IR 05000263/2007301
| ML070680188 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/08/2007 |
| From: | Hironori Peterson Division of Reactor Safety III |
| To: | Conway J Nuclear Management Co |
| References | |
| 50-263/07-301 ER-07-301, IR-07-301 | |
| Download: ML070680188 (32) | |
Text
rch 8, 2007
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT NRC INITIAL LICENSE EXAMINATION REPORT 05000263/2007301(DRS)
Dear Mr. Conway:
On February 19, 2007, the Nuclear Regulatory Commission (NRC) completed initial operator licensing examinations at your Monticello Nuclear Generating Plant. The enclosed report presents the results of the examination which were discussed on February 16 and March 1, 2007, with you and Mr. Earl, respectively, and with other members of your staff.
The NRC examiners administered initial license examination operating tests during the week of February 12, 2007. Members of the Monticello Nuclear Generating Plant Training Department administered an initial license written examination on February 19, 2007, to the applicants.
Four senior reactor operator and four reactor operator applicants were administered license examinations. The results of the examinations were finalized on March 6, 2007. Six applicants passed all sections of their examinations resulting in the issuance of two senior reactor operator and four reactor operator licenses. Two applicants failed the written examination and will not be issued licenses. The applicants who failed the NRC examination were issued proposed license denial letters.
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 Records System (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. 50-263 License No. DPR-22
Enclosures:
1. Operator Licensing Examination Report 05000263/2007301(DRS)
2. Simulation Facility Report 3. Post Examination Comments and Resolutions 4. Written Examinations and Answer Keys (RO/SRO)
REGION III==
Docket No: 50-263 License No: DPR-22 Report No: 05000263/2007301(DRS);
Licensee: Nuclear Management Company, LLC Facility: Monticello Nuclear Generating Plant Location: Monticello, Minnesota Dates: February 6 through February 19, 2007 Examiners: N. Valos, Chief Examiner B. Palagi, Examiner C. Zoia, Examiner Approved by: Hironori Peterson, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000263/2007301(DRS); 02/12/2007 - 02/19/2007; Nuclear Management Company, LLC;
Monticello Nuclear Generating Plant; Initial License Examination Report.
The announced initial operator licensing examination was conducted by regional NRC examiners in accordance with the guidance of NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9.
Examination Summary:
- Eight examinations were administered (four senior reactor operator and four reactor operator).
- Six applicants passed all sections of their examinations resulting in the issuance of two senior reactor operator and four reactor operator licenses.
- Two applicants failed the written examination and will not be issued licenses. The applicants who failed the NRC examination were issued proposed license denial letters.
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA5 Other
.1 Initial Licensing Examinations
a. Inspection Scope
The NRC examiners conducted an announced initial operator licensing examination during the week of February 12, 2007. The facility licensees training staff used the guidance prescribed in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, to prepare the outline and develop the written examination and operating test. The examiners administered the operating test, consisting of job performance measures and dynamic simulator scenarios, during the period of February 12 through February 16, 2007. The facility licensee administered the written examination on February 19, 2007. Four senior reactor operator and four reactor operator applicants were examined.
b. Findings
Written Examination The NRC examiners determined that the written examination, 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.
A total of ten post-examination comments (8 RO; 2 SRO comments) were submitted by the applicants to the facility training department. One of the post-examination comments was associated with a clarification made to a question by the facility during the administration of the examination. Of the ten post-examination comments, the facility agreed with two of the comments. The post-examination comments were submitted to the NRC on February 26, 2007. The results of the NRCs review of the comments are documented in Enclosure 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."
Examination Results Six applicants passed all sections of their examinations resulting in the issuance of two senior reactor operator and four reactor operator licenses. Two applicants failed the written examination and will not be issued licenses. The applicants who failed the NRC examination were issued proposed license denial letters.
.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) and integrity in accordance with 10 CFR 55.49, Integrity of Examinations and Tests, and NUREG-1021, Operator Licensing Examination Standard for Power Reactors. The examiners reviewed and observed the licensees implementation and controls of examination security and integrity measures (e.g.,
security agreements) throughout the examination process.
b. Findings
The licensees implementation of examination security requirements during examination preparation and administration were acceptable and met the guidelines provided in NUREG-1021, Operator Licensing Examination Standards for Power Reactors. No violations of 10 CFR 55.49 occurred during the examination preparation and administration.
