IR 05000336/2005301
| ML051530113 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/01/2005 |
| From: | Conte R Operations Branch I |
| To: | Christian D Dominion Resources |
| Shared Package | |
| ML042440181 | List: |
| References | |
| IR-05-301 | |
| Download: ML051530113 (22) | |
Text
June 1, 2005
SUBJECT:
MILLSTONE UNIT 2 REACTOR OPERATOR AND SENIOR REACTOR OPERATOR INITIAL EXAMINATION REPORT NO. 05000336/2005301
Dear Mr. Christian:
This report transmits the results of the Reactor Operator (RO) and Senior Reactor Operator (SRO) licensing examination conducted by the NRC during the period of March 18 - 21, 2005.
This examination addressed areas important to public health and safety and was developed and administered using the guidelines of the Examination Standards for Power Reactors (NUREG-1021, Revision 9).
Based on the results of the examination, both Senior Reactor Operator and six of the seven Reactor Operator applicants passed all portions of the examination. One Reactor Operator applicant failed the site-specific written examination. The nine applicants included seven ROs, and two instant SROs. Mr. DAntonio discussed performance insights observed during the examination with training department personnel on March 21, 2005. On April 21, 2005, final examination results, including an individual license number for the one individual who was not still pending completion of waiver requirements, were given during a telephone call between Mr. DAntonio and Mr. Trad Horner.
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 (PARS) component of NRCs document system (ADAMS). These records include the final examination and are available in ADAMS Package No. ML042440181 (RO and SRO Written - Accession Number ML051400118; RO and SRO Operating Section A - Accession Number ML051400126; RO and SRO Operating Section B -
Accession Number ML051400132; and RO and SRO Operating Section C - Accession Number ML051400134), and Facility Post Examination Comments on the Written Exams - Accession No. ML051390229. ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Mr. David Should you have any questions regarding this examination, please contact me at (610) 337-5183, or by e-mail at RJC@NRC.GOV.
Sincerely,
/RA/
Richard J. Conte, Chief Operations Branch Division of Reactor Safety Docket No. 50-336 License No. DPR-65 Enclosure: Initial Examination Report No. 05000336/2005301 cc w/encl:
J. A. Price, Site Vice President - Millstone C. L. Funderburk, Director, Nuclear Licensing and Operations Support D. W. Dodson, Supervisor, Station Licensing M. Wilson, Manager, Nuclear Training L. M. Cuoco, Senior Counsel C. Brinkman, Manager, Washington Nuclear Operations J. Roy, Massachusetts Municipal Wholesale Electric Company First Selectmen, Town of Waterford R. Rubinstein, Waterford Library J. Markowicz, Co-Chair, NEAC E. Woollacott, Co-Chair, NEAC E. Wilds, Director, State of Connecticut SLO Designee J. Buckingham, Department of Public Utility Control G. Proios, Suffolk County Planning Dept.
R. Shadis, New England Coalition Staff G. Winslow, Citizens Regulatory Commission (CRC)
S. Comley, We The People D. Katz, Citizens Awareness Network (CAN)
R. Bassilakis, CAN J. M. Block, Attorney, CAN S. Glenn, INPO
Mr. David
SUMMARY OF FINDINGS
IR 05000336/2005301; March 18-24, 2005; Millstone Point Unit 2; Initial Operator Licensing
Examination. Eight of nine applicants passed the examination (six of seven reactor operators, and two of two senior reactor operators).
The written examinations were administered by the facility and the operating tests were administered by three NRC region-based examiners. There were no inspection findings of significance associated with the examinations.
REPORT DETAILS
REACTOR SAFETY
Mitigating Systems - Reactor Operator (RO) and Senior Reactor Operator (SRO) Initial License Examination
a. Scope
of Review The NRC examination team developed the site-specific written examination; the facility developed the operating examination. Together with Millstone Unit 2 training and operations personnel, the examiners verified or ensured, as applicable, the following:
- The examination was prepared and developed in accordance with the guidelines of Revision 9 of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors. A review was conducted both in the Region I office and at the Millstone Unit 2 plant and training facility. Final resolution of comments and incorporation of test revisions were conducted during and following the onsite preparation week.
- Simulation facility operation was proper.
- A test item analysis was completed on the written examination for feedback into the systems approach to training program.
- Examination security requirements were met.
