IR 05000440/2007301
| ML071210359 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 04/26/2007 |
| From: | Hironori Peterson Division of Reactor Safety III |
| To: | Pearce L FirstEnergy Nuclear Operating Co |
| References | |
| 50-440/07-301 ER-07-301 | |
| Download: ML071210359 (29) | |
Text
ril 26, 2007
SUBJECT:
PERRY NUCLEAR POWER PLANT NRC INITIAL LICENSE EXAMINATION REPORT 05000440/2007301(DRS)
Dear Mr. Pearce:
On March 2, 2007, the Nuclear Regulatory Commission (NRC) completed administration of initial operator licensing examinations at your Perry Nuclear Power Plant. The enclosed report presents the results of the examination which were discussed on March 2 and March 15, 2007, with you and Mr. Evans, respectively, and with other members of your staff.
The NRC examiners administered initial license examination operating tests during the week of February 26, 2007. Members of the Perry Nuclear Power Plant Training Department administered an initial license written examination on March 5, 2007, to the applicants. Six senior reactor operator and three reactor operator applicants were administered license examinations. Two of the applicants were previously licensed reactor operators at the Perry Nuclear Power Plant. The results of the examinations were finalized on April 13, 2007. Four applicants failed the written examination and were issued proposed license denial letters. Five applicants passed all sections of their examinations resulting in the issuance of two senior reactor operator and one reactor operator licenses. In accordance with NRC policy, the licenses for the one applicant who passed the written examination with an overall score less than 82 percent, and one applicant who passed the senior reactor operator-only portion of the written examination with a score less than 74 percent, were withheld pending the outcome of any written examination appeal that may be initiated.
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. 50-440 License No. NPF-58
Enclosures:
1. Operator Licensing Examination Report 05000440/2007301(DRS)
2. Simulation Facility Report 3. Post Examination Comments and Resolutions 4. Written Examinations and Answer Keys (RO/SRO)
REGION III==
Docket No: 50-440 License No: NPF-58 Report No: 05000440/2007301(DRS);
Licensee: FirstEnergy Nuclear Operating Company (FENOC)
Facility: Perry Nuclear Power Plant, Unit 1 Location: Perry, Ohio Dates: February 26 through March 5, 2007 Examiners: M. Bielby, Chief Examiner C. Phillips, Examiner R. Walton, Examiner Approved by: Hironori Peterson, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000440/2007301(DRS); 02/26/2007 - 03/02/2007; FirstEnergy Nuclear Operating
Company; Perry Nuclear Power 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:
- Nine examinations were administered (six senior reactor operator and three reactor operator).
- Five applicants passed all sections of their examinations resulting in the issuance of two senior reactor operator and one reactor operator licenses. The licenses for two senior reactor operators who passed the written examination were withheld pending the outcome of any written examination appeal that may be initiated.
- Four 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 Perry Training Department trainers prepared the examination outline and developed the written examination and operating test. The NRC examiners validated the proposed examination during the week of January 29, 2007, at the Perry Nuclear Power Plant training building with the assistance of members of the facility licensee training staff.
Also, during the on-site validation week, the examiners audited at least 10% of the license applications for accuracy. The NRC examiners conducted an announced initial operator licensing examination during the week of February 26, 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 26 through March 2, 2007. The facility licensee administered the written examination on March 5, 2007. Six senior reactor operator and three 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 (9 RO; 1 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 March 12, 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 Five applicants passed all sections of their examinations resulting in the issuance of two senior reactor operator and one reactor operator licenses. In accordance with NRC policy, the licenses for the one applicant who passed the written examination with an overall score less than 82 percent, and one applicant who passed the senior reactor operator-only portion of the written examination with a score less than 74 percent, were withheld pending the outcome of any written examination appeal that may be initiated.
Four 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. However, during the time period of the development of the initial license written examination, the licensee notified the NRC of one incident which had the potential to affect the integrity of the written examination.
The incident associated with examination security was identified by the licensee on October 4, 2006, when they discovered that an email contained attachments that included the written examination, and were not password protected. During written examination development by the licensee and a contractor, the two individuals had been using password protected email attachments to resolve written examination comments.
