ML090720952
| ML090720952 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 03/12/2009 |
| From: | Hironori Peterson Operations Branch III |
| To: | Bezilla M FirstEnergy Nuclear Operating Co |
| References | |
| IR-09-301 | |
| Download: ML090720952 (32) | |
See also: IR 05000440/2009301
Text
March 12, 2009
Mr. Mark Bezilla
Site Vice President
FirstEnergy Nuclear Operating Company
Perry Nuclear Power Plant
P. O. Box 97, 10 Center Road, A-PY-A290
Perry, OH 44081-0097
SUBJECT: PERRY NUCLEAR POWER PLANT, UNIT 1
NRC INITIAL LICENSE EXAMINATION REPORT 05000440/2009301(DRS);
Dear Mr. Bezilla:
On February 27, 2009, the Nuclear Regulatory Commission (NRC) examiners completed initial
operator licensing examinations at your Perry Nuclear Power Plant. The enclosed report
documents the results of the examination which were discussed on January 16, 2009, with
Mr. A. Mueller Jr. and other members of your staff. An exit meeting was conducted by telephone
on March 4, 2009, between Mr. A. Mueller Jr. of your staff and Mr. Walton, of Operator
Licensing, to review the resolution of the station=s post examination comments and the proposed
final grading of the written examination for the license applicants.
The NRC examiners administered an initial license examination operating test during the week
of January 12, 2009. The written examination was administered by Perry Nuclear Power Plant
training department personnel on January 21, 2009. Eight Senior Reactor Operators and one
Reactor Operator applicant were administered license examinations. The results of the
examinations were finalized on February 27, 2009. All applicants passed all sections of their
respective examinations and seven were issued senior operator licenses and one was issued an
operator license. One senior operator license was withheld until the individual met experience
requirements.
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 NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
M. Bezilla -2-
We will gladly discuss any questions you have concerning this examination.
Sincerely,
/RA/
Hironori Peterson, Chief
Operations Branch
Division of Reactor Safety
Docket Nos. 50-440
License Nos. NPF-58
Enclosures: 1. Operator Licensing Examination
Report 05000440/2009301(DRS)
w/Attachment: Supplemental Information
2. Simulation Facility Report
3. Post Examination Comments w/NRC Resolution
4. Written Examinations and Answer
Keys (SRO)
cc w/encls 1 & 2: J. Hagan, President and Chief Nuclear Officer - FENOC
J. Lash, Senior Vice President of Operations and
Chief Operating Officer - FENOC
D. Pace, Senior Vice President, Fleet Engineering - FENOC
K. Fili, Vice President, Fleet Oversight - FENOC
P. Harden, Vice President, Nuclear Support
Director, Fleet Regulatory Affairs - FENOC
Manager, Fleet Licensing - FENOC
Manager, Site Regulatory Compliance - FENOC
D. Jenkins, Attorney, FirstEnergy Corp.
Public Utilities Commission of Ohio
C. OClaire, State Liaison Officer, Ohio Emergency Management Agency
R. Owen, Ohio Department of Health
cc w/encls 1, 2, 3, & 4: A. Mueller, Jr. Training Director, Perry Power Plant
M. Bezilla -2-
We will gladly discuss any questions you have concerning this examination.
Sincerely,
/RA/
Hironori Peterson, Chief
Operations Branch
Division of Reactor Safety
Docket Nos. 50-440
License Nos. NPF-58
Enclosures: 1. Operator Licensing Examination
Report 05000440/2009301(DRS)
w/Attachment: Supplemental Information
2. Simulation Facility Report
3. Post Examination Comments w/NRC Resolution
4. Written Examinations and Answer
Keys (SRO)
cc w/encls 1 & 2: J. Hagan, President and Chief Nuclear Officer - FENOC
J. Lash, Senior Vice President of Operations and
Chief Operating Officer - FENOC
D. Pace, Senior Vice President, Fleet Engineering - FENOC
K. Fili, Vice President, Fleet Oversight - FENOC
P. Harden, Vice President, Nuclear Support
Director, Fleet Regulatory Affairs - FENOC
Manager, Fleet Licensing - FENOC
Manager, Site Regulatory Compliance - FENOC
D. Jenkins, Attorney, FirstEnergy Corp.
Public Utilities Commission of Ohio
C. OClaire, State Liaison Officer, Ohio Emergency Management Agency
R. Owen, Ohio Department of Health
cc w/encls 1, 2, 3, & 4: A. Mueller, Jr. Training Director, Perry Power Plant
DISTRIBUTION:
RidsNrrPMPerry
RidsNrrDorlLpI3-2
RidsNrrDirsIrib Resource
Mark Satorius
Kenneth Obrien
Cynthia Pederson (hard copy - IRs only)
DRPIII
DRSIII
Patricia Buckley
ROPreports@nrc.gov
NLS
DOCUMENT NAME: G:\DRS\WORK IN PROGRESS\PER 2009 301 DRS OL.DOC
G Publicly Available G Non-Publicly Available G Sensitive G Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII
NAME KWalton:co HPeterson
DATE 03/06/09 03/12/09
OFFICIAL RECORD COPY
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket Nos: 50-440
License Nos: NPF-58
Report No: 05000440/2009301(DRS)
Licensee: First Energy Corporation
Facility: Perry Nuclear Power Plant, Unit 1
Location: Perry, Ohio
Dates: January 12 - January 21, 2009
Examiners: R. Walton, Senior Operations Engineer
D. Reeser, Operations Engineer
C. Zoia, Operations Engineer
Approved by: Hironori Peterson, Chief
Operations Branch
Division of Reactor Safety
Enclosure 1
SUMMARY OF FINDINGS
ER 05000440/2009301(DRS); 1/12/2009 - 1/21/2009; First Energy Corp., Perry Station Initial
License Examination Report.