The NRC examiners administered the operating portion of the examination to all applicants from March 21 - March 24, 2005. The written examination was previously administered by the Millstone Unit 2 training staff on March 18, 2005.
b. Findings
Grading and Results Eight of nine applicants (both SROs and six of seven ROs) passed all portions of the initial licensing examination.
Examination Administration and Performance The following errors occurred during administration of the examinations:
In one scenario, the simulator operator inserted an incorrect malfunction. The scenario was accepted as run because the examiners observed that the incorrect malfunction was somewhat more challenging to the applicants than the intended malfunction.
In one scenario, the simulator operator inadvertently omitted an ATWT malfunction.
The remainder of the scenario was not dependent on this ATWT and was completed.
However, this omission required an additional partial scenario to be run on this crew to capture the missing malfunction.
One JPM had not been corrected to incorporate comments from the validation week to run it as a normal rather than an alternate path JPM. This resulted in some confusion for the examiner and unnecessary stress for the applicant.
During the performance of one JPM, the examiners observed that attempting to synchronize an EDG by following the procedure exactly as written, would sometimes result in failure of the diesel breaker to close. The facility has determined that this is a procedural problem rather than a simulator fidelity issue. CR-05-02859 Simulator Several simulator performance discrepancies or procedural problems were noted during the preparation and administration of the examinations as noted in Attachment 3, simulation facility report. This area is unresolved pending further review by the facility in order for the NRC staff to determine if these issues are performance deficiencies.
URI 05000336/2005301-01 Technical Specifications A potential issue was noted with the Technical Specifications during development of the written examination as follows:
- Wide Range Nuclear Instruments (WRNIs) are required to be operable in modes 3, 4,
5. - Power Range Nuclear Instruments (PRNIs) are required to be operable in modes 1, 2.
- Mode 2 is defined as Keff >
.99 - Power Range Nuclear Instruments come on scale at.1% power.
It thus appears possible to be in compliance with Technical Specifications with the plant in Mode 2, with power still below the range of the PRNIs, but with no requirement for operable WRNIs. This issue has been entered into the facility corrective action process as CR-05-3176, and is unresolved pending further review to determine if the NRCs understanding is correct, if the issue is acceptable, or if the issue is a performance deficiency. URI05000336/2005301-02.
4OA6 Exit Meeting Summary
On April 21, 2005, the NRC provided conclusions and examination results to Millstone Unit 2 management representatives via telephone. License numbers for one of the eight applicants that passed all portions of the initial licensing examination were also provided during this time. License numbers for the remaining seven applicants were withheld pending completion of the requirements of their waivers. Mr. Trad Horner was informed that when the NRC is notified in writing that these requirements have been completed by these seven individuals, their licenses would be issued. The ninth applicant passed all sections of the operating portion, but failed the written portion of the initial licensing examination and, therefore, was denied a license at this time.
The NRC expressed appreciation for the cooperation and assistance that was provided during the preparation and administration of the examination by the licensees training staff.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
Michael Wilson, Manager, Operator Training
Jim Kunze, Unit Two Operations
Pete Strickland, Unit 2 Shift Manager
Richard Ashey, Unit 2 Instructor
Daniel Pantalone, Unit 2 Instructor
NRC
Joseph
- M. DAntonio, Operations Engineer
Donald
- E. Jackson, Senior Operations Engineer
Peter
- A. Presby, Operations Engineer
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
ITEM NUMBER
TYPE
DESCRIPTION
Simulator deficiencies identified
during NRC exam process. 05000336/2005301-02
Potential gap in Technical
Specification requirements for
Nuclear Instrument Operability.
Post Exam Comments
Facility Post Exam Comments and NRC Resolution
An NRC-developed written exam was administered to 7 RO applicants and 2 SROI applicants
on Monday, March 18, 2005. The facility-developed operational exam was administered to
these same applicants during the following week, March 21, 2005 - March 24, 2005).
The facility submitted post exam comments for the following 5 questions on the RO written
exam and also for the following job performance measures in the admin area:
Written Exam Question Number
Admin Job Performance Measure
A2SRO
A1RO
A2RO
The facility-proposed change to Question #32 on the written exam is accepted as proposed. All
other facility-proposed changes to written exam questions (#38, 66, 71 and 72) are not
accepted. The facility-proposed change to the Admin JPMs are accepted. The final facility
comment was received on April 18, 2005.