The final email to be exchanged with attachments, including the entire written examination with comments and corrections, was being returned to the licensee from the contractor when the licensee identified the lack of password protection. The NRC examiners were appropriately notified of the incident. The licensee subsequently developed another written examination outline and prepared a second written examination. The licensee documented this issue in their corrective action program as Condition Report Number CR 06-7253. 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. B. Allen, Plant Manager, and other members of the licensee management on March 2, 2007. A subsequent exit via teleconference was held on March 15, 2007, with Mr. T. Evans, Training Manager, 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
- B. Allen, Plant Manager
- M. Alfonso, Chemistry Manager
- N. Bonner, Oversight Manager
- A. Cayia, Performance Improvement Initiative Director
- K. Cimorelli, Work Management Director
- D. Evans, Operations Manager
- T. Evans, Training Manager
- K. Howard, Manager, Design, Perry
- B. Huck, Engineering
- H. Kelley, Emergency Preparedness Manager
- T. Kledzik, Regulator Affairs Engineer
- M. Kuntz, Supervisor
- J. Lausberg, Manager, Regulatory Compliance
- W. OMalley, Licensed Operator Requalification Training Supervisor
- K. Russell, Corporate Engineer
- J. Shaw, Director, Nuclear Engineering
- S. Thomas, Manager, Radiation Protection
- M. Wesley, Performance Improvement Initiative Manager
NRC
- M. Bielby, Chief Examiner
- C. Phillips, Examiner
- M. Wilk, Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and 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: Perry Nuclear Power Plant
Facility Licensee Docket No.: 50-440
Operating Tests Administered: February 26 through March 2, 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 26, 2007, to change a variable for reactor physics core
parameter initialization and re-boot the computer to restore the correct trending
of the core parameters. CR 07-15186 was written to document the problem.
Enclosure 2
Post Examination Comments and Resolutions
RO Question 11:
The plant is operating at 100% reactor power when one Reactor Recirculation Pump trips. All
systems respond as designed to this event.
How will RPV Water Level initially respond and what is the reason for this response?
RPV Water Level will:
A. INCREASE due to the displacement of water into the downcomer by increased
steam voiding.
_____ B. INCREASE due to the continuing addition of feedwater at 100% rated feedwater
flow.
_____ C. DECREASE due to the lack of coolant velocity to sweep voids into the steam
separator.
_____ D. DECREASE due to the runback of feedwater pumps to minimum speed.
Answer: A
Applicant Comment:
Candidates believed that answer B was also correct. No documentation was available to
support candidate position. Event was run in simulator, feedwater flow follows steam demand
immediately.
Facility Proposed Resolution:
Reject candidate comment. No supporting material to defend position. Question to remain as
is.
NRC Resolution:
Upon review of the question, applicant comment, and facility proposed resolution, the NRC
determined the original answer A was the only correct answer. This question was used on the
Perry 2001 NRC exam. Perry Updated Final Safety Analysis Report (UFSAR), Section 15.3.1,
Revision 12, describes the reactor water level response and reason for the response when one
reactor recirculation pump (RRP) trips during plant operation at 100% reactor power. After the
RRP trip, the void fraction increases with reduced core flow, and displaces water into the
downcomer that results in an initially higher reactor water level. In addition, the licensee ran the
event on the simulator to verify that feedwater flow followed steam flow, and decreased due to
decreasing power. Answer B is an incorrect statement because feedwater and steam flow
follow power such that when the RRP trips, power decreases, steam and feed flow decreases,
Enclosure 3
Post Examination Comments and Resolutions
and feedwater addition does not continue at 100% flow. Based on the reviewed information,
the NRC determined that A is the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
RO Question 19:
The plant is in MODE 4 with the following plant conditions:
- RHR B is operating in the Shutdown Cooling Mode when the pump trips.
- RHR A is not available.
- No Reactor Recirculation Pumps are operating.
- No RWCU Pumps are operating.
