IR 05000461/1997313
| ML20217D045 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 03/24/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20217C980 | List: |
| References | |
| 50-461-97-313OL, NUDOCS 9803270284 | |
| Download: ML20217D045 (154) | |
Text
{{#Wiki_filter:. O . U. S. NUCLEAR REGULATORY COMMISSION REGION lll Docket No: 50-461 License No: NPF-62 l Report No: 50461/97313(OL) Licensee: Illinois Power Company Facility: Clinton Power Station
Location: Route 54 West Clinton, IL 61727 Dates: January 26,1998 - February 5,1998 Examiners: D. R. McNeil, Chief Examiner J. D. Ellis, Examiner D. S. Muller, Examiner G. F. Larizza, Examiner (In training) Approved by: Melvyn N. Leach, Chief Operator Licensing Branch 9803270284 980324 PDR ADOCK 05000346 ' V PM
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~ EXECUTIVE SUMMARY ' Clinton Power Station NRC Examination Report 50-461/97313 i A licensee developed and NRC approved initial operator licensing examination was administered to nine license applicants [two Reactor Operators (RO) and seven Senior Reactor o '~ Operators (SRO)). BtEllt% Two SRO license applicants passed all portions of their respective examinations and - were issued SRO licenses.
- Two SRO license applicants failed portions of the examination and were denied operator licenses.
Three SRO and two RO license applicants passed all portions of their respective. - examinations but were not issued operating licenses.- Licenses will be issued upon. .. completion of 10 CFR 55 required reactivity manipulations and all Clinton Power Station training program requirements.
- Examination Summarv: Control room operators were observed monitoring control room instrumentation at acceptable time intervals.; Their demeanor was business like and professional during observed periods. (Section 01.1) , CPS No. 3213.01 Fire Detection and Protection, Rev.19, was inadequate in that it allowed emergency operation of the diesel fire pump with no Jacket cooling water flow to _ . cool the diesel. (Section O3.1) -' Contrary to the requirements of 10 CFR 55.49, unauthorized people' gained access to a copy of NUREG 1021, " Operator Licensing Examination Standards for Power Reactors," < interim Revision 8, January 1997, Form ES-301-2, Individual Walk-Through Test _ Outline, listing the proposed examination Job Performance Measures (JPMs) by title.
- 1 This resulted in a breach of examination security. Additional examples of examination security problems were noted. (Section 05.1)
, Training department personnel developed a written examination that proved to be a , good evaluation tool for determining applicant competence. However, the examination , !showed a lack of attention to detail, Applicants were well prepared to take the written . examination. (Section 05.2) Rigorous enforcement of motor operated valve test preparation switch use appeared to . be lacking and needed improvement. (Section 05.4) ,
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l.' ( Examination developers failed to meet the guidelines of NUREG 1021 when developing the JPM examination outline. Validation of the JPM examination by facility personnel l was weak. (Section 05.4.c) l
- individual communications practices of some of the applicants was poor and needed
' improvement. Applicants failed to comply with Emergency Operating Procedure (EOP) steps to initiate suppression pool cooling when required and used alternate injection - systems when preferred injection systems were available. One applicant unnecessarily delayed execution of an EOP step resulting in unnecessary core uncoverage.
(Section 05.5) s
. I ', . L Report Details < <
l. Ope' rations
' Conduct of Operations
' 01.1 - General Comments ' i i 'During the initial license examination validation, examiners observed control room operations. The observed period included a shift tumover. Control room operators
were noted to be observing control room instrumentation at acceptable time intervals.
During the observed period,' operator demeanor was business like and professional.
Operations Procedures and Documentation J
, . 03.1: General Comments l While validating a Job' Performance Measure (JPM) to conduct an emergency startup of a non-safety related diesel driven fire pump (CPS No. 3213.01, Fire Detection and Protection, Rev.19), it was determined by NRC examiners that the diesel engine would ' .have been started and operated without Jacket cooling water flow.' The' procedure directed operators to open DFP 3A to provide cooling water flow to the diesel engine.
DFP 3A was already open per the system lineup sheet. NRC examiners reviewed the cooling water flow path with DFP 3A open and discovered an upstream isolation valve,- DFP 9A was shut, preventing cooling water flow to the diesel engine. Examiners l expressed concem over this valve lineup. Changes were implemented by plant staff to correct the procedure error.
05.
Operator Training and. Qualification 05.1 General Comments - Operator initial license examinations were administered at Ehe Clinton Power Station . (CPS) to nine applicants during the' weeks of January 26 and February 4,1998. Two
- Senior Reactor Operator (SRO) applicants passed all portions of their examinations and t were issued operating licenses. _Two SRO applicants failed portions of the operating examination and were denied operating licenses.,The remaining five applicants passed '
all portions of their examinations, but were not issued operating licenses because they had not completed all requirements to obtain the license. Upon certification'of completion of the 10 CFR required reactivity manipulations and the CPS licensed o , . operator training program, the licenses will be issued, E " The CPS training depadment participated in an examination process in which the license examination was developed by CPS training department personnel and approved by the NRC in accordance with guidance prescribed by NUREG 1021, )
" Operator Licensing Examination Standards for Power Reactors," Interim Revision 8, .J ' y
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.. 4-l January 1997. As part of the pilot program, the NRC administered the operating test and members of the CPS training staff administered the written examination.
All materials developed by the licensee for the examination were submitted to the NRC ' on schedule. The submitted examination material was technically accurate, but the , . written examination required significant rework to properly format the questions using ' bullet format and correct typographical errors.
l During development of the examination a copy of NUREG 1021 Form ES-301-2, l l Individual Walk-Through Test Outline, listing the proposed examination JPMs by title, ) was not property controlled and was obtained by unauthorized personnel. This was brought to the attention of the Rlli examiners by training department staff members of the Vermont Yankee Nuclear Station. The Rlli examiners then informed CPS training p staff members of the loss of examination integrity. Facility instructors were able to determine the source of the unauthorized release of examination material and took " appropriate steps to prevent re-occurrence of the examination compromise. Suitable replacement JPMs were developed by facility instructors and accepted by the NRC , l examiners. Loss of examination integrity is considered a violation of 10 CFR 55.49 which states, " applicants, licensees, and facility licensees shall not engage in any l activity that compromises the integrity of any application, test, or examination required i by 10 CFR 55." Because the compromised JPM examination was not administered to a license applicant and adequate compensatory measures were initiated by training department personnel, this event is being treated as a Non-Cited Violation, consistent with Section IV of the NRC Enforcement Policy (NCV 50-461/97313-01).- Three additional events irivoNing examination security occurred during validation and administration of the examination. Event 1: it was determined that two contract examination developers worked on examination development for several days prior to L signing the examination security agreement. NRC expectations are delineated in NUREG 1021, ES-201 D.2.c, which states that personnel who will receive detailed knowledge of any portion of the NRC licensing examination must acknowledge their responsibilities by reading and signing an examination security agreement before they obtain detailed examination knowledge. Event 2: a maintenance contractor entered the simulator, bypassing signs on the simulator door that restrict entry to authorized personnel only. Examination materials were open; however, the contractor was removed from the simulator without viewing any portion of examination material. Event-3:' a maintenance contractor entered the simulator, bypassing signs on the simulator door that restrict entry to authorized personnel only. No examination material was
- available to be seen by the contractor. The contractor was removed from the simulator.
' ' These events and the above non-cited violation are examples of poor examination - security and indicate a need for improvement in the examination security area.
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t, O5.2' Wntien Examination l.
a. - Examination Scope a i
- Using NUREG-1021, the examiners reviewed each written examination question for comprehension, validity, and level of difficulty. The aggregate of the written examination was reviewed for compliance with the submitted test outline to ensure a comprehensive examination.
b.- Observations and Findings The submitted written examination questions were technically accurate, comprehensive, - i l.
valid and examined at the correct difficulty level. The questions needed a significant. amount of rework in the area of question construction (bullet format) and typographical. l ' . error correction.
' The examination developers met the guidelines of NUREG-1021 for construction of a ~ ' written examination by adhering to the NRC approved examination outline.
Examiners observed the start of the written examination. All examination rules were d read to the applicants and enforced by the training staff. The training staff conformed to all guidelines for administering the written examination.
All applicants passed the written examination with scores ranging from 85% to 97%. ) _ c.
Conclusions Training department personnel developed a written examination that proved to be a , good evaluation tool for determining applicant competence. Attention to details such as ~ typographical errors and examination construction (format) was weak. Applicants were well prepared to take the written examination.
. 05.3 Administrative Job Performance Measures , a.
Examination Scope
- Using NUREG-1021, examiners reviewed each administrative JPM for applicability,
- importance, and safety significance. The aggregate of the administrative JPMs was reviewed to ensure all required areas of the administrative JPM examination were
- represented.
.- b. - Observations and Findings The office review of the submitted administrative JPMs indicated they met all NRC i i guidelines. ' During validation of the administrative JPMs it appeared that the W . l-s l-s
e .j i ! :. l administrative JPMs were well planned and prepared. During administration of the _ I administrative JPMs, no significant weaknesses were noted in applicant preparation for this portion of the examination.
An administrative' JPM that required applicants to provide Protective Action r.
Recommendations (PARS) was submitted with an incorrect answer. After discussion l .with facility instructors a correct answer was determined and submitted to the NRC examiners.. One applicant failed the administrative JPM area of the examination, c.
Conclusions )
Training department personnel prepared an administrative JPM examination that met all NRC guidelines. The administration of the administrative JPMs revealed no significant ! weaknesses in applicant preparation for this portion of the examination.
05.4 Ooeratina Job Performance Measures p - a.
Examination Scooe Using NUREG-1021, examiners reviewed each JPM for applicability, importance, and ' safety significance. The aggregate of the JPMs was reviewed to ensure all required
areas of the JPM examination were represented. JPM follow up questions were , ! reviewed for applicability and to determine if they were considered direct look up or simple memory questions.
b.
Observations and Findinas The licensee submitted an outline of proposed JPMs that did not meet standards prescribed in NUREG-1021 for addressing the numbers of functional safety systems, i The submitted outline for control room JPMs would have required the applicants to i _ perform two tasks from Safety System Group V. NUREG-1021 guidelines stated that l each safety system group should be used only once per control room examination.
Facility trainers replaced one of the Group V JPMs when requested by NRC examiners.
One JPM had to be replaced as the JPM task was also being performed during a I ' dynamic simulator scenario. The guidelines for altemate path and shutdown plant JPMs were met. Critical tasks were correctly identified within the JPMs.
i L The system follow-up questions submitted with the JPMs were considered simple memory and/or direct lookup questions. The weak questions were identified to the " facility instructors. Replacement questions were provided to the examiners that were -_ acceptable and more closely met the guidelines of NUREG-1021 for JPM follow-up questions.' i
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l l-V l.1 h' During validation of the JPMs the following problems were encountered: (1) A procedure to emergency. start a Diesel Fire Pump resulted in an unacceptable system operating condition. See section 03,1 for details l L (2) While validating the diesel fire pump start JPM, examiners were told the correct knurled knob to tum to open the fuel oil solenoid was located on the front of the diesel engine govemor. The indicated knurled knob actually controlled the fuel
oil pressure to the diesel engine, and did not open the solenoid valve. The l correct knurled knob to use to open the fuel oil solenoid was on top of the diesel ! engine govemor. This error resulted in an incorrect cue being given to four appl l cants during the administration of the JPM. A replacement JPM was validated and administered to those applicants given the incorrect cue.
. L (3) The submitted materials contained a significant number of typographical errors.
(4) One follow-up question on a'JPM had three procedures in which to look to determine an answer. Two of the procedures provided an incomplete answer.
- The incomplete answers were found in an annunciator procedure, HVAC RAD MONITOR TRIP BYPASS DIV 1 and in CPS No. 4001.02. The complete, correct answer was found in another annunciator procedure, HVAC j RADIATION MONITOR TRIP BYPASSED DIV 2. Facility instructors - i initiated procedure change requests to correct the procedures.
During administration of the JPMs the following applicant weaknesses were noted: (1) Applicant use of the Motor Operated Valve (MOV) test preparation switches was inconsistent and in some cases left safety equipment in a reduced state of emergency readiness. While performing control room JPMs, applicants correctly . used the MOV test switches, failed to use the MOV. test switch, or left the MOV test switch in the TEST position which inserts thermal overload protection and torque switches into the control circuitry of the MOV. This protection is required to be removed during normal plant oparation as it may interfere with the MOV's
- ability to respond during an emergency.
'(2) The SRO applicants with no prior operations experience had difficulty in locating ) equipment and completing JPM tasks in the plant. One SRO applicant took 70 minutes to complete two JPMs designed to take 25 minutes.
! Two applicants failed the Operating JPM portion of their examination.
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Conclusions ' Examination developers failed to meet the guidelines of NUREG-1021 when developing the JPM examination outline. Additional attention was necessary to correct examination j development errors concerning content and validation. Validation of the JPM !
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. [ examination by facility personnel was weak as evidenced by the number of errors found ! ' during the NRC validation of the JPMs. Rigorous enforcement of MOV test preparation o switch use appeared to be lacking and needed improvement.
, l 05.5 Dynamic Simulator Scenarios l-a.
Examination Scooe Using NUREG-1021, examiners reviewed each dynamic simulator scenario. Each scenario was reviewed for content, applicability, and safety significance.
bl Observations and Findinas The integrated dynamic scenarios submitted by the licensee had the correct number of l required malfunctions for each applicant. The submitted scenarios were revised to reflect the change in the number and makeup of applicants during the examination i validation week. The initial conditions were different for each scenario and included a low power scenario.
During administration of the dynamic simulator scenario examination the following applicant weaknesses were noted:
(1) Communications standards were not always enforced. Three way communications were not used by some applicants. This did not result in any errors during the dynamic scenarios.
l (2) Applicants failed to correctly execute steps of the Emergency Operating Procedures (EOPs). Three significant failures to comply with EOP standards were noted by examiners.
CPS EOPs stated that when the suppression pool temperature reaches . 95'F, operators are to initiate suppression pool cooling. Two applicants failed to initiate suppression pool cooling when the suppression pool temperature reached, then exceeded 95'F. During follow-up questioning both applicants stated that suppression pool cooling doesn't provide much cooling and is not a top priority when executing EOP steps. One of the two applicants also stated that he was watching the heat capacity temperature limit and since he wasn't approaching that limit, he felt that it was not necessary to initiate suppression pool cooling.
CPS EOPs directed operators to initiate cooling water flow to the reactor . during an Anticipated Transient Without Scram (ATWS) using preferred makeup systems that inject coolant outside the core shroud. During the scenario all injection systems were failed except for high pressure core i spray (HPCS), low pressure coolant injection (LPCI) loop B and LPCI loop C. HPCS and LPCIloop C inject inside the shroud. LPCI loop B I can be lined up to inject outside the shroud by bypassing interlocks and f g L
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?: t p.= h-l . ). using the shutdown cooling flowpath. Two applicants failed to initiate !' LPCI loop B using the shutdown cooling flowpath when directed to do so [ by the EOPs. When asked why they failed to comply with that EOP _ l requirement, one applicant stated that he missed the step, the other l applicant stated that he had already restored the level with HPCS and l didn't see the need to use LPCI loop B.
CPS EOP-1 A, ATWS RPV Control, directed operators to blow down the . L ' reactor vessel if they cannot maintain vessel level above -193 inches.
. - . Operators are allowed to, and under EOP execution guidelines, should,
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execute this step as soon as they make the determination they cannot maintain vessel level above -193 inches. One applicant made the determination that he could not maintain vessel level above -193 inches before he reached -20 inches, then waited until vessel level lowered . down to -193 inches before taking the action to blow down the vessel.' One applicant failed the dynamic simulator scenario portion of the examination.
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Conclualona ' - Individual communications practices of some of the applicants was poor and'needed improvement. - Applicants failed to comply with Emergency Operating Procedure (EOP) steps to initiate suppression pool cooling when required and used altemate injection systems when preferred injection systems were available. One applicant unnecessarily delayed execution of an EOP step resulting in unnecessary core uncoverage.
'05.6 Post Examination Activities zThe licensee informed the Chief Examiner by telephone that there were no post examination comments for the written examination. There were several minor clarifications issued by the facility monitor while the written examination was being.
administered and one significant change. The change was approved by the examiners.
- i Question #91 (SRO), #83 (RO) was altered as there was no correct answer supplied for - the question. The question required the operator to recognize an entry condition into EOP-6, Primary Containment Control. All answers provided to the applicants were I incorrect. Facility instructors modified all four of the answers to prevent revealing the _ "7 . correct answer during question correction. Changing only the correct answer would have immediately provided the correct answer to the applicants.- j - A review of the written examination revealed that two questions were missed by more ' than 50% of applicants. The questions were reviewed by training department personnel and found.to be training deficiencies in the initial license operator training program.
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- Applicants were unable to identify the person (by title) allowed to verbally -
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authorize personnel radiation exposures up to legallimits. (RO/SRO #13)- . i
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' , . -(2) Applicants were unable to state the reason RHR "B" is preferred over RHR "A" when transferring water to Radwaste during shutdown cooling operation.
