IR 05000440/2004301
| ML080840356 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 02/17/2005 |
| From: | Lanksbury R Operations Branch III |
| To: | Richard Anderson FirstEnergy Nuclear Operating Co |
| References | |
| 50-440/04-301 | |
| Download: ML080840356 (39) | |
Text
February 17, 2005
SUBJECT:
PERRY NUCLEAR POWER PLANT NRC INITIAL LICENSE EXAMINATION REPORT 50-440/04-301(DRS)
Dear Mr. Anderson:
On December 9, 2004, the NRC completed initial operator licensing examinations at your Perry Nuclear Power Plant. The enclosed report presents the results of the examinations.
NRC examiners administered the operating test during the weeks of November 29, 2004, and December 6, 2004. NRC examiners and members of the Perry Nuclear Power Plant Training Department staff administered the written examination on December 9, 2004. Four Reactor Operator (RO) and seven Senior Reactor Operator (SRO) applicants were administered license examinations. The results of the examinations were finalized on January 18, 2005.
Six applicants passed all sections of their examinations and were issued respective operator or senior operator licenses. Two RO applicants and three SRO applicants failed the written examination and received proposed denial letters.
The initial license examination was developed by NRC examiners and reviewed by several Perry Nuclear Power Plant operators and trainers prior to examination administration.
Comments generated during this review were incorporated into the written examination in accordance with approved procedures. A letter detailing 18 post-written examination comments was submitted by Perry Nuclear Power Plant Training Department personnel on December 15, 2004. The NRCs review of these comments indicated an inadequate review prior to submittal to the NRC based on a lack of supporting information and technical inadequacies for a number of the comments. Following communication of this issue, a second letter modifying the original post-examination comments was received on January 10, 2005. The second letter requested withdrawal of eight comments and provided additional reference materials for the remaining 10 comments. The number of post examination comments was well above that expected for a 10 CFR Part 55 required examination and indicative of an inadequate pre-examination review of the written examination by station personnel. In accordance with 10 CFR Part 2.390 of the NRC's Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
We will gladly discuss any questions you have concerning this examination.
Sincerely,
/RA/
Roger D. Lanksbury, Chief Operations Branch Division of Reactor Safety Docket No. 50-440 License No. NPF-58
Enclosures:
1.
Operator Licensing Examination Report 50-440/04-301(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:
50-440/04-301(DRS)
Licensee:
FirstEnergy Nuclear Operating Company Facility:
Perry Nuclear Power Plant Location:
10 Center Road Perry, OH 44081 Dates:
November 29 through December 9, 2004 Examiners:
D. McNeil, Chief Examiner R. Walton, Examiner M. Bielby, Examiner D. Reeser, Examiner (In Training)
Approved by:
R. Lanksbury, Chief Operations Branch Division of Reactor Safety Enclosure
Enclosure
SUMMARY OF FINDINGS
ER 05000440/2004301(DRS); 11/29/2004-12/09/2004; Perry 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:
- Eleven examinations were administered (four Reactor Operator and seven Senior Reactor Operator).
- Six applicants passed all sections of their respective examinations and were issued applicable operator licenses. Two RO applicants and three SRO applicants failed the written examination and received proposed denials of their license applications. It was determined that 10 post examination comments were not within the range of expectations for a license examination required by 10 CFR Part 55. (Section 4OA5.1)
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA5 Other
.1 Initial Licensing Examinations
a. Examination Scope
The NRC examiners conducted an announced initial operator licensing examination during the weeks of November 29, 2004, and December 6, 2004. The NRC examiners used the guidance established in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, to prepare the examination outline and to develop the written examination and operating test. The NRC examiners administered the operating test during the weeks of November 29, 2004, and December 6, 2004. The NRC examiners and members of the Perry Nuclear Power Plant (PNPP) Training Department administered the written examination on December 9, 2004. Four Reactor Operator (RO) and seven Senior Reactor Operator (SRO) applicants were examined.
b. Findings
Written Examination The licensee reviewed the written examination developed by NRC examiners. Written examination comments developed during the examination validation week (November 1, 2004) were evaluated and incorporated into the written examination in accordance with the guidance contained in NUREG-1021.
During the post examination review of the examination results by NRC examiners, it was determined that one examination question (Question #007) was inappropriately administered to the reactor operator applicants. The question required the application of a technical specification which is an operating skill not required of reactor operators.
The question was deleted from the reactor operators examination, but retained in the 75 reactor operator questions administered in the senior reactor operator written examination. In addition, the examiners determined that another examination question (Question #040) required a change to the answer key. The question provided a set of conditions, then asked what would be the response of RPS (Reactor Protection System)to the given plant conditions. The apparent correct answer, [(b.), initiates a rod block]
was not correct. Control rod blocks are not initiated by RPS. Distractor (a.), however, stated that RPS would not do anything under the given conditions, which was the correct answer. The answer key was modified to accept only (a.) as the correct answer for Question #040. Since the examination had been graded and results mailed to the applicants, a new answer key and final scores were provided to the stations training department in a separate letter.
A total of 18 post-examination comments (15 RO; 3 SRO exam comments) were submitted by the stations training department personnel on December 15, 2004. The
NRCs review of these comments indicated an inadequate review prior to submittal to the NRC based on a lack of supporting information and technical inadequacies for a number of the comments. Following communication of this issue, a follow-up letter modifying the initial comments was provided on January 10, 2005. The results of the NRCs review of all 18 comments are documented in Attachment 3, Post Examination Comments and Resolutions. The total number of final licensee comments (ten) and required changes to the examination answer key was indicative of an inadequate pre-examination review of the examination by station personnel. The high number of comments withdrawn by the follow-up letter indicated an inadequate review of the original comments by station personnel prior to submission to the NRC for evaluation.