The announced initial operator licensing examination was conducted by regional Nuclear
Regulatory Commission examiners in accordance with the guidance of NUREG-1021, AOperator
Licensing Examination Standards for Power Reactors,@ Revision 9.
Examination Summary:
$ Nine of nine applicants passed all sections of their respective examinations. Seven
applicants were issued senior operator licenses and one applicant was issued an
operator license. One senior operator will be issued a license after experience
conditions have been met (Section 4OA5.1).
- The examiners identified that the licensee used software that incorporated two-phase
fluid flow for modeling feedwater in the Perry simulator. This software has been used in
some but not all BWR simulators. This condition is an unresolved item pending further
review by the NRC (Enclosure 2).
1 Enclosure 1
REPORT DETAILS
4. OTHER ACTIVITIES (OA)
4OA5 Other
.1 Initial Licensing Examinations
a. Examination Scope
The Nuclear Regulatory Commission=s examiners prepared the examination outline and
developed the written examination and operating test. The NRC examiners validated the
proposed examination during the week of December 15, 2008, at the Perry Nuclear
Power Station Training Building with the assistance of members of the licensee training
staff. During the on-site validation week on December 15, 2008, the examiners audited
two license applications for accuracy. The NRC examiners conducted the operating
portion of the initial license examination during the week of January 12, 2009. The NRC
examiners and members of the Perry Nuclear Power Station training department staff
administered the written examination on January 21, 2009. The NRC examiners used
the guidance established in NUREG-1021, AOperator Licensing Examination Standards
for Power Reactors,@ Revision 9, to prepare, validate, revise, administer, and grade the
examination.
b. Findings
Written Examination
During the validation of the written examination several questions were modified or
replaced. Changes made to the written examination were documented on Form
ES-401-9, AWritten Examination Review Worksheet@ which is available electronically in
the NRC Public Document Room or from the Publicly Available Records component of
NRC's document system (ADAMS). The licensee submitted four written examination
post-examination comments for consideration by the NRC examiners when grading the
written examination. The post-examination comments and the NRC resolution for the
post-examination comments are contained in Enclosure 3, APost Examination Comments
and Resolutions.@ The NRC examiners graded the written examination on
February 19, 2009, and conducted a review of each missed question to determine the
accuracy and validity of the examination questions.
Operating Test
During the validation of the operating test, two Job Performance Measures (JPMs) were
modified and changes were made to the dynamic simulator scenarios. The JPMs were
replaced since the JPMs were determined to be too simplistic in nature (inadequate
difficulty level). Changes made to the operating test were documented in a document
titled, AOperating Test Comments,@ which is available electronically in the NRC Public
Document Room or from the Publicly Available Records component of NRC's document
system (ADAMS). The NRC examiners completed operating test grading on
February 19, 2009.
2 Enclosure 1
Examination Results
Eight applicants at the Senior Reactor Operator (SRO) level and one applicant at the
Reactor Operator (RO) level were administered written and operating tests. Two of the
SRO applicants were previously licensed as ROs at Perry Power Station. Nine
applicants passed all portions of their examinations and eight applicants were issued
operating licenses. One applicants license was withheld until experience requirements
had been met.
.2 Examination Security
a. Scope
The NRC examiners reviewed and observed the licensee's implementation of
examination security requirements during the examination validation and administration
to assure compliance with 10 CFR 55.49, AIntegrity of Examinations and Tests.@ The
examiners used the guidelines provided in NUREG 1021, "Operator Licensing
Examination Standards for Power Reactors@ to determine acceptability of the licensee=s
examination security activities.
b. Findings
No Findings
4OA6 Meetings
Debrief
The chief examiner presented the examination team's preliminary observations
and findings on January 16, 2009, to A. Mueller, Jr., and other members of the
Perry Operations and Training Department staff.
Exit Meeting
The chief examiner conducted an exit meeting on March 4, 2009, with Mr. A. Mueller, Jr.,
Perry Station Training Director by telephone. The NRC=s final disposition of the station=s
post-examination comments were disclosed and revised preliminary written examination
results were provided to A. Mueller, Jr., during the telephone discussion. The examiners
asked the licensee whether any of the material used to develop or administer the
examination should be considered proprietary. No proprietary or sensitive information
was identified during the examination or debrief/exit meetings.