4OA6 Meetings
Exit Meeting The chief examiner presented the examination teams preliminary observations and findings with Mr. J. Conway, Site Vice President, and other members of the licensee management on February 16, 2007. A subsequent exit via teleconference was held on March 1, 2007, with Mr. J. Earl, General Supervisor Operations Training, following review of the site post-examination comments. No proprietary items were identified during the administration of the examination nor during the exit meeting with the licensee. The licensee acknowledged the observations and findings presented.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
Enclosure 1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- J. Conway, Site Vice President
- B. MacKissock, Operations Manager
- S. Halbert, Training Manager
- J. Earl, General Supervisor Operations Training
- G. Allex, Supervisor Operations Training - Continuing
- O. Olson, Supervisor Operations Training - Initial
- J. Ruth, Operations Training
NRC
- N. Valos, Chief Examiner
- B. Palagi, Examiner
- C. Zoia, Examiner
- S. Thomas, Senior Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None
Closed
None
Discussed
None
LIST OF ACRONYMS
ADAMS Agency-Wide Document Access and Management System
CFR Code of Federal Regulations
CR Condition Report
DRS Division of Reactor Safety
NRC Nuclear Regulatory Commission
PARS Publicly Available Records System
RO Reactor Operator
SRO Senior Reactor Operator
Attachment
SIMULATION FACILITY REPORT
Facility Licensee: Monticello Nuclear Generating Plant
Facility Licensee Docket No.: 50-263
Operating Tests Administered: February 12 through February 16, 2007
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 items were
observed:
ITEM DESCRIPTION
There was one simulator exam scenario delay of approximately 30 minutes on
the morning of February 14, 2007, until three Process Computer screens could
be restored in the simulator. CAP01077256 was written associated with the
issue.
Enclosure 2
Post Examination Comments and Resolutions
Question Number 4
A LOCA is in progress. Both Core Spray pumps are injecting at 3000 gpm each to maintain
RPV water level above the top of active fuel. Which of the below listed plant parameters may
be an indication of pump cavitation in the Core Spray pump(s)?
A. Steadily lowering of Core Spray pump amps
B. Steadily lowering of Core Spray discharge pressure
C. Annunciator 3-A-29 (CORE SPRAY 1 PRESS VLV LEAKING) in alarm
D. Repeated alarming and subsequent clearing of annunciator 3-A-41(AC INTERLOCK)
Answer: D
Applicant Comment:
An applicant commented that answer A should also be accepted as correct.
Answer A should also be accepted. The discharge head of the Core Spray (CS) pumps is
approximately 300 psig. The setpoint for the annunciator alarm AC Interlock is 100 psig. The
stem of the question states both CS pumps are in service. Both CS pumps must be below
100 psig to clear the alarm. Even with severe cavitation, the applicant does not believe that
both CS pumps would drop in discharge pressure at the same time to clear the alarm. Even
though the Bases for C.4-B.04.01G, ECCS Suction Control During LOCA stated that the
alarm may be an indication of ECCS suction plugging, this statement assumed only one CS
pump was running. That was NOT the case in the question stem. Step 1.f of C.4-B.04.01G
stated that the pump motor amperage would be erratic or decreasing for plugging strainer.
Therefore, answer A is the most correct answer.
Facility Proposed Resolution:
The question grading for the exam should not change. The effects of cavitation include
fluctuations in discharge pressure and motor current. This is supported in Lesson Plan
M-8120L-114, Fluid Statics and Dynamics, which was part of the Initial License Training (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.
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.
Enclosure 3
Post Examination Comments and Resolutions
The applicant stated that the Bases for C.4-B.04.01G assumed that only one CS pump was
running, when stating that the AC INTERLOCK alarm may be an indication of ECCS suction
plugging. However, the Bases for C.4-B.04.01G does not stipulate how many ECCS pumps
may be running. The applicant stated that pump motor amperage would be erratic or
decreasing for plugging strainer. Though this is valid, the question did not ask for an indication
of a plugging strainer, the question asked for an indication of pump cavitation in the Core Spray
pump(s). The Bases for C.4-B.04.01G stated that one indication of an ECCS suction strainer
plugging is: Erratic and dramatic fluctuations in pump discharge pressure. One alarm that
may clue the operator to this condition is the alarming and subsequent clearing of AC
INTERLOCK (3-A-41).
The indications that a pump is cavitating include fluctuations in discharge pressure and motor
current. These effects of cavitation are supported in Lesson Plan M-8120L-114, Fluid Statics
and Dynamics, which was part of the Initial License Training Generic Fundamentals Course.
