IR 05000315/2006301
| ML061020640 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/04/2006 |
| From: | Hironori Peterson Operations Branch I |
| To: | Nazar M Indiana Michigan Power Co |
| References | |
| IR-06-301 | |
| Download: ML061020640 (29) | |
Text
ril 4, 2006
SUBJECT:
DONALD C. COOK NUCLEAR POWER PLANT, UNITS 1 AND 2 NRC INITIAL LICENSE EXAMINATION REPORT 05000315/2006301(DRS); 05000316/2006301(DRS)
Dear Mr. Nazar:
On February 20, 2006, the NRC completed initial operator licensing examinations at your Donald C. Cook Nuclear Power Plant. The enclosed report presents the results of the examination which were discussed on February 17 and March 6, 2006, with Mr. Weber and Mr. Fleetwood, respectively, and with other members of your staff.
The NRC examiners administered initial license examination operating tests during the weeks of February 6 and 13, 2006. Members of the D.C. Cook Nuclear Plant Training Department administered an initial license written examination on February 20, 2006, to the applicants.
Nine Senior Reactor Operator-Instant (SRO-I) and five Senior Reactor Operator-Upgrade (SRO-U) applicants were administered license examinations. The results of the examinations were finalized on March 23, 2006. Ten applicants passed all sections of their corresponding examinations and were issued senior reactor operator (SRO) licenses. Four applicants failed the written examination and will not be issued licenses.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records 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 Nos. 50-315; 50-316 License Nos. DPR-58; DPR-74
Enclosures:
1. Operator Licensing Examination Report 05000315/2006301(DRS);
2. Simulation Facility Report 3. Post Examination Comments and Resolutions 4. Written Examinations and Answer Keys (RO/SRO)
REGION III==
Docket No: 50-315; 50-316 License No: DPR-58; DPR-74 Report No: 05000315/2006301(DRS);
Licensee: Indiana Michigan Power Company Facility: Donald C. Cook Nuclear Power Plant Location: 1 Cook Place Bridgman, MI 49106 Dates: February 6 through February 20, 2006 Examiners: N. Valos, Chief Examiner C. Moore, Examiner B. Palagi, Examiner Approved by: Hironori Peterson, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000315/2006301(DRS); 05000316/2006301(DRS); 02/06/06 - 02/20/06; Donald C. Cook
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:
- Fourteen examinations were administered (nine Senior Reactor Operator-Instant and five Senior Reactor Operator-Upgrade).
- Ten applicants passed all sections of their corresponding examinations and were issued senior reactor operator (SRO) licenses. Four applicants failed the written examination and will not be issued licenses.
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA5 Other
.1 Initial Licensing Examinations
a. Inspection Scope
The NRC examiners conducted an announced operator licensing initial examination during the weeks of February 6 and February 13, 2006. 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 6 through February 17, 2006. The facility licensee administered the written examination on February 20, 2006. Nine Senior Reactor Operator-Instant (SRO-I), and five Senior Reactor Operator-Upgrade (SRO-U) 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 eight post-examination comments (5 RO; 3 SRO comments) were submitted by the applicants and station training department personnel on February 28, 2006. In addition, there was one clarification made to a question during the administration of the examination. The results of the NRCs review of the comments are documented in 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."
One post-examination comment was identified by the NRC examiners during the administration of the operating test. The results of the NRCs review of the comment is documented in Enclosure 3, Post Examination Comments and Resolutions.
Examination Results Ten applicants passed all sections of their corresponding examinations and were issued senior reactor operator (SRO) licenses. Four applicants failed the written examination and will not be issued licenses.
.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.
However, during the time period of the development of the initial license examination, the licensee notified the NRC of one issue which had the potential to affect the integrity of the operating examination.
The issue associated with examination security was identified by the licensee on November 18, 2005, when it was discovered that one Operations person had failed to sign the NRC Examination Security Form ES-201-3. The individual was involved in the validation of four dynamic simulator scenarios on November 6, 2005, with two other Operations personnel. Prior to the validation, the three individuals were briefed on NRC security rules using form ES-201-3 and agreed to abide by the exam security requirements. Form ES-201-3 was made available at the exam security briefing to obtain the individuals signatures and acknowledgment of their participation in the briefing. However, verification was not performed that all three individuals had signed the exam security form. Following identification of the issue, the individual was immediately contacted and a signature on form ES-201-3 was obtained following a brief discussion and recognition of the omission. The individual stated that he abided by the requirements of the exam security agreement.