Operating Test The NRC examiners validated the operating test during the validation week and replaced or modified several items in the proposed operating test. Test changes, agreed upon between the NRC and the licensee, were made in accordance with NUREG-1021 guidelines.
Test/Examination Results Six applicants passed all sections of their respective examinations and were issued applicable operator licenses. Two RO applicants and three SRO applicants failed the written examination and received proposed denials of their license applications.
.2 Examination Security
a. Inspection Scope
The NRC examiners briefed the facility contact on the NRCs requirements and guidelines related to examination physical security (e.g., access restrictions and simulator considerations). The examiners observed the implementation of examination security and integrity measures (e.g., security agreements) throughout the examination process.
b. Findings
The NRC examiners determined that on two occasions examination security was inadequate. On the first occasion an unqualified NRC inspector observing the examination process inadvertently left a copy of an examination simulator scenario outside of the secure area of the simulator. Since the scenario was uncontrolled and it could not be verified that unauthorized personnel had not viewed the scenario, it was replaced with a new scenario. A Condition Report (CR) was written (CR 04-5774) to document and track this occurrence. On the second occasion a crew of candidates (one SRO(U) and two ROs) were left unescorted in their classroom when other candidates with examination specific knowledge (a complete dynamic simulator scenario) were uncontrolled. A follow-up investigation was conducted wherein it was determined that no compromise of the operating test occurred. This determination was made because:
- (1) the uncontrolled candidates were escorted to the training building door and directed to leave the site or a denial of their application would occur;
- (2) none
of the candidates in the classroom had a cell phone or text messenger;
- (3) the classroom telephone had been disconnected;
- (4) the candidates were left alone in the classroom for a short period of time (approximately 30-60 minutes); and
- (5) the candidates in the classroom stated that none of them had any contact with any other candidates that day. Likewise, the candidates that were uncontrolled stated that they had not had any contact with the candidates in the classroom until after the dynamic scenarios were completed for that examination day. A CR was written (CR 04-06413) to document and track this occurrence.
4OA6 Meetings
.1 Exit Meeting
The chief examiner presented the examination team's preliminary observations and findings on December 9, 2004, to Mr. F. Von Aun and other members of the Operations and Training Department staff. The licensee acknowledged the observations and findings presented. No proprietary information was identified by the stations staff during the exit meeting.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- F. von Ahn, Director, Site Operations
- P. Bordley, Operations Superintendent
- D. Bowen, Compliance Engineer
- R. Collings, Fleet Training Manager (FENOC)
- J. Lausberg, Regulatory Compliance Manager
- J. McHugh, Operations Training Supervisor
- K. Meade, Compliance Supervisor
- J. Messina, Director, Program Improvement
NRC
- D. McNeil, Chief Examiner
- M. Bielby, Examiner
- R. Walton, Examiner
- D. Reeser, Examiner Under Instruction,
- B. Grimmel, NRC Intern
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None
Closed
None
Discussed
None
LIST OF ACRONYMS
Agency-Wide Document Access and Management System
Division of Reactor Safety
NRC
Nuclear Regulatory Commission
Publicly Available Records
Reactor Operator
Senior Reactor Operator
SIMULATION FACILITY REPORT
Facility Licensee:
Perry Nuclear Power Plant
Facility Docket No.:
50-440
Operating Tests Administered:
November 29 - December 8, 2004
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 10 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
SLC Pump Run
When the SLC pumps run for a significant period of time, the simulator
program execution halts. This occurred at the end of one scenario
when the examiners were preparing to end the scenario. This was a
known problem and has already been documented.
Turbine Control
Valves
While performing a main turbine stop valve surveillance, turbine
bypass valves opened to control reactor pressure. This did not appear
to be a correct response. The condition may be correct because of the
stations up-rate and switch to partial arc emission. A simulator work
request was written to document the condition and subsequent follow-
up.
Post Examination Comments and Resolutions
Question No. 1:
During the up-shift of Reactor Recirculation Pumps to fast speed, the A pump was
successfully shifted to fast speed. However, during the up-shift of the B pump, Breaker 5B
did not close and Breaker 1B tripped open. You observe the following plant conditions:
-
Reactor Power:
34% RTP and stable
-
Core Flow:
Mlb/hr
-
Core Plate d/p:
1.8 psid
-
Reactor Recirc Pump A is in Fast Speed with its FCV at 9% VALVE TRAVEL
-
Reactor Recirc Pump B is Off with its FCV at 9% VALVE TRAVEL
As the Operator at the Controls, which of the following actions would be correct?
a.
Downshift the A Recirc Pump to slow speed.
b.
Close the A Flow Control Valve to minimum.
c.
Close the Suction Valve for Recirc Pump B.
d.
Open the B Flow Control Valve to 100%.
Answer: c.
Facility References:
SDM: B33
LP:
OT Combined B33, Obj. I
ONI-C51
Facility Comment:
There are 2 correct answers. Answer A is also correct as the A pump could be
downshifted to minimize loop flow mismatch. Reference - Tech Spec 3.4.1 loop flow
mismatch.
Subsequent Facility Comment:
Based upon further review by the Utility, the comment on this question is being
withdrawn.
NRC Resolution:
No change was made to the answer key for question #1.
Post Examination Comments and Resolutions
Question No. 2:
Given the following initial plant conditions:
-
Mode 3 with a plant cooldown in progress following an extended high power run.