1.
Written Exam Question #32
Question.
The plant is operating at full power with all equipment functional, except for the 'B' HPSI
Pump, which is OOS for maintenance.
Then, a large break LOCA occurs combined with a loss of Bus 24D (due to an electrical
fault on 24D).
Which one of the choices correctly completes the following statement regarding the
impact of the loss of ECCS pumps.
Ten hours after the event, a loss of the only available ___________ adversely affect
long term core cooling because the remaining ______________.
A
HPSI pump would, LPSI pump does NOT have a system flowpath for boron
precipitation control.
B
HPSI pump would, LPSI pump could NOT be procedurally realigned for boron
precipitation control via hot leg injection.
[Key answer]
C
LPSI pump would NOT, HPSI pump is preferred for boron precipitation control.
D
LPSI pump would NOT, HPSI pump could be procedurally realigned for boron
precipitation control via hot leg injection.
Facility Comments:
Comments:
The justification for the original correct choice B states that although it is physically
possible to align A LPSI pump for Hot Leg Injection, there is no procedural guidance to
do so. The reason there is no procedural guidance for this system alignment under the
given conditions (24D deenergized), is because without 24D power it is not possible to
align LPSI for Hot Leg Injection. One of the major valves in the flow path to achieve this
alignment (2-SI-652) is powered by Facility Two (24D) and is located in containment
(inaccessible with an LOCA). The procedure section the operators are referred to with
24D lost assumes Facility One HPSI & LPSI pumps are both available.
Therefore, as stated in Choice A, if the available HPSI is lost, core cooling would be in
jeopardy because there is no system flowpath for Hot Leg Injection with the available
LPSI pump.
Recommendation:
Based on the above explanation, we believe both Choice A and Choice B are
technically correct for the given information in the stem.
NRC Response:
Facility recommendation for multiple correct answers is ACCEPTE
- D.
The question stem states that Bus 24D is de-energized. Motor-operated gate valveProperty "Contact" (as page type) with input value "D.</br></br>The question stem states that Bus 24D is de-energized. Motor-operated gate valve" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.,
2-SI-652, is in the flowpath for LPSI hot leg recirculation and must be opened to set up
that flowpath. However, the valve operator is powered from the deenergized bus and,
therefore, 2-SI-652 cannot be opened during the given event. These facts establish the
conditions necessary for Choice A to be a correct answer for the question. Choice B
remains an equally valid correct answer. Choice A and Choice B are both correct.
References:
- EOP-2541, Appendix 18, Simultaneous Hot and Cold Leg Injection, Revision 000
- SDC-00-C, Shutdown Cooling System Lesson Material, Revision 3
2.
Written Exam Question #38
Question:
Given the following plant conditions:
- 100% power
- SG levels at setpoint
- Steam flow and feed flow matched
- SG2 Feed Flow Transmitter FT-5269A output fails high
With NO operator actions, which of the following describes the expected plant
response?
A
SG level lowers, but stabilizes above the low level reactor trip.
[Key answer]
B
SG level lowers to the low level reactor trip.
C
SG level rises, but stabilizes below the high level turbine trip.
D
SG level rises to the high level turbine trip.
Facility Comments:
Comments:
Our original concern with this question was that it required the Candidate to go beyond
the knowledge solicited by the K/A (i.e., how will the system respond to the failed
instrument), and make a quantitative judgment as to the amount the system will
respond to the failed instrument. That was the reason behind our suggested rewrite of
the question (see attached) to eliminate the choice of...low level reactor trip and
replace this distractor with one that was clearly wrong in its magnitude of response. In
disallowing this suggested change, the Candidate was forced into a quantitative
judgment, which depending on the specific tuning of the system for that operating cycle,
could result in either Choice A or Choice B being the correct response.
Recommendation:
Based on the above explanation, we believe both Choice A and Choice B are
acceptable answers for the given information in the stem.
NRC Response:
Facility recommendation for multiple correct answers is NOT ACCEPTED.
Question matches the selected K/A in that it tests an applicants knowledge of the
design response to a failure high of a feed flow input to the SG level control system,
which controls feed regulating valve position.