ATTACHMENT: SUPPLEMENTAL INFORMATION
3 Enclosure 1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
K. Krueger, Plant General Manager
D. Evans, Manager Operations
A. Cayia, Director Performance Improvement
R. Coad, Manager - Regulatory Compliance
A. Mueller, Jr., Manager Training
J. Pelcic, Nuclear Compliance
D. Zielinsky, Training Department
R. Torres, Training Department
J. Kelley, Training Department
D. Richmond, Training Department
NRC
R. Walton, Chief Examiner
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
UNR (00050/440-2009-301-01), Two-Phase Fluid Flow Modeling for Feedwater
Closed
None
Discussed
None
LIST OF DOCUMENTS REVIEWED
None
LIST OF ACRONYMS USED
ADS Automatic Depressurization System
ADAMS Agency-Wide Document Access and Management System
BWR Boiling Water Reactor
DRS Division of Reactor Safety
NRC Nuclear Regulatory Commission
IR Inspection Report
SPDS Safety Parameters Display System
1 Attachment
SIMULATION FACILITY REPORT
Facility Licensee: Perry Nuclear Power Station
Facility Docket No: 50-440
Operating Tests Administered: 1/12/2009 - 1/16/2009
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:
Incorporation of Two-Phase Fluid Flow Modeling for Feedwater
During onsite validation week of the initial license exam at the Perry station, the inspectors noted
that reactor vessel water level appeared to increase with no high pressure injection or operator
intervention after emergency depressurization. With reactor pressure lowering, the reactor
vessel water level swelled about 100 inches. The vessel water level then lowered due to the
ADS valves being opened.
The licensee informed the examiners that previously they had loaded computer software that
changed the high pressure feedwater injection from a single phase fluid flow model to a
two-phase fluid flow model. This resulted in a flashing of high temperature feedwater in
feedwater heaters #6. This condition produced flow into the reactor vessel after the vessel
pressure lowered to the high pressure feedwater heater #6 saturation pressure.
This simulator computer model was taught to the initial license class and requalification classes.
The initial license exam was administered during the week of January 12, 2009, with this
simulator software included.
The examiners determined that this simulator modeling had been included at two other facilities
in the industry. The examiners were uncertain of the pedigree and approval status of this
computer software modeling since it had not yet been approved by the BWR owners group, and
had not been widely accepted by other BWR utilities. This issue was considered an Unresolved
Item (50-440/2009301-01) pending further review by NRC Headquarters Operations staff.
Change in Simulator Modeling between On-Site Validation and Exam Administration
During the week of December 15, 2008, the examiners validated the Perry operating exam with
an operating crew. The operating crew used the Safety Parameters Display System (SPDS)
display screens in the overhead of the simulator to track and trend various parameters important
1 Enclosure 2
to equipment operation and plant monitoring. The SPDS computer received inputs from the
simulator computer. On January 9, 2009, the licensee implemented a change to the SPDS
process computer that was believed to be a graphics change - a change that would not alter
computer modeling.
The following week, on Monday, January 12, 2009, during the administration of the initial license
operating test, the examiners, examinees and licensee simulator operators noted that the SPDS
computer display panel did not accurately display reactor vessel wide range water level.
The following day, after running the first scenario, and seeing that the SPDS computer had
rejected wide range reactor vessel water level input from the simulator computer, the licensees
staff determined that a change to the SPDS computer software had occurred since onsite
validation. Specifically, a change to the SPDS computer included a file that inhibited the SPDS
computer from receiving wide range input from the simulator computer. As a result, the SPDS
displays for wide range level indication were erroneous. The scenarios were continued with this
software change included until the file was removed on Tuesday night, January 13, 2009.
NUREG 1021, ES-301-4, item 8 required that computer modeling not be changed between
onsite validation and exam administration. Since the SPDS computer software was changed
that affected important monitored parameters, the examiners believed there was a potential for
invalidating the Perry Initial operating exam for January 12 and 13, 2009.
The Operator Licensing Branch in Headquarters reviewed this event and determined that the
exam was not invalidated. The erroneous indications on the SPDS panel were clearly identified
by their color, the applicants had access to accurate wide range level indications on the main
control boards and that all other functions worked normally. There was no reason to treat this
any different than any other instrumentation malfunction or to invalidate the affected scenarios.
This event described illustrated the risk of making even simple changes that were not expected
to alter the simulators response.
2 Enclosure 2
POST EXAMINATION COMMENTS AND RESOLUTIONS
RO Question Number 17
A plant startup is in progress per IOI-0001 Cold Startup. The following plant conditions exist:
Reactor Pressure 200 psig
Main Condenser Vacuum 5.0 HgA
Mechanical Vacuum Pumps are being cycled to maintain vacuum
Main Turbine Warming is in progress
Motor Feed Pump is providing Reactor Level Control
TBCC Pumps A and B operating
The following alarm is received on 1H13-P870, TBCC PUMP SUCTION FLOW LOW. The
operator checks TBCC Parameters at 1H13-P870 with the following indications:
TBCC A Pump red and green light off, no discharge pressure indicated.
TBCC B Pump red light on, green light off, no discharge pressure indicated.
TBCC C Pump red light off, green light on, no discharge pressure indicated.
Per ONI-P44 Loss of Turbine Building Closed Cooling, an ____.