The licensee documented this issue in the corrective action program as Condition Reports Numbers 05326057 and 05326066. The NRC examiners were appropriately notified of the issue. The examiners reviewed the licensees investigation and assessed the issue for a possible violation of 10 CFR 55.49, Integrity of Examinations and Tests.
The examiners determined that no actual examination compromise had occurred. The violation was considered minor in nature and was not subject to enforcement action in accordance with NRC enforcement policy.
4OA6 Meetings
Exit Meeting The chief examiner presented the examination teams preliminary observations and findings with Mr. L. Weber and other members of the licensee management on February 17, 2006. A subsequent exit via teleconference was held on March 6, 2006, with Mr. I. Fleetwood 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
- L. Weber, Plant Manager
- R. Lingle, Operations Manager
- S. Stewart, Training Manager
- R. Gillespie, Operations Training Manager
- L. Bush, Site Senior License
- M. Scarpello, Nuclear Regulatory Assurance Supervisor
- I. Fleetwood, Initial License Training Supervisor
NRC
- N. Valos, Chief Examiner
- B. Palagi, Examiner
- C. Moore, Examiner
- B. Kemker, 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: Donald C. Cook Nuclear Plant
Facility Licensee Docket Nos. 50-315; 50-316
Operating Tests Administered: February 6 - February 17, 2006
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
None
Post Examination Comments and Resolutions
Written Examination
Question Number 6
Which of the following is the basis for establishing/maintaining S/G Narrow Range level
between 27%-50% (non-adverse containment) for small or intermediate LOCA's?
A. Maintains a static head of water to reduce any existing S/G tube leakage.
B. Ensures adequate feed flow or S/G inventory to ensure a secondary heat sink.
C. A RCP may have to be started if 02-OHP-4023-FR-C.1 is entered later in the event.
D. Maintains the water level above the top of the U-tubes to prevent depressurizing S/G.
Answer: B
Applicant Comment:
An applicant commented that answer D should also be accepted as correct.
Although answer B does represent the overall purpose of the entire EOP step (E-1, Loss of
Reactor or Secondary Coolant, Step 3: Check Intact SG Levels), the question was specific to
the basis for the 27-50% level established in the step. Per the Background Document for
2-OHP-4023-E-1, the setpoint for SG level (27-50%) was selected since it was above the
automatic Auxiliary Feedwater (AFW) actuation setpoint (and thus ensures no subsequent
automatic actuation of AFW components), and was also consistent with the setpoint used in the
E-3 series procedures to prevent AFW actuation during Steam Generator Tube Rupture
(SGTR) events.
Answer D should also be accepted, since this answer reflects the basis specified in the
Background Document (i.e., set to prevent auto actuation of AFW and be consistent with the
E-3 series procedures).
Facility Proposed Resolution:
The facility management disagreed with the applicant. The facility recommended that answer
B be retained as the only correct answer.
The SG level setpoint (27%) is conservatively higher than required to meet minimum heat sink
requirements. This value was chosen to prevent auto actuation of AFW. The upper level
(50%) ensures adequate inventory with level readings on span and provides for monitoring level
in the SGs to detect tube failures. 12-OHP-4023-E-3, Steam Generator Tube Rupture does
not require an upper SG level control band. The SG lower level setpoint (13%) was to minimize
the potential for SG overfill and to maintain water level in the ruptured SGs above the top of the
U-tubes in order to promote thermal stratification to prevent ruptured SG depressurization.
Post Examination Comments and Resolutions
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 purpose of the question was to ask the
basis for maintaining SG narrow range level between 27% and 50% for small or intermediate
size loss of coolant accidents (LOCAs). The applicant stated that distractor D should also be
accepted as a correct answer. Distractor D stated: Maintains the water level above the top
of the U-tubes to prevent depressurizing the S
- G. However, this statement was the purpose of
maintaining the water level in the ruptured SG(s) for a SGTR event, and not for a LOCA event.