Selected K/A: 059.K4.08
Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following:
Feedwater regulatory valve operation (on basis of steam flow, feed flow mismatch)
The facility concern would be understandable if whether or not the plant tripped for this
malfunction a borderline situation, but it is not - this malfunction does not approach the
trip setpoint. It is reasonable and appropriate to expect a license applicant to be able to
have some idea of the magnitude of pant response to an instrument failure, and in this
case to identify whether or not this particular SGWLC instrument failure will result in an
automatic trip. In order to correctly answer the question, the applicant must understand
that the outputs of two feed flow transmitters are averaged to become a single input to
the steam / feed mismatch portion of the three-element level control system. Further,
the applicant must realize that at 100% reactor power each feed flow transmitter output
is near the high end of its instrument range (100% power reading 5.9 E6 lbs/hr, with full
range reading 6.0E6 lbs/hr). With this information an applicant can easily determine
that a failure high of a feed flow transmitter will cause a relatively small change to the
averaged feed flow signal. The feed regulating valve will begin to throttle closed in
response to the change. However, the level input signal will cause the feed regulating
valve to reopen with only a small change in actual level.
The facility staff ran a failure high malfunction on one feed flow transmitter for #2 SG on
the simulator from 100% power for demonstration purposes. No operator action was
taken. #2 SG level was observed to drop from 70% NR level to 67.5% NR level over
approximately 2 minutes before turning and beginning to trend back to setpoint. No
reactor trip occurred. The low level reactor trip is set to occur when SG level lowers to
49.5% NR level (TS requires setpoint >48.5%).
Given that the failure only results in a 3% change in SG level and that a 20.5% to 21.5%
level change would be required to initiate the reactor trip, the distractor B answer (that
level will lower to the low level reactor trip) is clearly wrong in its magnitude of response
and could not be a correct response regardless of specific tuning for that operating
cycle.
References:
- MP2 Tech Specs, Section 2.2, Limiting Safety System Settings
- RPS-01-C, Reactor Protection System Lesson Material Revision 6
- FWC-01-C, "Feedwater Control System, Lesson Material Revision 2
3.
Written Exam Question #66
Question:
Unit 2 is conducting a reactor start up. Given the following events and conditions:
- Wide range (WR) logarithmic nuclear instrument (NI) channels C and D are out of
service
- The reactor is not yet critical
- The ECP expected critical rod height is 100 steps on Regulating Group 6
- Regulating Group 4 is withdrawn to 60 steps
- WR NI Channel A failed low
WRL NI Channel A <1.0E-1 CPS
WRL NI Channel B 6.2E2 CPS
Which one of the following statements correctly describes the required action (if any)
required to comply with TECHNICAL SPECIFICATIONS?
A
Immediately trip the reactor.
B
Insert all control rods and shutdown the reactor.
C
Stop the startup until WRL NI Channel A has been repaired. NO other actions
are required.
D
Immediately ensure adequate shutdown margin.
[Key answer]
Facility Comments:
Comments:
Choice D (the correct answer) is directly related to ACTION 4 for verifying compliance
with TS 3.1.1.1 SHUTDOWN MARGIN, which is applicable for the current mode of
operation (MODE 3). Choice D is an acceptable answer but is not a complete answer,
taken by itself, because it implies that startup may continue.
Choice B Insert all control rods and shutdown the reactor is also an acceptable
response because it is a proceduralized action of OP 2202 Reactor Startup IPTE and
is the conservative philosophy of operations in DNAP-1410 Reactivity Management.
These actions prevent non-compliance with TS LCO 3.0.4 and its BASIS, which requires
exercise of good practice in restoring systems or components to OPERABLE status
before plant startup.
The Examinee may have been more compelled to place emphasis on compliance with
reactor startup termination than shutdown margin verification because the startup
procedure had just previously verified adequate SDM and has controls in place to
continuously monitor conditions that could affect SDM (e.g., ECP). It would be
unacceptable to maintain current plant conditions considering the time required to re-
verify SDM by obtaining and analyzing an RCS boron sample.
Recommendation:
Accept Choice D and Choice B
NRC Response:
Facility recommendation for multiple correct answers is NOT ACCEPTED.