-
RHR loop "B" is in Shutdown Cooling (SDC) mode
-
Coolant temperature is 335°F
-
RPV pressure is 110 psig
Select the statement that describes the effect on the SDC Suction Isolation Inboard and
Outboard Valves (1E12-F009 and 1E12-F008) if Bus EH12 (4.16 KV) trips.
a.
1E12-F008 and 1E12-F009 will shut.
b.
1E12-F008 and 1E12-F009 will NOT shut.
c.
1E12-F008 will shut, 1E12-F009 will NOT shut.
d.
1E12-F008 will NOT shut, 1E12-F009 will shut.
Answer: c.
Facility Reference:
ONI-R22-1, attachment 1.
Facility Comment:
There are 2 correct answers. Answer A is also correct assuming all automatic actions
occur. George Lesiak was not given the information that the bus EH12 stayed de-
energized.
Subsequent Facility Comment:
correct answers - Mr. Lesiak indicated he did not receive information supplied by the
proctor that bus EH12 stayed de-energized and did not energize on the diesel supply.
Answer A is also correct assuming the diesel re-energizes the bus and power is restored
to the 1E12F009 valve.
NRC Resolution:
The intent of the question was that Bus EH12 trips indicating the bus had a fault and
would not be re-energized by the diesel generator when it started. This should not be
confused with a breaker trip where an alternate power supply could re-energize the bus.
ONI-R22-1 refers to the bus as de-energized, or a loss of the bus when it can be re-
energized by another power source and does not use the word tripped. Since the bus
tripped, it could not be re-energized and answer (a.) would not be correct. Concerning
Mr. Lesiaks indication that he did not receive information supplied by the proctor that
bus EH12 stayed de-energized, a review of the questions asked by the applicants during
the examination found no clarifying statements given to any applicants concerning this
question. One SRO(U) applicants asked one of the NRC examiners for a clarification
for this question. He was told the bus tripped (given in the stem), that he should re-read
Post Examination Comments and Resolutions
the question and not make any assumptions that could not be supported by the
questions stem. There was nothing in the stem of the question that would imply the bus
was re-energized, and as such, no one could assume the bus re-energized. Providing a
clarification or additional information that EH12 stayed de-energized was not necessary.
Since no clarifying information was provided, the SRO(U)s question was not
documented by the examiner. Additionally, there was no comment by the examination
review team indicating the question was improperly worded. The comment was
rejected; no change was made to the answer key for question #2.
Post Examination Comments and Resolutions
Question No. 10
During refueling operations, a fuel bundle is being lifted from the core for movement to the
spent fuel pool, the following events occur:
-
Containment Ventilation exhaust radiation monitors alarm.
-
Bubbles are observed coming from the bundle being moved.
Select the statement that correctly describes the IMMEDIATE ACTIONS to be performed:
a.
Immediately stop all fuel movement, evacuate all personnel from the Refuel
Floor, and suspend all Core Alterations.
b.
Immediately stop all fuel movement, evacuate unnecessary personnel from the
Refuel Floor, and suspend all Core Alterations.
c.
Place the bundle in a safe condition, evacuate unnecessary personnel from the
Refuel Floor, and suspend all Core Alterations.
d.
Place the bundle in a safe condition, evacuate all personnel from the Refuel
Floor, and suspend all Core Alterations.
Answer: c.
Facility Reference:
ONI-J11-2
Facility Comment:
There are two correct answers. Answer D is also correct because the radiation alarm
requires entry into ONI-D17. Reference: ONI-D17
Subsequent Facility Comment:
Based upon further review by the Utility, the comment on this question is being
withdrawn. Condition Report 04-06852 was written to ensure the wording in the two
procedures is made correct.
NRC Resolution:
Although the comment was withdrawn, the NRC examiners believe the correct answer
to this question is distractor (d.), rather than distractor (c.). The stations examination
validators did not recognize there was an entry into ONI-D17 and answer (d.) was
missed. The two referenced procedures appear to have conflicting answers in that one
procedure calls for evacuation of unnecessary personnel while the second procedure
call for evacuation of the refuel floor, indicating that all personnel should be evacuated.
An operator in the control room would initially receive the containment ventilation high
radiation alarms referred to in the initial conditions of the question. The alarm response
instructions associated with these alarms directs the operator to enter ONI-D17 which
required an evacuation of the refuel floor. Subsequent actions direct the entry into
ONI-J11-2 for the refuel accident with bubbles coming from the bundle referenced in the
Post Examination Comments and Resolutions
initial conditions. This directs the evacuation of non-essential personnel. The NRC
examiners believe that the conservative (and first) response to this condition was to
evacuate all personnel from the refuel floor under these conditions. As a result of this
question a Condition Report was developed directing modification of ONI-J11-2 to
evacuate all personnel from the refuel floor. The NRC has determined that an
evacuation of all personnel from the refuel floor as outlined in ONI-D17 is the only
correct response. The answer key was amended to accept answer (d.) as the only
correct answer for question #10.
Post Examination Comments and Resolutions
Question No. 14
Following a DBA LOCA (assume all systems operated as designed), which one of the following
modes of RHR operation has the most significant long term impact on maintaining the
Containment integrity?
a.
Low Pressure Coolant Injection Mode
b.
Shutdown Cooling Mode
c.
Containment Spray Mode
d.
Suppression Pool Cooling Mode
Answer: d.
Facility Reference
PEI Bases Document
SDM T23/P53
USAR Chapter 6
Facility Comment:
Toss out for Reactor Operators. This is beyond the scope of what is required
knowledge.
Subsequent Facility Comment:
Based upon further review by the Utility, the comment on this question is being
withdrawn. The initial request was to remove this question based on being beyond the
JTA requirements for ROs. Feedback from the Lead Examiner indicates this was a 3.8
K/A requirement for the R
- O. Utility review of the K/A Catalog, assuming this was a
knowledge under 230000 A4.14, makes this comment basis invalid.