(RO #61, SRO_#54) 05.7, Simulator Fidelitv Examiners observed some simulator modeling deficiencies during the examination administration.~ Examiners and facility instructors were able to provide cues to the applicants to disregard the erroneous indications where applicable. The examiners > concluded the identified deficiencies did not preclude completion of valid evaluations of license applicant performance. Simulator deficiencies are documented in Enclosure 2, Simulation Facility Report.
V. Management Meetings X1 Exit Meetina Summarv The chief examiner presented the examination team's observations and findings to members of the licensee's management on February 4,1998. The licensee acknowledged the findings
presented. No proprietary information was identified during the examination or at the exit meeting.
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,, , PARTIAL LIST OF PERSONS CONTACTED .- Licensee q J. Christensen, Director Nuclear Training G. Hunger, Plant Manager a M. Lyon, Operations Manager W. MacFarland, Senior Vice President and Chief Nuclear Officer . ! W. Maguire, Operations Manager J. Palchak, Training Manager ! P. Telthorst, Assistant Director, Operations J. Wemlinger, Supervisor, Operations Training NBC
- K. Stroedter, Resident inspector (*) Personnel not in attendance at the exit meeting on July 10,1997.
' ITEMS OPENED, CLOSED, AND DISCUSSED j Ooened 50-461/97313-01 NCV Examination security compromise Closed Nme t Discussed None p
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., y.b, . Enclosure 2 SIMULATION FACILITY REPORT Facility Licensee: 'Clinton Power Station - ~ Facility Licensee Docket No: 50-461 l Operating Tests Administered: January 26 - February 4,1998 . The following documents observations made by the NRC examination team during the i ' January / February 1998, initial 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.. 'l 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.
. D'uring the conduct of the simulator portion of the operating tests, the following items were;
- observed
i l l ITEM :- DESCRIPTION =1 During administration of a JPM involving placing RHR A in shutdown cooling from the '. remote shutdown panel, it was discovered that-i . the Division i Emergency Diesel Generator.
n tripped when HS502 was placed in the" emergency position.- (CPS Condition Report
. No.1-98-02-304) 2.
During administration of a JPM involving
- j restoration of instrument air to the drywell, it :
. was discovered that 1SA031 (Drywell isolation L valve) did not open. (CPS Condition Report , ' No. 1-98-02-308) ' j t > !- h i Lu u - ' ' i <. u - ~ r ., - ! , ! o
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. . . .... . . . .. _ _ _ _ _ _ _ _ _ , ES-401 Site-Specific Written Examination Form ES-401-7 Cover Sheet ~
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS 1.
After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing ' the examination.
2.
To pass the examination, you must achieve a grade of 80.00 percent or greater. Every question is worth one point.
3.
For an initial examination, the time limit for completlag the examination is four hours.
For a requalification examination, the time limit for completing both sections of the examination is three hours. If both sections are administered in the simulator during a single three-hour period, you may retum to a section of the examination that was already completed or retain both sections of the examination until the allotted time has expired.
4.
You may bring pens and calculators into the examination room. Use only black ink to [ ensure legible copies.
5.
Print your name in the blank provided on the examination cover sheet and the answer sheet. You may be asked to provide the examiner with some form of positive identification.
6.
Mark your answers on the answer sheet provided and do not leave any question blank.
., ' Use only the paper provided and do not write on the back side of the pages. If you i decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change.- 7.
If the intent of a question is unclear, ask questions of the NRC examiner or the designated facility instructor only.
8. ' Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
) ' Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheating.
i g.
When you complete the examination, assemble a package including the examination ! questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received , assistance in completing the examination. The scrap paper will be disposed of immediately after the examination.
10.
After you have tumed in your examination, leave the examination area as defined by the ' proctor or NRC examiner. If you are found in this area while ths examination is still in , progress, your license may be denied or revoked.
11.
Do you have any questions? i
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REACTOR OPERATOR Page 3 of 57.
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. This page intentionally blank (replaces answer sheets) .t a , ,
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, p i REACTOR OPERATOR ' Page 4 of 57 1; l , ~ QUESTION: 001 (1".00) - 10CFR55 lists the requirements that a licensed operator MUST meet to maintain his/her license in an " active status." The operator shall actively perform the functions of a licensed operator for a MINIMUM of:- a~ four - 8 hour shifts per calender queder.
b.
five - 8 hour shifts por calendar month.
c.
seven - 8 hour shifts per calendar quarter.
, - d.
three - 8 hour shifts per calendar month.
QUESTION:002 (1.00) Given the following conditions:
The plant is in Mode 5 with irradiated fuel in the RPV - water level is 23'2" feet above the top of the RPV flange - No refueling activities are in progress, and no one is on the refuel floor - The plant has been in cold shutdown for several months.
- RHR "B" has been inoperable for a week, for planned outage maintenance.
- RHR "A" has been running in shutdown cooling for six weeks.
- Then, RHR "A" pump trips on pump motor overcurrent.
CHOOSE the statement that completes this sentence: The REQUIRED action is to... a.
within 72 hours, restore one RHR shutdown cooling subsystem to service . b.
within 4 hours, restore one RHR shutdown cooling subsystem to service ' c.
- within 12 hours, verify an altemate method of decay heat removalis available d.
within 1 hour, verify an altemate method of decay heat removal is available
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REACTOR OPERATOR Page 5 of 57 a.. QUESTION: 003 (1.00). Given the following: A normal plant startup is in progress with the unit at 23%. -
Then, the Reactor Operator reports: increasing reactor power with no change in RR flow or rod motion - Further checks show that feedwater temperature is decreasing and a loss of -- feedwater heating is in progress.
What immediate Operator Action is required? ' a.
Immediately reduce power to 20 MWe (approx 60 MWt) by inserting control rods, b.
Immediately reduce power by 20 MWe (approx. 60 MWt) by inserting control rods.
c.
No immediate operator action is required based on current power level.
d.
Immediately reduce power by 20 MWe (approx. 60 MWt) by reducing recirculation flow.
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( , u . REACTOR OPERATOR ~ Page 6 of 57 , QUESTION: 004 (1.00) You are on duty when an area operator calls you regarding a valve lineup he is conducting. He - asks you for the~ proper method to set a manual throttle valve to 25 tums open, as instructed on - the lineup.
- SELECT your response to the area operator from the following.
a.
Move the handwheel counter-clockwise to the full open travel position, then - move the handwheel clockwise 25 tums, b.
' Move the handwheel clockwise to the full closed travel position, then move the handwheel counter-clockwise 25 tums.
c.
Move the handwheel counter-clockwise in the open direction, from full closed , until you just hear system flow through the valve, then open the valve 25 tums.
d.
Move the handwheel clockwise in the closed direction, from full open until you just hear system flow through the valve stops, then open the valve 25 tums.
QUESTION: 005 (1.00) . CPS No. 3310.01, Reactor Core isclation Cooling (RI) cautions that when starting RCIC, . always use the maximum starting rate by setting RCIC Pump Flow Cont,1E51-R600 to 600 gpm until the turbine has reached operating speed.
WHICH of the following is the explanation for this precaution? a.
This is to build up bearing oil pressure as quickly as possible to prevent bearing damage.
l b.' This is to build up steam seal pressure as quickly as possible to maintain i radiation levels and temperatures in the RCIC pump room as low as possible.
i c.
This is to prevent dsmage to the RCIC turbine exhaust check valve.
. . d.
This is to minimize injection time as required by the USAR.
j s
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REACTOR OPERATOR Page 7 of 57 ) QUESTION: 006 (1.00) An annunciator has multiple inputs, but only one of the inputs is providing nuisance annunciation. Once the single nuisance input has been disabled per CPS No.1406.01 (ANNUNCIATOR TRACKING PROGRAM), which of the following describes the process used to track the inoperable alarm? a.
A Temporary Modification should be used along with an entry in the Out of Service / Disabled Annunciator Log.
b.
A Plant Modifiation should be used along with an entry in the Control Room Operator's Log.
c.
An Engineering Work Request should be used along with a entry in the Out of ) Service / Disabled Annunciator Log.
' d.
A red arrow entry should be placed in the Control Room Operators Log along with an entry in the Shift Supervisors Log.
l l j '
I l
s' i , ' REACTOR OPERATOR Page 8 of 57 QUESTION:007 (1,00) i During the performance of a surveillance, the performer encounters a step with "QV HOLD" . ritten. WHAT is the expect 3d response in accordance with CPS 1005.01, CPS Procedures w , , ' . and Documents? a.
The performer is required to stop and notify inspection personnel to allow for planned inspections. Once notification han been accomplished and the agreed to time (or a reasonable amount of time) has passed, the work activity may continue.
b.
The performer is required to stop and notify inspection personnel to allow for - planned inspections. The work activity shall not proceed without the point being signed by inspection personnel, or inspection personnel being notified and authorizing the activity to proceed, or the point waived / reclassified.
c.
The performer is required to notify inspection personnel in order for them to plan - when they will perform the verification activity. The work shall be capable of being verified after work completion.
,.; l d.
The performer is required to continue work on the surveillance until completion.
After the work is done, the performer shall call for inspection personnel to ' '
l perform the inspection.
I h i i I i , t , ,
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REACTOR OPERATOR Page 9 of 57 QUESTION: 008 (1.00) I WHICH of the following is allowed when hanging tags in accordance with CPS 1014.01 SAFETY TAGGING 7 a.
Two caution tags may be hung on an open drain valve only if the same contact person has responsibility for both tags.
b.
A danger tag may be hung on a closed breaker inside an energized electrical I cabinet. The tag can be attached to the breakor with string and a small section of tygon hose.
. c.
A danger tag may be hung for a removed breaker. The tag can be attached to - the breaker cubicle door.
j d.
The Area Operator may change a required position by initialing and dating the handwntten correction on a danger tag.
. I QUESTION: 00g (1.00) The Assistant SS is designated as Fire Brigade Leader. The Fire Brigade Leader is responsible I for WHICH of the following during a fire? a.
Make emergency plan notifications b.
Notify off site fire protection agencies for assistance ' c.
Taking charge of the at-the-scene fire-fighting operations d.
Taking charge of the fire-fighting operations from the control room i , . ' s U
y - REACTOR OPERATOR Page 10 cf 57; . ~ ' QUESTION: 010 (1.00) , Given the following: The plant is in Mode 5 with refueling operations in progress, _.
- The Core Alterations Surveillance Log shows that the refuel position one-rod-out - interlock surveillance was last completed at 0800.
Then, when performed at 2130 by operations, the one-rod-out interlock surveillance fails.
. WHAT actions are required in accordance with CPS Technical Specifications? j a.
Immediately suspend loading of irradiated fuel into the RPV; initiate schon 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.
i d.
Immediately initiate action to insert all insertable control rods and place the mode - switch in the SHUTDOWN position in 1 hour.
QUESTION: 011 (1.00) ) What are the minimum number of licensed operator (s) required At the Controls' during fuel handling,' per 1401.01 (CONDUCT OF OPERATIONS)? j a.
One Licensed Reactor Operator b.
Two Licensed Reactor Operators c.
One Licensed Senior Reactor Operator d.
Two Licensed Senior Reactor Operator
i i .
. REACTOR OPERATOR ' Page 11 of 57 ' ,
' QUESTION: 012 (1.00) E Conditions in a recently surveyed area are: - - 50 dpm/100cm2 alpha loose surface - 50%DAC airborne radmactwity : - WHICH of the following describes the posting requirements for the area? a.
' Contaminated area - required Airbome radioactivity area - required Contaminated area - required b.
. ' Airborne radioactivity area - NOT required c.
- Contaminated area - NOT required Airbome radioactwity area _ required - d.
Contaminated area - NOT required' Airbome radioactivity area - NOT required ' QUESTION: 013 (1.00) -- SELECT the choice that completes the following statement.
. - During an emergency the ~ may verbally authorize personnel exposure up to legal limits, per 10CFR20. (MUST be recorded in logs) a.
Shift Supervisor i b.
Emergency Manager ] ) n - c.
Emergency Director ' , d.' TSC Radiological Controls Supervisor .
u
m ! , . REACTOR OPERATOR Page 12 of 57 , .. l t QUESTION: 014. (1.00). The solenoids for tho' Scram Discharge Volume vent & drain Air Operated Valves reposition to ...[ CHOOSE ONE]. i a.
. supply pressurized water to the bottora of the drive piston.
b.
. exhaust water from the top of the drive piston.- c.
exhaust air from the diaphragm operated SDV vent and drain valves.
s - d. ' exhaust air from the diaphragm operator of the scram inlet and outlet valves.
QUESTION: 015 (1.00) Given the following: Scram signals are present on Division I and Division 11 of RPS.
- The plant is stable with the turbine on line at 50% load.- - Then, the C Area operator reports that the scram air header has NOT depressurized, and tho' scram pilot solenoids on the HCU's checked all feel warm.
! - WHICH of the following would depressurize the scram air header? I ! a.
- De-energize the backup scram valve pilot solenoids by opening their breakers, j
! b.
Roset the scram to energize the scram valve pilot solenoids and allow the SDV ) to drain.
c.
De-energize the ARI solenoids by opening their breakers.
d.
De-energize the scram pilot solenoids by opening their breakers.
l
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, REACTOR OPERATOR Page 13 of 57 . QUESTION: 016 (1.00) < - WHICH of the following effects would occur as a result of throttling closed 1C11-F003, CRD i DRIVE PRESSURE CONTROL VALVE? a.
DECREASES control rods insertion time on a scram.
- b.
INCREASES control rod withdrawal speed, c.
INCREASES cooling water flow to each CRDM.
d.
DECREASES seal flow to the Recirculation Pumps.
QUESTION: 017 (1.00) I Given the following: The reactor is operating at 100% power.
- Channel C is in sensor bypass for maintenance.
- Then, the APRM DOWNSCALE annunciator alarms due to APRM Channel A failing downscale.
All other APRM channels are OPERABLE. As a result of this failure: a.
The reactor WILL scram k b.
There WILL be no scram nor rod blocks , , c.
control rod withdrawal motion is blocked by RC&lS. Control rod insertion is l permitted.
, d.
control rod insertion and withdrawal motion is blocked by RC&lS.
- ,
[ m ' i L: ' REACTOR OPERATO'R-Page 14 of 57 QUESTION: 018 (1.00) After Anticipated Transient Without Scram (ATWS) condition which required the injection of ' Standby Liquid Control (SLC) System solution into the corn, the oporstors MUST refill the SLC Storage Tank.
the containment is accessible - SLC Storage tank contains approximately 1500 gallons - .. SLC is shutdown - SLC Storage Tank solution is 79 degrees F - -- SLC Storage Tank sodium pentaborate concentration is 13.5%. - CHOOSE the operator action, prior to involving Chemistry, to be taken to prepare the SLC System for future injection, if needed.
_ a.
Use the Makeup Condensate System and fill the tank to 5000 to 5150 gallons.
b.
~ Place local Mixing Heater control switch in the AUTO position.
c.
INCREASE the SLC Storage Tank solution concentration to greater than 14.5%. d.
Fill the tank to clear the low level alarm using the Makeup Condensate System.
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REACTOR OPERATOR Page 15 of 57 , . QUESTION: 019 (1.00) How WILL INDICATED (on P680 panel) loop flow and core flow change if one of the two calibrated A-loop jet pump flow signals fails to zero? i s.
"A" Loop flow WILL DECREASE 50 percent, "B" loop flow WILL INCREASE 50 percent and core flow WILL remain unchanged.
b.
"A" Loop flow WILL DECREASE 10 percent, "B" loop flow WILL INCREASE 10 percent and core flow WILL remain unchanged.
I ' A" Loop flow WILL remain unchanged "B" loop flow WILL remain unchanged " c.
, and core flow WILL DECREASE 25 percent.
d.
"A" Loop flow WILL remain unchanged, "B" loop flow WILL remain unchanged and core flow WILL DECREASE 5 percent.
' " . QUESTION: 020 (1.00) The plant is operating at 100 percent power. WHICH of the following describes the response of '- ) the Recirculation Flow Control System to a Feedwater pump trip if the vessel level falls to 28 inches? - l ' a.
Recire pumps downshift to slow speed.
b.
Recirc pumps trip to OFF - c.
Flow control valves run back to - 54% indicated valve position, i d.
Flow control valves run back to - 19% indicated valve position.
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, REACTOR OPERATOR Page 16 of 57 QUESTION: 021 (1.00) While in EOP-4 (Steam Cooling) adequate core cooling is assured by WHICH of the following?
a.
Core submergence with LPCI Injection, level above TAF, - 162 inches b.
Steam Cooling when level is above -193 inches c.
Steam Cooling when level is above -205 inches d.
Natural circulation with LPCI Injection, level above -45.5 inches QUESTION: 022 (1.00)
The plant is operating at 60% reactor power with Low Pressure Core Spray System (LPCS) surveillance, CPS No. 9052.01, LPCS PUMP OPERABILITY, is being performed.. The following conditions occur.
A high drywell pressure signal is received and the reactor scrams - RX water level stabilizes just below level 8 - RX pressure is 975 psig and stable on the Turb. bypass valves - Drywell pressure is 2.3 psig.