The facility is reading implications into the distractors which are not stated, and also
appear to be reading implications of required actions ( insert rods to shutdown the
reactor) into their procedures when such actions are not stated. This question
specifically asks the required actions to satisfy TECHNICAL SPECIFICATIONS. The
facility desired second correct answer insert all rods and shutdown the reactor does
not satisfy the Technical Specification requirement to verify shutdown margin. Inserting
rods is most certainly a prudent action, but that action is not explicitly required by any
procedure referenced by the licensee as a response to loss of WRNIs, does NOT
satisfy the TS, and recognition of that fact was the whole point of this question. The
requirement for wide range logarithmic nuclear instruments in MODE 3 per Technical
Specifications is for a minimum of 2 OPERABLE channels. With less than 2
OPERABLE channels, TS 3.3.1 Action 4 directs immediate verification of shutdown
margin requirements. While there are additional actions that may be taken as a prudent
response to inadequate instrumentation during a reactor startup, these actions are not
required to comply with technical specifications, and are not explicitly required by any
facility procedure. Further, Choice B cannot be considered a correct answer to the
question because it does not include the actions required by technical specifications.
References:
- MP2 Technical Specifications, Table 3.3-1, Reactor Protective Instrumentation
4.
Written Exam Question #71
Question:
A transfer of a new fuel assembly is in progress from one location in the spent fuel pool
to another using OP-2303B, "SFP Fuel Handling Operations. The operator raises the
hoist with the desired assembly grappled until upward motion is stopped by the upper
limit switch interlock.
What must be done next?
A
Release hoist raise switch, use the bridge/trolley controls to move to destination.
B
Stop all hoist and crane movement and notify Reactor Engineering immediately.
[Key answer]
C
Lower assembly into initial location and contact Reactor Engineering for
resolution.
D
Slowly lower hoist until load cell indicates 250 to 290 pounds, then continue
move.
Facility Comments:
Comments:
Choice B (the correct answer) is an acceptable answer. It correctly describes the
required actions if the SFP Platform Crane Operator fails to stop when the stainless
steel hose clamp on fuel handling tool is level with the top of SFP platform crane safety
rail.
Choice A is also an acceptable response. The stem of the question does not provide
any information regarding a human performance error on the part of the SFP Platform
Crane Operator (i.e., failure to STOP when the stainless steel hose clamp on fuel
handling tool is level with the top of SFP platform crane safety rail). It is reasonable for
the examinee to assume that operations are proceeding as expected and that the next
action would be to move the bridge/trolley to position the fuel over its final rack location
(a move that is not prevented by interlock).
Other considerations of OP 2303B SFP Fuel Handling Operations:
Level of Use Reference; the procedure shall be readily available to the user, in
the area where the work activity is being performed, such that the user can
obtain a copy of the document as needed to perform the procedure.
Prerequisite 2.1.2: All personnel participating in fuel handling have been briefed
and are thoroughly familiar with this procedure and individual responsibilities.
Examinees were required to answer without the use of reference material.
Recommendation:
Accept Choice B and Choice A
Justification:
Examinees who selected Choice B exhibited knowledge of the new and spent fuel
movement procedures and also knowledge of fuel handling equipment interlocks.
Examinees who selected Choice A exhibited knowledge of the new and spent fuel
movement procedures by correctly identifying the next procedurally directed step,
considering that they were not cued that the SFP Platform Crane Operator had
incorrectly performed the previous step.
NRC Response:
Facility recommendation for multiple correct answers is NOT ACCEPTED.
Question matches the selected K/A in that it tests an applicants knowledge of an
important operational restriction contained within spent fuel handling procedures.
Selected K/A: Generic 2.2.28
Knowledge of new and spent fuel movement procedures.
Facility comments that it is reasonable for an applicant to assume that operations are
proceeding as expected since the question stem does not state that a human
performance error has occurred. However, the applicants ability to identify that fuel
movement being stopped by an interlock is an abnormal situation is central to the
knowledge being tested. It is not an assumption, but a determination the applicant was
expected to make based on the conditions in the stem of the question. An applicant
with the expected level of knowledge regarding fuel handling operations and, in
particular, knowledge of important precautions contained in OP-2303B, "SFP Fuel
Handling Operations, would recognize that the described actions are prohibited and
that the procedure requires immediate notification of Reactor Engineering.
Further, the statement do not attempt to lower the assembly into the upender or
storage rack is contained in both the precaution (3.1) and in the caution prior to the
step for raising the assembly (4.2.12). Step 4.2.14 directs the operator to stop all hoist
and crane movement and notify reactor engineering immediately if the upper limit switch
interlock stops hoist motion.