- RPV Water Level is currently 230 inches on SHUTDOWN RANGE
ONI-E12-2, Loss of Decay Heat Removal would require (1) and reactor water temperature
monitoring would be from (2) per IOI-12 Maintaining Cold Shutdown.
_____ A. (1) raising RPV Water Level to greater than 250 inches
(2) bottom head drain
B. (1) starting a Reactor Recirculation Pump
(2) recirc suction
_____ C. (1) starting a Reactor Water Cleanup Pump
(2) bottom head drain
_____ D. (1) starting a Reactor Water Cleanup Pump
(2) recirc suction
Answer: B
Applicant Comment:
Candidates argue that Off Normal Instruction (ONI) supplemental actions are not required to be
known from memory. First part of answer is supplemental action memory question.
Facility Proposed Resolution:
Reject candidate comment. Supplemental actions are not required to be known from memory.
However, to answer this question students needed to know what are available temperature
monitoring methods in Mode 4. Once identified, all methods listed for ONI-E12 are viable, but
only one would provide for temperature monitoring.
Facility Analysis for Validity (Failure Greater Than 50%):
Seven of nine applicants missed the question. The question was evaluated as valid with no
changes required. Applicants missed the question due to a knowledge weakness regarding the
determination of reactor water temperature during Cold Shutdown conditions.
Enclosure 3
Post Examination Comments and Resolutions
NRC Resolution:
Upon review of the question, applicant comment, and facility proposed resolution, the NRC
determined the original answer B was the only correct answer. The applicants statement that
the first part of the answer is a supplemental action, and Off Normal Instruction (ONI)
supplemental actions are not required to be memorized, is not all together a correct statement.
The sequence of steps in the performance of supplemental actions is not required to be
memorized, but applicants are required to have a working knowledge of supplemental actions
based on their knowledge of the plants systems and integrated operating procedures. In other
words, an operator must have a conceptual working knowledge of procedural actions on what
and why an action is performed, but not the actual sequential performance of supplemental
actions. The related learning objective, Objective Link OT-3046-01(LP)-B, stated that the
applicant will be able to use an updated copy of IOI-12, Maintaining Cold Shutdown, discuss
general guidelines, and determine reactor water temperature. The question information is
taken from an abnormal (ie, ONI) procedure; however, the question does not require recall of
sequential procedure steps or actions, rather it requires an understanding of the concept that
temperature monitoring is only valid if a reactor recirculating pump is operating. The applicant
also needs to know what temperature monitoring methods are available in Mode 4, Cold
Shutdown. As a result, the question is valid, and answer B is the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
RO Question 22:
With the reactor operating at 100% power the Main Steam Isolation Valves closed.
No Control Rod motion occurred. No operator actions have been taken.
Which one of the following correctly describes the initiating signal and actuation(s) produced by
the Redundant Reactivity Control System (C22) which will reduce reactor power?
_____ A. High Reactor Pressure will immediately trip the Reactor Recirculation Pumps off.
B. High Reactor Pressure will downshift the Reactor Recirculation Pumps to Slow
Speed and then trip the Reactor Recirculation Pumps off.
_____ C. RPV Water Level 2 will immediately trip the Reactor Recirculation Pumps off.
D. RPV Water Level 2 will downshift the Reactor Recirculation Pumps to Slow
Speed and then trip the Reactor Recirculation Pumps off.
Answer: B
Applicant Comment:
Candidates argued that answer C is also correct since the stem of the question does not ask
which will come first.
Facility Proposed Resolution:
Candidates comment is accurate. This question was raised during the exam and the proctor
placed on the board: Based on given conditions, which will occur first? Since this additional
information was put out to all candidates, only answer B is correct.
NRC Resolution:
Upon review of the question, applicant comment, and facility proposed resolution, the NRC
determined the original answer B was the only correct answer. As the question was originally
written, the applicants comment that answer C was also correct was accurate. However, the
question was addressed and documented by the licensee proctors for all applicants during the
examination administration. The proctors provided the following additional information to the
question stem: Based on given conditions, which will occur first? Based on the additional
information provided to all applicants during the examination administration, the only correct
answer is B.