NRC Resolution:
No modification was made to the answer key for Question #14.
Post Examination Comments and Resolutions
Question No. 27
A LOCA has occurred resulting in significant Hydrogen generation. One division of Hydrogen
Igniters is in operation and one Combustible Gas Mixing Compressor is operating. Both
Hydrogen Recombiners are shutdown due to Hydrogen concentration exceeding 6% in
containment. Hydrogen concentration is continuing to increase. Which one of the following
statements best explains why Hydrogen concentration is continuing to increase?
a.
Hydrogen generation has exceeded the operational capability of the one division
of Hydrogen Igniters that are in service.
b.
A continuing increase in hydrogen concentration is indicative of a steam inert or
Oxygen starved environment.
c.
Hydrogen concentration will continue to increase until the Hydrogen Igniters
reach their operating temperature which can take several hours.
d.
The indicated increase must be due to a malfunction of the Hydrogen Analyzer
since actual concentration cannot exceed 6% as long as the Hydrogen Igniters
are in operation.
Answer: b.
Facility Reference:
EOP Bases
Facility Comment:
There are two correct answers. Answer A is also correct per Tech Spec 3.6.3.2 Bases,
which states one train can handle hydrogen generated from 75% of the fuel clad.
Reference TS 3.6.3.2
Subsequent Facility Comment:
correct answers - answer A could also be correct, Tech Spec Bases 3.6.3.2 states that
the H2 igniters are installed to accommodate the amount of hydrogen equivalent to that
generated from the reaction of 75% of the fuel cladding with water. The question stem
provides the information that only one train of H2 igniters is in operation and that there is
significant H2 generation. No specifics are provided as to the actual amount of H2
generated, and 1/2 of the required igniters are not functioning. The trainee could infer
that the potential exists that it has exceeded the capability of the single H2 igniter train in
operation, before an inert environment has occurred.
NRC Resolution:
The examiners reviewed Technical Specification 3.6.3.1 and 3.6.3.2 Bases and found
that the Bases document stated that: In MODES 1 and 2, the hydrogen igniter is
required to control hydrogen concentration to near the flammability limit of 4.0 v/o
following a degraded core event that would generate hydrogen in amounts equivalent to
a metal water reaction of 75% of the core cladding. This would indicate that there is an
upper limit of hydrogen that the igniters can successfully burn and would appear to
Post Examination Comments and Resolutions
make Answer (a.) correct. The Bases indicated that operation of only one train of
hydrogen igniters is required to limit hydrogen concentration to the 4.0 v/o limit. The
Bases further stated that: The hydrogen igniters are not included for mitigation of a
Design Basis Accident (DBA) because an amount of hydrogen equivalent to that
generated from the reaction of 75% of the fuel cladding with water is far in excess of the
hydrogen calculated for the limiting DBA loss of coolant accident (LOCA). The question
stem indicated that a LOCA had occurred. Therefore, the amount of hydrogen available
to burn would be significantly less than that amount of hydrogen generated by reacting
75% of the fuel clad with water. However, the stem stated that hydrogen concentration
was 6%, indicating excess hydrogen is being generated and the hydrogen recombiners
were unable to keep up with hydrogen production. Since the amount of hydrogen
generated was higher than the LOCA values, one could assume that the capacity of one
train of hydrogen igniters had been exceeded and the rise in hydrogen concentration in
containment must be caused by excessive hydrogen generation and distractor (a.) is
correct. The answer key for question #27 was amended to accept distractor (a.) and
distractor (b.) as correct answers.
Post Examination Comments and Resolutions
Question No. 34
The plant was operating at 100% reactor power when the plant experienced an earthquake. A
medium break LOCA occurred and RPV Level 2 was reached. All ECCS systems responded
correctly. RPV level is currently 180 inches and slowly increasing. The reactor failed to scram
and all efforts to manually insert control rods have failed. Standby liquid control has failed to
correctly initiate and shut down the reactor (failed SQUIBB valves). The Unit Supervisor has
decided to initiate Alternate Boron Injection (ABI) in accordance with PEI-SPI 1.8. What needs
to be done in order to successfully initiate Alternate Boron Injection?
a.
Secure HPCS.
b.
Close E22-F004 (HPCS Injection Valve)
c.
Secure both SLC pumps
d.
Connect a low pressure hose from the SLC storage tank to the suction of the
ABI pump; start the ABI pump, open the ABI pump discharge valve.
Answer: a.
Facility Reference:
PEI-SPI, Alternate Boron Injection, 1.8, R2, P4
Facility Comment:
There are two correct answers. Answer D is also correct because both answers are
parts of performing SPI 1.8. Reference: PEI-SPI 1.8
Subsequent Facility Comment:
correct answers - answer D is also correct. Both answers are parts of performing SPI
1.8. Verbal discussions with the lead examiner indicates that there is no correct answer
and that both of the above are partial answers. The question will be removed from the
test. The utility concurs with this decision.
NRC Resolution:
Upon review of the question it was determined that there are no correct answers
provided for this question. The stem asks what needs to be done in order to
successfully initiate AB
- I. The correct answer (a.), does not provide all the alignment
steps that are required to set up ABI in order to inject boron, only the step to secure
HPCS. This makes distractor (a.) incorrect. The alignment steps are partially outlined
in distractor (d.), but distractor (d.) does not contain the requirement to secure HPCS
and insert another hose, making distractor (d.) incomplete, and therefore, incorrect.
Since there was no correct answer provided for this question, the answer key was
modified to delete question #34.