- WHICH of the following describes the status of the Low Pressure Core Spray system? l a.
LPCS WILL realign and inject into the vessel.
l l b.
LPCS WILL realign and operate on flow.
' c.
LPCS pump WILL trip.
d.
LPCS WILL remain in full flow test mode.
i
REACTOR OPERATOR Page 17 of 57 - QUESTION: 023 (1.00) WHICH of the following describes the normal operation of the HPCS Flow Valve,1E22-F012, during an automatic HPCS initiation? a.
Normally closed. Automatically opens after HPCS initiation and remains open until manually closed when system flow reaches 625 gpm and discharge pressure is less than 145 psig.
b.
Normally open. Automatically closes when system flow reaches 625 gpm and discharge pressure is less than 145 psig.
c.
Normally closed. Automatically opens after HPCS initiation and automatically closes when system flow reaches 625 gpm and discharge pressure over 145 psig.
d.
Normally open. MUST be manually closed from the contml room after system flow reaches 625 gpm and discharge pressure is over 15 psig.
% QUESTION: 024 (1.00) WHICH of the following identifies the starting sequence for the Standby Liquid Control (SLC) System when the control switch SLC PUMP B is placed in the RUN position? Assume all-appropriate actions NOT listed occur as expected.
a.
One squib valve fires and the RWCU Outboard Suction isolation Valve (1G33-F004) closes, b.
Two squib valves fire and the RWCU Outboard Suction isolation Valve (1G33-F004) closes.
c.
One squib valve fires and the RWCU Inboard Suction isolation Valve (1G33-F001) closes.
' 'd.
Two squib valves fire and the RWCU Inboard Suction isolation Valve (1G33-F001) closes.
.
REACTOR OPERATOR
- Page 18 of 57
. QUESTION: 025 (1.00) ' Durir[.) steady state power operations, the PRM Gain Adjustment Factor (AGAF) for APRM-C roads 1.04 on a current 00-3.' Under these conditions, the operation of APRM C is because . SELECT the choice that fills in the blanks.
' a.
conservative / APRM reading is GREATER THAN actual power b.
NON-conservative / actual power is GREATER THAN APRM reading c.
conservative / APRM reading is LESS THAN actual power , d.
NON-conservative / actual power is LESS THAN APRM reading QUESTION: 026 (1.00), "' From the following, choose how the Intermediate Range Monitor System (IRM) is interrelated with the Rod Control and information System (RC&lS).
! a.
IRM signal level amplifier provides flux indication to the RC&lE Rod Interface l Display Module.
b.
IRM trip units provide control rod insert blocks to the RC&lS.
. c. - IRM voltage preamplifier provides rod withdrawal blocks to the RC&lS.
d.
IRM trip units provide control rod withdrawal blocks to the RC&lS.
, .- , l;> l= >gr
e .- j.
c REACTOR OPERATOR ' Page 19 of 57 l I-- .. QUESTION: 027 '(1.00) WHICH of the following statements describes how the shorting links are used in the reactor protection system (RPS)? a.
Installation of the shorting links enables'a scram if any single SRM, IRM or. APRM channel trips.
b.
' Removal of the shorting links enable the SRM scrams in a coincidence of one-out-of-two-twice logic scheme.
' c.
Installation of the shorting links enable the SRM scrams in a coincidence of one-out-of-two-twice logic scheme.
.r d.
Removal of the shorting links enables a scram if any single SRM, IRM or APRM
channel trips.
QUESTION: 028 (1.00) A Local Power Range Monitor (LPRM) detector has been bypassed in APRM "A". WHICH of the following identifies what APRM "A" meter would indicate in the Count position if this is the only bypassed LPRM? a.
b.
l c.
i d.
i m l i ' < , . [ - '
' & i.) - , REACTOR OPERATOR Page 20 of 57 l l . QUESTION: 029 (1.00) - The plant is operating at 85% power. The following are the indications received when the APRM -
meter function switches in the backpanels are placed in the AVERAGE, COUNT, and FLOW I positions.
' AVERAGE COUNT FLOW APRM A
30 69% ~ APRM B 84 %
68% ' APRM C 86 %
69% APRM D - 89 %
61 % , WHICH of the following are the expected annunciators for these conditione? a.
ROD OUT BLOCK only . b.
DIV 2 OR 3 NMS TRIP only j
c.
ROD OUT BLOCK and DIV 2 OR 3 NMS TRIP . d.
ROD OUT BLOCK and DIV 1 OR 4 NMS TRIP i g-QOESTION: 030 (1.00) The following plant conditions exist:
Recirc pumps have tripped - Reactor has scrammed - A steem leak has occurred - The drywellis at 212F and 2.4 psig. - The RPV has been emergency depressurized j - J Which of the following level instruments would most accurately indicate RPV water level if level ) e w-- is on scale for that level instrument? a.
Narrow Range b. - Wufe Range c.
Shutdown Range i
d.
Fuel Zone Range ' ,
, REACTOR OPERATOR Page 21 of 57 " - QUESTION: 031 :(1.00) - Given the following:
- The plant is operating at 100% reactor power with 2 turbine driven reactor feed pumps (TDRFPs) running in automatic on the Master Water Level Controller with the tape set at 35 inches.
Then, the reactor water level transmitter selected for input to the Feedwater Level Control . System fails to a level of 33 inches.
Assuming no operator action, this WILL cause:
a.
RCIC initiation as water level DECREASES to Level 2.
b.
reactor water level to remain at 35 inbhes because the steam and feedwater flow
signals overcome the level control signal.
j c.
the motor driven feedwater pump to start.
. d.
a reactor scram as water level reaches Level 8.
, QUESTION: 032 (1.00) < Following an automatic initiation, RCIC speed is observed to be zero, the RCIC Trip-Throttle Valve (TTV) is open and the steam supply shutoff valve (1E51-F045) is closed. WHICH of the following conditions could have caused this RCIC response.
a.- RCIC Pump suction low pressure b.~ RCIC ISOLATION has been armed and depressed c.
Low reactor pressure < ~ d.
. Reactor waterlevel high (L8) i , . . y ,
,,
REACTOR OPERATOR Page 22 of 57 ,
1 - . QUESTION: 033 -(1.00)' Given the following plant conditions:' , / PLANT CONDITIONS RPV water level: -150 inches for 105 seconds.
' Drywell pressure: 2.0 psig.
(RHR) pumps: NOT running - LPCS Pump: NOT running SELECT the statement that describes the effect that pump motor breakers failing to close for the Low Pressure Core Spray (LPCS) pump and the Low Pressure Coolant injection (LPCI) pumps would have on the Automatic Depressurization System (ADS).
l a.
ADS automatically actuates at this time.
b.
ADS WILL NOT automatically actuate.
c.
- ADS WILL actuate 105 seconds after the ADS "A" manual initiation pushbutton is
' depressed.
' d.
ADS WILL actuate when the ADS "B" manual initiation pushbutton is depressed..
QUESTION: 034 (1.00) i During a LOCA, Containment Pressure is 7.6 psig. CHOOSE the statement that would result-l In the automatic initiation of the Containment Spray Mode of RHR.
j a.
LPCI has been in operation for 90 seconds and Reactor Vessel Water Level is greater than +30.8".-
- b.
LPCI has been in operation for 10 minutes and 10 seconds Reactor Vessel Water Level is greater than +45.5".
c.
Reactor Vessel Water Level is greater than +30.8" and Drywell Pressure greater than 1.68 psig.
) ' d.
' LPCI has been injecting for 10 minutes and 10 seconds and Drywell Pressure a greater than 1.68 psig.
!
REACTOR OPERATOR Page 23 of 57-QUESTION: 035 (1.00)' - During shutdown testing with RHR "A" in shutdown cooling a Group 3 high RPV pressure isolation signal is inadvertently generated.- WHICH of the following describes the expected system response to this signal? - a.
"A" pump trips and both SDC suction isolation valves (1E12-F008/F009) close.
. b. - "A" pump remains running and the neither SDC suction isolation valve ' (1E12-F008/F009) closes.
c.
"A" pump trips and the both SDC suction isolation valves (1E12-F008/F009) remain open.
d.
"A" pump trips and Division l SDC suction isolation valve (1 E12-F008) closes.
- QUESTION:036 (1.00) ' SELECT the statement that describes the operation of the Safyy Relief Valves operated from y../.;,0, "in EMERGENCY.
.a - the Remote Shutdown Panel (RSP) if the associated Transfer 25vlfdes(6,6) 4re, ' a.
All automatic functions except Safety mode are disabled.
'b.
Relief Mode and Safety Mode disabled. All other automatic functions operable.
c.
Relief Mode and low-low Set Mode disabled. All other automatic functions operable.- d.- NO automatic functions are disabled. Low-low set WILL NOT initiate when valves are opened manually from the RSP.
i . ^ 1l sy
, ' REACTOR OPERATOR ! Page 24 of 57 ,, , t QUESTION:037 (1.00)- Which of the following causes the Steam Bypass Valves to open if the main turbine should trip - with the reactor at 60% power? ' ' a.
' Stsam line pressure WILL be greater than the PRESSURE SET setpoint.
- b.
The BYPASS JACK signal WILL be greater than the control valve demand signal c.
The STEAM BYPASS AND PRESSURE REGULATOR pressure signal WILL be ~ less than the load limit setting ' d.
Turbine control valve demand signal WILL be less than the MAXIMUM ' - COMBINED FLOW limiter signal.
' QUESTION: 038 (1.00).
Given the following: ' The plant is operating at 100% -- both Turbine Driven Reactor Feed Pump (TDRFPs) are in' service.
- Narrow range level instrument N004A is failed high - feedwater level control is in automatic and selected to "B" level instrument.
- - - Then, narrow range level instrument N004C suddenly fails high.
, WHICH of the following is the response of the TDRFPs? a.
The RFPTs WILL continue to control level in the normal band.- ' .. b.
The RFPTs WILL trip.
i c.
The RFPTs WILL DECREASE speed and control at a lower level.
,. d.
The RFPT flow controllers WILL shift to manual.
, ' 'i
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- REACTOR OPERATOR Page 25 of 57 QUESTION: 039 (1.00)
The purpose of Steam Programming is to: s.
- Prevent RPV level from reaching the RPV Main Steam Line nozzles.
, b.
. INCREASE the efficiency of the RPV Moisture Separators at less than rated steam flow, c.
INCREASE neutron population by providing adequate RPV level to reduce leakage.
d.
. Maintain adequate NPSH to the Reactor Recirculation pumps at high steam flow conditions.
QUESTION: 040 (1.00) l ' Given the following:
SGTS TRN A EXH FAN has automatically started.
- WHICH of the following signals ALONE could have directly caused the SGTS TRN A EXH FAN to start? a.
Drywell temperature of 135 degrees F i b.
Containment CCP Exhaust High Radiation c.
Low flow on the SGTS TRN A EXH FAN ~d.
Level 8 .
' . REACTOR OPERATOR Page 26 of 57 \\ - QUESTION: 041 (1.00) ,, - Emergency Diesel 1 A started on a LOCA signal and closed into bus 1 A1. TRIPPED DIESEL " y ' ~ GENERATOR 1 A annunciator was received. The diesel generator breaker opened and the diesel shutdown. What is the likely cause of the diesel generator breaker trip and subsequent
- diesel shutdown?'
a.
A generator overcurrent condition b.
A generator differential overcurrent c.
- A underfrequency condition on bus 1 A1 d.
A generator neutral winding overcurrent fault QUESTION: 042 (1.00) On the initiation of a reactor scram from 100% power, the insert line o' e CRDM breaks off completely at the weld on the pipe connected to the insert port at the CRDM.
] .
- WHICH of the following describes the expected response of the rod? a.
WILL immediately start to insert and WILL fully insert with reactor pressure only , b.
- WILL immediately start to insert and WILL fully insert with accumulator pressure only.
, ! c.
WILL NOT insert until reactor and accumulator pressures equalize and WILL insert with reactor pressure only.
. d.
WILL hydraulically lock and WILL NOT insert l . t mm.
-
REACTOR OPERATOR Page 27 of 57 QUESTION: 043 (1.00) .Given the following: The plant is at 100% power, EOL, when a reactor scram occurs.- - Rod 24-29 was selected just prior to the scram.
- No operator actions have been taken.
- -The Green LED for each rod is on at the full core display.
- Then, OD-7 is selected on the process computer and indication for rod 24-29 is NOT 00.
WHAT is the reason for this indication? a.
The scram has NOT been reset b.
It is receiving abnormal data from Rod Position indication system c.
A selected rod is automatically deselected on a scram d.
A full core ATWS is in progress.
- QUESTION: 044 (1.00) A total loss of component cooling water occurs and is unrecoverable. A Recirc Pump trip is required... a.
immediately b. ' - within 1 minute c.
within 5 minutes d.
when motor winding temperature reaches 225 degrees F.
i , ! . c
.s I - p - l' . . . . REACTOR OPERATOR Page 28 of 57 .
l ' QUESTION: 045 (1.00) , WHICH of the following describes the expected RWCU system lineup as a result of a high Filter
.- ! -- Domineralizer inlet temperature ?.- ! a.
One pump running, F/D's on hold, both containment stu: tion isolation valves
open.
I' b.'- No pumps running, F/D's online, both containment suction isolation valves shut.
c.
. One pump running, F/D's online, both containment suction isolation valves open.-
d.
No pumps running, F/D's on hold, outboard containment' suction isolation valve shut.
QUESTION: 046 (1.00) Given the following: The plant is shutdown with core alterations in progress.
- The operators have flushed and warmed "A" RHR Loop for Shutdown Coonng.
.- - Then, the operator starts the "A" RHR Pump and immediately opens the 1E12-F037A, RHR A to Containment Pool Cooling Shutoff Valve.
- - CHOOSE the reason this is done.
L a.
Prevent RHR A Pump damage due to low flow, b.
Reduce system pressure quickly to prevent damsge to CY flush lines.
, c.
Prevent the RHR A Pump Min Flow Valve from opening on low flow.
. d.
Establish flow to prevent Feedwater Check Valve oscillations or fluttering.
-
r; ' u-Lp V REACTOR OPERATOR Page 29 of 57.
I-p . l QUESTION: 047 - (1.00) WHICH of the following interiock conditions WILL alkw the suppression pool cooling valve , '(1E12 F024A/B) to be opened during a LOCA? { [ - L a . a.
' The valve can be opened to allow for SPC with no interlock conditions.
b.
The respective LPCI Injection Valve (1E12-F042NB) is shut.
l c.
The respective LPCI Injection valve (1E12-F042NB) seal-in circuitry is reset.
d.
The respective LPCI Injection Valve (1E12-F042NB) is shut and a high drywell . pressure exists.
l l: QUESTION: 048 (1.00): k l' . Given the following: , During a plant transient in which Residual Heat Removal (RHR) A&B initiated in - the LPCI mode ~a valid Containment Spray mode automatic initiation signal is received - l-tho' operators confirm RHR Loop "A" has properly initiated in the Containment - L Spray Mode.
Then,' approximately one (1) minute after "A" Loop initiates, the "B" Loop still has NOT initiated.' , Choose from the following the cause for this condition: . ! a.
For the Containment Spray Mode of operation, the RHR "B" Loop is manual initiation only.
b.- - For the Containment Spray Mode of operation, the RHR "B" Loop lags "A" Loop ' by ninety (90) seconds.
' , _ c.
RHR Loop "B" has NOT operated in the Low Pressure Coolant. injection Mode i for greater than ten (10) minutes.
l d.
RHR' Loop "B" should have shifted to the Containment Spray Mode of operation , 90 seconds prior to RHR Loop "A", therefore it is inoperable.
. ..
' . -
- REACTOR OPERATOR Page 30 of 57- ' L i
[ -- t ' QUESTION: 049 '(1.00) H . - HIGH RADIATION CONTROL ROOM HVAC SYSTEM DIVISION 2 annunciator is received.
l The Control Room HVAC System WILL automatically go into High Radiation isolation mode if !' the condition is detected by: s " L a.. .'Any one of the four detectors (PR009A/B/C/D).
f.. b.
' Any two of the four detectors (PR009A/B/C/D).
c.
- PR009A and B detectors- , d.
PR009A and C detectors .g < L QUESTION: 050 (1.00) ' While operating at 100% power conditions the following plant conditions exist: A" RHR running in suppression pool cooling " - Diesel Generator 1A is tagged out - A plant transient occurs with a concurrent loss of both off-site power sources. Drywell pressure. - rises to 9.7 psig and the RPV depressurizes_to 200 psig - WHICH of the following identifies the expected RHR "A" configuration as a result of this' ' - transient? ~ a.
running in suppression pool cooling mode.
. b.
Running and injecting.
c.
NOT running d.. Running on MINIMUM flow . > ' [
' . V REACTOR OPERATOR Page 31 of 57 o-I L L - QUESTION: 051 - (1.00) The plant is 'at 40% power with 2 CD/CB pumps running. Both running condensate booster pumps trip.
, " - SELECT the plant response.
I- . a.
Both running condensate pumps WILL trip.
, b.
Standby condensate pumps auto start l c.