Per Lesson Plan M-8120L-114, the indication specified in distractor A, Steadily lowering of
Core Spray pump amps, is not an indication of cavitation of the CS pumps. Since distractor
D, Repeated alarming and subsequent clearing of annunciator 3-A-41(AC INTERLOCK),
provided the only indication of pump cavitation of the CS pumps, distractor D was retained as
the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 17
During a Reactor startup, when do plant conditions support the design limitations of the Reactor
Water Level Control System allowing it to be placed in 3 Element Control?
A. When the first Feedwater Control Valve is placed in automatic.
B. When the Master Feedwater Level Controller is placed in automatic.
C. When the second Feedwater Control Valve is placed in automatic.
D. When feedwater flow is sufficient to clear Reactor Recirc pump low flow interlock.
Answer: C
Applicant Comment:
An applicant commented that answer D should also be accepted as correct.
Answer D should also be accepted as correct. Ops Manual B.05.07-01, Reactor Level
Control states that the steam and feedwater flow signals lose their accuracy below 30% power.
Also, on page 7 of B.05.07-01, it states that a separate single element control scheme is used
when reactor power is less than 20% of rated. The question asked per design limits of the
water level control system, when can the Digital Feedwater Control System (DFCS) be placed
in 3 element control. The 3 element would work at greater than 20% feed flow and with one
Feedwater Fegulating Valve (FRV) in service. Although the start up procedure C.1, 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 greater than 20%
per its design.
Facility Proposed Resolution:
The question grading for the exam should not change. Per Ops Man B.05.07-01, 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 greater than 20% are for transition from the Feedwater Low Flow
Regulating Valve to the Main Regulating Valves.
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. Ops Manual B.05.07-01, Reactor Level
Control states that the steam and feedwater flow signals lose their accuracy below 30% power.
Both the steam and feedwater flow signals are used when in 3 Element Control.
Enclosure 3
Post Examination Comments and Resolutions
The design limitations of the Reactor Water Level Control System would not allow placing 3
Element Control in service with inaccurate steam and feedwater flow signals. The applicant
also stated that a separate single element control scheme is used when reactor power is less
than 20% of rated power. However, this statement is related to control of the Feedwater Low
Flow Regulating Valve, not to 3 Element Control of the Main Feedwater Regulating Valves.
Thus, distractor D, When feedwater flow is sufficient to clear Reactor Recirc pump low flow
interlock, which occurs at approximately 20% reactor power would not be a correct answer.
Since, per C.1, Reactor Startup, at approximately 40% power, the second Main Feedwater
Regulating Valve is placed in service, and then Reactor Level Control is transferred to 3
Element Control, distractor C was retained as the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 20
What would be the effect on 4.16 KV breaker operation if all D.C. control power is lost?
Breaker operation with the control switch would be lost to
A. all 4.16 KV breakers.
B. all 4.16 KV breakers EXCEPT for buses 13 and/or 14.
C. all 4.16 KV breakers EXCEPT for buses 15 and/or 16.
D. all 4.16 KV breakers EXCEPT for buses 17 and/or 18.
Answer: D
Applicant Comment:
An applicant commented that answer A should also be accepted as correct.
A clarification was made to question 20 which changed the acceptable answer. Without the
clarification, answer D would be correct since the 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 no bus 17 or bus 18 control
switches in the control room, there would be no 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 Proposed Resolution:
The question grading for the exam should be changed to accept both A and D as correct
answers. 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.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to accept the facilitys comment and accept both answer A and D as correct
answers.
During the administration of the examination, a question was asked by an applicant as to which
control switch the question stem was referring to (i.e., the control switch in the Control Room or
Enclosure 3
Post Examination Comments and Resolutions
at the breaker). The facility provided a clarification that the question stem was asking Breaker
operation with the Control Room control switch would be lost to, with the words underlined
added for clarification. This clarification of the question was communicated to all of the
applicants. The clarification provided changed the intent of the question. Buses 17 and 18 are
in the plant discharge structure, with control power supplied alternating current (AC) from one of
two control power transformers. There are no bus 17 or bus 18 control switches in the control
room.
In answering the clarified question, since there are no bus 17 or bus 18 control switches in the
control room, there would be no control switches that could operate 4.16 KV breakers in the
control room due to the loss of all DC control power. This would make distractor A, all
4.16 KV breakers, a correct answer. However, in answering the clarified question, distractor
D is still correct, since there are no control room switches for bus 17/18 breakers. Distractors
A and D thus reduce to the same answer all 4.16 KV breakers. Therefore, the answer key
was modified to accept both A and D as correct answers.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 31
A plant transient was in progress with the following conditions present:
- D/W pressure 15 psig and rising
- D/W temperature 285°F and rising
- RPV water level -175 inches and slowly lowering
- RPV pressure 850 psig
- All Control rods are fully inserted
- EOPs 1100 and 1200 have been entered and are being executed
- The SAMGs have not been entered
- Drywell spray is directed to be placed in service
Which of the following predicts the impact on Drywell temperature and why?