The purpose of 12-OHP-4023-E-3, Steam Generator Tube Rupture, step 4: Check Ruptured
SG(s) Level is To establish and maintain a water level in the ruptured SGs above the top of
the U-tubes to promote thermal stratification to prevent ruptured SG depressurization. For a
LOCA, the purpose of 12-OHP-4023-E-1, Loss of Reactor or Secondary Coolant, step 3:
Check Intact SG Levels is, in part, To ensure adequate ... SG inventory to ensure a
secondary heat sink for small and intermediate size LOCAs. Since distractor D described
the purpose of maintaining the water level in the ruptured SG(s) for a SGTR event, and not for
a LOCA event, distractor D will not be accepted as a correct answer; distractor B was
retained as the only correct answer.
Post Examination Comments and Resolutions
Question Number 7
Unit 2 Reactor Startup is in progress with Reactor Power at 2% and rising.
The following conditions exist:
- RCP No. 3 Lower Bearing water temperature is 208°F and stable.
- RCP No. 3 Motor Bearing temperature is 187°F and stable.
- RCP No. 3 Seal Leakoff temperature is 179°F and stable.
- RCP No. 3 Motor Temperature is 148°C and rising.
- Annunciator 107 Drop 40, RCP Motor Overheated - LIT
Which ONE of the following operator actions MUST be taken based upon these
conditions?
A. Do NOT trip the reactor. Trip the No. 3 RCP. Power must be maintained less than 5%
(Mode 2).
- B. Initiate reactor shutdown per 02 OHP 4021.001.003, Power Reduction and trip the
No. 3 RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- C. Manually trip the reactor, Enter 02 OHP 4023.E 0, Reactor Trip or Safety Injection, perform
immediate actions, then trip the No. 3 RC
- P.
D. Immediately Trip the No. 3 RCP. Initiate reactor shutdown per 02 OHP 4021.001.003,
Power Reduction and be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Answer: B
Applicant Comment:
An applicant commented that answer C should also be accepted as correct.
Answer B and the associated Knowledge and Abilities (K/A) statement When to secure
Reactor Coolant Pumps (RCPs) on high stator temperatures is the design of this question.
The conditions specified in the question stem also include elevated temperatures on RCP motor
bearings, which makes the determination of the RCP status more complex than just an elevated
stator temperature on the RCP motor.
Answer B is based on the specific procedure step requirements contained in procedure
2-OHP-4022-002-001, Malfunction of a Reactor Coolant Pump, step 10. However, step 8 of
the procedure deals with elevated RCP lower pump bearing temperature (i.e., greater than
175oF) and in conjunction with the step 9 criteria for determining RCP motor bearing damage
(i.e., greater than 185oF), requires a Shift Manager (SM) determination to allow for continued
RCP operation. Without SM permission for continued operation, the procedure required a plant
shutdown with no time requirement given for tripping the RCP.
Post Examination Comments and Resolutions
Based on the fact that the RCP was showing evidence of both pump problems and motor
problems, both conditions needed to be considered when making a decision on what actions to
take. Using the information provided in the question stem, and knowledge of pump and motor
characteristics, RCP vibrations would be expected to be rising as well. Therefore, although no
single immediate RCP Trip criteria was met (per step 1 of 02-OHP-4022-002-001), trip criteria
was met per step 9 of the procedure, based on multiple RCP parameters being elevated.
Without having the procedure in hand, based on training and Conservative Operational
Decision Making, the prudent decision was to trip the Reactor, then trip RCP No. 3.
Facility Proposed Resolution:
The facility management disagreed with the applicant. The facility recommended that answer
B be retained as the only correct answer.
The question stem stated that the parameters associated with the RCP were stable with the
exception of Motor temperature which was 148ºC and rising. All parameters were below values
which required an immediate Reactor trip and trip of the affected RCP. Increased Motor
temperatures do not require an immediate trip of the Reactor and trip of the RCP. Since the
plant was not analyzed/licensed to operate with less than 4 RCPs in operation, distractor B
was the correct answer. The facility therefore recommended that answer B be retained as 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 made the assumption that
RCP No. 3 vibrations were rising to support his proposed resolution to manually trip the
Reactor and then trip the affected RCP. This condition was not specified in the question stem.
In NUREG-1021, Appendix E, Policies and Guidelines for Taking NRC Examinations, it was
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 rising RCP No. 3 vibrations, distractor B was retained as the
only correct answer.
Post Examination Comments and Resolutions
Question Number 24
The following conditions exist:
A plant startup is in progress.