A. immediate scram may not be necessary because the Main Turbine is not in operation
B. immediate scram may not be necessary because reactor pressure control is on the
Bypass Valves
C. immediate scram is required because the Motor Feed Pump is providing level control
D. immediate scram is required because the Mechanical Vacuum Pumps can not be cycled
Answer: A
Reference: ONI-P44, Loss of Turbine Building Closed Cooling, Revision 7, Page 5
1 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Applicant Comment:
An applicant asserted that the answer key should be changed so that distractor C was the only
correct answer.
The applicant provided the following justification:
1. The question asks what actions are required for a loss of Turbine Building Closed Cooling
(TBCC) per ONI-P44, Loss of Turbine Building Closed Cooling. The stem of the question
contains the status of the TBCC pumps with indication of NO discharge pressure for any of
the 3 TBCC pumps. Lack of discharge pressure is symptomatic of a system leak and a total
loss of TBCC.
2. Plant TBCC pump discharge pressure indicates 16-17 psig when in standby due to the
height of water from the expansion tank. The bottom of surge tank is at elevation 66010
and pump suction is at elevation 6259.250.
3. The procedure directs that for a total loss of TBCC, the reactor be scrammed.
(ONI-P44 immediate action 3.4).
4. Core flow is < 58 mlbm during plant start-up at 200 psig. (No core flow reduction required.)
The applicant provided the following justifications for the distractors:
A. Incorrect answer - Based on a total loss of TBCC an immediate scram is required no
standby TBCC pump is available - no discharge pressure indicated which signifies a leak
in the system.
B. Incorrect answer - Based on a total loss of TBCC an immediate scram is required no
standby TBCC pump is available - no discharge pressure indicated which signifies a leak
in the system. Bypass valve HPUs require shutdown at 150 degrees in sump.
C. Correct answer - Based on a total loss of TBCC an immediate scram is required and the
Feed and Condensate system will be shut down when temperature limits are reached.
D. Incorrect answer - Based on a total loss of TBCC an immediate scram is required
however the Mechanical Vacuum Pumps CAN be cycled. The shutdown limit at 102
degrees F can be exceeded (reference ONI-P44 Attachment 1 limits).
Reference: ONI-P44, Loss of Turbine Building Closed Cooling, pages 5, 8, 10, 11.
2 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Facility Proposed Resolution:
The facility agreed with the applicant and stated that the answer key should be changed so that
distractor C was the only correct answer. The facility also stated that the question asked what
actions were required for a loss of Turbine Building Closed Cooling (TBCC) per ONI-P44. The
stem of the question contained the status of the TBCC pumps with indication of no discharge
pressure for any of the 3 TBCC pumps. The procedure directed that for a total loss of TBCC the
reactor be scrammed. The facility referred to the applicants comments for details.
Reference: ONI-P44, Loss of Turbine Building Closed Cooling, pages 5, 8, 10, 11.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
resolved to delete the question from the examination.
The question asked what actions were required for a loss of Turbine Building Closed Cooling
(TBCC) per ONI-P44, Loss of Turbine Building Closed Cooling. The stem of the question
contained the status of the TBCC pumps with indication of no discharge pressure for any of the
3 TBCC pumps. Lack of discharge pressure would be symptomatic of a system leak and a
complete loss of TBCC.
Facility procedure ONI-P44, Loss of Turbine Building Closed Cooling, Step 3.4, required that
for a complete loss of TBCC, the reactor be scrammed. However, a NOTE preceding this step
qualified this step by stating:
The Reactor is shutdown in anticipation of loss of cooling to various loads, e.g.,
Generator Stator. An immediate shutdown may NOT be necessary if the main turbine is
NOT in operation.
Step 4.3.8 of ONI-P44 also required shutdown of TBCC components that reached their
temperature limit. Attachment 1 of ONI-P44, TBCC Served Component Limitations, provided
the temperature limits for the Motor Feedwater Pump and the Mechanical Vacuum Pumps. In
Attachment 1, the temperature limitation of 102°F for the Mechanical Vacuum Pumps had an
asterisk that stated:
This limit for vacuum considerations only and may be exceeded.
However, a NOTE preceding Step 4.3.8 stated:
The Mechanical Vacuum Pumps should NOT be used due to the loss of cooling water to
the seal water coolers.
Based on the above information, the following distractor evaluation was performed:
3 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Distractor A was considered a correct answer based on the NOTE preceding Step 3.4 of
ONI-P44, because it stated that an immediate scram may NOT be necessary if the main turbine
is not in operation. For the conditions stated in the question stem, the main turbine was not in
operation, and thus the NOTE was applicable. Thus, distractor A, immediate scram may not be
necessary because the Main Turbine is not in operation, was a correct answer.
Distractor B was an incorrect answer based on the following:
- ONI-P44, Step 3.4, required that for a complete loss of TBCC, the reactor be scrammed.
- With the temperature limitation of 150°F provided in Attachment 1 of ONI-P44, for the Steam
Bypass Hydraulic Power Unit (HPU) Reservoir, the Bypass Valves could not be used for any
significant time period before its temperature limit was reached and a scram was required.
Thus, distractor B, immediate scram may not be necessary because reactor pressure control is
on the Bypass Valves, remained an incorrect answer.