A. Drywell temperature would continue to rise; Drywell Spray is NOT ALLOWED as plant
conditions are outside the Drywell Spray Limit curve.
B. Drywell temperature would lower; as Drywell Spray IS ALLOWED to be initiated by
placing the Containment Spray/Cooling LPCI Initiation switch in Bypass and opening the
Drywell Spray inboard and outboard isolation valves.
C. Drywell temperature would continue to rise since neither Drywell Spray inboard or
outboard isolation valve is allowed to be opened under these plant conditions.
D. Drywell temperature would lower as drywell spray is allowed to be initiated by placing
both the Containment Spray 2/3 Core Height Bypass and Containment Spray/Cooling
LPCI Initiation switches in BYPASS, and then opening the Drywell Spray inboard and
outboard isolation valves.
Answer: C
Applicant Comment:
An applicant commented that answer D should also be accepted as correct.
Answer D should also be accepted. The RHR logic allows drywell sprays to be placed in
service after bypassing the 2/3 core height interlock and the Containment Spray(CS)/Cooling
Low Pressure Coolant Injection (LPCI) initiation switch. Although the Drywell (DW) spray
procedure (Part C of procedure Ops Man C.5-3502, Drywell Spray - RHR-A) stated that this
should be done when in the Severe Accident Management Guidelines (SAMGs), the logic would
allow answer D to be correct. The applicant assumed other pumps (like High Pressure
Coolant Injection and CS) were available to recover level. The basis did state DW spray and
injection can be alternated to provided adequate cooling. Therefore, answer D also correctly
answered the question.
Facility Proposed Resolution:
Enclosure 3
Post Examination Comments and Resolutions
The question grading for the exam should not change. Site procedures, other than SAMGs, do
not allow bypassing of the interlock for DW spray; the question states The SAMGs 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.
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 question asks, for a given set of conditions during a plant transient, to predict the impact on
Drywell temperature and why? Distractor D states Drywell temperature would lower as
drywell spray is allowed to be initiated by placing both the Containment Spray 2/3 Core Height
Bypass and Containment Spray/Cooling LPCI Initiation switches in BYPASS, and then opening
the Drywell Spray inboard and outboard isolation valves. Procedure Ops Man C.5-3502 only
allows placing the Containment Spray 2/3 Core Height Bypass switch to MANUAL OVERRIDE
and the Containment Spray/Cooling LPCI Initiation switch in BYPASS, if required to spray the
containment by the SAMGs. The question stem states The SAMGs have not been entered.
Although system design provides the means to bypass the interlock with the given conditions,
bypass of the interlock is clearly not allowed by site procedure. Therefore, distractor C was
retained as the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 46
With the reactor operating steady state in MODE 1, what effect, if any, will a loss of instrument
air have on the Reactor Building to Suppression Pool Vacuum Relief Dampers, AO-2379 and
AO-2380?
A. These valves would fail OPEN.
B. These valves would fail CLOSED.
C. No effect as these valves are normally aligned to air and use nitrogen as a backup
supply.
D. No effect as these valves are normally aligned to nitrogen and use bottled nitrogen as a
backup supply.
Answer: D
Applicant Comment:
An applicant commented that answer C should also be accepted as correct.
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 was true. As to the alignment, instrument air is
normally aligned to the system, but pressure set points have this valve closed (CV-7478). The
valve would automatically open if a problem existed with the Instrument Nitrogen system. The
second part of answer C states that nitrogen is used as a backup supply. This is true.
Alternate Nitrogen is used as the backup supply for these valves. Therefore, answer C is
technically correct also.