Reactor power is 25%.
All control systems are in AUTOMATIC.
Control Bank D rods are at 150 steps.
The lower detector for power range nuclear instrument channel N41 fails HIGH, causing
channel total power indication to increase to 70%.
Assuming no operator action is taken, what will be the response of the rod control system?
A. Rods will not move.
B. Rods will drive all the way in.
C. Rods will drive in some distance then stop.
- D. Rods will insert, and then withdraw.
Answer: D
Applicant Comment:
An applicant commented that the question should be deleted from the examination.
An applicant commented that the question should be eliminated, since the question was invalid.
The applicant stated that the Reactor does not respond as specified in the question stem, and
none of the answers were correct.
A scenario was run on the D.C. Cook simulator under the following conditions:
Power was raised to 24% power with Control bank D at 179 steps and Tavg/Tref was matched
(simulator setup IC15 was used as the starting point, the simulator setup was then saved to
IC282).
The specified failure (lower detector for power range nuclear instrument channel N41 failed
high) was then inserted. The simulator responded as follows:
- N41 channel total power indication failed to 120% (full scale HIGH)
- Control rods started stepping in, and Tavg started lowering.
- Rods were still stepping in (less than 128 steps on Control Bank D, and Control Bank C
was also stepping in) when the Reactor tripped on Low Pressurizer Pressure.
Facility Proposed Resolution:
The facility management disagreed with the applicant. The facility recommended that answer
D be retained as the only correct answer.
Post Examination Comments and Resolutions
The N41 power range failure was re-performed on the simulator to more closely match the
conditions in the question stem (i.e., the N41 lower detector was failed high, causing channel
total power indication to increase to 70% power). The simulator response indicated the control
rods initially stepped inward due to the power mismatch created by the failure. Once this signal
decayed, the control rods withdrew. The cycle of control rods stepping inward and then
withdrawing occurred again at a slower rate but with less magnitude. The question did not ask
for an integrated plant response, but only the response of the control rod system. The facility
therefore recommended that answer D be retained as 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 question stem did not state that the
N41 lower detector was failed offscale high, instead the question was qualified to state that the
lower detector was failed high to cause channel total power indication to increase to 70%
power. The N41 power range failure was performed on the simulator to closely match the
conditions in the question stem (i.e., the N41 lower detector was failed high, causing channel
total power indication to increase to 70% power). The simulator response indicated the control
rods initially stepped inward due to the power mismatch created by the failure, and once this
signal decayed, the control rods withdrew. This sequence of rod movement following the N41
failure corresponded to that stated in distractor
- D. Therefore, the question will not be deleted
from the examination, and distractor D was retained as the only correct answer.
Post Examination Comments and Resolutions
Question Number 43
Fire in the Unit #1 Control Room Cable Vault has resulted in loss of equipment control and
normal habitability. The Control Room has been evacuated.
Which ONE of the following would be used to monitor Reactor Power from outside the control
room?
A. Source Range - N21
B. Source Range - N23
C. Wide Range - N21
D. Wide Range - N23
Answer: B
Facility Comment:
The question required the applicant to know the specific nuclear instrumentation (NI) located
outside the Control Room to monitor Reactor Power in the event of a Control Room evacuation.
There were two choices: (1) N21 or (2) N23. N23 provided indication outside the control room.
The question also required the applicant to know the scale of the NI: (1) Source Range or
(2) Wide Range. This portion of the question was irrelevant. In the event of an actual event,
knowing that Reactor Power can be read outside the Control Room is all that is relevant. The
job can still be performed safely and effectively without knowing the specific scale of the
instrument.
Facility Proposed Resolution:
The facility management recommended deleting this question from the examination.
NRC Resolution:
Upon review of the question, the facility comment, and the facility proposed resolution, it was
decided to accept only the original correct answer for the following reasons: (1) requiring an
applicant to know what NI was available outside the Control Room is conceptually no different
than requiring the applicant to know what NI is available in the Control Room, (2) if the applicant
is performing activities in the plant that could affect reactivity during a control room evacuation
(e.g., dumping steam, adjusting auxiliary feedwater, boration of the reactor coolant system), the
applicant is expected to know what NI was available to the operator who is monitoring reactor
power locally (e.g., what the NI at the panel is capable of detecting), (3) it was expected that the
applicant would understand that the local NI would not have a range that would include
indication up to 100% power (as does Wide Range) for a Control Room evacuation event, in
which the reactor is expected to be shutdown, and (4) the question fully satisfies the Knowledge
and Abilities (K/A) statement of Knowledge of NI system design feature(s) and/or interlocks
which provide for the following: Reading of source range/intermediate range/power range
outside control room (with a high K/A importance rating of 3.9/4.0 for RO/SRO). Therefore, the
question will not be deleted from the examination and distractor B was retained as the only
correct answer.