Enclosure 3
Post Examination Comments and Resolutions
RO Question 24:
PEI-T23, Containment Control has been entered due to a scram discharge volume rupture.
Containment Sprays have been initiated due to rising Containment temperature and pressure.
Containment Spray operation is required to be terminated:
_____ A. before Containment Pressure is reduced below 0.0 psig.
_____ B. before Containment Pressure is reduced below 0.5 psig.
_____ C. after Containment Temperature is reduced below 95 degrees F.
_____ D. after Containment Temperature is reduced below 185 degrees F.
Answer: A
Applicant Comment:
Candidates argue that answer B is an Operation expectation and is taught and practiced.
Facility Proposed Resolution:
Reject candidate comment. What is emphasized is that Containment sprays are secured at
0.5 psig so that they are off prior to 0.0 psig. Question to remain as is.
NRC Resolution:
Upon review of the question, applicant comment, and facility proposed resolution, the NRC
determined the original answer A was the only correct answer. Perry emergency operating
procedure flowchart, PEI-T23 [Plant Emergency Instruction], Containment Control (Containment
Temperature Control), Revision G, and PEI Bases, Revision 7, require containment sprays to
be terminated if containment pressure cannot be maintained greater than 0 psig. In addition,
lesson plan OT-3402-09, Drywell and Containment Pressure Control, Revision 4,Section I.C.2,
states that Containment Spray operation must be terminated before containment pressure
drops below 0.0 psig. During initial license training, the operations training staff emphasize that
containment sprays are secured at 0.5 psig to ensure they are off prior to 0.0 psig. The
licensee could not identify documentation that 0.5 psig was an operations expectation. Based
on the reviewed information, only answer A is correct.
Enclosure 3
Post Examination Comments and Resolutions
RO Question 28:
Certain ATWS conditions require the operators to deliberately lower RPV Water Level in order
to lower Reactor Power.
Which of the following describes why Reactor Power decreases as RPV Water Level is
lowered?
_____ A. Reducing natural circulation causes increased void formation in the core.
_____ B. Uncovering the feedwater spargers reduces the core inlet subcooling to the core.
_____ C. Equalizing the density of the water inside and outside the shroud lowers the void
distribution throughout the core.
_____ D. Reducing the NPSH available to the Jet Pumps reduces the natural circulation
flow the Jet Pumps deliver to the core.
Answer: A
Applicant Comment:
Candidates argued that answer B is also correct. Excerpt from PEI Bases page 5:
(PEI Bases Document, Revision 7, Page 5 of 393)
Lowering RPV water level to two feet below the feedwater spargers (100 in.). This is done to
reduce core-inlet subcooling. This is an effective action to mitigate the consequences of large
power oscillations. This action reduces core flow and inlet subcooling by preheating cold
feedwater with vessel steam, which reduces reactor power and the likelihood of instabilities.
Facility Proposed Resolution:
Candidate comment valid. Answer B would add negative reactivity and cause Reactor Power to
decrease.
The licensee provided the following additional information to support Answer B as a correct
response: The bases for this step, besides discussing the prevention of oscillations, also
mentions that power will decrease due to the lowering level. Both address the why.
Response A states that there is an increase in void formation. Response B states that there is
a decrease in core inlet cooling. Both of these effects are correct based on the action in the
response, i.e., natural circulation and uncovering feedwater spargers; and both of the effects
add negative reactivity and would lower power. Therefore, we agree that there are two correct
answers and that they both address the why.
Enclosure 3
Post Examination Comments and Resolutions
Facility Analysis for Validity (Failure Greater Than 50%):
Five of nine applicants missed the question. The question was evaluated as a faulty question.
Question had two correct answers. If two correct answers are accepted all applicants would
receive credit for the question. Four applicants picked answer A and the other five picked the
new correct answer B.