Facility Proposed Resolution:
The question grading for the exam should not change. As stated in the applicant comment
above, valve CV-7478, in its 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 stated. Answer D 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 normal supply to the Reactor Building
to Suppression Pool Vacuum Relief Dampers (AO-2379 and AO-2380) is the Instrument
Nitrogen supply from the Liquid Nitrogen Tank (reference drawing NH-36049-12). Bottled
nitrogen from the Alternate Nitrogen System is the first backup supply to the Vacuum Relief
Dampers (reference drawing NH-36049-10). The Instrument Air system supplies a secondary
Enclosure 3
Post Examination Comments and Resolutions
backup supply to the Vacuum Relief Dampers via valve CV-7478, which opens on low nitrogen
header supply pressure (reference drawing NH-36049-12). Therefore, the Vacuum Relief
Dampers are not normally aligned to air as distractor C stated. Distractor D, No effect as
these valves are normally aligned to nitrogen and use bottled nitrogen as a backup supply,
properly described the pneumatic supply to the Vacuum Relief Dampers, and distractor D was
retained as the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 59
The reactor was operating at rated conditions when the following events occurred: (Assume
SPDS is NOT available)
- 0700: 5-B-46 (CONDENSER LOW VACUUM) alarm was received
- 0705: 7-B-16 (VAC 24 IN TRIP #1) alarm was received
- 0705: 7-B-17 (VAC 24 IN TRIP #2) alarm was received
- 0708: PR-1264, CONDENSERS A AND B VACUUM, on C-07 indicated 6.1 Hg Abs.
- 0710: PR-1264, CONDENSERS A AND B VACUUM, on C-07 indicates 7 Hg Abs. and
remains stable
Given the above information, when is the EARLIEST time a manual Reactor Scram is required
to be initiated?
A. 0721
B. 0726
C. 0729
D. 0731
Answer: C
Applicant Comment:
An applicant commented that answer B should also be accepted as correct.
Answer B should also be accepted. Annunciators 7-B-16 and 7-B-17 alarm at 24 inches. By
typical engineering convention atmospheric pressure is 30 inches. Therefore, absolute
pressure in the condenser would be 6 inches Hg Abs, which would place one in the ALERT
region of Figure 1 at time 0705, requiring a scram 20 minutes later. Although procedure
Ops Man C.4-B.06.03.A, Decreasing Condenser Vacuum, stated that the operator should use
recorder PR-1264 when the Safety Parameter Display System (SPDS) was not available,
scramming the reactor at time 0729 would be 21 minutes later, which would not be in
accordance with the C.4-B.06.03.A procedure.
Facility Proposed Resolution:
The question grading for the exam should not change. The referenced procedure,
C.4-B.06.03.A, Decreasing Condenser Vacuum, under INDICATIONS, listed the annunciator 7-B-16 and 7-B-17 alarm setpoints of 24 Hg Vacuum as being equivalent to approximately
5" Hg Abs on recorder PR-1264; this procedure would override the stated typical engineering
convention. The procedure also required a reactor scram when in the ALERT region for
Enclosure 3
Post Examination Comments and Resolutions
longer than 20 minutes, therefore scramming at T+21 minutes is in accordance with the
C.4-B.06.03.A procedure.
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. Procedure Ops Man C.4-B.06.03.A,
Decreasing Condenser Vacuum, step 2.a under INDICATIONS, listed the annunciator 7-B-16 (VAC 24 IN TRIP #1) and 7-B-17 (VAC 24 IN TRIP #2) alarm setpoints of 24 Hg
(inches of Mercury) Vacuum as being equivalent to approximately 5 Hg Abs (Absolute) on
recorder PR-1264. Thus at time 0705, when the alarms for annunciators 7-B-16 and 7-B-17
were received, the plant was not in the ALERT region of Figure 1, Turbine Exhaust Pressure
Limits, of Ops Man C.4-B.06.03.
- A. For the reactor at rated conditions, as stated in the
question stem, the ALERT region of Figure 1 extended from a Turbine Exhaust Pressure of
6.0" Hg Abs to 7.5" Hg Abs. (NOTE: Figure 1 was provided to the applicant with the
examination.)
Procedure C.4-B.06.03.A stated that the operator should use recorder PR-1264 when the
Safety Parameter Display System (SPDS) was not available (the question stem stated that
SPDS was not available). The first instance at which the plant was in the ALERT region was
at time 0708 when it was given in the question stem that recorder PR-1264, CONDENSERS A
AND B VACUUM, on C-07 indicated 6.1 Hg Abs. Since the C.4-B.06.03.A procedure required
a reactor scram when the plant was in the ALERT region for longer than 20 minutes, time
29 (as specified in distractor C) was the earliest time at which a manual Reactor Scram was
required to be initiated in accordance with the procedure (when the plant was in the ALERT
region for 21 minutes).