Standby condensate booster pump auto starts l d. - Feedpumps WILL trip on low suction pressure.
- QUESTION: 052 (1.00) , Diesel Generator "1A" is running and is paralleled with 4160 volt bus "1A1". A Loss of Coolant Accident occurs followed 3 minutes later by a loss of voltage on bus "1A1".
SELECT the expected sequence of events for the above conditions.
a.
The diesel WILL trip on the LOCA signal, then restart and automatically reenergize the bus when the loss of voltage occurs.
b.
. The diesel WILL trip on the LOCA signal and MUST be placed in the normal standby lineup to allow an auto start and the output breaker to close on loss of bus voltage.
,,., - c.
After the diesel output breaker opens on the LOCA signal, the operator MUST take the Output Breaker Control Switch to " Trip" and then release allowing the l Diesel to reenergize the bus.
' d.
The LOCA signal WILL trip the diesel output breaker, then the operator MUST - take the Output Breaker Control Switch to " Trip" and then "Close" to allow the Diesel to reenergize the bus.
H l u ,! I ' . _
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g , i " REACTOR OPERATOR Page 32 of 57
l QUESTICN: 053 (1.00) UPS 1 A has been transferred to its attemate source. WHICH of the following is a concem for operating in this configuration for long periods of time.
a.
damage may result to the computer b.
RC&lS WILL become INOP c.
overheating of the UPS inverter may occur d.
this may cause a self test failure QUESTION: 054 (1.00) .. WHICH water fire suppression system is ONLY MANUALLY initiated ? l a.
wet sprinkler system in the RCIC pump area.
b.
Pre-action sprinkler system in the Plant Service Water (WS) pump area.
c.
Water spray / deluge system for the Main Power transformer, d.
Deluge system for the SGTS charcoal filter.
i < - i e e
._ > { L: REACTOR OPERATOR Page 33 of 57 A QUESTION: 055 (1.00) . The plant was ' operating at 100% power with EHC in control of reactor pressure when the Bypass Jack was taken to the INCREASE position and held.
WHICH of the following describes the expected plant response if the Jack is held in the . INCREASE position? a.
The bypass valves WILL open full. The control valves WILL close as the bypass valves open. Reactor pressure WILL remain at pressure set setpoint.
! b.
The bypass valves WILL open full. The control valves WILL open as the bypass valves open until the' control valve limit stop is reached. Reactor pressure WILL- ' DECREASE and stabilize at a lower value than before the transient.
. c.
The control valves WILL open until the control valve limit stop is reached. The. bypass valves WILL open fully. Reactor pressure WILL DECREASE until the MSlVs close.
, l d.
There WILL be no change since bypass valves MUST be placed in TESTING to enable the Bypass Jack l l k-p , .. O ! .. ij .
REACTOR OPERATOR , Page 34 of 57 QUESTION: 056 (1.00) Given the following: i I f You have been directed to initiate SLC.
. You tum the keylock switch for the SLC Pump 'A' to RUN.
- Then, you note that the explosive valve fires, but SLC Pump 'A' indicates tripped.
Why hasn't SLC Pump 'A' started? a.
RWCU Outboard Isolation Valve (1G33-F004) has NOT yet closed.
b.
RWCU Inboard Isolation Valve (1G33-F001) has NOT yet closed.
c.
SLC Storage Tank Outlet Valve (1C41-F001A) has NOT yet fully opened.
d.
SLC Pump B keylock switch has NOT been turned to RUN.
i
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REACTOR OPERATOR Page 35 of 57 i i - o . QUESTION: 057 (1.00) Given the following: ' The plant is operating at 100% power when a loss of the 4160V bus 1B1 occurs.
- ' Then, following a scram due to Main Steam Isolation Valves (MSIV) closure, the operator notes
- .
L that the Rod Control & Information System (RC&lS) indication is blinking ON and OFF.
i CHOOSE from the following the action the operator could take to verify ALL RODS IN using the RC&lS display: a.
Depressing the DATA SOURCE pushbutton to select the operable channel.
l.
' b.
Depressing the RAW DATA and SCRAM VALVES pushbuttons to deterinine rod f positions.
c.
Acknowledging the ACCUMULATOR FAULT to allow the rods to settle into the full in notch.
l-d.
Depressing the DATA MODE and DATA SOURCE pushbuttons to select the ! operable channel.
l l i QUESTION: 058:(1.00) I: - The plant is operating at 100% power when NORMAL OFF GAS FLOW LOW RANGE LOW j ! - annunciatoris received.
J WHICH of the following statements is correct? - a. -
- Off gas discharge valve 1N06-F060 has closed and WILL cause a loss of condenser vacuum b.
. Precooler iniet Valves, has closed and WILL cause a loss of condenser vacuum 'i c.
Recombiner.1 A/1B Condenser Level Controller closes and WILL NOT cause a foss of condenser vacuum j d.
The Charcoal Adsorbers are automatically placed into service after the alarm is received, and WILL NOT cause a loss of condenser vacuum.
i l . i .
L._ m
Ly ' ) - s ' 1: REACTOR OPERATOR Page 36 of 57
, _ QUESTION: 059 (1.00)- ' Work in the area of the Containment Building ventilation radiation monitors has resulted a loss ~ f all of the Containment Building ventilation exhaust radiation monitor channels. The crew o . should verify:- E a.
Containment Building ventilation has isolated, j , b.
Containment Building ventilation continues to operate normally.
c.
Only the inboard Containment Building ventilation dampers have shut.
d.
Only the outboard Containment Building ventilation dampers have shut.
QUESTION: 060 (1.00) The following conditions exist: A routine reactor startup is in progress - The mode switch is in STARTUP ' - - The main turbine is tripped.
- A valid MSIVisolation has occurred.
- The reactor did NOT scram (No ATWS conditions exists).
- Which of the following was the only signal that could have generated the MSIV isolation? a.
. Low reactor water level b.
High main steam line flow. c.
High main steam line radiation d.
Low main steam line pressure .
.- ' REACTOR OPERATOR Page 37 of 57 ' QUESTION: 061 (1.00) - ,, f ' WHICH of the following describes the concern with transferring water to Radwaste from RHR "A" verses RHR "B" during shutdown cooling operation 7 a. - high conductivity.
b.
High temperature.
c.
High radiation.
" d.
High fiowrate.
. QUESTION: 062 (1.00)
- WHICH of the following correctly describes reactor core flow orificing?.
a.
Orificing is used do provide INCREASED coolant flow in the lower power fuel bundles which experience a higher resistance to flow than the higher power fuel bundles, y b.. Orificing is used to provide INCREASED coolant flow in higher power fuel bun @s which experience a lower resistance to flow than the lower power fuel bute&.:4.
c.
Onficing is used to prevent DECREASED coolant flow in the higher power fuel bundles which experienca a higher resistance to flow than the peripheral fuel bundles.
d.
. Orificing is used to prevent DECREASED coolant flow in the peripheral fuel. bundles which experience a lower resistance to flow than the central fuel bundles.
.
.. REACTOR OPERATOR Page 38 of 57 QUESTION: 063 (1.00) WHICH of the following is a Safety Limit violation? al Steam dome pressure reaches 1310 psig for a hydrostatic test during cold shutdown.
> b.
MCPR reaches 1.08 during a loss of feedwater heating transient from full power.
c.
Reactor mode switch is placed in RUN with steam dome pressure at 0 psig.
d.
RPV water level momentarily drops to -10 inches on Wide Range during Refueling.- , ' QUESTION: 064 (1.00) ' Given the following: ' Refueling'is in progress - < - . Mode switch in REFUEL - Main Hoist loaded with a fuel bundle - t ~ Refuel bridge is over the containment transfer pool - WHICH of the following conditions WILL generate a Rod Block? a.
A control rod is selected from the Rod Select Matrix.
b.
The refuel bridge is moved over the core.
' c.
A control rod is not selected from the Rod Select Matrix.
' l_ d.
The Fuel Grapple control is placed in the RAISE position.
.
l I
REACTOR OPERATOR . Page 39 of 57 QUESTION: 065 (1.00) . Given the following: TURB TRIP EHC SYS has alarmed.
WHICH of the following describes the response of the MSR 1A/1B DRN VLVs (1TD-MSR(1-4))? a.
Do not Auto close, but can be opened by the control switches.
I b.
Auto open and are interlocked to prevent closing by the control switches.
c.
Auto close and are interlocked to prevent opening by the control switches.
d.
Do not Auto open, but can be closed by the control switches.
QUESTION: 066 (1.00) Given the following: The plant is in a transient condition. The following indications are noted: The Mode Switch is in Shutdown.
- 18 rods are between notch 04 and 36 -. , Reactor poweris approximately 40%. l - The MSIVs are open; the main turbine is on line.
- Reactor pressure is being maintained at 920 psig.
- Reactor water levelis 46 inches.
- Which of the following WILL initially cause reactor level to DECREASE? a.
Emergency depressurization.
. Boron injection.
b.
, c.- Turbine Trip.
d.
Recirc Pump Tri \\ REACTOR OPERATOR. Page 40 of 57 ' l' ' ' ! .' QUESTION: 067 (1.00) Given the following: The plant is operating at 100% power - feedwater controlis in automatic.
l -- E 1 Then, a transient occurs and RPV level reaches +7" on the narrow range instruments.
l WHICH of the following represent the level demand 11 seconds after level passed through +8.9 - inches decreasing? a.
Level dem?nd is 1/4 of the value on the level controller setpoint.
b.
Level demand is +18" c.
Level demand is +36" d.
Level demand +40" QUESTION:068 (1.00) Given the following:
The plant is operating at 60 % power.
- i Then, a generator trip occurs.
! WHICH of the following describes the condition of the Turbine Stop and Control valves in , ~ response to this event ? E a.
TSVs remain open, TCVs remain ope:1.
b.- TSVs remain open, TCVs close.
ic.
TSVs close, TCVs remain open.
! d.
TSVs close, TCVs close.
.
- REACTOR OPERATOR - Page 41 of 57 . QUESTION: 06g (1.00)' Givwn the following:, ' The plant is in Mode 4 -: reactor recirculation pumps are shutdown - RHR is in the Shutdown Cooling Mode of loop "B" -- reactor water level is 47 inches on Shutdown Range instruments.
- CHOOSE why the water level is to be maintained at this level.
' a.
Ensure proper NPSH for the RHR pump operating in the Shutdown Cooling Mode.
. b.
To bring indicated level on scale for Upset Range, allowing work on the other levelindicators.
c. ' Provide sufficient suction head for the Reactor Water Cleanup Pumps.
d.
Assist in natural circulation when RR shutdown to remove decay heat from the fuel . , QUESTION: 070 (1.00) . CHOOSE the method (s) of venting the containment to maintain pressure below the Primary Containment Pressure Limit? a.
Place Continuous Containment Purge in service L
- b.
Vent to the spent fuel pool using the FC retum header.
L l c.
Place Containment Atmosphere Monitoring system in service L i , ! .d.
Place Hydrogen Recombiners in service i < ! , i ! i l
.. .+ f-sREACTOR OPERATOR : Page 42 of 57 j . . L l ) QUESTION: 071 (1.00) a Given the followng conditions: , During plant heatup and pressurization reactor pressure is 460 psig and steady -- all bypass valves are closed - reactor water level is +34" (Narrow Range) - l~ the feedwater startup controller is in automatic
.- ' reactor poweris constant.
- - Then, the reactor water level control system fails, causing level to INCREASE. Assume no operator action is taken. SELECT the statement below that describes plant response to the transient. Reactor power WILL.. a.
INCREASE due to cold water addition i . ' b.
DECREASE due to the shrink from the cold water.
' - c.
INCREASES as pressure INCREASE due to collapsing voids.
d.
remain constant due to overall power coefficient.
j o ~ QUESTION: 072 (1.00) Given the following: -! The reactoris in power operation.
- WHAT is the potential consequence of prolonged isolaton of a control rod (>2 hours)? a.
. Overheating of the associated HCU Accumulator Piston Seals reducing the ability to insert or SCRAM the control rod b.
Overheating of the associated CRDM Drive Piston Seals reducing the ability to insert or SCRAM the control rod
' c.. Erosion of the other CRDM Drive Piston Seals from excessive Cooling Water ' Flow reducing the ability to insert or SCRAM the control rod
- d.
Erosion of the HCU SCRAM inlet and Outlet Valve graphitar seats due to overheating . ,.
. , i i
i ! O REACTOR OPERATOR Page 43 of 57 QUESTION:073 (1.00) In EOP-1, RPV Control, which of the following are you directed to do if you are to use SRVs to , - control pressure?. . -a.
Enter STEAM COOLING, EOP4.
t b.
Ensure Suppression Pool level is at least 8 feet.
c.
Open all Main Turbine Bypass Valves.
d.
You CAN NOT use SRVs unless there is a danger of overpressurization.
! ! ~ QUESTION: 074 (1.0J).
What two (2) plant parameters are required to determine the Primary Containment Level Limit? a.
Containment Pressure and Suppression Pool level b.
Suppression Pool Level and Reactor Pressure c.
Suppression pool temperature and Reactor Pressure d.
Suppression Pool temperature and Containment pressure j !
i l I-l L.
'
I l l
- REACTOR OPERATOR Page 44 of 57 - QUESTION: 075 (1.00) You are in EOP-1 RPV Control and you determine that Blow Down WILL most likely be - required. ' Following directions in EOP-1, you anticipate Blow Down and open the Main Turbine - . - ' Bypass Valves to the Main Condenser. As a result of this action, the conditions which were leading you to Blow Down are NOT as severe, and it is obvious that the EOP WILL no longer require Blow Down You should; a.
Enter EOP-3 RPV Depressurization and Blow Down since the determination to anticipate Blow Down was made
' b.
Open some SRV's to INCREASE the cooldown rate . c.
Recognize that your action mitigated the severity of the condition, and determine ' that Blow Down is no longer required d.
Defer Blow Down until the depressurization into the Main Condenser is complete I' r - j ' .j ' ,
REACTOR OPERATOR Page 45 of 57 l QUESTION: 076 (1.00)' '. Given the following: The plant is operating at 100%. -
- .
i' Then, both TDRFPs trip for an unknown reason. Plant conditions stabilize with: the mode switch in shutdown - all rods fully inserted - o ' . suppression pool water temperature is 93 degrees F - ' suppression pool level is 19'1" L - ' reactor water level is at -50" on wide range instruments - _, reactor pressure is 800 psig.
J -- Based on the information provided above, what operator action, is required? a.
Enter EOP-3, Blowdown.
b.
Enter EOP-1 RPV Control - cc Enter EOP-6, Containment Control d.
Enter EOP-7, Hydrogen Control > - I' )
r % REACTOR OPERATOR Page 46 of 57 ., i ' QUESTION: 077 (1.00) The plant is in an ATWS condition and is operating in EOP-1 A, A1WS RPV Control, Level Leg directs the operator to hold RPV water level between -162" and +52" using ONLY the following systems:
" Condensate /Feedwater - - ., CRD ) - RCIC ' J
.- RHR (through shutdown cooling) , - Why are these systems specifically designated for use in these conditions? 1 a.
These systems all have a reactor grade water source to inject into the reactor.
b.
At this point in the ATWS, reactor pressure precludes use of other systems.
c.
They provide the operator with much more precise level control during the ATWS.
d.
Their point of injection outside the shroud provides some " pre-heating" of the water.
QUESTION: 078 (1.00) The Plant is operabng at 90% Rated Power when suddenly RR Pump "B" trips to off. From the following choose the appropriate operator response.
y a.- Scram the Reactor if the Restricted Zone is entered.
.
b.
Shut the associated 1B33-F023B "B" Loop Suction Valve.
j c.
If the Restricted Region is' entered, wait until core instabilities are observed, then j . Scram the Reactor.
d.
Open the FCV on the "A" RR pump as required to keep flow from causing a entry into the Restricted Zone.
I i
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~l- , i: -l L - i g L REACTOR OPERATOR Page 47 of 57 ' il L , c , iQUESTION: 07g (1.00)
- WHICH of the following would be the most likely consequence of failing to re-open the / discharge valve of a tripped Reactor Recirculation (RR) Pump 5 minutes after shutting it? E, s.
Erroneous core flow data would be indicated on the core flow instruments due to - l the instruments automatically subtracting the reverse core data based on the
position of the RR Pump breakerc i '- b.
Strabfication and cooldown of the RR loop would occur due to failing to re-open - the discharge valve.
> c.
The tripped RR Pump would NOT receive any recirculation flow into it's seal area and the pump seal could ultimately fail.
d.
The differential pressure resulting from the loop cooldown could hydraulically lock the valve closed.
j 'l QUESTION: 080 (1.00) The Unit was on line at 100% power when condenser vacuum began to decay. Presently, condenser vacuum is 7.5 inches Hg. NO Operator actions have been taken. The current status - of the plant is: a.
Main turbine is tripped, MSIVs are open, BPVs are controlling reactor pressure.
a
- b. - Main turbine is on-line, MSlVs are open, BPVs are controlling reactor pressure.
. _ Main turbine is on line, MSIVs are open, TCVs are controlling reactor pressure.
l c.
- d.