Post Examination Comments and Resolutions
Question No. 45
With the reactor at 100% power, which ONE of the following conditions would be an indication
of an open Safety Relief Valve (SRV)? (Assume no other plant problems.)
SRV tailpipe temperature is...
a.
dependent upon drywell pressure and would be in a range from 320°F to 547°F.
b.
stable at approximately 547°F.
c.
stable at approximately 345°F.
d.
less than or equal to 240°F.
Answer: c.
Facility Reference:
Steam Tables
Facility Comment:
There are two correct answers. Answer A is also correct because Drywell pressure has
an effect on suppression pool level. This changes the pressure above SRV riser.
Therefore, SRV temperature will be a range dependent upon Drywell pressure.
Subsequent Facility Comment:
Based on further review by the Utility, the comment on this question is being withdrawn.
NRC Resolution:
No change was made to the answer key for question #45.
Post Examination Comments and Resolutions
Question No. 55
While attempting to insert a control rod, the operator depresses the INSERT pushbutton and
observes the following:
-
No rod motion
-
CRD DRIVE WATER HEADER FLOW at 0 gpm
-
CRD COOLING WATER FLOW at 60 gpm
Which ONE of the following is the possible cause of these indications?
a.
CRD Flow Control Valve failed closed.
b.
Associated drive water stabilizing valves failed closed.
c.
Associated Insert Exhaust Directional Control Valve (DCV 121) failed closed.
d.
Associated Insert Drive Directional Control Valve (DCV 123) failed closed.
Answer: d.
Reference:
Facility Comment:
There are two correct answers. Answer C is also correct because there will be a dp
developed and the rod may possibly move due to the differences in surface areas of p
under and p over. Reference GEK - says possible cause can be inoperable directional
control valve.
Subsequent Facility Comment:
The two correct answers was based upon the fact that there may be no delta p
developed without the exhaust valve being open. Although the rod may move due to the
differences in surface areas of p under and p over and leakage through the seals, the
potential exists that no flow will be developed with the exhaust valve closed, as there
would be no flow path. This would make answer C also correct.
NRC Resolution:
The question was developed using the past operating experience of the NRC
examiners. When the reason for the correct answer was presented to the stations
validation team, they accepted the reasoning and distractor (c.) was not modified.
However, in accordance with the information provided in the comments, it is possible no
flow would be developed under the given conditions and distractor (c.) could be a
correct answer. The examination answer key was modified to accept distractors (c.)
and (d.) as correct answers to question #55.
Post Examination Comments and Resolutions
Question No. 58
A Main Steam line break (18 minutes ago) has resulted in the following plant conditions:
-
Drywell pressure is 4.0 psig and decreasing slowly.
-
Suppression Pool Temperature is 150°F
-
RPV water level is being maintained at the Main Steam lines with LPCS due to
exceeding RPV Saturation Temperature in the Drywell.
-
RHR B is operating in the Suppression Pool Cooling mode.
-
RHR A is operating in the Containment Spray mode.
-
Containment pressure is approaching 0 psig.
You have been directed to secure Containment Spray. While shutting the Containment Spray
Shutoff Valve (F028A) you observe that the Minimum Flow Valve (F064A) did NOT open. You
should...
a.
Reopen the Containment Spray Shutoff Valve (F028A)
b.
Open the LPCI A Injection Valve (F042A)
c.
Open the RHR A Test Valve to Supp Pool (F024A)
d.
Shutdown RHR Pump A
Answer: d.
Facility Reference:
PEI-3.1
SOI-E12
PEI-T23
Facility Comment:
There are two correct answers. Answer C is also correct because at a suppression pool
temperature of 150° F, RHR A can be placed in suppression pool cooling. Reference -
SPI 3.2 directs the following: 3.7.3.1 OPEN RHR A(B) TEST VALVE TO SUPR POOL
E12-F024A(B). Attachment 1, PY-CEI/OIE-0628L, Page 2 of 2.
Subsequent Facility Comment:
This question has two correct answers. Due to suppression pool temperature of 150°F,
if securing from containment spray per PEI-SPI 3.1, you may go in to suppression pool
cooling on the RHR A Loop which would open the E12-F024A. In order to maintain the
loop available, the SRO could direct performance of step 3.7.3.1. This would make
answer C also correct, as opening the valve would prevent the pump from operating
without minimum flow protection. The utility believes there are two correct answers.
NRC Resolution:
The NRC examiners reviewed PEI-T23, Containment Control and determined that if
Suppression Pool Temperature cannot be maintained less than 95°F, then Except for
Post Examination Comments and Resolutions
any RHR Pump required for adequate core cooling, operate all available Suppression
Pool cooling. As noted in the post-examination comment, the steps necessary to place
suppression pool cooling in service correspond with distractor (c.), making distractor (c.)
a correct answer. This correct answer was not identified during the pre-examination
review. The answer key was modified to accept answers (c.) and (d.) as correct
answers for question #58.
Post Examination Comments and Resolutions
Question No. 60
The plant is initially operating steady state at 75% RT
- P. If one(1) MSIV drifts closed, reactor
power will (1) due to (2).
a.
(1)
drop to approximately 0%
(2)
a reactor scram caused by a high steam flow Group 1 isolation.
b.
(1)
decrease to approximately 60% RTP
(2)
the loss of steam flow from the associated Main Steam line.
c.
(1)
increase to approximately 90% RTP
(2)
the increased differential pressure need to push the same amount of
steam through three steam lines.
d.
(1)
remain the same
(2)
the response of the Steam Bypass/Pressure Control system to maintain a
constant reactor pressure.
Answer: c.