The facilitys contention that Choice A is also an acceptable response is incorrect.
There is no ambiguity within the procedure regarding required actions when upward
motion of a grappled assembly is stopped by the upper limit switch interlock.
The facility also challenges this question because fuel handling operations would be
briefed and the applicant would have the procedure readily available. This question
does not require an applicant to have the procedure memorized. This question tests
understanding of an important operating limitation appropriately emphasized within a
procedure. To answer this question, the applicants must recognize that actuation of the
interlock was an abnormal situation, and that the response to abnormal occurrences
during fuel handling is to notify RE. This is not an unreasonable level of detail to expect
from an applicant on a closed reference question. While the examination standard does
allow use of reference materials on a selective basis as attachments to the written
exam, it cautions that the references must not improve the applicants chances of
guessing the correct answer by eliminating incorrect distractors. Use of a reference in
this instance would have reduced this question level of difficulty to that of a direct
lookup.
5.
RO Written Exam Question #72
Question:
Refueling is in progress. A new fuel assembly has just been lowered into core location
(core map attached). You are the PPO and have noted the following before and
after readings on the wide range logarithmic power channels:
BEFORE AFTER
Based on these indications, which of the following is required?
A
Suspend all core alterations and positive reactivity additions.
B
Commence boration per AOP-2558, "Emergency Boration."
C
Continue to monitor nuclear instruments, NO immediate action required.
[Key answer]
D
Withdraw the fuel assembly and contact Reactor Engineering for guidance.
Facility Comments:
Comments:
Choice C (the correct answer) is an acceptable answer for an anticipated count rate
multiplication due to the loading of a new fuel assembly in a location adjacent to the CH
B Wide Range detector. Since no information is supplied in the stem of the question as
to the refueling method (e.g., Full Core Reload) it is reasonable to expect this change in
some instances.
Choice B is also an acceptable response. The stem of the question provides no
additional information regarding the method of refueling and status of refueling.
Assumptions could be made as to the refueling method (e.g., fuel shuffle) and previous
data trends of a 1/M plot and/or count rate changes.
Historical data shows that during a fuel shuffle the 1/M value rarely dips below (.8).
Using the provided count rate data from the stem of this question 1/M values (
CRinitial/CRfinal ) are as follows; CH A (.95), CH B (.56 almost a doubling), CH C (.84), CH
D (.83). If an assumption is made that initial count rates at the start of this fuel shuffle
was even lower, then its effect on a 1/M plot would be greater and a doubling may be
evident. If the operator believes that an unanticipated count rate multiplication has
occurred he/she is compelled by OP 2209A to commence an emergency boration.
Recommendation:
Accept Choice C and Choice B
Justification:
Choice B correctly describes the required operator action of OP 2209A and is a
conservative response to a situation that required judgement.
4.5.15 IF, ant any time, unanticipated count rate multiplication, (i.e., doubling), is
indicated, PERFORM the following:
a.
SUSPEND refuel operations.
b.
Refer To AOP 2558, Emergency Boration and PERFORM applicable
actions to initiate boration to RCS.
c.
Immediately NOTIFY Reactor Engineering and SM.
d.
REQUEST evaluation be completed prior to restart of fuel handling
activities.
e.
INITIATE CR.
NRC Response:
Facility recommendation for multiple correct answers is NOT ACCEPTE
- D.
As stated in the question and the facility commentProperty "Contact" (as page type) with input value "D.</br></br>As stated in the question and the facility comment" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., the procedural requirement is to
commence an emergency boration if an unanticipated doubling of count rate occurs. A
core map was included with this question for the applicants to use as an attached
reference. Using the core map, applicants could determine from information in the
question stem that the inserted assembly was adjacent to the wide range logarithmic
power instrument that is showing increased counts. The count rate increase indicated
on this channel has not doubled from the initial readings given. The other channels are
indicating only very slight rise in counts. For the given conditions, emergency boration is
not required. In addition, if we were to accept the facility argument the B was
appropriate, then answer A would be correct as well, a third correct answer requiring
deletion of this question rather than acceptance of a second answer.
Applicants are cautioned at the start of the written exam per NUREG 1021, Appendix E,
to not make assumptions regarding conditions that are not specified in the question
unless they occur as a consequence of other conditions that are stated in the question.
Presuming that initial counts were even lower at the start of refueling requires the
applicant to make assumptions not stated in the question.