' Main turbine is tripped, MSIVs are closed, SRVs are controlling pressure.
' , \\ -~ , [. .
n l, REACTOR OPERATOR Page 48 of 57 ,
r .. - QUESTION: 081 (1.00) l ? CHOOSE the reason for securing the RCIC Gland Seal Compressor prior to resotoring the Div 1 Desel Generator to service when restoring loads following a loss of AC power event.
a.
To prevent a compressor shunt trip' due to low voltage when the DG is started.' b.
To ensure sufficient DG field flashing current is available during startup.- c.
To reduce the starting load on the DG which could cause a DG trip on ! undervoltage.' l d.
To prevent the compressor from being load shed after the DG is started by the l: DG logic.
l l-QUESTION: 082 (1.00) Given the following: [ The plant is operating at 100% power - feedwater level control is in automatic.
l - = Then, the B Steam flow signal to the Reactor Water Level Control System fails upscale.
,, Assume NO operator action. WHAT is the effect on the ACTUAL Reactor Vessel water level? Reactor water level WILL... l-a.
continuously DECREASE until low level scram setpoint is reached.
l
b.
INCREASE until terminated by a high level feed pump trip.- c.
DECREASE to a lower level and stabilize.
d.
INCREASE to a higher level and stabilize.
p . t
, REACTOR OPERATOR Page 4g of 57 ,
~ QUESTIONi 083 (1.00) - WHICH of the following requires entry into EOP-6, PRIMARY CONTAINMENT CONTROL 7 Suppression Pool Temper'ature is[ rees F.
4gy a.
Suppression Pool L'evelis [eet ' W' Otti usve<s b.
Drywell Temperature is hrees F.
Cite 9) 1p yumf ISS c.
g . p d.
Drywell Hydrogen Concentration i 4mwgr.
QUESTION: 084 (1.00)- Given the following: The plant is at full power - , , one loop of RHR is in service in the Suppression Pool cooling mode for routine - f cooldown.
w Then, an SRV has inadve'rtently opened. The other loop of RHR was immediately started in . Suppression " 1 cooling.
o ~ WHICH of the fvViLL result it the SRV CAN NOT be reciosed?
- s.
.., suppression pool WILL heat up a bit, but when the second loop is maximized
the heat-up WILL stop.
< b.
continued cooldown of the pool with both loops in operation, but at a reduced '; . ~ rate due to the open SRV.
. c.
cooldown of the pool. Each heat exchanger is sized to equal the heat rate of , , L one open SRV.
,
- --
~ d.
heat-up of the' suppression pool as the two heat exchangers CAN NOT equal L heat rate of an open SRV.
- .
. . , .
- f
,..
,i.
s - REACTOR OPERATOR ' Page 50 of 57 . QUESTION: 085 (1.00) WHICH of the following systems / modes of operation is available from the Remote Shutdown f -- Panel? , ~! . a.' Div 1 RHR in Fuel Pool Cooling Assist i b.- Div 2 RHR in the Shutdown Cooling Mode J c.
Inboard and Outboard Main Steam isolation Valves s d.
Div 1 RHR in the Suppression Pool Cooling Mode QUESTION: 086 (1.00) Given the following: . A refueling accident in the Containment Building Fuel Transfer Pool resulted in a fuel bundle being dropped and is damaged.
a ' . Then, the control room annunciator HI RAD INITIATION SGTS is received.
! SELECT the statement that describes the expected plant response that prevents a High - Off-Site Release Rate.
, s.
VF (Fuel Building) supply and exhaust dampers remain open, the supply and exhaust fans trip on the high radiation signal.- > ~ b.- ' VF (Fuel Building) supply and exhaust dampers isolate and the supply and exhaust fans trip.
c.
VF (Fuel Building) exhaust dampers isolate and the exhaust fans trip when the damprs isolate, the supply fans continue to operate.
d.
VF (Fuel Building) supply dampers isolate and the supply fans trip on the high radiation signal, the exhaust fans continue to operate.
<
t ' v {y . [: ',.s - REACTOR OPERATOR ' Page 51 of 57 o - QUESTION: 087 (1.00)- 'While operating at 75 percent reactor power, the following alarms are received: . CCW EXPANSION TANK LVL Hl/LO-CCW PMP AUTO TRIP .. In addition to the CCW System, WHICH of the following should be monitored / tracked during this event? . ' a.
Reactor Recirculation Pumps b. ' - Fuel Building HVAC cooling coils c.
Drywell Chillers d.
' Auxiliary Building HVAC cooling coils QUESTION: 088 (1.00) - , Given the following: The plant is at 100% power.
= - Then, two (2) SRVs fail open. A short time later, reactor water level is steady and bulk
Esuppression pool water temperature indicates 97 degrees F.- - CHOOSE the correct action to be taken:
- a.
Place the Reactor mode switch in SHUTDOWN due to elevated suppression ~ f ~ pool temperature, b.
Enter EOP-1, RPV Control.
L.
c.
Enter EOP-6, Primary Containment Control.
! v.
. Start up a Standby Gas Treatment train on the secondary containment and shutdown the Fuel Building HVAC.
- ..
r .y - REACTOR OPERATOR Page _52 of 57 , , <
QUESTION: 089_ (1.00) Given the following: The Containment Compressed Gas Supply isolation Valves, ilA012B and ilA013B have automatically isolated. WHICH of the following signals can cause this condition? 'a; . Low IA/SA header pressure of 70 psig.
b.
High Drywell pressure and/or Level 1.
c.' An ADS automatic actuation signal.
d.
High Drywell pressure and/or Level 2.
] - QUESTION: 090 (1.00) . Given the following: A Main Control Room evacuation has occured " - plant control has been established at the Remote Shutdown Panel.
- .WHICH Safety Relief Valve (SRV) has the MOST air available for operatum following a loss of instrument air, (Assume no operator actions have been taken for the loss of instrument air.)
, ! a.
F051G , b. - F051D p
c.
F051C , . ' d.
F051B . . .h'
m REACTOR OPERATOR Page 53 of 57 , QUESTION: 091 (1.00) s Given the following:
- ~
Reactor pressure is 855 psig .. . - control rod 22-11 is at position 00 with the it's nitrogen accumulator isolated for - , repair, j Then, the operating Control Rod Drive (CRD) pump trips and the ACCUM TROUBLE annunicator is received for the rods 18-27 (at position 00) and 38-23 (at position 48).
WHAT achon required? a.
Immediately place the reactor mode switch to " SHUTDOWN".
. b.. If any other accumulator becomes inoperable for a withdrawn rod, immediately place the reactor mode switch in " SHUTDOWN."
c.
If any other accumulator becomes inoperable for a withdrawn rod, start a CRD i i pump within 20 minutes or place the reactor mode switch in " SHUTDOWN."
d.
Start a CRD pump within 20 minutes or place the reactor mode switch in : " SHUTDOWN."
.,. QUESTION: 092;(1.00) While operstmg in EOP-6, " Primary Containment Control", the operator is directed to perform ' an emergency RPV depressurization when plant conditions "CAN NOT be held below the Heat Capacity Limit".
' ~ IDENTIFY plant conditions that MUST be evaluated to make the decision to blowdown.
,, .'s.
Primary Containment water level and containment pressure b.
RPV pressure ~and suppression poo_l temperature . c. ' Drywell pressure and drywelliemperature - d.
. RPV pressure and drywell temperature q.. ,
m ! ! REACTOR OPERATOR Page 54 of 57
QUESTION!093 (1.00) ' ) -. ' Given the following: ATWS actions are in progress.
- drywell temperature has INCREASED to the unsafe region of Detail B, RPV- - Saturation Temperature.
A loss of alllevelinstruments has been declared - , ' RPV flooding has been entered.
- ' .\\ L ' WHY are the RPV Water Level Instruments taken to be unusable? The combination of... ' a.
drywell pressure and temperature makes the reference legs unreliable (- b.
- drywell pressure and RPV temperature makes the reference legs unreliable l c.
drywell pressure and temperature makes the variable legs unreliable.
l-d.
drywell temperature and RPV pressure make the reference legs unreliable l-l QUESTION: 094 (1.00) The initiating signals for a valid suppression pool dump actuation have occurred and are still
L present.
' WHAT actions are necessary to close the Division i Suppression Pool Dump Valves (1SM001 A 'and 1SM002A) and have them remain closed.
The Division i Suppression Pool Dump Valves may be closed... ,.; , , L a.
after the " Suppression Pool Dump Valve Mode Selector Div I" is placed in I ' " Disable".
b.
after the 25 minute delay timer from the initiation signal has timed out.
, l i' L c.
after the "SM System Div i in Test" switch is placed in the " Test" position.
L d.
after the "LCPS/LPCI FM RHR A Seal in Roset" pushbutton on P601 is depressed.
l e - ,
.- \\~ ! . REACTOR OPERATOR Page 55 of 57 h QUESTION: 095 (1.00) A valid Fuel Building Exhaust High Radiation signal comes in. WHAT is the ret.ponse of the o Standby Gas Treatment System (VG).
! ,-- s.
- BOTH trains of VG start. The following ventilation systems are running
- VRNQ, VF, CCP -
' - BOTH trains of VG start. VF is secured. The following ventilation systems are b.-
running: VRNQ, CCP.
c.
ONE train of VG starts. The following ventilation systems are secured: VRNQ,: VF, CCP.
d.
ONE train of VG starts. VF is secured. The following ventilation systems are running: VRNQ, CCP. Only
, QUESTION: 096 (1.00)- Given the following: A plant transient has occurred resulting in a primary system discharging into secondary ) containment.
Then, RP reports the Fuel Building Vent Plenum Monitors,1RIX-PR006A-D reads 18 mr/hr and increasing.
j L WHICH of the following operator actions is requ; red? a.
Rapidly depressurize the RPV using the main turbine bypass valves; OK to exceed 100 degree F/hr cooldown rate.
' b.
. Verify Fuel Building Ventilation (VF) isolation and SGTS (VG) starts.
, _ < . c.
Isolata the primary system that is discharging into the area even if it is needed i ' for core cooling.
l-h d.
Defeat VF isolatic'is and restart V j .c R J p.
j.
[l . k REACTOR OPERATOR' Page 56 of 57 l: - ! QUESTION: 097, (1.00) ( Plant conditions are as follows: l.
The reactor has been in COLD SHUTDOWN for two (2) days following power - . ! operation.
(~ Reactor vessel water level is 35 inches.
- L Neither reactor recirculation pump is available.
-- l' 1 Shutdown coolmg has isolated and the shutdown cooling suction valves CAN ,- , l: NOT be opened.
i ' H WHICH of the following operator actions WILL prevent reactor vessel stratification AND provide j 7 ecay heat removal? i d , ! a.
Place Reactor Water Cleanup System in service.
' l b.
Insert a manual scram to maximize CRD flow to the bottom head region.
c.
Start a second CRD pump and maximize CRD cooling water differential Pressure (d/p).
d.
Raise reactor vessel water level until the HIGH REACTOR LEVEL Hl/ LOW annunciatoris received.
j ' QUESTION: 098 (1.00) A primary system is discharging into the secondary containment. WHICH of the following conditions would require BLOW DOWN in accordance with EOP-3, EMERGENCY RPV ' DEPRESSURIZATION? j a.
LPCS Area water level above max safe and LPCS floor drain sump levels above max normal.
L b.
RWCU Pump Room B area temperature above max normal and RWCU Pump j l Room C area temperature above max normal.
c.
LPCS Pump Room area radiation above max safe and RHR Pump Room C area radiation above max safe.
d.' RHR Eqpt Area cooler delta T above max normal and RCIC Eqpt Area cooler delta T above max normal.' . d .
i i l I REACTOR OPERATOR Page 57 of 57 l ' QUESTION: 099 (1.00) CHOOSE the immediate Operator Action for CPS. No. 4011.02, Spent Fuel Pool Abno: mal Water Level DECREASE.
a.
Suspend any fuel / core component lift operations after placing loads in a safe condition.
b.
Sound the containment evacuation alarm.
c.
Verify all appropriate automatic actions occur and manually perform any that have NOT.
d.
Start standby gas treatment and establish secondary containment integrity.
> QUESTION:100 (1.00) Given the following: ] The plant is shutdown to mode 4 - shutdown cooling is placed in service.
- Then, shutdown cooling becomes unavailable.There are procedures in place to establish - I alternate heat removal flow paths should shutdown cooling become unavailable.
WHATis the reason for implementing these procedures? These procedures are available,.. a.
because there are no incore temperature detectors.
b.
to prevent an inadvertent mode change.
c.
to prevent ORDM from over heating d.
to protect the incore neutron detectors l ("*"*"" END OF EXAMINATION *"*"****) . I
REACTOR OPERATOR Page 1 of 11 l l REFERENCE DATA ANSWER: 006 (1.00) l
- a.
ANSWER: 001_-(1.00) REFERENCE: c.
CPS 1406.01, Annunciator Tracking - REFERENCE: . . Program, Rev.10, page 4, . 10CFR55.53, Conditions of licenses.
Section 8.1.3 & 8.1.4.b l LP 87460, CPS No.1401.01, Objective 1.4 LP87468, Objective.1.5 L 2.1.2 ..(KA's). 2.1.11 ..(KA's) l . ANSWER: 002 (1.00) ANSWER: 007 (1.00)
y d b.
~ REFERENCE:. REFERENCE: LP87205.1.19 - , _. . CPS No.1005.01, CPS PROCEDURES L ' Technical Specifications 3.9.8 action A AND DOCUMENTS, Rev 38, page 7.
2.1.11 ..(KA's) LP85132, Objective.1.1.1
2.2.12 ..(KA's) i ! ANSWER: 003 (1.00) ! c' ANSWER: 008 (1.00) l REFERENCE: . c CPS No. 4005.01, Rev.13, Loss Of REFERENCE: j Feedwater Heating note and immediate CPS 1014.01 SAFETY TAGGING,' Rev. 22,
actions on page 2: ' page 14, NOTE above step L LP87505, Objective.1.2 8.3.1.4 2.1.20 ..(KA's) LP85140, Objective.1.2.8 2.2.13 ..(KA's) ANSWER: - 004 (1.00) c-ANSWER: ' 009 (1.00) - REFERENCE: c ( CPS No.- 1052.01, CONDUCT OF REFERENCE: l SYSTEM LINEUPS, Rev 7, page 14, LP88602, Objective.1.18 Section 8.5.2.1.2/.3 ' CPS No.1893.00 CPS OPERATIONAL l-LP85110, Objective.1.1 FIRE PROTECTION PROGRAM, Rev. 6, ' 2.1.29 ..(KA's) Section 3.0 2.4.25 ..(KA's) ANSWER: 005 (1.00) l l-C ' REFERENCE l .3310.01,' Reactor Core Isolatior, Cooling (RI), Rev.15, page 4, Section 4.4 ' LP87217. Objective 1.17-2.1.32 ..(KA's) . ' l'
- e:
REACTOR OPERATOR Page 2 of 11 ANSWER:'010 (1.00) ANSWER: 015 (1.00) - c.
d REFERENCE: ' REFERENCE: . Technical SpecWications SR 3.9.2.2: LP 87212 ~ Perform CHANNEL FUNCTIONAL TEST,7 201001K404 ..(KA's) days LP87630, Objectvie.1.2.2 . 2.2.24 ..(KA's) ANSWER: 016 (1.00) b .' REFERENCE: ANSWER: 011 (1.00) LP 85201, Control Rod Drive Hydraulics ~ a .. 201001K408 ..(KA's) ' REFERENCE:- CPS No.1401.01, CONDUCT OF OPERATIONS, Rev.' 28, page 22, step ANSWER: 017 (1.00) 8.3.1.2. ORM 6.2, Table 6.2.2-1 c LP87633,' Objective.1.7 REFERENCE: 2.2.32_ ..(KA's) 5004-1L, Annunciator Response procedure for APRM Downscale LP87411, Objective 1.4.1 201002K103 ..(KA's) ' ANSWER: 012 (1.00) s.- REFERENCE: ANSWER: - 018 (1.00) ' LP88602, Objective.1.13 d-2.3.1 ..(KA's) REFERENCE: See CPS No. 3314.02, Section 8.1.3 and - Figure 1 to find tank level less than low level ANSWER: 013 (1.00) alarm.
d LP87201, Objective.1.6.12 REFERENCE: 211000A208:..(KA's) EPIP RA-03, Radiological Exposure - Guidelines, Section 4.1.2 - LP 11016 ANSWER: 019 (1.00) ' LP85234, Objective.1.1 d 2.3.4 ..(KA's) ' REFERENCE: LP85202 . . . LP85202, Objective '.1.18.2 ANSWER: - 014 (1.00) 202002K103 ,.(KA's) ' 'C, REFERENCE: LP87212 Section 1.1.1 E.4
- . 201001 A105...(KA's)
. f REACTOR OPERATOR Page 3 of 11 1 ANSWER: 020 (1.00) ANSWER: 026 (1.00) _ d-d ' REFERENCE: ' REFERENCE: . CPS No. 4008.01 Abnormal Reactor ' LP85482, Objective 1.4.2 Coolant Flow, Rev.15, page 5.