Distractor C is a correct answer based on the following:
- ONI-P44, Step 3.4 required that for a complete loss of TBCC the reactor be scrammed.
- The question stem stated that the Motor Feedwater Pump was providing Reactor Level
Control. With the temperature limitations provided in Attachment 1 of ONI-P44, for the Motor
Feedwater Pump, the Motor Feedwater Pump could not be run for any significant time period
before its temperature limit was reached and a scram was required.
Thus, distractor C, immediate scram is required because the Motor Feed Pump is providing
level control, was a correct answer.
Distractor D is a correct answer based on the following:
- ONI-P44, Step 3.4, required that for a complete loss of TBCC, the reactor be scrammed.
- The NOTE preceding Step 4.3.8 stated that the Mechanical Vacuum Pumps should NOT be
used due to the loss of cooling water to the seal water coolers.
- With the temperature limitations provided in Attachment 1 of ONI-P44, for the Mechanical
Vacuum Pumps, the Mechanical Vacuum Pumps could not be run for any significant time
period before its temperature limit was reached and a scram was required.
Thus, distractor D, immediate scram is required because the Mechanical Vacuum Pumps can
not be cycled, was a correct answer.
Finally, the Examiners Standard, ES-403, part D.1.c stated:
If it is determined that there are two correct answers, both answers will be
accepted as correct. If, however, both answers contain conflicting information,
the question will likely be deleted. For example, if part of one answer states that
operators are required to insert a manual reactor scram, and part of another
answer states that a manual scram is not required, then it is unlikely that both
answers will be accepted as correct, and the question will probably be deleted.
4 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
If three or more answers could be considered correct or there is no correct
answer, the question shall be deleted.
Since there were three correct answers (A, C, and D) identified for the question, and two
combinations of these answers could not logically be true at the same time (A and C or D), it was
resolved to delete the question from the examination.
5 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
SRO Question Number 12
The plant was operating at 100% reactor power when a grid disturbance caused a generator
load rejection. This resulted in a reactor scram. All plant equipment responded as designed.
Per RPS Instrumentation Tech Spec Bases, the primary scram signal analyzed to provide
protection from a generator load rejection event is __(1)__.
As the Unit Supervisor you direct a reactor level band of __(2)__ per EOP-1 RPV Control.
1 2
A. reactor vessel steam dome pressure high 130 to 219
B. reactor vessel steam dome pressure high 178 to 219
C. turbine control valve fast closure, trip oil pressure low 130 to 219
D. turbine control valve fast closure, trip oil pressure low 178 to 219
Answer: A
References: Technical Specification (TS) 3.3.1.1 Bases
Updated Safety Analysis Report (USAR), Revision 12
EOP-1 Guideline, Revision 0
6 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Applicant Comment:
An applicant asserted that the answer key should be changed so that distractor C should also be
accepted as correct, in addition to distractor D.
The applicant provided the following basis for this reasoning:
1. The questions asks the bases for the generator load rejection scram and the level band the
Unit Supervisor would direct following a grid disturbance that results in a generator load
rejection.
2. Turbine Control Valve - Fast Closure, Trip Oil Pressure Low is the bases for the generator
load rejection.
3. A level band of 130-219 inches is correctly given per the EOP bases for step RLC-4 in
EOP-1, which states the wide RPV water level control band permitted by this step is
sufficient to assure adequate core cooling yet avoid unwarranted demands on an operators
attention.
a. If unnecessarily constrained within narrower limits, an operator may be less effective in
performing concurrent duties.
4. The narrower band of 178-219 inches is suggested per Guideline 2 in EOP-1. Per the EOP
bases guidelines provide supplemental information to the operator.
5.
a. A guideline flag in the flowpath refers to the guideline text.
b. The Guideline text provides supplemental bases information that the operator can call
upon if need to help in decision making and performance of the flowcharts.
6. Since these steps are at the same relative location in the level leg of EOP-1 either band
would be correct to order with the given conditions within the question and might be
amended as plant conditions and Control Room work load changes.
7. The recommendation is that two answers (C and D) are correct because the two level bands
given would be correct if directed.
The applicant provided the following distractor analysis:
A. Incorrect answer - Reactor vessel steam dome pressure high is not the bases for the
trip.
B. Incorrect answer - Reactor vessel steam dome pressure high is not the bases for the
trip.
C. Correct answer - Turbine Control Valve - Fast Closure, Trip Oil Pressure Low and a
Level Band of 130-219 for the wider level band is also correct.
7 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
D. Correct answer - Turbine Control Valve - Fast Closure, Trip Oil Pressure Low and a
Level Band of 178-219 for the narrow level band is still correct.
References: Technical Specification 3.3.1.1 Bases, pages 3.3-11 and 3.3-18
EOP Bases, page 22
EOP-1, RPV Control Bases, pages 25 and 26
Facility Proposed Resolution:
The facility agreed with the applicant and commented that the answer key should be changed so
that distractor C should also be accepted as correct, in addition to distractor D. The facility
stated that the question asked the basis for the generator load rejection scram and the level
band the Unit Supervisor would direct following a grid disturbance that results in a generator
load rejection. Turbine Control Valve Fast Closure Trip Oil Pressure Low is the bases for the
generator load rejection, a level band of 130-219 inches is correctly given per the EOP basis and
the narrower band of 178-219 inches is suggested per Guideline 2. The facility referred to the
applicants comments for details.
