwt.
When the signal is restored, the operator will click the APPLY UFMD LIMIT button
on the PPCS UFMD Correction Factors screen to return both limits to 1772 Mwt.
B. The UFMD Operating Limit will read 1749 MWt and RTO Operating Limit will read
1772 Mwt.
Power will be reduced to less than 1749 Mwt, and when the signal is restored, the
operator will click the APPLY UFMD LIMIT button on the PPCS UFMD Correction
Factors screen to return the UFMD limit to 1772 Mwt.
C. Both the UFMD and RTO Operating limits will read 1769 Mwt.
When the signal is restored, the operator will click the APPLY UFMD LIMIT button
on the PPCS UFMD Correction Factors screen to return both limits to 1772 Mwt.
D. The UFMD Operating Limit will read 1769 MWt and RTO Operating Limit will read
1772 Mwt.
Power will be reduced to less than 1769 Mwt, and when the signal is restored, the
operator will click the APPLY UFMD LIMIT button on the PPCS UFMD Correction
Factors screen to return the UFMD limit to 1772 Mwt.
Answer: A
Applicant Comment:
An applicant commented that there was no correct answer to this question. He argued
that when the flow channel is returned, the UFMD limit will go to 1772. Then, when the
APPLY UFMD LIMIT button is clicked only the RTO Limit changes, not both the RTO
and the UFMD Limit.
Facility Proposed Resolution:
The facility management maintained that answer A is correct. They agree that when
the flow channel is returned, the UFMD limit will go to 1772. However, to have both the
the RTO and the UFMD Limit read 1772 the APPLY UFMD LIMIT button must be
clicked. Therefore, answer A remains correct.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to retain the question. This question had a two part answer.
First, for the stem conditions what would the UFMD and RTO Operating Limit read, and
then how to reset the limits? Only answer A supplied the correct answer for the first
part of the question. The applicant commented that the second part of answer A was
incorrect because the UFMD limit would automatically reset. While the second part of
the answer could have been worded differently, it is correct. Both the UFMD and RTO
Operating Limits would not read 1772 MWt until the APPLY UFMD LIMIT button was
pushed. Because A is a correct answer the question was retained.
Question Number 67
Given the following:
- Waste Gas Decay Tank A is in service.
- The relief valve for that tank, WG-14A, lifts and fails to reseat.
What is the effect on the plant?
A. The release will be automatically isolated when the Waste Gas Analyzer senses the
pressure drop in the Waste Gas Decay Tank.
B. The release will be automatically isolated when the Waste Gas Decay Tank
pressure reducing control valve WG-201 closes.
- C. The release will NOT be automatically isolated, but will be monitored by the Aux.
Building Vent radiation monitors R-13 and R-14.
- D. The release will NOT be automatically isolated, but will be detected by the Charging
Pump Room Area Monitor R-4.
Answer: C
Applicant Comment:
An applicant commented that the stem of the question lists the wrong valve number for
the Waste Gas Decay Tank A relief valve. He argues that if the question is answered
using the valve with the valve number given in the stem, answer D could be a possible
correct answer, if a pipe failure was also assumed.
Facility Proposed Resolution:
The facility management maintained that this question had only a typographical that did
not invalidate the question. They agreed that in the question stem The relief valve for
that tank, WG-14A, lifts and fails to reseat. should have read The relief valve for that
tank, WG-13A, lifts and fails to reseat. However, they argued that during the
examination no individual questioned the difference between the valve number and the
valve name. Additionally, they argued that even if someone attempted to answer the
question for valve WG-14A, which is a valve that connects the Waste Gas Decay Tank
to a vent header, answer D would not be correct.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to accept only the original correct answer. This question asks
the effect on the plant of the relief valve on Waste Gas Decay Tank A lifting and failing
to reseat. An applicant commented that the wrong valve number was given for the
relief valve (which is true), and if one attempted to answer the question using the valve
with the provided valve number (and ignoring the fact that the stem clearly states it is
the relief valve that opens), answer D could be right. In fact D would not be a correct
answer even if the provided valve number is assumed to be correct. Additionally, none
of the candidates asked a question during the examination about the difference
between the valve number and the valve description. This was a minor typographical
error that does not appear to have confused any applicant or invalidated the question.
Therefore, C was retained as the only correct answer.
Question Number 81
Given the following:
- The plant is at 100% power.
- Surveillance Test SP-42-312B, Diesel Generator B Availability Test, is in
progress with DG B running.
- Component Cooling Pump B trips on overcurrent.
- BRB-104, ckt 10 supplying DG B tripped open and CANNOT be closed.
What is the effect of this condition?
A. DG B is inoperable and must have its fuel supply locally isolated.
A plant shutdown must commence within one hour using the Standard Shutdown
Sequence (Technical Specification 3.0.c)
B. DG B is Degraded but Operable and can be controlled locally.
A 24-hour LCO is applicable for restoration of the DC Distribution System and a
2-hour LCO is applicable for restoration of Component Cooling Pump B.
C. DG B is inoperable and must have its fuel supply locally isolated.
A 72-hour LCO is applicable for restoration of Component Cooling Pump B and a
7-day LCO is applicable for restoration of DG B.
D. DG B is Degraded but Operable and can be controlled locally.
Only the 72-hour LCO is applicable for restoration of Component Cooling Pump B.