NRC Resolution:
Upon review of the question, the NRC disagreed with the applicants comment and the facility
proposed resolution that the question has two correct answers. The NRC determined that the
original answer A was the only correct answer. The question asks why reactor power
decreases as reactor pressure vessel (RPV) water level is lowered. Answer A states that
reducing natural circulation causes increased void formation in the core. The purpose (i.e.,
why) this is done is to reduce reactor power as stated in PEI Bases Document, PEI-B13 RPV
Control (ATWS), Revision 7, Discussion, page 165 of 393. Answer A is supported by Perry
emergency operating procedure flowchart PEI-B13, RPV Control (ATWS), Revision K, and PEI
Bases Document, which requires deliberately lowering RPV water level by terminating and
preventing injection to the RPV to reduce natural circulation which causes increased void
formation in the core resulting in addition of negative reactivity (answer A).
The applicants and licensee also supported a second correct answer (B) by providing additional
information from PEI Bases Document, Revision 7, page 5 of 393, for lowering RPV water level
below the feedwater spargers which reduces core flow (increases voids which decreases
power) and core-inlet subcooling by preheating cold feedwater with vessel steam which reduces
the likelihood of instabilities. However, lowering reactor water level, even while above the
spargers, will cause reactor power to decrease due to the lowered water level creating reduced
natural circulation through the core creating increased voids which adds negative reactivity
resulting in decreased reactor power. As written, answer B does not address the reason why
reactor power decreases. Based on the reviewed information, only answer A is correct.
Enclosure 3
Post Examination Comments and Resolutions
RO Question 45:
During a reactor startup, the following conditions exist:
- ALL Source Range Monitor (SRM) detectors are fully inserted.
- ALL SRM Channels indicate 9 x 104 cps.
Per IOI-1, Cold Startup, which one of the following describes the correct operation of SRM
detectors?
SRM detectors are:
_____ A. fully inserted until the reactor is at the Point of Adding Heat. Then the SRM
detectors are fully withdrawn.
_____
the SRM detectors are fully withdrawn.
_____
the SRM detectors are fully withdrawn.
D. partially withdrawn to maintain 102 to 105 cps until IRM overlap is confirmed and
the reactor is at the Point of Adding Heat. Then the SRM detectors are fully
withdrawn.
Answer: C
Applicant Comment:
Candidates argue that answer D is also correct. Can perform IOI steps in any order and IOI
allows withdrawal of SRMs to maintain 100 to 1.0e5 cps.
Facility Proposed Resolution:
Reject candidate comment. Answer D is not correct. SRMs can not be withdrawn until
IRM/SRM overlap is confirmed per Surveillance requirement 3.3.1.1.6. Question to remain as
is.
Note: The licensee provided additional information during a phone discussion. They stated
that the applicants comment is tied to the following items contained in the scope of the IOI, and
it reads:
This instruction is designed to be followed step-by-step. However, the sequence may
be altered or the steps may be annotated N/A in accordance with station administrative
procedures to suit existing plant conditions and time requirements. The reason for a
step being annotated N/A shall be noted in the Remarks blank. For steps that are
already fulfilled by initial conditions, FIC may be entered in the Remarks blank.
Enclosure 3
Post Examination Comments and Resolutions
Substeps within a step should not be performed out of sequence. Initial spaces for each
step are provided as an operator aid.
The licensee stated that Technical Specification (TS) Surveillance Requirement (SR) 3.3.1.1.6
states that SRM / IRM overlap is to be performed prior to withdrawing SRMs from the fully
inserted position.
The licensee further stated that the IOI-1 step to perform the SRM / IRM Overlap is Step 4.3.21,
and the step to maintain the SRMs 102 to 105 cps by withdrawing the SRMs, is Step 4.3.28
(answer C). If the order of these two steps were reversed (answer D), the TS SR 3.3.1.1.6
would be violated. The order for these two steps is determined by the TS surveillance
requirement to perform the overlap check. Thus, it would NOT be appropriate to perform
step 4.3.28 before step 4.3.21 in this particular case. The flexibility to perform steps in a
different order than specified by the IOI must be done without violating any other administrative
requirement. Technical Specifications would be violated in this case, hence, the utility does
NOT support the applicant's position.
Facility Analysis for Validity (Failure Greater Than 50%):
Five of nine applicants missed the question. The question was evaluated as valid with no
changes required. Applicants missed the question due to a knowledge weakness regarding the
conditions to be met before SRM detectors can be fully withdrawn from the core during a
reactor startup.