Facility Reference:
USAR Chapter 15
SDM N32/C85
Facility Comment:
There are two correct answers. Answer D could also be correct because
it depends on the speed of the MSIV closure. If slow enough, Answer D
would also be correct.
Subsequent Facility Comment:
Based upon further review by the Utility, the comment on this question is being withdrawn.
NRC Resolution:
No change was made to the answer key for question #60.
Post Examination Comments and Resolutions
Question No. 65
Given the following:
-
The plant is in Mode 5 with refueling operations in progress.
-
The refuel position one-rod-out interlock surveillance was last completed
satisfactory at 0800.
-
Then, when performed again at 2130 by operations, the one-rod-out interlock
surveillance failed.
WHAT actions are required in accordance with PNPP Technical Specifications?
a.
Immediately suspend loading of irradiated fuel into the RPV; initiate action to
restore Secondary Containment to operable.
b.
Immediately suspend in-vessel fuel movement with equipment associated with
the inoperable interlock and insert all insertable control rods.
c.
Immediately suspend control rod withdrawal and initiate actions to fully insert all
insertable control rods in cells containing one or more fuel assemblies.
d.
Immediately initiate action to insert all insertable control rods and place the mode
switch in the SHUTDOWN position in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Answer: c.
Plant Reference:
PNPP Technical Specification 3.9.2
Facility Comment:
There are two correct answers. Answer B is also correct because the surveillance
covers 2 Tech Specs - 3.9.1.1 and 3.9.2.2, making Answer B and C correct. The
question states the surveillance fails not just the one rod out interlock.
Subsequent Facility Comment:
The question states the surveillance fails and both specifications are in the
surveillance, not just the one rod out interlock. The surveillance SVI-C71-T0427, covers
SR 3.9.1.1 and 3.9.2.2 making B and C correct. The utility believes that there are two
correct answers.
NRC Resolution:
The NRC examiners reviewed Technical Specification 3.9.1 and 3.9.2 and considered
the initial conditions created within the question stem. The stem only references the
one-rod-out interlock (Technical Specification 3.9.2) and does not reference the
refueling interlocks (Technical Specification 3.9.1). The frequency of the interlock test
also implied that only the one-rod-out interlock was tested. The one-rod-out interlock is
tested every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while the refueling interlocks are only tested once every 7 days.
Even if it is interpreted that Technical Specification 3.9.1 is entered when only the one-
Post Examination Comments and Resolutions
rod-out interlock failed, the immediate actions required by Action Statement A1 are:
suspend in-vessel fuel movement with equipment associated with the inoperable
interlock(s). OR (Action Statement A2.1, A2.2) Insert a control rod withdrawal block
AND Verify all control rods are fully inserted. Distractor (b.) has more actions than
required by Action Statement A1 (insert all insertable control rods), and does not state
all of the actions required by Action Statement A2 (Insert a control rod withdrawal block).
Since distractor (b.) does not contain the correct actions for an entry into Technical Specification 3.9.1, the distractor was considered incorrect and the comment was
rejected.
Technical Specification 3.9.2 is the Refuel Position One-Rod-Out Interlock
specification. The specification states: The refuel position one-rod-out interlock shall
be OPERABLE. The stem of the question stated: the one-rod-out interlock surveillance
failed. This indicated an entry into Technical Specification 3.9.2, not Technical Specification 3.9.1 which refers to refueling equipment interlocks. This material is
required to be memorized since the technical specification completion time is
Immediately. Because the material is required to be memorized and only the
one-rod-out interlock was referenced, the comment was rejected; no change was made
to the answer key for question #65.
Post Examination Comments and Resolutions
Question No. 66
Assume that you receive your license on March 1, 2005, but because of vacation and required
training you do not start standing watches (RO or SRO as applicable) until Monday March 28,
2005 and are scheduled to stand watch through Sunday April 3, 2005. Your shifts are
scheduled for eight hours each day. Select the statement below that describes your license
status on April 1, 2005.
a.
Your license is considered active and you can assume the watch on April 1,
2005. If you stand watches through Sunday, you will not need to stand any more
watches until the July-September quarter to maintain proficiency.
b.
Your license is considered active and you can assume the watch on April 1,
2005. If you stand watches through Sunday, you will need to stand at least four
additional watches before July 1, 2005 to maintain proficiency.
c.
Your license will be considered inactive and you cannot assume the watch on
April 1, 2005. You must complete a minimum of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of shift functions,
under the direction of a licensed RO or SRO as applicable, in the position to
which you are assigned in order to regain active status.
d.
Your license will be considered inactive and you cannot assume the watch on
April 1, 2005. You may regain active status by completing your Friday through
Sunday shifts, under the direction of a licensed RO or SRO as applicable, in the
position to which you are assigned.
Answer: b.
Facility Reference
Facility Comment:
Throw out. These are certification requirements they receive prior to their first watch. No
INPO ACAD requirements require this to be taught prior to receiving a license.
Subsequent Facility Comment:
Based upon further review by the Utility, the comment on this question is being
withdrawn.
NRC Resolution:
No change was made to the answer key for question #66.
Post Examination Comments and Resolutions
Question No. 73
Select the statement below that reflects an Operations Section expectation for TRANSIENT
ALARM RESPONSE during implementation of Plant Emergency Instructions (PEI).
a.
Entry into the TRANSIENT ALARM RESPONSE mode shall be announced by
the Unit Supervisor.
b.
Locked in alarms that are abnormal for the present plant status should be
communicated to the Unit Supervisor.
c.
Recurring alarms that annunciate ONI or PEI entry conditions do NOT need to
be re-announced.
d.