The applicant stated that distractor B should also be accepted as a correct answer. He stated
that by typical engineering convention atmospheric pressure is 30" Hg. Since annunciators 7-B-16 and 7-B-17 alarm at 24" Hg Vacuum, the absolute pressure in the condenser would be
6" Hg Abs when annunciators 7-B-16 and 7-B-17 alarmed. This would place one in the
ALERT region of Figure 1 at time 0705, requiring a scram 20 minutes later. The applicant
stated that since distractor B was time 0726 (21 minutes after time 0705), this would make
distractor B a correct answer.
However, procedure C.4-B.06.03.A listed the annunciator 7-B-16 and 7-B-17 alarm setpoints of
Hg Vacuum as being equivalent to approximately 5 Hg Abs on recorder PR-1264 (and not
Hg Abs), which provided a more accurate representation of the absolute pressure
corresponding to 24" Hg Vacuum on recorder PR-1264 than that provided by the estimate of 6"
Hg Abs per typical engineering convention. The applicants were also provided with a book of
Steam Tables in which the conversion factor of 2.0360 was given for obtaining pressure in
inches of Hg from a pressure specified in pounds per square inch (psi). Using this
conversion factor, a pressure of 14.7 psig would correspond to 29.93 Hg Abs (since Monticello
is at approximately the 935 foot elevation above sea level, the nominal atmospheric pressure at
the site would actually be reduced to approximately 14.25 psig or 29.0" Hg Abs per Marks
Enclosure 3
Post Examination Comments and Resolutions
Standard Handbook for Mechanical Engineers). Using the high end value of 14.7 psig
(29.93" Hg Abs) for the pressure corresponding to atmospheric pressure at the Monticello site,
when annunciators 7-B-16 and 7-B-17 alarmed at 24" Hg Vacuum, this would correspond to a
pressure of 5.93" Hg Abs (NOTE: 24" Hg Vacuum would correspond to 5.0" Hg Abs for the
elevation at the Monticello site). Both the values of 5 Hg Abs (on recorder PR-1264) and 5.93"
Hg Abs (from the calculation described above) would place one below the ALERT region entry
of 6" Hg Abs (i.e., in the ALLOWABLE OPERATION region) of Figure 1. Therefore, the
procedural requirement to scram when in the ALERT region for longer than 20 minutes would
not begin when annunciators 7-B-16 and 7-B-17 alarmed at time 0705. Therefore, distractor
B is not a correct answer, and distractor C was retained as the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 62
While performing test 0081 (CONTROL ROD DRIVE SCRAM INSERTION TIME TEST) the
incorrect control rod was withdrawn from position 14 to 48. The crew recognized the error.
What action is required to ensure adequate margin to fuel thermal limits exists?
A. Notify Reactor Engineering to run a predictor and provide a recovery plan.
B. Obtain fuel thermal limit information before any rod recovery is performed.
C. Insert the control rod to its previous position without obtaining information on fuel
thermal limits.
D. Fully insert the control rod to position 00 and obtain further guidance from Reactor
Engineering.
Answer: C
Applicant Comment:
An applicant commented that answer D should also be accepted as correct.
Answer D should also be accepted as correct. Ops Manual B.01.03-05 H.10, Inadvertent
Control Rod Withdrawal/Rod Drift Out is used for an inadvertent control rod withdrawal/rod drift
out which is the condition in the question stem. This procedure stated that if annunciator 5-A-19, ROD SELECT BLOCK TIMER MALFUNCTION was not in alarm (which would be the
case in accordance with the question stem), then to immediately insert the rod to position 00
using normal or emergency insert.
Facility Proposed Resolution:
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 C was 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 stated 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.
Enclosure 3
Post Examination Comments and Resolutions
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.
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 stated that the correct
procedure to use for the conditions stated in the question stem was Ops Manual
B.01.03-05 H.10, Inadvertent Control Rod Withdrawal/Rod Drift Out, since this procedure is
used for an inadvertent control rod withdrawal/rod drift. However, the Bases for the procedure
states that the procedure is used for inadvertent control rod withdrawal or rod drifting during
notch-out movement, for the situation where the rod continues to withdraw past the desired
notch at the normal speed or drifts out with a slow erratic movement. These conditions were
not those stated in the question stem.
The question was associated with a situation in which the incorrect control rod was withdrawn
from position 14 to 48, with the crew subsequently recognizing that the wrong control rod was
withdrawn. For this situation, procedure Ops Manual B.01.05-05, Section 3, Recovery From an
Inadvertent Control Rod Withdrawal was the correct procedure to use. In this procedure, the
control rod that was inadvertently withdrawn is inserted to its previous position without obtaining
information on fuel thermal limits. The Bases for this procedure states that check thermal
margins prior to inserting the mispositioned rod would result in an unnecessary delay in
returning the reactor to its desired configuration. Therefore, distractor C, Insert the control
rod to its previous position without obtaining information on fuel thermal limits, was retained as
the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 70
Which of the following describes the MCPR safety limit?