Post Examination Comments and Resolutions
Question Number 50
The following conditions exist:
- Unit 1 is at 100% power and stable.
- Steam Generator Level Controls are in AUTOMATI
- C.
- Steam Generator #12 Steam Flow Channel 1, 1-MFC-121, is selected to the Steam
Generator Level Control System.
A blown fuse causes 1-MFC-121 to fail offscale low.
Which ONE of the following describes the expected plant response?
(Assuming no operator action)
The Steam Generator Level Control system will:
A. initially lower feed flow and then slowly return #12 SG level to approximately program level.
B. automatically transfer the # 12 FW Regulating Valve Controller to Manual to maintain the
current valve position.
C. initially raise feed flow and then slowly return #12 SG level to approximately program level.
- D. lower feed flow to #12 SG to 0 pph, resulting in a Reactor Trip.
Answer: B
Facility Comment:
After a review of associated electrical diagrams and discussion with the System Engineer, it
was determined that there are multiple fuses in the control system that could fail. These fuses
may be upstream or downstream of the square root extractor for the Steam Flow indication.
The System Engineer and electrical diagrams support that the location of the blown fuse will
have an impact on the system response. A failure of a fuse downstream of the square root
extractor will result in answer B being correct (with the Taylor Controller in Manual and
maintaining the current (prior to failure) conditions). A failure of a fuse upstream of the square
root extractor will not shift the Taylor controller to Manual at the current (prior to the failure)
position, and will result in Steam Flow going to 0 pph. The result of Steam Flow at 0 pph will
drive the Feedwater Regulating Valve closed and result in a Steam Generator (SG) Low-Low
Level Reactor Trip.
Facility Proposed Resolution:
The facility management recommended that two answers be accepted, both B and D. The
stem of the question did not designate whether the fuse was upstream or downstream of the
square root extractor for Steam Flow. As such, the potential existed for two possible answers
depending on the assumptions of the applicant. If the applicant assumed that the blown fuse
was downstream of the square root extractor for Steam Flow, then answer B was correct. If
the applicant assumed that the blown fuse was upstream, then answer D was correct.
Post Examination Comments and Resolutions
NRC Resolution:
Upon review of the question, the facility comment, and the facility proposed resolution, it was
decided to accept the facilitys comment and accept both answer B and D as correct
answers.
From additional information obtained from the facility, it was determined that the fuses both
upstream and downstream of the square root extractor (SRE) for the Steam Flow channel were
actually microfuses located on solid state circuit cards. The facility had documented via
Condition Report 03206022 (written on July 25, 2003) that the output of the SRE was clamped
to prevent the output signal from going below zero (0%) no matter what the input signal was.
The controller fail-over circuitry, that would automatically transfer the Feedwater (FW)
Regulating Valve controller to MANUAL to maintain the current valve position, required a signal
of minus 8% (-8%) on the output of the SRE to actuate. A blown fuse upstream of the SRE,
which would cause the input signal to the SRE signal to fail offscale low, would not cause the
controller fail-over circuitry to actuate, since the output signal from the SRE would be 0% for
this case. Thus, for the case of a blown fuse upstream of the SRE, distractor D (i.e., lower
feed flow to #12 SG to 0 pph, resulting in a Reactor Trip) would be the correct answer.
If the blown fuse was downstream of the SRE, the controller fail-over circuitry would sense a
signal offscale low (i.e., -8% input), which would actuate the controller fail-over circuitry, and
would automatically transfer the #12 FW Regulating Valve controller to MANUAL to maintain
the current valve position. Thus, for the case of a blown fuse downstream of the SRE,
distractor D would be the correct answer.