The TRANSIENT ALARM RESPONSE mode will remain in effect until the PEIs
are exited.
Answer: b.
Facility Reference:
Perry Operations Section Expectations Handbook
Facility Comment:
There are two correct answers. Answer A is correct because the alarm response mode
shall be given at the next brief. The answer does not state when the announcement
should be made.
Subsequent Facility Comment:
Reference for transient response (PYBP-POS-2-3) to this question. Answer A is correct
because the alarm response mode shall be given at the next brief. This would make
answer A correct. Answer B is also correct during transient response, as it is the ROs
responsibility to report alarms that are not normal for the situation.
NRC Resolution:
The NRC examiners reviewed the post-examination comment and agree that the stem
of the question was not worded such that distractor (a.) could be eliminated as an
incorrect answer. This condition was not identified during the pre-examination
validation. The answer key was modified to accept distractors (a.) and (b.) as correct
answers to question #73.
Post Examination Comments and Resolutions
Question No. 74
Select the statement below that correctly describes a requirement related to Fire Brigade
composition.
a.
The Fire Brigade Leader must have a Reactor Operator or Senior Reactor
Operators license.
b.
If the Fire Brigade composition drops below the minimum number of five (5), it
must be restored to at least the minimum number within one (1) hour.
c.
Any member of the Operations shift crew may be assigned to the Fire Brigade.
d.
Any site employee who is knowledgeable, trained, and skilled in fire fighting
operations may be a member of the Fire Brigade.
Answer: d.
Facility Reference:
PAP-0126
PAP-1910
Facility Comment:
There are two correct answers. Answer C is correct because all positions on a shift
crew can be fire brigade members. There is no reference that disqualifies any position.
Subsequent Facility Comment:
Based upon further review by the Utility, the comment on this question is being
withdrawn.
NRC Resolution:
No change was made to the answer key for question #74.
Post Examination Comments and Resolutions
Question No. 75
Select the statement below that describes the PAP-0528 sequence adherence requirement
when utilizing Alarm Response Instructions.
a.
Immediate Actions shall be performed in sequence.
b.
Subsequent Actions shall be performed in sequence.
c.
Initiation of a Condition Report is required if Subsequent Actions are not
performed as written.
d.
Initiation of a Condition Report is required if Immediate Actions are performed
out of sequence.
Answer: c.
Facility Reference:
PAP-0528
Facility Comment:
There are two correct answers. Answer B is correct as stated in PAP-0528 as it directs
Subsequent actions should be performed in order along with the site expectation that
should is considered shall. The old procedure used to say the steps may be performed
out of order. This statement has been removed.
Subsequent Facility Comment:
Based upon further review by the Utility, the comment on this question is being
withdrawn.
NRC Resolution:
No change was made to the answer key for question #75.
Post Examination Comments and Resolutions
Question No. 79
The plant was operating at 100% power when a LOCA occurred. All control rods are fully
inserted. LPCS and LPCI A are both injecting into the RPV. NO other ECCS pumps are
available. As long as both pumps are injecting, RPV water level can be maintained above TAF.
Suppression Pool temperature is 130°F and rising. Select the statement below that correctly
describes the use of LPCI A for Suppression Pool cooling.
a.
LPCI A must be diverted to Suppression Pool Cooling to ensure that
Suppression Pool temperature is maintained below the Heat Capacity Limit,
since LPCS can maintain adequate core cooling through spray cooling alone.
b.
LPCI A may be diverted to Suppression Pool Cooling as long as LPCS is able to
maintain RPV water level above -25 inches (the Minimum Steam Cooling RPV
water level).
c.
LPCI A must be diverted to Suppression Pool Cooling, irrespective of adequate
core cooling, when neither Suppression Pool temperature nor Reactor pressure
can be maintained below the Heat Capacity Limit (HCL)
d.
LPCI A may be diverted to Suppression Pool Cooling only if additional injection
sources become available to be used with LPCS to maintain RPV water level
above 0 inches
Answer: d.
Facility Reference
PEI Bases
PEI-B13 and PEI-T23
Facility Comment:
There are two correct answers. Answer B is correct because minimum steam cooling
water level with injection meets the definition of adequate core cooling. Reference: PEI
Bases Definition of Adequate Core Cooling
Subsequent Facility Comment:
PEI bases defines adequate core cooling as level being above the Minimum Zero
Injection Water Level. As Answer B states that LPCI may be diverted to Suppression
Pool Cooling as long as LPCS can maintain RPV water level above -25 inches.
(consistent with the non ATWS flow chart not requiring ED until level cannot be restored
and maintained above -25") answer B is also correct. Answer D is correct because
maintaining level above zero inches ensures the core is cooled.
NRC Resolution:
The NRC examiners reviewed the PEI Bases and found the following information:
Within the EPGs, three viable mechanisms for establishing adequate core cooling are
definedcore submergence, spray cooling, and steam cooling. Submergence is the
preferred method for cooling the core. The core is adequately cooled by submergence
Post Examination Comments and Resolutions
when it can be determined that RPV water level is at or above the top of the active fuel.
All fuel nodes are then assumed to be covered with water and heat is removed by
boiling heat transfer. Adequate spray cooling is provided in BWR/3 through BWR/6
designs. The covered portion of the core is cooled by submergence while the uncovered
portion is cooled by the spray flow. Currently this method is not used and is under
design review to ensure that it is acceptable for use at Perry. Steam cooling is relied
upon only if RPV water level cannot be restored and maintained above the top of the
active fuel, cannot be determined, or must be intentionally lowered below the top of the
active fuel.