215003K102 ..(KA's) LP 85402, Reactor Recirculation Flow : Control 202002A207'..(KA's) ANSWERi 027 (1.00) d-REFERENCE: ANSWER: 021 (1.00) LP 85215 SRMs, Objective.1.1.10 c-215004K101 ..(KA's) REFERENCE: EOP-4 > LP87556l Objective.1.2.3. ' ' ANSWER: 028 (1.00) 203000K502...(KA's) b-REFERENCE: . . LP 87411, Objective.1.1.2 ANSWER: 022 (1.00) 215005A404 ..(KA's) b REFERENCE: LP 85209, Objective 1.7 ANSWER: ;029 (1.00) 209001K408 ..(KA's) c REFERENCE: LP 87411.1.1.2 Annunciator procedures ANSWER: 023 (1.00) 5006-2H and 5005-3B c 215005A104 ..(KA's) REFERENCE: LP85380, Objective.1.2.7 209002A301 ..(KA's) ANSWER: 030 (1.00) d l REFERENCE: ! ANSWER: - 024 (1.00) LP 85423, Objective.1.6.5 ' i , ' c.
REFERENCE:~ 216000K501 ..(KA's) j LP 85211, Objective.1.4.2 ' 211000A308 ..(KA's) ANSWER: 031 (1.00) -d ANSWER: 025 (1.00) REFERENCE: l b~ LP87570, Objective 1.8.2 l.
REFERENCE: 259001A207 ..(KA's) l LP87411, Objective 1.1.3 . ' 215005A107 ..(KA's) t . , ' ,
REACTOR OPERATOR-Page 4 of 11 ANSWER:. 032 (1.00) ANSWER: 037 (1.00) 'd a
- REFERENCE
REFERENCE: 5063-20: RCIC PUMP SUCTION LP 87241, section 1.4.1 and LP 87248 ' PRESSURE LOW: AUTO ACTIONS:RCIC 241000K306 ..(KA's) Turbine trip.
LP 85217'.1.5.6, Reactor Core Isolation Cooling * REFERENCE . ANSWER: 038 (1.00) CPS No. 3310.01, Rev.15,'pg 30, Appendix - b B.
REFERENCE: 217000A301 ..(KA's) CPS No. 3103.01, FEEDWATER, Rev.15, page 8, section 4.8 LP87570, Objective.1.5.2 ANSWER:' 033 (1.00) 259001K607 ..(KA's) b ' REFERENCE: ^ ' - LP 87218, Objectives.1.2.1 &.2, ANSWER: 039 (1.00) Engineered Safety Systems Actuation b 218000K501 ..(KA's) REFERENCE: LP87570, Section 1.1.1 259002K605 ..(KA's) L . ANSWERf 034 (1.00) d . REFERENCE: ANSWER: 040 (1.00) LP85205, Objective 1.4.6 b CPS No. 3312.01, Residual Heat Removal, REFERENCE: ' Rev. 28, pg 17, Note at top of page LP 85261, Objective.1.4.1 ' 223001A304 ..(KA's) 319.01, STANDBY. GAS TREATMENT.
4 SYSTEM, Rev.13, pg. 6, note for step 8.1.2.1' ANSWER: 035 (1.00) 261000K101 ..(KA's) a L REFERENCE: LP 85205, RHR, Obj 1.4.5 ANSWER:' 041 (1.00) ! '223002K108 ..(KA's) b - REFERENCE: l.
.. . LP87264, Objective.1.5.6.3.c L ~ ANSWER: 036 (1.00) 5060-7E TRIPPED DIESEL GENERATOR [. a.. 1A, CPS No. 3506.01, Diesel Generator .. ' i.
REFERENCE: and Support Systems, Rev. 23 Section LP 85433, index item 1.2.2 2.1.7 239002K405 ..(KA's) - 264000K402 ..(KA's) . N
J l L REACTOR OPERATOR Page 5 of 11 !. ! ANSWER:1 042 (1.00) L ANSWER: 048 (1.00) a b REFERENCE:. REFERENCE: LP87400, Section 1.4. 2 . CPS No. 3312.01, RESIDUAL HEAT '201004K404 ..(KA's) REMOVAL (RHR), Rev. 28; ' LP85205, Objective.1.4.7 226001A408 ..(KA's) - ANSWER: 043 (1.00) e a.
REFERENCE: ANSWER: 049 (1.00) LP 87401, Objective.1.1.5 d 201003K405-..(KA's). REFERENCE: LP 85447, Objective.1.4.1 -
- Control Room HVAC System, CPS No.
ANSWER: 044 (1.00) 5052-7M b ' , . 290003K101 ..(KA's) l REFERENCE: L LP85202, Objective.1.6.1 . CPS No. 3302.01, Rev. 21,~ pg. 40, step ANSWER: : 050 (1.00) 8.2.4.2 c , 202001 A218_'..(KA's) REFERENCE: ( 85205,.1.11 - RHR l 230000K606-..(KA's) ANSWER: 045 (1.00) Lj l l d.
I REFERENCE: ANSWER: 051 (1.00) LP85204, Objective.1.4.1 d 5000-1C,' Annunciator Response procedure REFERENCE: 204000A213 ..(KA's) - LP 85256, Objective.1.5.4 , l 5002.-1J & 1C, Annuciator Response
. procedures ' E _. ANSWER: 046 (1.00) 256000A201 ..(KA's) a REFERENCE:- i.
CPS No. 3312.01, RESIDUAL HEAT ANSWER: 052 (1.00) ' ' REMOVAL, Rev. 28, pg 41, Section c l 8.1.12.4.15 REFERENCE: " . LP86205, Objective.1.5.7 CPS No. 3506.01,' Rev. 23, pg 8, Step 4.18 205000K201 ..(KA's) LP87264, Objective.1 262001K101 ..(KA's) r ANSWER: 047'(1.00) ' ' a REFERENCE: 87205,.1.7.1 (page 36). 219000K101 ..(KA's) ~ ( b I
REACTOR OPERATOR Page 6 of 11 ANSWER: 053 (1.00) ANSWER: 058 (1.00) a a REFERENCE: REFERENCE: LP86576, Objective.1.8 LP 85255, Objective.1.1.2 CPS No. 3509.01,4.6: 5130.06, Annunciator Response Procedure, 262002K401 ..(KA's) Rev. 26, pg 5, step 3.
271000K301 ..(KA's) ANSWER: 054 (1.00) d ANSWER: 059 (1.00) REFERENCE: a LP85261, Objective.1.2.3 REFERENCE: CPS No.1893.04, M310,719' Control LP 85273, Objective.1.1.3 Control and HVAC Equipment Area, Rev. 4, Pre-fire interlocks plan 272000K403 ..(KA's) ' 286000A304 ..(KA's) ANSWER: 060 (1.00) ANSWER: 055 (1.00) b a REFERENCE: REFERENCE: LP85239, Objective.1.6.1 LP85248, Objective.1.1.2 CPS No. 4001.02, Automaticisolation CPS No. 3105.04, Rev 7, pg 4, Section Group I, Rev.12, pg 2, 2.5,2.2 and pg 20, Section 8.2.9 Technical Specifications Table 3.3.1.1 ~ KA's) 239001K116 ..(KA's) 245000K302 ..( . ANSWER: OSS (1.00) ANSWER: 061 (1.00) c.
c ' REFERENCE: REFERENCE: CPS No. 4411.10, EOP SLC Operations, LP85205, Objective.1.2.5 Rev.1, page 3, section 2.1, Note CPS No. 3312.01 RESIDUAL HEAT l LP85211, Objective.1.4.3 REMOVAL ' '204000K607 ..(KA's) 268000K502 ..(KA's) ANSWER: 057 (1.00) ANSWER: 062 (1.00) I d c ' REFERENCE: REFERENCE: LP8701, Objective.1.1.5 LP87448, Section 10.1.1.6 CPS No. 3304.02, ROD CONTROL AND 290002K403 ..(KA's) l lNFORMATION SYSTEM ' ' (RC&lS), Sections 8.1.7.1 & 8.1.7.2.
LER, Scram April 9,1996 ' 214000K105 .'.(KA's) I . I ,
- REACTOR OPERATOR Page 7 of 11 . ANSWER:.063 (1.00) ANSWER: 068 (1.00) b d REFERENCE: REFERENCE: . LP87621, Objective.1.1 LP85245, Main Turbine / Generator, Technical Specifications 2.1.1.2 states in ll.A.1.1.3 part, 295007A105 ..(KA's) ' l' 290002K507 ..(KA's) c . ANSWER: 069 (1.00) l ANSWER: 064 (1.00) d
b REFERENCE: l REFERENCE: CPS No. 3303.01, REACTOR WATER l LP87401, Objective.1.2.5, RC&lS, pg 19, CLEANUP, Rev. 20, pg 8, Step 6.7 l d) 2 - LP87442, Objective.1.4 234000K104-..(KA's) 295009K105 .,(KA's) ANSWER:. 065 (1.00) ' ANSWER:' 070 (1.00) b b l REFERENCE: REFERENCE: 5007-1B TURB TRIP EHC SYS, Rev. 24, 4411.06 Emergency Containment Venting, pg 3, Auto Action Step 1 Purging, And Vacuum Relief Section 2.5: LP85245, Objective.1.5.2 Vent to Spent fuel Pool Using FC Return 295005A203 ..(KA's) Header 4402.01, EOP-6, 87558 LP87558, Objective.1.2.3 ANSWER: 066 (1.00) 295010A101 ..(KA's) c REFERENCE: L LP874981.9.7.1.10.1.11, Step 4 ANSWER: 071 (1.00) a 95005K103-..(KA's) REFERENCE: LP 87498,.1.9.2 Transient and Event worksheet ANSWER: 067 (1.00) 295014K206 ..(KA's) b REFERENCE: LP87570,.1.7.1 ANSWER: 072 (1.00) 295006A102 ..(KA's) b REFERENCE: CPS No. 3304.01, Control Rod Drive Hydraulics, Rev. 24, pg 36, Note to step 8.2.3.5 LP86201, Objective.1.8.4 , j 295005K201 ..(KA's) .
.
-i e REACTOR OPERATOR Page 8 of 11 ANSWER: 073 (1.00) ANSWER: 078 (1.00) b: a ! REFERENCE: REFERENCE: LP87552, Objective.1.2.4 &.5 CPS No. 4008.01, Abnormal Reactor , CPS No. 4401.01, EOP-1, RPV Control Coolant Flow, Rev.15, pg 3, step 3.1 .295024K101 ..(KA's) LP87508, Objective.1.2 ] 295001K102 ..(KA's) ! ANSWER: - 074 (1.00) a ANSWER: 079 (1.00)
REFERENCE: -b.
i
EOP-6 REFERENCE: l ~ LP87558, Objective.1.10 CPS No. 4008.01, Abnormal Reactor
- 295024A209
..(KA's) Coolant Flow, Rev.15, pg 5, step 6.1 l LP87508, Objective.1.3.2 295001A101 ..(KA's) ANSWER: 075 (1.00) ] c REFERENCE: ANSWER: 080 (1.00) ' LP87552, Objective.1.9.5 d , RPV Control (EOP-1) REFERENCE: ) 295025A201 ..(KA's) LP87498, Transient Analysis, Section l ! .1.9.11 295002K305 ..(KA's) ANSWER: 076 (1.00) b.- REFERENCE: ANSWER: 081 (1.00) LP87552, Objective.1.1 b ' RPV Control (EOP-1).
REFERENCE: 295031A204 ..(KA's) LP 87513, Sections 1.7 and 1.9 ' 295003K106 ..(KA's) . ANSWER: 077:(1.00) d.
ANSWER: 082 (1.00) REFERENCE: d ATWS RPV Control (60P 1A) REFERENCE: LP87553, Objective 1.3.2 LP87570, Feedwater Control, Section 295037A202 ..(KA's) ' IV.1.8.2 295008K203 ..(KA's)
REACTOR OPERATOR Page 9 of 11 t ' ~ ANSWER: 083 (1.00) ANSWER: 089 (1.00) l c d (, REFERENCE: REFERENCE: . l LP87558, Section 1.1 CPS No. 4001.02, Automatic isolation, Rev.
i 295012K101-..(KA's) 13, checklist C001 group 8 ! LP85239, Objective.1.8.9 295019K303 ..(KA's)
L ' ANSWER: 084-(1.00) .d i-REFERENCE: ANSWER: 090 (1.00) LP87509, Suppression Pool, Step 1.7 a.
! l LP85234, Objective.1 REFERENCE: 295013K201 ..(KA's) LP85239, Section 1.1.15 295019K214 ..(KA's) ANSWER: 085 (1.00) ' d.. ANSWER: 091 (1.00)-
!- REFERENCE: d LP85433, Section 1.4 REFERENCE: , L 295018K201 .'.(KA's) LP85201,.1.8.17 Technical Specifications 3.1.5.b.
295022K207 ..(KA's) l ANSWER: 088 (1.00)
i b-REFERENCE: ANSWER: 092 (1.00) LP 85449, Section 1.3.4 b l 295017K204 ..(KA's) _ REFERENCE: ! LP87558 ' LP87558, Objective.1 ANSWER: 087 (1.00) 295028K301 ..(KA's) a l REFERENCE: l LP 85208, Objective.1.9 ANSWER: 093 (1.00) l. 5040.02, Annunciator Response Procedure, d L Rev. 25, step 2.d and 3203.01, REFERENCE: 295018A102 ..(KA's) LP87553,.1.1, EOP-1 A, ATWS RPV ' Control, Detail B.
295028K303 ..(KA's) l-ANSWER: 088 (1.00) l c l; REFERENCE: ANSWER: 094 (1.00) t LP87558,.1.1 Suppression pool water a . temperature of 95 degrees F is an entry REFERENCE: condition for EOP 8.- 87408 page 8 295027K301 ..(KA's) LP87408, Objective.1 295029A201 ..(KA's)
l
l- . REACTOR OPERATOR : Page 10 of 11- . ANSWER: 095 (1.00) ANSWER: 100 (1.00)- -b b- - REFERENCE: REFERENCE:- - LP85261 STANDBY GAS TREATMENT LP 87299, Objective.1.1.
j SYSTEM, item 1.4.1_ CPS No. 4006.01, Loss of Shutdown 295033K204...(KA's) - Cooling, Note to Step 4.5 .295021K305 ..(KA's) L ANSWER: 096 (1.00)- b' REFERENCE: J L87559, Objective.1.1 Secondary Containment Control (EOP-8) 295034K201 ..(KA's) ANSWER: 097 (1.00)' a REFERENCE: LP 87299, Objective'.1.5, . CPS No. 3303.01, Reactor Water Cleanup,
- Rev.-20, pg 38, Note beginning section 8.2.6, item b 295021K102
..(KA's) ANSWER: ' 098 (1.00) c REFERENCE: LP87559, Objective.1.1 ' EOP-8, SECONDARY CONTAINMENT CONTROL.
295032A201'..(KA's) i 1Y ~ ANSWER: 099 (1.00)
-a.
i REFERENCE: i ' CPS No. 4011.02, Spent Fuel Pool Abnormal Water Level Decrease, Rev. 3; pg 2, step 3.1 ' LP87297, Objective '.1.2 l 295023K101:..(KA's) . . -l (*."""*" END OF EXAMINATION """"")
REACTOR OPERATOR Page 11 of 11 A N S \\N E R.K E Y MULTIPLE CHOICE 001 - c 021 c 041 b 061 c-081 b 002 d - 022 b 042 a - 062.c 082 d 003 c-023 c'- 043 a 063 b 083 c 004 c 024 c 044 b '064 b 084 d 005 c 025 b 045 d 065 b 085 d 006 - a 026 d 046 a 066 c 086 b 007. b 027 d 047 a 067 b 087. a 008 c-028 b 048 b 068 d 088 c 009 c 029 c 049 d - 069 d 089 d 010 c 030 d 050 c 070 b. 090 a 011 a 031 d 051 d-071 a 091 d 012 a 032 d 052 c 072 b 092 b. 013 d 033 b 053 a 073 b 093 d 014 c 034 d 054 - d 074 a 094 a 015 ~ d 035 a 055 a 075 c 095 b 016 b 036 a 056 c 076 b 096 b 017 c 037 a 057 d 077 d 097 a ' 018 d 038 b ~ - 058 a 078 a 098 c 019 - d 039 b 059 a 079 b 099 a 020 d ' 040 b' 060 b 080 d 100 b - (""*""* END OF EXAMINATION ""*""*) .
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u ~ >> l l NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS !' 1.
After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.
i 2.
To pass the examination, you must achieve a grade of 80.00 percent or greater. Every question is worth one point.
3.
For an initial examination, the time limit for completing the examination is four hours.
l For a requalification examination, the time limit for completing both sections of the i examination is three hours. If both sections are administered in the simulator during a single three-hour period, you may retum to a section of the examination that was already completed or retain both sections of the examination until the allotted time has expired.
4.
You may bring pens and calculators into the examination room. Use only black ink to ensure legible copies.
5.
Print your name in the blank provided on the examination cover sheet and the answer sheet. You may be asked to provide the examiner with some form of positive identification.