References: Technical Specification 3.3.1.1 Bases, pages 3.3-11 and 3.3-18
EOP Bases, page 22
EOP-1, RPV Control Bases, pages 25 and 26
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to accept the facilitys comment and accept both distractors C and D as correct
answers.
The question asked for the Technical Specification (TS) Bases (from the Reactor Protection
System (RPS) Instrumentation section of the TS) for the generator load rejection event and the
level band the Unit Supervisor would direct following a grid disturbance that resulted in a
generator load rejection and resultant reactor scram.
From page B.3.3-18 of the RPS Instrumentation TS Bases, the Turbine Control Valve Fast
Closure, Trip Oil Pressure Low function is the primary scram signal for the generator load
rejection event.
From EOP-01, RPV Control, Revision A, step RLC-4 stated to Restore and Maintain RPV level
between 130 inches and 219 inches. The EOP Bases document for this step stated:
The wide RPV water level control band permitted by this step is sufficient to assure
adequate core cooling yet avoid unwarranted demands on an operators attention.
If unnecessarily constrained within narrower limits, an operator may be less effective in
performing concurrent duties.
In EOP-01, prior to step RLC-4, there was a guideline flag, which referred to Guideline 2. In
EOP-01, a list of General Guidelines was provided in a text box at the bottom of the flowchart.
General Guideline 2 stated:
8 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
If other EOP actions have a higher priority allow HPCS or RCIC to operate in automatic
between Level 2 and Level 8. Closely monitor system operation. The level band should be
expanded above Level 8 and below Level 2 to allow the system to operate within the level
band. Maintain level above 100 inches to 260 inches. RPV level should be maintained 178
to 219 inches whenever possible and shall be greater than 178 inches whenever shutdown
cooling is in service.
The EOP Bases document associated with the Guideline text stated:
Guidelines provide supplemental information to the operator. A guideline flag in
the flowpath refers to the guideline text. The Guideline text provides
supplemental bases information that the operator can call upon if needed to help
in decision making and performance of the flowcharts.
Based on the above information, the following distractor evaluation was performed:
Distractors A and B were incorrect in that the TS Bases for the generator load rejection event
was turbine control valve fast closure, trip oil pressure low, and not reactor vessel steam dome
pressure high. In addition, from page B.3.3-11 of the RPS Instrumentation TS Bases, no
specific safety analysis took credit for the Reactor Vessel Steam Dome Pressure High function.
Distractor C was a correct answer based on the following:
- The TS Bases for the generator load rejection event was turbine control valve fast closure,
trip oil pressure low, and
- Step RLC-4 of EOP-01, which stated to restore and maintain RPV level between 130 inches
and 219 inches. In addition, the EOP Bases document for this step stated that the wide RPV
water level control band permitted by this step was sufficient to assure adequate core
cooling yet avoid unwarranted demands on an operators attention, and if unnecessarily
constrained within narrower limits, an operator may be less effective in performing
concurrent duties.
Distractor D is a correct answer based on the following:
- The TS Bases for the generator load rejection event is turbine control valve fast closure, trip
oil pressure low, and
- Step RLC-4 of EOP-01, which stated to restore and maintain RPV level between 130 inches
and 219 inches. The level band of 178 to 219 inches was encompassed by the level band of
130 to 219 inches stated in step RLC-4. In addition, the narrower band of 178 to 219 inches
was suggested per Guideline 2 in EOP-01. Per the EOP Bases, the guidelines provided
supplemental information to the operator that the operator could call upon if needed to help
in decision making and performance of the flowcharts.
Therefore, the answer key was modified to accept both distractors C and D as correct answers.
9 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
SRO Question Number 16
A plant startup is in progress with reactor power at 29%.
Number 1 Turbine Bypass Valve fails open.
Full Core Display, RPC MODE light is between GP1-4 Full Out and LO Power Set PT marks.
The __(1)__ and the Unit Supervisor would suspend control rod __(2)__.
(1) (2)
A. Rod Withdrawal Limiter is Inoperable Withdrawal
B. Rod Withdrawal Limiter is Inoperable movement except by scram
C. Rod Pattern Controller is Inoperable Withdrawal
D. Rod Pattern Controller is Inoperable movement except by scram
Answer: B
References: Technical Specification 3.3.2.1
ARI-H13-P680-0005-C9, Revision 11
Applicant Comment:
An applicant asserted that the question should be deleted from the examination since none of
the distractors were correct.
The applicant provided the following justification:
1. The question asks for the operator to make a declaration of Operability in accordance with
Technical Specifications for either the Rod Pattern Controller or the Rod Withdrawal limiter
and then determine the required actions for control rod movement following a failed bypass
valve with reactor power at 29%.
2. PNPP Technical Specifications do not require the Rod Withdrawal Limiter (RWL) to be
Operable until greater than 33.3% RTP and the Rod Pattern Controller (RPC) is only
required to be operable when less than or equal to 19% RTP.