Because the question stem did not provide the specific location of the fuse that had failed, an
applicant could assume the failed fuse was upstream or downstream of the SRE and arrive at
B or D as the correct answer. Therefore, the answer key was modified to accept both B
and D as correct answers.
Post Examination Comments and Resolutions
Question Number 57
The operator incorrectly opens the breaker labeled "7.5 KVA Static Inverter Channel IV" on 250
VDC distribution panel "MCAB." The operator realizes the mistake and immediately recloses
the breaker.
Which ONE of the following describes the effect of these actions, if any?
A. The alternate power source to the CRID Inverter will be lost when the breaker is reclosed.
The CRID will transfer to the 120 VAC from the Regulating Transformer.
B. The alternate power source to the CRID Inverter will be lost. No automatic action will occur
when the breaker is reclosed. The auto transfer lockout must be reset at the inverter.
C. The normal power source to the CRID Inverter will be lost so it will auto transfer to the
alternate source. When the breaker is reclosed, it will auto transfer to the normal source.
D. The normal power source to the CRID Inverter will be lost so it will auto transfer to the
alternate source. When the breaker is reclosed, the auto transfer lockout must be reset at
the inverter.
Answer: C
Facility Comment:
During administration of the written examination, an applicant asked the facility a question for
clarification of distractor D. The applicant asked if auto transfer occurred after the breaker
was reclosed (then auto transfer was locally reset).
Facility Proposed Resolution:
The facility provided a clarification that no automatic action would occur when the breaker was
reclosed. Distractor D was clarified to state: The normal power source to the CRID Inverter
will be lost so it will auto transfer to the alternate source. No automatic action will occur when
the breaker is reclosed. The auto transfer lockout must be reset at the inverter.
NRC Resolution:
Upon review of the question, the facility comment, and the facility proposed resolution, it was
decided to accept the clarification to distractor D. The clarification to distractor D still
retained distractor D as an incorrect answer. Distractor C was retained as the only correct
answer.
Post Examination Comments and Resolutions
Question Number 83
The plant is at 95% power.
- 1200 on April 3, the 1S SI pump is declared inoperable.
- 1430 on April 4, the 1W Centrifugal Charging pump is declared inoperable.
- 1430 on April 5, the 1S SI pump is restored to OPERABLE status.
Including any extensions that are permitted by TS, which one of the following describes the
LATEST time and date to restore the 1W Centrifugal Charging pump to OPERABLE status
without requiring a unit shutdown?
A. 1200 on April 6
B. 1200 on April 7
C. 1430 on April 7
D. 1430 on April 8
Answer: A
Facility Comment:
This question should have provided references for use in answering the question (e.g., a copy
of the applicable Technical Specifications (TS)). Failure to provide references is inconsistent
with expectations for memorization of Limiting Condition for Operation (LCO) Required Action
items with long time periods such as 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and rules of usage for application of LCO
extensions. It is also inconsistent with the operator Lesson Plan objective which states, in part,
Given a specific plant system or component use Technical Specifications to determine
(emphasis added).
This question requires the applicant to apply rules of usage regarding LCO extension time for
subsequent inoperability of components from memory. Per Operating Licensing Program
Feedback question 401.11 regarding application level questions the following answer is found.
The NRC does not expect operators to memorize the TS, nor does it endorse operating
the plant from memory. However, the NRC does expect operators to recognize TS
entry conditions immediate actions and (in the case of senior operators) bases when
presented in a multiple choice format on the written examination. If they do not
compromise the integrity of other questions on the exam, it is acceptable to provide
extracts from the TS to the license applicants for use in answering application-level
questions.
Facility Proposed Resolution:
The facility management recommended deleting this question from the examination. Failure to
provide the necessary references placed the applicant at an unfair disadvantage and caused
Post Examination Comments and Resolutions
excessive reliance on memory. Since the references were not provided, the facility
recommended deleting this question from the examination.
NRC Resolution:
Upon review of the question, the facility comment, and the facility proposed resolution, it was
decided to delete the question from the examination. The NRC agreed that applicable
Technical Specifications should have been provided to the applicants to answer the question
(1) due to the long time periods (such as 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) associated with the LCO Required
Actions, and (2) so that application (as opposed to memorization) of the rules of usage for LCO
extensions was tested.
Post Examination Comments and Resolutions
Question Number 84
The following plant conditions exist on Unit 2:
- The East CCW HX is in service with the West CCW Pump running.