Answer: C
Applicant Comment:
An applicant commented that answer A should also be accepted as correct. He
argued that, this question has two answers dependent on the time frame. He agrees
that answer C is correct initially. However, he argues that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the
conditions given in the stem A Diesel Generator testing is required, because all B
train emergency equipment is not OPERABLE this testing can not be performed.
Therefore, at that time a plant shutdown must commence within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Facility Proposed Resolution:
The facility management maintained that there was no reason, based on the question
stem, to extrapolate what the answer to this question could be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later. They
argued that based on the conditions given in the question stem C is the only correct
answer.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to accept only the original correct answer. The applicant
correctly points out that under the condition of the stem a once per 24-hour test of the
A Diesel Generator is required and cannot be performed. Therefore he believes that if
the conditions still existed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later a shutdown must be commenced in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,
making answer A correct. While site management contends it is unreasonable to
answer this question by speculating what the answer could be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, in that the
24-hour period until Diesel Generator testing is required would be used to attempt to
return equipment to service, thereby preventing the need to shutdown. Technical Specifications 3.0 c, 3.3.d.2, 3.7.b.2, and 3.7.c were reviewed and it was found that
there was no requirement that a plant shutdown must commence within 1-hour of the
conditions given in the question stem. Therefore, C was retained as the only correct
answer.
Question Number 88
What is the reason for the Feedwater Isolation signal generated from High-High SG
level?
A. Preclude excessive SG tilts due to cooler feedwater supplied to ONE SG.
B. Prevent overfill of the SG that may result in damage to secondary components.
C. Ensure containment pressure remains within maximum internal pressure limit with
the affected SG faulted inside containment.
A/B Main Valve, fully open.
Answer: B
Applicant Comment:
An applicant commented that answer C should also be accepted as correct. He
argued that per the Update Safety Analysis Report, one of the 4 major factors that
influence release of mass and energy following a steam line break is Steam Generator
(SG) water level. Having too much inventory could challenge containment integrity, and
the Feedwater Isolation signal generated from High-High SG level prevents this.
Therefore, C should also be considered a correct answer.
Facility Proposed Resolution:
The facility management disagreed with the applicant. They argued that the specific
reason for Feedwater Isolation signal on Steam Generator (SG) high level was to
prevent damage to secondary components as stated in the Technical Specification
Basis.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to accept only the original correct answer. The applicant
contended that there are two correct answers given for this question. He believes that
in addition to the reason given in the Technical Specification Basis, the High-High
feedwater isolation is provided to limit containment pressure in the event of a faulted
Steam Generator. This is a misconception on the part of the candidate. The USAR
analysis for a faulted Steam Generator is based on normal level (not level at the High-
High limit) and a feedwater isolation from the resulting safety injection signal. The
Technical Specification 3.5 Basis states Main feedwater isolation actuation occurs as a
result of a Hi-Hi steam generator water level to prevent steam generator overfill
conditions. Steam generator overfill may result in damage to secondary components;
for example, high moisture steam could erode the turbine blades at an accelerated
rate. no other reason for this trip is mentioned. Therefore, B was retained as the only
correct answer.
Question Number 90
Given the following:
- The plant has experienced a fire in the Control Room.
- The actions of E-O-06, Fire In Alternate Fire Zone, are being performed.
What are the indications of an air dryer filter failure, and what are the procedural
actions performed by the Control Room Supervisor to mitigate the consequences of this
failure.
A. Instrument Air Drier/Filter 1A/1B differential presure will rise above 5 psid.
The CRS will verify SA-121, Air Drier/Filter Bypass CV, opens.
B. Instrument Air Drier/Filter 1c differential pressure will rise above 10 psid.
The CRS will open SA-100A, Air Drier 1A Supply, and IA-300, 11/2 Alt IA, and then
request the Control Operator A to start Air Compressor B.
C. Instrument air header pressure will drop below 95 psig.
The CRS will depress the Air Drier 1C RESET pushbutton to confirm problem, and
then align flow through Instrument Air Drier 1B.
D. Instrument air header pressure will drop below 100 psig.
The CRS will open SA-70 and SA-71, 11/2" Dedicated Instrument Air Header
Isolations, that bypass the Instrument Air Driers.
Answer: D
Applicant Comment:
An applicant commented that answer A should also be accepted as correct. He
argued that high D/P across IA-121 is an early indication of an air dryer filter failure and
that the normal automatic action for that condition is opening of the bypass valve.
Therefore, A is a correct answer.
Facility Proposed Resolution:
The facility management disagreed with the applicant. They argued that answer A
gives the correct actions under normal conditions. However, in the situation given in
the question stem the control room is evacuated and E-O-06 is in progress. Under
these conditions the Air Dryers are bypassed by opening the Dedicated Air Header
isolations, which supply the air directly through a separate in-pipe air dryer, making D
the only correct answer.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed
resolution it was decided to accept only the original correct answer. The applicant
contended that there are two correct answers given for this question. He argues that
answer A contains the normal action for air dryer failure and is therefore correct.
However, the question stem states that there is a fire in the control room and the
actions of procedure E-O-06 are being carried out, these are the action described in
answer D not the action described in answer
- A. Therefore, D was retained as the
only correct answer.
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession # ML060060460.