References:
- OP-2209A, "Refueling Operations, Revision 24
6.
JPM A2SRO
A step identified as critical in this SRO JPM required applicants to specify the procedure
that contains the necessary post-maintenance testing guidance.
Facility Comments:
Justification for accepting 2604AO and 2308X11 as a correct answer to JPM-A2SRO
SRO AWO Acceptance:
Millstone Unit 2 is developing a new set of procedures designed to simplify component
maintenance and retesting. These new procedures, called Maintenance Operating
Procedures (MOPs), are designated as 2300X. The MOPs include the steps of the
2600 procedure and additional sections for venting, draining and tagging for the
component in question. When a MOP is approved, it becomes the preferred post-
maintenance procedure and replaces the previously used 2600.
At the time the SRO candidates were trained and when this JPM was developed, the
MOPs did not exist. At the time of the NRC Exam, a MOP was approved for the A
HPSI Pp. I believe this is the only MOP that is approved for use at this time.
Due to the present state of transition from the 2600 procedures to the 2300X
procedures, Millstone Unit 2 requests both 2604AO and 2308X11 be considered correct.
NRC Response:
Recommendation to allow both 2604AO and 2308X11" as correct procedure
references is ACCEPTED.
The 2300 procedure covering post-maintenance testing of HPSI Pump A (relating to
this JPM) has been issued, and is the correct procedure for the evolution. However, the
license applicants were trained to use the 2600 plant equipment TS surveillance
procedures as post-maintenance test procedures. An applicant, not yet trained on
recent procedure changes, would be expected to identify the 2600 procedure as the
correct procedure. The 2600 procedures will continue to exist and are used for conduct
of periodic scheduled surveillances. The 2300 series procedures provide more
comprehensive guidance, including steps for drain, refill and testing to restore
operability.
References:
- 2604AO, HPSI PUMP INSERVICE TESTING, 1,750 PSIA, FACILITY 1, Revision
000-00
- 2308X11, A HPSI PUMP MAINTENANCE, Revision 000-01
7.
JPM A1RO
Several steps identified as critical in this SRO JPM appeared to have incorrect error
bands.
Facility Comments:
During grading of this JPM, the NRC noted several applicants who obtained the correct
end result despite errors in intermediate steps identified as JPM critical steps. The NRC
grader determined that there were errors in the answer key for this JPM and requested
a corrected key.
NRC Response:
The NRC verified that the resubmitted key contained the corrections identified by the
NRC grader.
8.
JPM A2RO
Facility Comments:
Tagging of the chiller compressor breaker in this JPM should not have been a critical
step. The compressor is interlocked with the associated chilled water pump, so tagging
the chill water pump prevents the compressor from starting.
NRC Response:
The facility requested change to the answer key is ACCEPTE
- D.
The NRC reviewed the chilled water system OP 2330CProperty "Contact" (as page type) with input value "D.</br></br>The NRC reviewed the chilled water system OP 2330C" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., which addresses system
maintenance and does not require the chiller compressors to be tagged under the
conditions in this JPM. The NRC also reviewed the tagging procedure WC-2 and noted
that this procedure does not preclude dependence on interlocks for equipment
protection. No procedural basis was found for requiring this breaker to be tagged.
ES 501 Simulation Facility Report
Facility Licensee: Millstone Unit 2
Facility Docket No. 50-336
Operating Tests Administered on: March 21 - March 24, 2005
This form is to be used only to report observations. These observations do not constitute audit
or inspection findings and, without further verification and review, are not indicative of
noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or
approval of the simulation facility other than to provide information that may be used in future
evaluations. No licensee action is required in response to these observations.
While conducting simulator scenario validation and during examination administration,
examiners observed the following simulator performance discrepancies:
During validation of the scenarios, Facility 1 and 2 EBFAS could not be used cross-train
with Facility 2 and 1 H2 PURGE valves to depressurize containment, although there is
no reason physically why this should not work. DR# 2005-2-0013.
During validation of the scenarios, CH-501 VCT Outlet Valve control switch did not
appear to operate as indicated per the schematic diagram. DR# 2005-2-0021
During the administration of one scenario, the PMW flow controller and flow totalizer
indicated approximately.1 gpm with the flow control valve closed. DR#2005-2-0023
NUREG-1021 Revision 9