The test question indicated that RPV level was only being maintained because two
ECCS pumps were injecting, implying RPV level would go below top of active fuel if
RHR A was diverted to cool the suppression pool. This led the examiners to believe
that only distractor (d.) was a correct answer. However, at the PEI decision step
referenced by this question, the PEI basis stated that it was permissible to alternate the
use of RHR pumps between the RPV injection mode and suppression pool cooling
modes, as the need for each occurs, and so long as adequate core cooling can be
maintained. This implied that it was acceptable to divert RHR A pump to suppression
pool cooling until adequate core cooling was no longer maintained. This information
indicated that distractor (b.) was correct and answer (d.) was not correct. Since the
PEIs allow an operator to divert RHR A without waiting for additional pumps to become
available to support LPCS, then distractor (d.) cannot be correct. This was not identified
during examination validation. The answer key was modified to accept answer (b.) as
the correct answer for question #79.
Post Examination Comments and Resolutions
Question No. 86
HPCS, LPCS, and all three RHR pumps started on High Drywell Pressure. The following
conditions are observed:
-
All control rods are fully inserted
-
RPV Level is 150 in. WR and decreasing slowly
-
RPV Press is 800 psig and decreasing slowly
The BOP Operator reports that RHR Pump A Minimum Flow Valve (F064A) is shut and will not
open. Given the current plant condition which of the following actions would be most
appropriate to assign the balance of plant operator?
a.
Declare RHR Pump A INOPERABLE. Shutdown RHR Pump A and pull its
control power fuses.
b.
Declare RHR Pump A OPERABLE; open the Test Return Valve to Suppression
Pool (F024A) to establish > 1650 gpm.
c.
Declare RHR Pump A INOPERABLE but available. Dispatch a plant operator to
attempt to manually open the Minimum Flow Valve (F064A).
d.
Declare RHR Pump A INOPERABLE, but available. Shutdown RHR Pump A
until reactor pressure is low enough for the injection valve (F042A) to open, then
restart the pump.
Answer: a.
Facility Reference
SOI-E12
Facility Comment:
There are two correct answers. Answer D is correct because the pump
may be required for adequate core cooling and should not be completely
removed from service until it is known whether it is needed or not.
Subsequent Facility Comment:
Based upon further review by the Utility, the comment on this question is being
withdrawn.
NRC Resolution:
No change was made to the answer key for question #86.
Post Examination Comments and Resolutions
Question No. 92
Given the following conditions:
-
Reactor Plant at 100% RTP
-
The following annunciators are in alarm:
-
HOT SURGE TANK LEVEL HI
-
HTR 4 ISOL HOT SRG TK LEVEL HI
-
The Extraction Steam supply and Steam Seal Evaporator drains to Heater 4
have automatically isolated
-
N21-F220, Hot Surge Tank Level Control Bypass Valve indicates closed
-
N21-F230, Hot Surge Tank Level Control Valve is partially open and is
unresponsive to the Hot Surge Tank Level Controller signals (in either AUTO or
MANUAL)
-
Local manual control of N21-F230, Hot Surge Tank Level Control Valve was
unsuccessful
-
Hot Surge Tank level is 150" and increasing slowly
Which one of the following actions should you direct the ATC Operator to perform while
maintaining current power level?
a.
Shutdown one of the Condensate Booster Pumps
b.
Perform the Securing Flow to the Hot Surge Tank section of SOI-N21
c.
Throttle open Condensate Minimum Flow Recirculation Valve (N21-F245, Short
Cycle Clean-Up Valve)
d.
Manually trip all Hotwell and Condensate Booster Pumps
Answer: c.
Facility Reference
SDM N32/85
Facility Comment:
Throw out the question. The only correct answer is non-conservative and
would not be in the best interest of the health and safety of the public. In
a later, clarifying communication the following was provided: with the hot
surge tank level at 150 inches and heater 4 isolated the operator would
enter ONI-N36 for a loss of feedwater heating. This requires reactor
power be lowered, making the stem of the question non-conservative
(Remaining at 100% power). Additionally, three subsequent action steps
should have been executed prior to reaching this step. Subsequent
action steps are not required to be memorized.
Subsequent Facility Comment:
There is no correct answer listed. With a Heater 4 isolation, the ATC is required to take
the immediate actions of ONI-N36, which require the operator to reduce reactor power
to less than 95%. The answer listed as correct is in the subsequent actions of the ARI,
that would not be done until after the immediate actions are completed and power is
decreased. This is asking the SRO to direct a supplemental action, prior to the required
immediate action. If he did order supplemental actions, answers B (ARI H13-P680-
0002-E2, action 4.7.3) C (ARI H13-P680-0002-E2, action 4.5), and D (ARI H13-P680-
0002-E2, action 4.7.4) are all subsequent actions in the ARI that could be used for
mitigating the Hot Surge Tank High level. In this situation, there would be three
potential actions to mitigate the problem. The Utility believes the question should be
deleted as the immediate actions of ONI-N36 require the operator to lower reactor
power.
NRC Resolution:
The pre-examination review did not elicit comments of this nature, but only requested
that the N21-F245 valve be identified as the Short Cycle Clean-Up Valve. The NRC
examiners agree this question would require a violation of the stations procedures to
force the question conditions and should be deleted from the examination. The follow-
up information also indicated that subsequent action steps are not required to be
memorized as this step is a subsequent action in the ONI. The NRC examiners agree
that subsequent action steps are not required to be memorized, however, this question
can be answered with an application of system knowledge. This also explains why three
steps of the procedure were skipped - their memorization was not required. Because
the examination question did not provide correct initial conditions that comply with the
stations procedures in the test question, the question was deleted and the answer key
amended to reflect a deletion of question #92.
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession #ML050480645