NRC Resolution:
Upon review of the question, applicant comment, and facility proposed resolution, the NRC
determined the original answer C was the only correct answer. The applicants comment that
IOI steps can be performed in any order is partially correct; however, TS requirements cannot
be violated. The TS SR 3.3.1.1.6 requires that SRM / IRM overlap be performed prior to
withdrawing SRMs from the fully inserted position. Answer C agrees with requirements of TS
SR 3.3.1.1.6 and IOI-1, Cold Startup, Revision 20, Steps 4.3.21-25, to verify SRM / IRM overlap
per TS SR 3.3.1.1.6 with SRMs fully inserted, and Step 4.3.30, to fully withdraw SRMs when
IRMs are on Range 3 (or higher). Answer D would violate TS SR 3.3.1.1.6 because it would
allow SRMs to be partially withdrawn before overlap with IRMs was confirmed. Based on the
additional information, the original answer C is the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
RO Question 63:
Following a transient, INST AIR CNTMT ISOL 1P52-F200 and INST AIR DRYWELL ISOL
1P52-F646 have isolated.
Which one of the following describes the effect of this isolation on the Safety Relief Valves
(SRVs)?
_____ A. ONLY the ADS SRVs can be opened.
_____ B. ALL SRVs can be opened at least once in the Relief Mode.
_____ C. ALL SRVs will open ONLY in the Safety Mode.
_____ D. ALL SRVs cannot be opened.
Answer: B
Applicant Comment:
Candidates argue that answer A is also correct. Excerpt from PEI Bases page 120 is provided
below. PEI Bases page 120 is attached for review. Defeating all isolation interlocks for the
SRV pneumatic supply and restoring pneumatics promotes more stable pressure control should
SRVs be required to augment pressure stabilization. These actions are performed after the
system isolation dependent upon time, manpower, and the need or anticipated need for SRV
use. Loss of the continuous pneumatic supply to the SRVs limits the number of times that
SRVs can be cycled manually since this mode of valve operation requires pneumatic pressure.
While the SRV accumulators provide a reserve pneumatic supply, leakage through in-line
valves, fittings, and actuators may deplete the reserve capacity. There is thus no assurance as
to the number of SRV operating cycles remaining following loss of the continuous SRV
pneumatic supply. For these reasons, if the continuous pneumatic supply is or becomes
unavailable when SRVs must be used to augment RPV pressure control, the valves should be
closed to limit the number of cycles and conserve pneumatic pressure so that the SRVs will be
available if emergency RPV depressurization is later required. If other pressure control
systems are not capable of maintaining RPV pressure below the lowest SRV lift setpoint, SRVs
will still open when the lift setpoint is reached.
Facility Proposed Resolution:
Reject candidate comment. Additional reference supports argument that answer B is correct.
Isolation of air limits the number of times an SRV will cycle, but all SRVs have an accumulator
and will open at least once. Question to remain as is.
Enclosure 3
Post Examination Comments and Resolutions
Facility Analysis for Validity (Failure Greater Than 50%):
Five of nine applicants missed the question. The question was evaluated as valid with no
changes required. Applicants missed the question due to a knowledge weakness regarding the
effect that a loss of the Instrument Air System will have on the Safety Relief Valves (SRVs).
NRC Resolution:
Upon review of the question, applicant comment, and facility proposed resolution, the NRC
determined the original answer B was the only correct answer. The following is taken from our
USAR (page 5.2-19) concerning SRV accumulator capacity:
Actuation of either solenoid A or solenoid B on the safety/relief valve will cause the
safety/relief valve to open; hence, there is no single failure of a logic component or
safety/relief valve solenoid valve which would result in failure of the safety/relief valve to
open. The trip units for each safety/relief valve within each division are in series, and
failure of one of the transmitters will not cause the safety/relief valves to open. Each
safety/relief valve is provided with its own pneumatic accumulator and inlet check valve.
The accumulator capacity is sufficient to provide one safety/relief valve actuation, all that
is required for overpressure protection. Subsequent actuations for an overpressure
event can be spring actuations to limit reactor pressure to acceptable levels.