6.
Mark your answers on the answer sheet provided and do not leave any question blank.
Use only the paper provided and do not write on the back side of the pages. If you decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change, 7.
If the intent of a question is unclear, ask questions of the NRC examiner or the designated facility instructor only.
8.
Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
Avoid all contact with anyone outside the examination room to eliminate even the ! appearance or possibility of cheating, g.
When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on u examination cover sheet indicating that the work is your own and that you have neither given nor received
assistance in completing the examination. The scrap paper will be disposed of - immediately after the examination.
10.
After you have tumed in your examination, leave the examination area as defined by the proctor or NRC examiner, if you are found in this area while the examination is still in progress, your license may be denied or revoked.
11.- Do you have any questions? L !
.. SENIOR REACTOR OPERATOR Page 3 of 58 This page intentionally blank (replaces answer sheets)
)
m - I - N; , 2 SENIOR REACTOR OPERATOR Page.4 of 58
L l ' QUESTION: 001 (1.00) ' ,9 ' 10CFR55 lists the requirements that a licensed operator MUST meet to maintain his/her license in an " active status." The operator shall actively perform the functions of a licensed operator for~ a MINIMUM of: p - a.
. four - 8 hour shifts per calendar quarter, - b.
five - 8 hour shifts per calendar month.
y c.
seven - 8 hour shifts per calendar quarter.
! ' d.
three - 8 hour shifts per calendar month.
' QUESTION: 002 (1.00) Given the following: ' ' The plant is in mode 2 with a startup in progress. Shift tumover is normally scheduled at 0700 - hours.
- .
Then, at 0400 hours, one of the two on-duty Reactor Operators passes out and is taken to an ' area hospital.
Choose the required action for replacing the Reactor Operator in accordance with CPS Operational Requirements Manual.
) a.
SS and LASS both MUST remain in the control room until the position is filled.
b; immediate action MUST be taken to fill the position, and it MUST be filled by 0600.
c.
Immediate action MUST be taken to assure the ill RO's relief is coming in on ., time.
- d.
One position may be vacant until the next shift comes on duty, if planned ' activities are canceled.
I , ?
SENIOR REACTOR OPERATOR Page 5 of 58 QUESTION:003 (1.00) ~ Given the following: A normal plant startup is in progress with the unit at 23% - Then, the Reactor Operator reports: Increasing reactor power with no change in RR flow or rod motion - Further checks show that feedwater temperature is decreasing and a loss of - feedwater heating is in progress.
What immediate Operator Action is required? a.
Immediately reduce power to 20 MWe (approx 60 MWt) by inserting control rods.
b.
Immediately reduce power by 20 MWe (approx. 60 MWt) by inserting control rods.
c.
No immediate operator action is required based on current power level.
d.
Immediately reduce power by 20 MWe (approx. 60 MWt) by reducing recirculation flow.
. L
. ' SENIOR' REACTOR OPERATOR Page 6 of 58
- QUESTION
- 004 (1.00)
,You are on duty when an area operator calls you regarding a valve lineup he is conducting. He asks you for the proper method to set a manual throttle valve to 25 tums open, as instructed on. .the lineup.
i ' ' SELECT your response to the area operator from the following.
- a.
Move the handwheel counter-clockwise to the full open travel position, then move the handwheel clockwise 25 tums.
b.
Move the handwheel clockwise to the full closed travel position, then move the handwheel counter-clockwise 25 tums.
. c.
Move the handwheel counter-clock;.i= in the open direction, from full closed until you just hear system flow through the valve, then open the valve 25 tums.
d.
Move i s handwheel clockwise in the closed direction, from full open until you just b' Jystem flow through the valve stops, then open the valve 25 tums.
-
. b __ SENIOR REACTOR OPERATOR Page 7 of 58 QUESTION: 005 (1.00) -. Given the following: Shortly after shift change, Station Nuclear Engineering arrives to support the performance of CPS No. 2202.02, Control Rod Sequence Exchange at Power procedure.
Then, while reviewing the procedure, you note the previous shift made a temporary change signed by an Qualified Nuclear Engineer and by an SRO Licensed Nuclear Engineer who last ' stood watch more than six months ago.
- Which action should be taken in accordance with CPS Technical Specifications /ORM7 - a.
This procedure CAN NOT be performed as written because the change MUST be approved by the On-Duty Shift Supervisor and the Plant Manager and initialed by the t. ASS.
b.- This procedure can be performed as written provided the On-Duty Shift Supervisor and the LASS approves the change and it is initialed by the "A" - Control Room Operator.
c.
This procedure can be performed as written after verbal permission is received from the Plant Manager and approved by the On-Duty SS, LASS and the "A" Control Room Operator.
d.
This procedure CAN NOT be performed as written because the change MUST be approved by two members of the plant staff management, one of whom MUST hold a current SRO licens r I.
L !- i f SENIOR REACTOR OPERATOR Page 8 of 58 l l-QUESTION: 006 (1.00) - An annunciator has multiple inputs, but only one of the inputs is providing nuisance annunciation. Once the single nuisance input has been disabled per CPS No.1406.01 (ANNUNCIATOR TRACKING PROGRAM), which of the following describes the process used to track the inoperable alarm? a.
A Temporary Modification should be used along with an entry in the Out of Service / Disabled Annunciator Log.
b.
A Plant Modification should be used along with an entry in the Control Room Operator's Log.
c.
An Engineering Work Request should be used along with a entry in the Out of Service / Disabled Annunciator Log.
d.
A red arrow entry should be placed hi the Control Room Operator's Log along with an entry in the Shift Superviscr's Log.
, QUESTION: 007 (1.00) During the performance of a surveillance, the performer encounter 0 a step with CV HOLD" written. WHAT is the expected response in accordance vith CPS 1005.01, CPS Ptocedures and Documents? a.
The performer is required to stop and nuMy inspectior, personc el to allott for planned inspections. Once notification hae 21:.1 accomolished and the agreed to time (or a reasonable amount of time) hu pFssed, ma w ork activi'y may continue.
b.
The performer is required to stop and notify inspudion pcrom el to al;as.or planned inspections. The work activity shall not procesd withgui f v noinf being signed by inspection personnel, or inspection per.oned he'4 not'fis c. an:f authorizing the activity to proceed, or the point vra:Vod/reckwiM1 c.
The performer is required to notify inspection personne' hVkr for them to plan when they will perform the verification activity The wc4 shall be capable of being verified after work completion.
d.
The performer is required to continue work on the survei!!ance until completion.
After the work is done, the performer shall call for inspection personnel to perform the inspection,
~ SENIOR REACTOR OPERATOR Page g of 58 i QUESTION:008 (1.00) , WHICH of the following is allowed when hanging tags in accordance with CPS 1014.01
- SAFETY TAGGING?.
a.
Two caution tags may be hung on an open drain valve only if the same contact person has responsibility for both tags.
b.
A danger tag may be hung on a closed breaker inside an energized electrical cabinet. The tag can be attached to the breaker with string and a small section of tygon hose.
. . c.
A danger tag may be hung for a removed breaker. The tag can be attached to q the breaker cubicie door.
d.
The Area Operator may change a required position by initialing and dating the handwntten correction on a danger tag.
QUESTION: 00g (1.00) Who is responsible for reviewing the results of Post Maintenance Testing (PMT) and completeness of PMT documentation? a.
Supervisor - Mechanical Maintenance i b.
Work Authority / Shift Technical Advisor ] c.
Supervisor - Operations Support Coordinator d.
Supervisor - Controls and Instruments
i ,
_ _ _ _. SENIOR REACTOR OPERATOR Page 10 of 58 QUESTION: 010 (1.00) Which of the following describes the bases for the Refueling Equipment interlocks LCO in accordance with CPS Technical Specifications? Ensures no more than one control rod may be withdrawn at a time.
a.
b.
Provides protection should a prompt reactivity excursion occur.
Ensure that fuel assemblies are NOT loaded with any control rod withdrawn, c.
d.
Reduces the radiological consequences of a fuel handling accident to within limits.
QUESTION: 011 (1.00) Given the following: The plant is in Mode 5 with refueling operations in progress.
- The Core Alterations Surveillance Log shows that the refuel position one-rod-out - interlock surveillance was last completed at 0800.
Then, when performed at 2130 by operations, the one-rod-out interlock surveillance fails.
WHAT actions are required in accordance with CPS Technical Specifications? Immediately suspend loading of irradiated fuelinto the RPV; initiate action to a.
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.
Immediately suspend control rod withdrawal and initiate actions to fully insert all c.
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 hour.
__
- - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _. _ _ _ _ _ _ ! SENIOR REACTOR OPERATOR Page 11 of 58 QUESTION: 012 (1.00) Conditions in a recently surveyed area are: 50 dpm/100cm2 alpha loose surface 50% DAC airbome radioactivity - Which of the following describes the posting requirements for the area? a.
Contaminated area - required Airborne radioactivity area - required b.
Contaminated area - required Airborne radioactivity area - not required c.
Contaminated area - not required Airbome radioactivity area - required d.
Contaminated area - not required Airbome radioactivity area - not required QUESTION: 013. (1.00) ' SELECT the choice that completes the following statement.
During an emergency the may verbally authorize personnel exposure up to legal limits, per 10CFR20. (MUST be recorded in logs) a.
Shift Supervisor ' b.
Emergency Manager l c.
Emergency Director d.
TSC Radiological Controls Supervisor ..... . _ - - _ _ _ _ _
. . . SENIOR REACTOR OPERATOR Page 12 of 58
- .
QUESTlON: 014 (1.00) A valve lineup is being performed to venfy the position of a safety related valve in a High Radiation area following maintenance on the system. It is expected to take 10 minutes to verify the position of the valve.
Which of the following describes the requirements for verification of the valve's position? a.
Valve position shall be verified using normal venfication methods. Independent verification is required.
b.
Valve position shall be verified using normal verification methods. Double - verification is required.
c.
An altemate verification method may be used in place of normal verification methods. Independent verification is required.
d.
An attemate verification method may be used in place of normal verification methods. Double verification is required.
, QUESTION: 015 (1.00) The Assistant SS is designated as Fire Brigade Leader. The Fire' Brigade Leader'is responsible for WHICH of the following during a fire? a.
Make emergency plan notifications b.
Notify off site fire protection agencies for assistance c.
- Taking charge of the at-the-scene fire-fighting operations d.
Taking charge of the fire-fighting operations from the control room !- - _ _ -
_ _ _ _ _ _ . _ ____ __ _ __ __ _ _ ___ _ _ i ! , l . . SENIOR REACTOR OPERATOR Page 13 of 58 - QUESTION: 016 (1.00) , . Plant conditions warrant declaration of Emergency Classification at the SITE AREA EMERGENCY level.' From the following, choose who would normally have Command Authority from the Emergency Operations Facility during the event.
a.
On-Duty Shift Supervisor b.
~ Emergency Operations Supervisor c.
EOF Director d.
Emergency Manager QUESTION: 017 (1.00) SELECT the choice that completes the following statement.
The is the attemate location for OSC operations.
- s.
- Service Building Lunchroom b.
Craft Security Portal c.
Emergency Operations Facility d.
737' Radwaste Building, machine shop area .. - _ _ _ _ _ _ _
, ...... . .. . . _ _ _ _ - _ _ _ _ - _ _ - _ _ _. _ - _ - -. _ - - _ -. - _ _ , - SENIOR REACTOR OPERATOR Page 14 of 58 QUESTION: 018 ~ (1.00) WHICH of the following effects would occur as a result of throttling closed 1C11-F003, CRD ! DRIVE PRESSURE CONTROL VALVE 7 ' l a.
DECREASES control rods insertion time on a scram.
'b.
INCREASES control rod withdrawal speed.
c.
INCREASES cooling water flow to each CRDM.
d. ' DECREASES seal flow to the Recirculation Pumps.
QUESTION: 019 (1.00) The plant is operating at 100% RTP. The Transmission Services Group provided the STA/SE with the following information at 0930 this moming: Projected MINIMUM Transmission System Predicted MINIMUM Operable Offsite Source Voltages - Voltages 138kV Source = 138.4kV Expected at 1700 138kV Source = 137.5kV 345kV Source = 346.0kV Expected at 1830 345kV Source = 346.1kV The appropriate actions have been taken in response to the above information.
At 1500 the South bus voltage bottoms out at 345.9kV. What are the MCR actions required? a.
Verify correct breaker alignment for each offsite circuit within 1 hour.
b.
Restore offsite circuit to OPERABLE status within 1 hour - c.
Restore offsite circuit to OPERABLE status within 24 hours d.
No actions are required l l
i-L SENIOR REACTOR OPERATOR - Page 15 of 58 l QUESTION: 020 (1.00) ' R'esctor power is 50%. A failure of the main turbine first stage pressure transmitters results in a l zero pressure signal being sont the RC&lS system.
! u WHICH of the following describes the effect of this failure on the RC&lS system? : a.
No effect on rod control since the Rod Pattem Controller receives its input from the APRMs.
b.
No effect on rod control since the Rod Pattom Controller receives its input from crossover header pressure.
c.
Continuous control rod withdrawal WILL be limited to two notches.
d.
Rod Pattem Controller WILL enforce its preprogrammed rod pattem criteria.
QUESTION: 021 (1.00) Following a Low Pressure Coolant injection (LPCI) initiation signal, the operator attempts to shift from LPCI mode of RHR operation to the Suppression Pool Cooling mode of RHR . operation.
SELECT the statement below which best describes the impact on Suppression Pool Cooling with the LPCI initiation signal still present? a.
1E12-F024A, Test Valve to Suppression Pool, WILL NOT stay open for 10 minutes.
- b.
1E12-F064A, RHR MINIMUM Flow Valve, WILL NOT stay clossd for 10 minutes, c.
1E12-F048A, Heat Exchanger Bypass Valve, WILL NOT stay closed for 10 minutes.
i d.
1E12-F014A, SSW Inlet RHR Heat Exchanger, WILL NOT stay open for 10 minutes.
,
) L i4 SENIOR REACTOR OPERATOR Page 16 of 58 ' , l ,
!- ! QUESTION: 022 (1.00). i After Anticipated Transient Without Scram (APNS) condition which required the injection of Standby Liquid Control (SLC) System solution into the core, the operators MUST refill the SLC Storage Tank.
the containment is accessible - SLC Storage tank contains approximately 1500 gallons - ' SLC is shutdown - SLC Storage Tank solution is 79 degrees F - SLC Storage Tank sodium pentaborate concentration is 13.5%. - CHOOSE the operator action, prior to involving Chemistry, to be taken to prepare the SLC System for future injection, if needed a.
Use the Makeup Condensate System and fill the tank to 5000 to 5150 gallons.
b.
Place local Mixing Heater control switch in the AUTO position.
c.
INCREASE the SLC Storage Tank solution concentration to greater than 14.5%. d.
Fill the tank to clear the low level alarm using the Makeup Condensate System.
- QUESTION: 023 (1.03) _ How WILL INDICATED (on P680 panel) loop flow and core flow change if one of the two calibrated A-loopjet pump flow signals fails to zero? a.
"A" Loop flow WILL DECREASE 50 percent, "B" loop flow WILL INCREASE 50 percent and core flow WILL remain unchanged.
b.
"A" Loop flow WILL DECREASE 10 percent, "B" loop flow WILL INCREASE 10 , ! percent and core flow WILL remain unchanged ' , c.
"A" Loop flow WILL remain unchanged, "B" loop flow WILL remain unchanged and core flow WILL DECREASE 25 percent.
! ' d.
"A" Loop flow WILL remain unchanged, "B" loop flow WILL remain unchanged and ccre flow WILL DECREASE 5 percent.
.
j
,. SENIOR REACTOR OPERATOR Page 17 of 58 I i= i , QUESTION: 024 (1.00) The plant is operating at 100 percent power. WHICH of the following describes the response of l . the Recirculation Flow Control System to a Feedwater pump trip if the vessel level falls to 28 i inches? ! a.
Recire pumps downshift to slow speed b.
Recire pumps trip to OFF c.
Flow control valves run back to - 54% indicated valve position.
. d.
Flow control valves run back to - 19% indicated valve position.
QUESTION: 025 (1.00) ..While in EOP-4 (Steam Cooling) adequate core cooling is assured by WHICH of the following? a.
Core submergence with LPCI Injection, level above TAF, - 162 inches b.
Steam Cooling when level is above -193 inches c.
Steam Cooling when level is above -205 inches d.
Natural circulation with LPCI Injection, level above -45.5 inches .
SENIOR REACTOR OPERATOR Page 18 of 58 QUESTION:026 (1.00) The plant is operating at 60% reactor power with Low Pressure Core Spray System (LPCS) surveillance, CPS No. 9052.01, LPCS PUMP OPERABILITY, is being performed.. The following conditions occur:- A high drywell pressure signal is received and the reactor scrams -: RX water level stabilizes just below level 8 - RX pressure is 975 psig and stable on the Turb. bypass valves - _ Drywell pressure 4 2.3 psig.
- WHICH of the following describes the status of the Low Pressure Core Spray system? a.
LPCS WILL realign and inject into the vessel.
b.