3. The question asks what to do if at 29% RTP. In accordance with TS 3.3.2.1 APPLICABILITY:
According to Table 3.3.2.1-1, there is none since current power is between 19% and 33.3%.
Since there is no applicability there is no operability requirement and no required action until
10 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
RTP is either greater than 33.3% or less than 19% RTP. This condition would be addressed
by a Potential LCO and administratively tracked.
4. Therefore, Answer A (i.e. Rod Withdrawal Limiter is Inoperable and the Unit Supervisor
would suspend control rod withdrawal) is incorrect.
The applicant provided the following distractor analysis:
A. Incorrect answer - RWL is NOT required to be operable therefore no required action.
B. Incorrect answer - RWL is NOT required to be operable therefore no required action.
C. Incorrect answer - RPC is NOT required to be operable therefore no required action.
D. Incorrect answer - RPC is NOT required to be operable therefore no required action.
References: Technical Specification 3.3.2.1, Control Rod Block Instrumentation and
Table 3.3.2.1-1
Facility Proposed Resolution:
The facility agreed with the applicant and commented that the question should be deleted from
the examination since none of the distractors were correct. The facility stated that the question
asked for the operator to make a declaration of Operability for either the Rod Pattern Controller
or the Rod Withdrawal Limiter and determine the required actions for control rod movement
following a failed bypass valve with reactor power at 29%. The facility commented that the Perry
Nuclear Power Plant Technical Specifications do not require the Rod Withdrawal Limiter to be
Operable until greater than 33.3% reactor thermal power (RTP) and required the Rod Pattern
Controller only operable at less than or equal to 19% RTP. The facility referred to the applicants
comments for details.
References: Technical Specifications 3.3.2.1, page 3.3-15 and Table 3.3.2.1-1
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to it was decided to delete the question from the examination.
The question asked for the Unit Supervisor to make a declaration of operability for either the Rod
Pattern Controller or the Rod Withdrawal Limiter, and to determine the required actions for
control rod movement following a failed open #1 turbine bypass valve with reactor power at 29%
and a plant startup in progress.
Technical Specification (TS) 3.3.2.1, Control Rod Block Instrumentation, defined the TS
operability requirements for both the Rod Withdrawal Limiter (RWL) and the Rod Pattern
Controller (RPC). The Applicability of the Control Rod Block Instrumentation is according to
Table 3.3.2.1-1 of the TS. In Table 3.3.2.1-1 of the TS, the Applicable Modes or Other
11 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Specified Conditions, was given as:
- For the RWL, either Reactor Thermal Power (RTP) greater than 66.7% or between
greater than 33.3% and less than or equal to 66.7%.
- For the RPC, RTP less than or equal to 19.0%, except during the reactor shutdown
process if the coupling of each withdrawn control rod has been confirmed.
Based on the above information, the Technical Specifications did not require either the Rod
Withdrawal Limiter to be operable until greater than 33.3% RTP, and the Rod Pattern Controller
was only required to be operable when less than or equal to 19% RTP. Since the question
asked for actions required with the plant at 29% reactor power, the TS did not apply, and there
was no operability requirement and no required action for this plant condition. Thus, there were
no correct answers, and it was decided to delete the question from the examination.
12 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
SRO Question Number 25
Alert JA1 was declared.
Emergency Coordinator duties remain with the Shift Manager.
When the Shift Manager is ready to terminate from event, the Shift Manager is responsible to
terminate the event ____.
Reference Provided: EPI-A1 Attachments 1 & 2
A. after consulting with the NRC, State and local counties
B. after consulting with the State and local counties
C. after consulting with the NRC
D. without consultation
Answer: D
References: EPI-A1, page 11
EPI-A2, pages 13 and 17
Applicant Comment:
An applicant asserted that the question should be deleted from the examination since none of
the distractors were correct.
The applicant provided the following justification:
The question asks for the required consultations when terminating from an Alert but does not
specify a specific instruction. There are two references listed which include conflicting and in
one case, vague guidance on correct execution.
a. In accordance with EPI-A-2, Emergency Action Levels, Section 5.3.1.6 and on Event
Termination Actions Checklist, item # A.4 - For events classified as an Alert or above,
the NRC, State of Ohio, and local counties have been consulted regarding event
termination.
b. In accordance with EPI-A-1, Emergency Actions Based on Event Classification,
Section 5.5.1.11 - Consult with NRC, State of Ohio, and local county officials regarding
the decision to terminate the emergency. The intent of this action is to involve the NRC,
State and local counties in event decision making; however, this action is not intended to
13 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
delay or hinder the Perry Plants ability to terminate from an Unusual Event or Alert that
no longer meets the criteria for any event at the time of declaration.
1) The last line of the step may given the correct circumstances contradict the first
line in the step.
2) This issue has been identified and a Condition Report is to be written to resolve
the differences between the above two procedures
The applicant provided the following distractor analysis:
A. Correct answer - IAW EPI A2 page 14 and 17, PNPP is to consult with the NRC, State
and local county officials regarding event termination, Answer A lists all three per the
EPI.