The applicant reference, PEI Bases Document, PEI 13, RPV Control (Non-ATWS), Revision 7,
i.e., page 120 of 393, does not state or support the answer A statement that ONLY the ADS
SRVs can be opened when the continuous pneumatic supply (i.e., containment and drywell
instrument air supply) is lost to the SRVs. The reference discusses that the loss of continuous
pneumatic air supply will limit the number of manual SRV cycles, and the accumulator will
provide an additional limited reserve pneumatic supply until depleted. Both of these statements
support answer B as the only correct answer.
Enclosure 3
Post Examination Comments and Resolutions
RO Question 64:
The plant is in MODE 5 with the following conditions:
- INST VOL NOT DRAINED and RPS INST VOL LEVEL HI are in alarm due to
maintenance on the Division 1 and 2 trip units for SDV high level.
in BYPASS.
- RPS Trip Systems A and B are reset.
Which one of the following describes the plant response when the Reactor Mode Switch is
placed in the STARTUP/STANDBY position?
_____ A. The SDV Valves remain open and no RPS actuation occurs.
_____ B. The SDV Valves remain closed and no RPS actuation occurs.
_____ C. The SDV Valves remain open and a full RPS actuation occurs.
_____ D. The SDV Valves remain closed and a full RPS actuation occurs.
Answer: D
Applicant Comment:
Candidates argue that the word remain in all distractors is confusing.
Facility Proposed Resolution:
Candidate comment valid. Question to change as shown below.
A. The SDV Valves remain open and no RPS actuation occurs.
_____ B. The SDV Valves close and no RPS actuation occurs.
_____ C. The SDV Valves remain open and a full RPS actuation occurs.
_____ D. The SDV Valves close and a full RPS actuation occurs.
NRC Resolution:
Upon review of the question, applicant comment, and facility proposed resolution, the NRC
determined that wording of the distractors was confusing and the change proposed to
distractor B and correct answer D by the licensee was appropriate. The proposed changes do
not affect the original correct answer as written. This was only an editorial comment to enhance
the question.
Enclosure 3
Post Examination Comments and Resolutions
RO Question 66
Following a transponder card failure, Control Rod 22-55 had its rod drive transponder bypassed
in the Rod Gang Drive Cabinet.
In this condition, Control Rod 22-55 __(1)__ be moved in the Individual Drive Mode and
__(2)__ be moved in the Gang Drive Mode when moving the same gang with a Control Rod
other than 22-55 selected on panel H13-P680.
(1) (2)
_____ A. can cannot
_____ B. can can
C. cannot cannot
_____ D. cannot can
Answer: D
Applicant Comment:
Candidates argue that C is also correct since the transponder card failure has not been
corrected or a specific transponder card failure is not in the stem.
Facility Proposed Resolution:
Candidate comment valid. References are attached for review. Depending on specific
transponder card failure answer C could also be correct.
Facility Analysis for Validity (Failure Greater Than 50%):
Seven of nine applicants missed the question. The question was evaluated as a faulty question
with two correct answers. If two correct answers are accepted seven of the applicants would
receive credit for the question. Two applicants picked answer D and the other five picked the
new correct answer C.
NRC Resolution:
Upon review of the question, the NRC agreed with the applicants comment and the facility
proposed resolution to accept two correct answers (C and D). The NRC reviewed applicable
portions of the General Electric vendor manual, GEK-75602A, paragraphs 2-108 and 2-109,
entitled TRANSPONDER and The Transponder (functions) respectively, and System Operation
Instruction, SOI-C11(RCIS), Section 7.17, Rod Drive Bypass in RGDC, Revision 22. In
Enclosure 3
Post Examination Comments and Resolutions
addition, the NRC discussed the operation of the Rod Gang Drive cabinet, rod drive bypass
function for each control rod, rod transponder cards, Individual Drive and Gang Drive Modes,
and integrated operation of the overall system with the licensee. Based on discussions with the
system Instrument and Control engineer, the licensee explained that each control rod has a
dedicated transponder card that contains withdrawal and insertion circuits providing drive
signals to the associated control rod hydraulic control unit (HCU) directional control valves
(DCVs) to control the motion of the control rod. The transponder receives drive signals from
the rod gang drive cabinet and can drive a single rod when in the Individual Drive Mode. When
in the Gang Drive Mode a single rod in the gang is selected to provide the drive signal to the
respective transponder card of each control rod in the gang. When a rod is placed in bypass,
the transponder is also bypassed in the Individual Drive Mode and the rod will not move.