A. With the reactor steam dome pressure < 785 psig AND core flow < 10% rated core flow:
Thermal Power shall be 25% RTP.
B. With the reactor steam dome pressure < 785 psig OR core flow < 10% rated core flow:
Thermal Power shall be 25% RTP.
C. With the reactor steam dome pressure 785 psig AND core flow 10% rated core flow:
MCPR shall be 1.12 for two recirculation operation or 1.10 for single recirculation
loop operation.
D. With the reactor steam dome pressure 785 psig OR core flow 10% rated core flow:
MCPR shall be 1.12 for two recirculation operation or 1.10 for single recirculation
loop operation.
Answer: B
Applicant Comment:
An applicant commented that there is no correct answer.
There was no correct answer. Answer B is a Reactor Core Safety Limit (SL) (or conditions) at
which Minimum Critical Power Ratio (MCPR) limits do not apply. It is NOT the MCPR limit. The
MCPR limits are 1.10 for two loop operations or 1.12 for single loops operations. The first two
distractors (A and B) only identifed when MCPR limits may or may not apply. The question
specifically asked which of the following described the MCPR safety limits, NOT when the
MCPR limits did not apply. The applicant thought the values were reversed in the chosen
answer, and went with it, because the first two distractors (A and B) did not answer the
question.
Facility Proposed Resolution:
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 = 1.00).
In the APPLICABLE SAFETY ANALYSES of Technical Specification 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 Anticipated Operational Occurrences
(AOOs). The reactor core SLs are established to preclude violation of the fuel design criterion
Enclosure 3
Post Examination Comments and Resolutions
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.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to delete the question from the examination. The question asked Which of the
following describes the MCPR safety limit? In the Bases for the Reactor Core SLs
(Section B 2.1.1), the terminology MCPR SL is only associated with reactor core SL 2.1.1.2,
which is:
2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated
core flow:
MCPR shall be 1.10 for two recirculation operation or 1.12 for single
recirculation loop operation.
The above definition of the MCPR SL is not associated with any of the distractors.
In Section B 2.2.1.1, distractor B (the original correct answer) is associated with reactor core
SL 2.1.1.1, Fuel Cladding Integrity. The BACKGROUND for the Bases for the Reactor Core
SLs, does state that the fuel cladding SL is defined with a margin to the conditions that would
produce onset of transition boiling (i.e., MCPR = 1.00). However, the Bases does not define
Section 2.1.1.1 as representing the MPCR SL. Only section 2.1.1.2 in the Bases is defined as
representing the MPCR S
- L. Therefore, there is no correct answer, and it was decided to
delete the question from the examination.
NOTE: The facility was contacted about the NRC Resolution stated above. The licensee
acknowledged the NRC Resolution.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 79
The reactor was operating in MODE 1 with the following conditions present:
- 11 Instrument Air Compressor, K-1A, is in LEAD
- 14 Instrument Air Compressor, K-1D, is in STBY
- 13 Instrument Air Compressor, K-1C, is taggout out for routine maintenance
- It is the 1900-0700 shift
- The Clearance Order Holder is NOT on site
- Instrument air pressure is 95 psig
If K-1A fails and K-1D is determined to NOT be able to maintain long term instrument air header
pressure above the low pressure alarm point, what actions are required for the CRS to direct
the removal of Danger Tags and place K-1C in service? (Assume instrument air pressure is
lowering 1 psig per 30 minutes.)
A. Immediately sign off the Clearance Order Holder(s) and direct the tags be removed.
- B. With concurrence of the Shift Manager, immediately sign off the Clearance Order
Holder(s) and direct the tags be removed.
the tags be removed.
with permission from the WCC, direct the tags be removed.
Answer: C
Applicant Comment:
An applicant commented that answer D should also be accepted as correct.
Per procedure FP-OP-TAG-01, Fleet Tagging, the Note in section 3.7 stated that Ops Shift
Supervision may be carried out by an Senior reactor Operator (SRO) or persons designated by
Ops Management in the Work Control center (WCC) acting for Ops shift supervision, provided
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, the applicant asked for the WCC permission, this told the applicant
the applicant that a WCC person was available to perform the task. For normal day to day
operation at the plant, as the CRS, the applicant stated that he would direct the WCC person to
perform the necessary actions and have the tags removed; at a minimum, since most work
started and ended with the WCC, the applicant stated that he would seek their concurrence
and/or permission.