B. Correct answer - IAW EPI A2 page 14 and 17, PNPP is to consult with the NRC, State
and local county officials regarding event termination, Answer B just lists the State and
local county officials however it does not say only, even though the NRC is also to be
consulted, Answer B is not wrong, Answer B is a subset of Answer A.
C. Correct answer - IAW EPI A2 page 14 and 17, PNPP is to consult with the NRC, State
and local county officials regarding event termination, Answer C just lists the NRC
however it does not say only, even though the State and local county officials are also to
be consulted, Answer C is not wrong, Answer C is a subset of Answer A.
D. Correct answer - EPI A1 page 11 (this is a different reference than for Answers A, B, C),
PNPP is to consult with the NRC, State and local county officials regarding event
termination however during an Unusual Event or an Alert this action is not intended to
delay or hinder PNPPs ability to terminate from an Unusual Event or an Alert. Therefore
given the correct circumstances consultation is not required.
References: EPI A1 page 11
EPI A2 pages 14 and 17
Facility Proposed Resolution:
The facility agreed with the applicant and commented that the question should be deleted from
the examination since none of the distractors were correct. The facility stated that the question
asked for the required consultations when terminating from an Alert in accordance with
EPI-A-0001, Emergency Actions Based on Event Classification, Section 5.5.1.11, Answers A, B,
C are correct. In accordance with EPI-A-0002, Emergency Action Levels, Section 5.3.6,
Answer D is correct. The facility referred to the applicants comments for details.
References: EPI A1 page 11
EPI A2 pages 14 and 17
14 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to it was decided to delete the question from the examination.
The question asked for the required consultations when terminating from an Alert condition in
accordance with the emergency plan procedures.
In accordance with EPI-A2, step 5.3.1.6, it stated:
For events classified as an Alert or above, the NRC, State of Ohio, and local counties
have been consulted regarding event termination.
In accordance with EPI-A2, Attachment 2, Section A, Event Termination Actions, step 6, it
stated:
[ALERT OR ABOVE ONLY] NRC, State of Ohio, and local counties consulted regarding
the decision to terminate the emergency.
In accordance with EPI-A1, step 5.5.1.11, it stated:
Consult with NRC, State of Ohio, and local county officials regarding the decision to
terminate the emergency. The intent of this action is to involve the NRC, State and local
counties in event decision making; however, this action is not intended to delay or hinder
the Perry Plants ability to terminate from an Unusual Event or Alert that no longer meets
the criteria for any event at the time of declaration.
Based on the above information, the following distractor evaluation was performed:
Distractor A was a correct answer based on EPI-A2, step 5.3.1.6, EPI-A1, step 5.5.1.11, and
EPI-A2, Attachment 2, Section A, Event Termination Actions, step 6, which stated that for
events classified as an Alert or above, the NRC, State of Ohio, and local counties have been
consulted regarding event termination.
Distractor B was a correct answer based on EPI-A1, step 5.5.1.11; EPI-A2, step 5.3.1.6; and
EPI-A2, Attachment 2, Section A, Event Termination Actions, step 6, which stated that for
events classified as an Alert or above, the NRC, State of Ohio, and local counties have been
consulted regarding event termination. Since distractor B required consulting with the State and
local counties, and did not specifically state that these were the ONLY organizations to be
consulted, this distractor was also a correct answer.
Distractor C was a correct answer based on EPI-A1, step 5.5.1.11; EPI-A2, step 5.3.1.6; and
EPI-A2, Attachment 2, Section A, Event Termination Actions, step 6, which stated that for
events classified as an Alert or above, the NRC, State of Ohio, and local counties have been
consulted regarding event termination. Since distractor C required consulting with the NRC, and
did not specifically state that this was the ONLY organization to be consulted, this distractor was
also a correct answer.
15 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Distractor D was an incorrect answer based on EPI-A1, step 5.5.1.11; EPI-A2, step 5.3.1.6; and
EPI-A2, Attachment 2, Section A, Event Termination Actions, step 6, which stated that for
events classified as an Alert or above, the NRC, State of Ohio, and local counties have been
consulted regarding event termination. Per EPI-A1, the intent is to involve the NRC, state, and
local counties in the decision making process. EPI-A1, step 5.5.1.11, goes on to state that this
action is not intended to delay or hinder the plants ability to terminate from an Unusual Event or
an Alert that no longer meets the criteria for any event at the time of declaration. As stated in
Section 5.3 of EPI-1A, an event that no longer meets the criteria for any event at time of
declaration, need not even be classified.
The applicant asserted that, given the correct circumstances, consultation with any organizations
was not required. However, applicants are instructed prior to the test per NUREG 1021,
Appendix E not to make assumptions regarding conditions that were not specified in the
question unless they occurred as a consequence of other conditions that were stated in the
question. Nothing in the stem of the question supported an assumption that the event no longer
met the EAL criteria at the time of the declaration. Therefore, the applicants assertion is invalid
and distractor D remained an incorrect answer.
Since three of the four distractors were correct answers, it was determined to delete the question
from the examination.
16 Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Post Examination Comments and Resolutions
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession # ML090650492
1 Enclosure 4