However, when in the Gang Drive Mode and the rod is bypassed, the associated transponder
still receives drive command signals from the selected rod in the gang. Based on a
transponder circuit card failure, the withdrawal, insertion or both circuit(s) could possibly be
affected. In summary, if the affected rod was bypassed, it would not move when in the
Individual Drive Mode; however, if in the Gang Drive Mode, and another rod was selected, the
bypassed rod would receive a drive signal and could possibly move, or not move, depending on
the type of transponder card circuit failure, i.e., failure of the insert, withdrawal, or both (upscale
or downscale). Based on the additional information, and the indeterminate and undefined
transponder card failure, the NRC determined that both answers C and D would be accepted as
correct answers.
Prior to submission to the NRC, the question did not contain the question prefix Following a
transponder card failure,.., and as written, answer D would have been the only correct answer
based on a CAUTION statement in SOI-C11(RCIS), Section 7.17 regarding bypassing a rod in
the rod gang drive system (with a fully functioning transponder card). However, as a method of
providing the applicants a reason for placing the rod in bypass, the licensee added the question
prefix without analyzing the affect it would have on the question answer. As a result, the NRC
will consider this question as Unsatisfactory in their evaluation of the written examination
submittal because the question had two correct answers.
Enclosure 3
Post Examination Comments and Resolutions
SRO Question 18:
The plant is operating at 100% power with RHR Loop A in Standby Readiness.
RHR PUMP A DISCHARGE PRESSURE HI/LO alarm is momentarily received on panel
H13-P601.
The operator reports RHR Pump A Discharge Pressure indication was low at 25 psig and has
returned to normal at 45 psig on panel H13-P629.
This condition is indicative of potential leakage from the (1) and the Unit Supervisor will (2) .
_____
- A. (1) RHR Pump A, 1E12-C001A pump seals
(2) direct performance of RHR A High Point Vent per SOI-E12, Residual Heat
Removal System to confirm the system is OPERABL
- E.
_____ B. (1) RHR Pump A, 1E12-C001A pump seals
(2) declare the system inoperable IMMEDIATEL
- Y.
_____ C. (1) LPCS & RHR A Water Leg Pump, 1E21-C002 pump seals
(2) direct performance of RHR A High Point Vent per SOI-E12, Residual Heat
Removal System to confirm the system is OPERABL
- E.
D. (1) LPCS & RHR A Water Leg Pump, 1E21-C002 pump seals
(2) declare the system inoperable IMMEDIATELY.
Answer: C
Applicant Comment:
Candidates argue that the LPCS Hi/Lo Pressure alarm should be added to the stem of the
question.
Facility Proposed Resolution:
Reject candidate comment. This additional information would help with the diagnosis of the
problem, but is not required to arrive at the correct answer. In addition if this additional
information would be added then RHR A Pump Seal leak would not be plausible. Question to
remain as is.
Facility Analysis for Validity (Failure Greater Than 50%):
Five of six applicants missed the question. The question was evaluated as valid with no
changes required. Applicants missed the question due to a knowledge weakness regarding the
location within the RHR system where the RHR Pump discharge pressure is sensed for the
high/low discharge pressure alarm.
Enclosure 3
NRC Resolution:
Upon review of the question, applicant comment, and facility proposed resolution, the NRC
determined the original answer C was the only correct answer. The applicant comment was a
suggested enhancement that the licensee rejected because it would add more information to
the stem and decrease the discriminatory value of the distractors A and B by making them
implausible. Based on the reviewed information, the NRC rejected the suggested
enhancement.
Enclosure 3
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession # ML071210394.
Enclosure 4