Enclosure 3
Post Examination Comments and Resolutions
Facility Proposed Resolution:
The question grading for the exam should not change. The question specifically asked 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.
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 question asked what actions are
required for the CRS to direct the removal of Danger Tags for the K-1C Instrument Air
Compressor, in order to place the K-1C compressor in service, during a condition where
instrument air pressure is lowering, no other Instrument Air Compressor is available, and the
Clearance Order Holder is not on site.
An applicant commented that distractor D should also be accepted as the correct answer.
The applicant stated that he would seek the WCC concurrence and/or permission prior to
directing that the Danger Tags be removed, since for normal day to day operation at the plant,
as the CRS, the applicant stated that he would direct the WCC person to perform the necessary
actions and have the tags removed, since most work started and ended with the WCC. The
applicant also stated that he assumed (based on the the description provided in distractor D
about obtaining permission from the WCC) that a WCC person was available to perform the
task.
However, the question specifically asked for the actions required for the CRS to direct removal
of the tags and place the compressor in service. Although the WCC, if available, may also
perform the actions to contact the Clearance Order Holder(s) and if they cannot be reached,
direct the tags be removed, 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 to perform these actions. The applicant also stated that he assumed (based on the
the description provided in distractor D about obtaining permission from the WCC) that a
WCC person was available to perform the task. The question stem stated that it was the
1900-0700 shift. The availability of the WCC was a condition not specified in the question
stem. In NUREG-1021, Appendix E, Policies and Guidelines for Taking NRC Examinations, it
stated, in part, that When answering a question, do not make assumptions regarding
conditions that are not specified in the question .... The applicants were briefed verbatim on
the contents of NUREG-1021, Appendix E prior to the administration of the written examination,
and were provided a copy of Appendix E. The applicants did not ask for a clarification of the
question during the administration of the written examination. Since there was no discussion in
the question stem concerning the availability of the WCC, and since permission of the WCC is
not required to perform these actions (to contact the Clearance Order Holder(s) and if they
cannot be reached, to direct the tags be removed), distractor B was retained as the only
correct answer.
Enclosure 3
Post Examination Comments and Resolutions
Question Number 87
During an ATWS event with SBLC injecting, the OATC reports SBLC tank level indicates
400 gallons. Predict the impact of this SBLC system condition on the plant and what actions
you would direct as CRS associated with this SBLC condition?
tank with boron solution.
secure the SBLC pump.
- C. The reactor will remain shutdown under all conditions, secure the SBLC pump when
level indicates 0 inches.
D. RPV water level must be retained between -126 inches and -33 inches until Cold
Shutdown Boron Weight has been achieved, secure the SBLC pump when level
indicates 0 inches.
Answer: C
Applicant Comment:
The applicant recommended removal of this question from the examination due to there being
no correct answers.
The applicant stated that Answer A was the most correct, since there is procedural direction to
refill the Standby Liquid Control (SBLC) tank with boron. Answer B was not correct, because
Cold S/D Boron weight is required to depressurize. Answer C was not correct, because
securing SBLC pump when reactor level indicates 0 inches was wrong. Securing the SBLC
pump had nothing to do with Reactor Level. Answer D was not correct, because Hot S/D
Boron weight is required to raise level and securing the SLBC pump had nothing to do with
Reactor Level at 0 inches.
Facility Proposed Resolution:
The question should be removed from the exam due to there being no correct answer as
worded. The correct answer C, as stated on the exam, stated secure SBLC pump when
level indicates 0 inches. This was not specific as to which level indicated 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 was referencing reactor water level. In this
case, there was 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.
Enclosure 3
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to delete the question from the examination. The original correct answer (distractor
C) stated secure SBLC pump when level indicates 0 inches. However, as described by
the applicant and the facility, this statement was not specific as to which level indicated
inches. The SBLC level instruments are in units of gallons, not inches. This inconsistency
of units could lead the applicants (as per the applicant comment) to assume that in distractors
C and D, the level indication specified of 0 inches was referencing reactor water level, and
not a SBLC tank level of 0 gallons. In any case, distractor C is incorrect as written to state to
secure the SBLC pump when level indicates 0 inches instead of secure the SBLC pump
when level indicates 0 gallons. Therefore, there is no correct answer, and it was decided to
delete the question from the examination.
Enclosure 3
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession #ML070660539.
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