- CCW Surge Tank level is stable.
- CRS-4301, East CCW HX Radiation Monitor, generates an External Failure Alarm due to a
faulty flow switch
Which ONE of the following describes the response of the CCW system and the required
actions, if any, for this condition?
A. No automatic actions will occur since the West CCW pump is running. No Lineup changes
are required, operation in this condition is allowed indefinitely. The CCW system remains
- B. No automatic actions will occur since the CRS-4401, West CCW HX Radiation Monitor is
still functioning. The West CCW HX must be aligned so the 2-CRV-412 Vent Valve will
automatically close on a high radiation signal. The East CCW HX must be declared
- C. 2-CRV-412, CCW Surge Tank Vent Valve, will automatically close. The West CCW HX
must be aligned so the 2-CRV-412 Vent Valve may be reopened. The East CCW HX must
be declared Inoperable.
- D. 2-CRV-412, CCW Surge Tank Vent Valve, will automatically close. No Lineup changes are
required, operation in this condition is allowed indefinitely. The CCW system remains
Answer: D
Applicant Comment:
An applicant commented that answer C should also be accepted as correct OR that the
question should be deleted from the examination.
Answer C should also be accepted as correct, since with valve 2-CRV-412 closed, the
procedural requirements for declaring the East CCW Loop OPERABLE (including the East
CCW Heat Exchanger) are not met.
The question should be deleted from the examination, since the stated answer was incorrect
since it specifically stated that operation with 2-CRV-412 closed was allowed indefinitely, and
that the Component Cooling Water (CCW) system would remain OPERABLE. Without a lineup
change (re-opening of 2-CRV-412), the requirements of 2-OHP-4030-216-020E, East
Component Cooling Water Loop Surveillance Test would not be met and the East CCW Loop
would have to be declared INOPERABL
- E. Therefore, operation in this condition (with
2-CRV-412 closed) was not allowed indefinitely, and the key answer D was incorrect.
The purpose of 2-OHP-4030-216-020E, East Component Cooling Water Loop Surveillance
Post Examination Comments and Resolutions
Test is to demonstrate OPERABILITY of East CCW Loop in accordance with Technical
Specification (TS) Surveillance Requirement (SR) 3.7.7.1 and 5.5.6.
The Acceptance Criteria of the procedure (Section 5, step 5.2) specifically stated the following:
Step 5.2: East (spare) CCW Loop has been demonstrated OPERABLE per Tech Spec
SR 3.7.7.1 by completion of Lineup Sheet 1 (Step 4.54).
Step 4.54 stated:
Step 4.54: Verify Lineup Sheet 1, Unit Two East CCW Loop Flow Path Verification has
been completed by an individual that was independent from performance of the
test.
Lineup Sheet 1 (page 42 of 42, last valve in lineup) required 2-CRV-412, CCW Surge Tank
Vent valve to be in the OPEN position. This step may be N/A in Modes 5, 6 and defueled
when the Valve Lineup Sheet was used to satisfy Appendix R requirements. There was no
allowance to N/A this step/position when in Modes 1-4.
If a Lineup change is not made such that 2-CRV-412 is reopened, the requirements of
2-OHP-4030-216-020E, East Component Cooling Water Loop Surveillance Test (which is the
procedure used to verify compliance with SR 3.7.7.1) would not be met, and the East Train of
CCW would have to be declared inoperable.
Facility Proposed Resolution:
The facility management disagreed with the applicant. The facility recommended that answer
D be retained as the only correct answer.
Answer D stated that 2-CRV-412 may be left closed indefinitely. This was true as far as the
functionality of the system was concerned. Technical Specification S SR 3.7.7.1 requires each
CCW manual, power operated, and automatic valve in the flow path servicing safety related
equipment, that is not locked, sealed, or otherwise secured in position, be in the correct position
and has a note to specify isolation of CCW flow to individual components does not render the
CCW System inoperable. The CCW vent valve 2-CRV-412 is not in the flow path servicing
safety related equipment. Therefore, 2-CRV-412 could be left in the closed position and still
have the system remain OPERABLE. Discussions with the system engineers indicate that
there are additional tank protection features that protect the tank from collapse and
overpressurization. These features are shown on drawing OP-2-5135-37, Flow Diagram CCW
Pumps and CCW Heat Exchangers. The features are a check valve (which acts as a vacuum
breaker) and a safety valve (to prevent overpressurization) that are on separate connections to
the CCW Surge Tank. Therefore closure of CCW vent valve 2-CRV-412 does not cause the
CCW system to be inoperable and does not preclude indefinite operation. The facility therefore
recommended that answer D be retained as the only correct answer.
Post Examination Comments and Resolutions
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 Technical Specification definition of
OPERABLE requires, in part, that A system, subsystem, train, component, or device shall be
OPERABLE or have OPERABILITY when it is capable of performing its specified safety
function(s) .... The closure of CCW vent valve 2-CRV-412 does not cause the CCW system to
be incapable of performing any of its safety functions. In addition, although the Acceptance
Criteria of procedure 2-OHP-4030-216-020E, East Component Cooling Water Loop
Surveillance Test stated that the East (spare) CCW Loop has been demonstrated OPERABLE
per Tech Spec SR 3.7.7.1 by completion of Lineup Sheet 1 (Step 4.54), the converse is not
necessarily true. That is, if any component is not aligned as per Lineup Sheet 1 of the
procedure, this does not necessarily imply that the East CCW Loop is inoperable. Each specific
component that is not aligned as per Lineup Sheet 1 would need to be evaluated as to its affect
on the OPERABILITY of the CCW system. For the specific example in this question of CCW
vent valve 2-CRV-412 being in the closed position (contrary to the normally open position
required by Lineup Sheet 1), this alignment does not cause the CCW system to be incapable of
performing any of its safety functions and the CCW system remains OPERABL
distractor C will not also be accepted as a correct answer, the question will not be deleted
from the examination, and distractor D was retained as the only correct answer.
Post Examination Comments and Resolutions
Question Number 95
This question contains security-related information that is not publicly available.
Post Examination Comments and Resolutions
Operating Test
Administrative JPM A, Review Completed Shutdown Margin Calculation In Mode 1"
NRC Comment:
One post-examination comment was identified by the NRC examiners during the administration
of the operating test. The examiners identified that Administrative Job Performance Measure
(JPM) A, Review Completed Shutdown Margin Calculation In Mode 1" was required to be
changed to have the applicant identify the following as Critical Steps during a review of a
manual shutdown margin calculation for Mode 1: (1) the number of misaligned rods to be used
in the SHUTDOWN MARGIN calculation was zero (instead of one), (2) the total unavailable rod
worth was 825 (800 to 850) pcm [instead of 946 (920-972) pcm], and
(3) the Net Excess SHUTDOWN MARGIN was 1325 (1300 to 1350) pcm [instead of 1204
(1178-1230) pcm].
NRC Resolution:
The NRC determined that the number of misaligned rods to be used in the SHUTDOWN
MARGIN calculation was zero. Control rod D-8 in Control Bank D was misaligned, since rod
D-8 was at 200 steps, which was more than 18 steps from the Control Bank D step counter
position of 219 steps. However, procedure 01-OHP-4021-001-012, Determination of Reactor
Shutdown Margin, Attachment 2, step 4.2.2 stated: Enter the total number of rods which are
misaligned, including dropped rods, AND which violate the insertion limits of the CORE
OPERATING LIMITS REPORT (COLR). The insertion limit for Control Bank D at the given
100% power was 189 steps (from Figure 2, Rod Bank Insertion Limits Versus Thermal Power
(Four - Loop Operation) of the COLR). Since the given rod D-8 position of 200 steps did not
violate the insertion limit of 189 steps, the number of misaligned rods to be used in the
SHUTDOWN MARGIN calculation was zero (and not one as originally required by the JPM).
This change to zero for the number of misaligned rods to be used in the SHUTDOWN MARGIN
calculation cascaded throughout the JPM, such that the (2) the total unavailable rod worth was
25 (800 to 850) pcm [instead of 946 (920-972) pcm], and (3) the Net Excess SHUTDOWN
MARGIN was 1325 (1300 to 1350) pcm [instead of 1204 (1178 to 1230) pcm]. In addition, the
step to determine that the Net Excess SHUTDOWN MARGIN was 1325 (1300 to 1350) pcm
was made a Critical Step for the JPM to ensure that the Task Standard for the JPM was met.
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession #ML060860245.