LPCS WILL res!ign and operate on flow.
c.
LPCS pump WILL trip, d.
LPCS WILL remain in full flow test mode.
QUESTION: 027 (1.00) _WHICH of the following describes the normal operation'of the HPCS Flow Valve,1E22-F012, during an automatic HPCS initiation 7 s.
Normally closed. Automatically opens after HPCS initiation and remains open ~ until manually closed when system flow reaches 625 gpm and discharge pressure is less than 145 psig.
, b.
Normally open. Automatically closes when system flow reaches 625 gpm and discharge pressure is less than 145 psig.
I c.
Normally closed. Automatically opens after HPCS initiation and automatically closos when system flow reaches 625 gpm and discharge pressure over 145
pag.
d.
Normally open. MUST be manually closed from the control room after system
flow reaches 625 gpm and discharge pressure is over 15 psig.
j i I
SENIOR REACTOR OPERATOR Page 19 of 58 QUESTION: 028 (1.00) WHICH of the following identifies the starting sequence for the Standby Liquid Control (SLC) System when the control switch SLC PUMP B is placed in the RUN position? Assume all appropriate actions NOT listed occur as expected.
a.
One squib valve fires and the RWCU Outboard Suction isolation Valve (1G33-F004) closes.
b.
Two squib valves fire and the RWCU Outboard Suction isolation Valve (1G33-F004) closes.
c.
One squib valve fires and the RWCU Inboard Suction isolation Valve (1G33-F001) closes.
d.
Two squib valves fire and the RWCU inboard Suction isolation Valve (1G33-F001) closes.
QUESTION: 029 (1.00) Given the following conditions: i The "B" Recirculation Loop is isolated.
- Reactor poweris 43%. - Loop recirculation flow is 55% of rated.
- WHICH of the following is the APRM Flow biased Thermal Power Trip setpoint. (Choices are rounded to the nearest whole number.)
j a.
04 percent b.
84 percent c.
79 percent d.
73 percent
SENIOR REACTOR OPERATOR Page 20 of 58 QUESTION: 030 (1.00) From the following, choose how the intermediate Range Monitor System (IRM) is interrelated ~ with the Rod Control and information System (RC&lS).
a.
IRM signal level amplifier provides flux indication to the RC&lS Rod Interface Display Module.
b.
IRM trip units provide control rod insert blocks to the RC&lS.
c.
IRM voltage preamplifier provides rod withdrawal blocks to the RC&lS.
d.
IRM trip units provide control rod withdrawal blocks to the RC&lS.
QUESTION: 031 (1.00) WHICH of the following statements describes how the shorting links are used in the reactor protection system (RPS)? a.
Installation of the shoiting links enables a scram if any single dRM, IRM or APRM channel trips.
b.
Removal of the shorting links enable the SRM scrams in a coincidence of one-out-of-two-twice logic scheme.
c.
Installation of the shorting links enable the SRM scrams in a coincidence of one- ' out-of-two-twice logic scheme.
d.
Removal of the shorting links enables a scram if any single SRM, IRM or APRM - channel trips.
l l
i l L i
F l SENIOR REACTOR OPERATOR Page 21 of 58 ) QUESTION: 032 (1.00) ,l' \\ A Local Power Range Monitor (LPRM) detector has been bypassed in APRM "A". WHICH of - the following identifies what APRM "A" meter would indicate in the Count position if this is the j only bypassed LPRM7 a.
33-b.
i c.
, d.
l TRIPS ALARMS Teammat FLOW URM MRM j gypgg3 RESET - _upsci upset upscL
Ns5L E" "0" DNS L
~ NORMA 1_ . I VOLTS u a %F _ r1 e is H Avs % 1xx , f FLOW %- 2KK H - II COUNT OTY SKX
10 CAL MA 4KK M
H ' EXT %
SENIOR REACTOR OPERATOR Page 22 of 58 - QUESTION:033 (1.00) The plant is operating at 85% power. The following are the' indications received when the APRM meter function switches in the backpanels are placed in the AVERAGE, COUNT, and FLOW positions.
' AVERAGE ' COUNT.
FLOW - APRM A 94 %
69% APRMB 84 %
68 % APRM C 86 %
69% '~ A P R M D 89 %
61 % { WHICH of the following are the expected annunciators for these conditions? a.
~ ROD OUT BLOCK only b.
DIV 2 OR 3 NMS TRIP only , c.
ROD OUT BLOCK and DIV 2 OR 3 NMS TRIP d.
ROD OUT BLOCK and DIV 1 OR 4 NMS TRIP , , I I i ,
SENIOR REACTOR OPERATOR Page 23 of 58 l l f QUESTION: 034 (1.00) The following plant conditions exist: ! Recirc pumps have tripped - ~ Reactor has scrammed - A steam ieak has occurred - , The drywellis at 212F and 2.4 psig - ' The RPV has been emergency deptessurized - Which of the following level instruments would most accurately indicate RPV water level if level is on scale for that level instrument? a.
Narrow Range b.
Wide Range c.
Shutdown Range d.
Fuel Zone Range QUESTION: 035 (1.00) Which of the following causes the Steam Bypass Vsives to open if the main turbine should trip ' with the reactor at 60% power? a.
Steam line pressure WILL be greater than the PRESSURE SET setpoint.
b.
The BYPASS JACK signal WILL be greater than the control valve demand signal c.
The STEAM BYPASS AND PRESSURE REGULATOR pressure signal WILL be . less than the load limit setting d.
Turbine control valve demand signal WILL be less than the MAXIMUM " COMBINED FLOWlimiter signal.
l
l ! l
L.
l t-L . , SENIOR REACTOR OPERATOR Page 24 of 58 L
l . QUESTION: 036 (1.00)' l .. . Following an automatic initiation, RCIC speed is observed to be zero, the RCIC Trip-Throttle ~ Valve (TTV) is open and the steam supply shutoff valve (1E51-F045)is closed. WHICH of the following conditions could have caused this RCIC_ response.
. a.
' RCIC Pump suction low pressure b.
RCIC ISOLATION has been armed and depressed c.
Low reactor pressure 'd.
Reactor water level high (L8) ' QUESTION: 037 (1.00) Given the following plant conditions: , PLANT CONDITIONS
- RPV waterlevel:
-150 inches for 105 seconds.
Drywell pressure: 2.0 psig (RHR) pumps: NOT running LPCS Pump . NOT running L.
' SELECT the statement that describes the effect that pump motor breakers failing to close for '!
the Low Pressure Core Spray (LPCS) pump and the Low Pressure Coolant Injection (LPCI) . pumps. would have on the Automatic Depressurization System (ADS).
I l
' a.
ADS automatically actuates at this time.
J b.
. ADS WILL NOT automatically actuate.
c.
ADS WILL actuate 105 seconds after the ADS "A" manual initiation pushbutton !s depressed.. j d.
ADS WILL actuate when the ADS "B" manual initiation pushbutton is depressed.
! I l ! I-l
L
SENIOR REACTOR OPERATOR _ Page 25 of 58 I
QUESTION: 038 '(1.00) During a LOCA, Containment Pressure is 7.6 psig.
CHOOSE the statement that would result in the automatic initiation of the Containment Spray - Mode of RHR.
I a.
LPCI has been in operation for 90 seconds and Reactor Vessel Water Level is - greater than +30.8".
b.
LPCI has been in operation for 10 minutes and 10 seconds Reactor Vessel Water Level is greater than +45.5" c.
Reactor Vessel Water Level is greater than +30.8" and Drywell Pressure greater than 1.68 psig.
d.
LPCI has been injecting for 10 minutes and 10 seconds and Drywell Pressure greater than 1.68 psig.
' QUESTION: 039 (1.00) i During shutdown testing with RHR "A"in shutdown cooling a Group 3 high RPV pressure ] isolation signal is inadvertently generated.
WHICH of the following describes the expected system response to this signal? a.
"A" pump trips and both SDC suction isolation valves (1E12-F008/F009) close.
b.
"A" pump remains running and the neither SDC suction isolation valve (1E12-F008/F009) closes.
j c.
"A" pump trips and the both SDC suction isolation valves (1 E12-F008/F009) remain open.
! d.
- A" pump trips and Division i SDC suction isolation valve (1E12-F008) closes.
"
, ! I L.
SENIOR REACTOR OPERATOR Page 26 of 58 t - p QUESTION: 040 (1.00) L , SELECT the statement that describes the operation of the Safety Relief Valves operated from the Remote Shutdown Panel (RSP)if the associat in EMERGENCY.
a.
All automatic functions except Safety mode are disabled, b.
Relief Mode and Safety Mode disabled. All other automatic functions operable, c.
Relief Mode and low-low Set Mode disabled. All other automatic functions operable.
. d.
NO automatic functions are disabled. Low-low set WILL NOT initiate when valves are opened manually from the RSP.
QUESTION: 041 (1.00) SELECT the reason for requiring a MINIMUM water level above the top of irradiated fuel seated in the spent fuel storage and upper containment fuel pool racks.
a.
To ensure sufficient removal of fission products released if an irradiated fuel assembly ruptures.
b.
To ensure the fuel bundle remains covered with water when rechanneling c.
To ensure sufficient heat removal for the irradiated fuel assemblies.
i d.
To ensure sufficient volume of water is available in case it is necessary to dump the upper Containment Pool to the Suppression Pool.
i '
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, SENIOR REACTOR OPERATOR Page 27 of 58 ' QUESTION: 042 (1.00) - ! When DC control power to the 6900V AC switchgear is lost, what is the impact on the CW - pump motor breaker with a breaker-trip condition present? {
a.
The motor breaker WILL NOT trip automatically but can be tripped at the breaker.
without endangering personnel.
L b.
The motor breaker WILL automatically trip to place the motor on the bus.
L . . . c.
The motor breaker WILL NOT trip automatically and CAN NOT be tipped at the breaker without endangering personnel.
a d.
Indication WILL be lost for breaker position but the breaker WILL operate. l I normally otherwise.
! l QUESTION: 043 (1.00)
- .
A total loss of component cooling water occurs and is unrecoverable. A Recirc Pump trip is required... a.
immediately b.. within 1 minute c.
within 5 minutes-L d.' - when motor winding temperature reaches 225 degrees F.
- -
, l l. s {
j, - i ' L-
SENIOR REACTOR OPERATOR Page 28 of 58 t ' t QUESTION: 044 (1.00) WHICH of the following describes the expected RWCU system lineup as a result of a high Filter Domineralizer inlet temperature 7 , i a.
One pump running, F/D's on hold, both containment suction isolation valves - open.
b.
No pumps running, F/D's online, both containment suction isolation valves shut.
' c.
One pump running, F/D's online, both containment suction isolation valves open.
d.
No pumps running, F/D's on hold, outboard containment suction isolation valve shut.
< QUESTION: 045 (1.00) - WHICH of the following identifies the functional isolation signals to the SDC suction isolation valves with shutdown cooling established at the Remote Shutdown Panel? a.
automatically isolate on low RPV water level only.
b.
automatically isolate on high RPV pressure only.
c.
automatically isolate on either low RPV water level or high RPV pressure.
d.- WILL NOT automatically isolate on either low RPV water level or high RPV
pressure.
!
l ' SENIOR REACTOR OPERATOR Page 29 of 58 l l l l< QUESTION:046 (1.00) V . . WHICH of the following interlock conditions WILL allow the suppression pool cooling valve
- (1E12-F024A/B) to be opened during a LOCA?
< , The valve can be opened to allow for SPC with no interlock conditions.
a'. ! b.
The respective LPCI Injection Valve (1E12-F042A/B) is shut.
E c.
- The respective LPCI Injection valve (1E12-F042A/B) seal-in circuitry is reset.
I.
d.
The respective LPCl injection Valve (1E12-F042A/D) is shut and a high drywell l pressure exists.
i ! 'l l QUESTION: 047 (1.00)~ Given the following:
The CRD system was operating normally with the flow control valve in automatic.
- ' Then, the flow control valve became mechanically bound.
- WHICH of the follow!ag describes the expected effect on CRD system response?
- .
a.
During RPV depressurization, there could be difficulty with rod motion, affecting ' rod speeds.
' O b.
During RPV depressurization, thero could be outward rod drifts and excessive rod speeds.
c.
During a scram, CRD pump runout could occur - d.
During a scram, there could be slow scram times.
l l .
p l
- .
y l-SENIOR REACTOR OPERATOR Page 30 of 58 ., - QUESTION: 048 - (1.00) - L - l - HIGH RADIATION CONTROL ROOM HVAC SYSTEM DIVISION 2 annunciator is received.
< The Control Room HVAC System WILL automatically go into High Radiation isolation mode if i the condition is detected by: l s a.
. Any one of the four detectors (PR009A/B/C/D).
b.
' Any two of the four detectors (PR009A/B/C/D).
f c.
PR009A and B detectors d.
PR009A and C detectors ' QUESTION: 049 (1.00) While operating at 100% power conditions the following plant conditions exist: "A" RHR running in suppression pool cooling
- Diesel Generator 1A is tagged out - ' A plant transient occurs with a concurrent loss of both off-site power sources. Drywell pressure ' rises to 9.7 psig and the RPV depressurizes to 200 psig ' WHICH of the following identifies the expected RHR "A" configuration as a result of this ' transient?- i a.
running in suppression pool cooling mode.
b.
Running and injecting.
c.
NOT running 1 d.
Running on MINIMUM flow L L
r- , SENIOR REACTOR OPERATOR Page 31 of 58 l I , QUESTION: 050 (1.00)
Given the following: ! The plant is operating at 100% power when a loss of the 4160V bus 181 occurs.
-- Then, following a scram due to Main Steam isolation Valves (MSIV) closure, the operator notes that the Rod Control & Information System (RC&lS) indication is blinking ON and OFF.
CHOOSE from the following the action the operator could take to verify ALL RODS IN using the RC&lS display: a.
Depressing the DATA SOURCE pushbutton to select the operable channel.
b.
Depressing the RAW DATA and SCRAM VALVES pushbuttons to determine rod positions, c.
Acknowledging the ACCUMULATOR FAULT to allow the rods to settle into the fullin notch.
, I L d.
' Depressing the DATA MODE and DATA SOURCE pushbuttons to select the i operable channel.
l i ,
.~ SENIOR REACTOR OPERATOR Page 32 of 58 QUESTION: 051 (1.00) Given the following: The plant is operating at 100% reactor power with 2 turbine driven reactor feed pumps. ' (TDRFPs) running in automatic on the Master Water Level Controller with the tape set at 35 inches.
j i Then, the reactor water level transmitter selected for input to the Feedwater Level Control i System fails to a level of 33 inches.
) Assuming no operator action, this WILL cause: ) a.
RCIC initiation as water level DECREASES to Level 2.
b.
reactor water level to remain at 35 inches because the steam and feedwater flow signals overcome the level control signal.
c.
the motor driven feedwater pump to start.
d.' a reactor scram as water level reaches Level 8.
QUESTION: 052 (1.00).
UPS 1 A has been transferred to its attemate source. WHICH of the following is a concem for operating in this configuration for long periods of time, a.
damage may result to the computer , b.
RC&lS WILL become INOP
c.
overheating of the UPS inverter may occur d.. this may cause a self test failure .
SENIOR REACTOR OPERATOR Page 33 of 58 QUESTION:053 (1.00) WHICH water fire suppression system is ONLY MANUALLY initiated 7 a.
. wet sprinkler system in the RCIC pump area.
b.
Pre-action sprinkler system in the Plant Service Water (WS) pump area.
c.
Water spray / deluge system for the Main Power transformer.
d.
Deluge system for the SGTS charcoal filter, !-
.
> ~ SENIOR REACTOR OPERATOR Page 34 of 58 QUESTION: 054 (1.00)' WHl'CH of the following describes the concern with transferrl.sg water to Radwaste from RHR ' "A" verses RHR "B" during shutdown cooling operation 7 a.
- high conductivity.
b.
High temperature.
c.
High radiation.
d.
High flowrote.
QUESTION: 055 (1.00) -WHICH of the following correctly describes reactor core flow orificing? - s.
Orificing is used do provide INCREASED coolant flow in the icwer power fuel bundles which experience a higher resistance to flow than the higher power fuel bundles.
b.
Orificing is used to provide INCREASED coolant flow in higher power fuel bundles which experience a lower resistance to flow than the lower power fuel bundles.
c.
Orificing is used to prevent DECREASED coolant flow in the higher power fuel bundles which experience a higher resistance to flow than the peripheral fuel _ bundles.
d.
Orificing is used to prevent DECREASED coolant flow in the peripheral fuel - bundles which experience a lower resistance to flow than the central fuel ~ bundles.
, SENIOR REACTOR OPERATOR Page 35 of 58 l QUESTION: 056 (1.00) WHICH of the following is a Safety Limit violation? a.
Steam dome pressure reaches 1310 psig for a hydrostatic test during cold - shutdown.
b.
MCPR reaches 1.08 during a loss of feedwater heating transient from full power.
c.
Reactor mode switch is placed in RUN with steam dome pressure at 0 psig.
d RPV water level momentarily drops to -10 inches on Wide Range during. Refueling.
, QUESTION: 057 (1.00) On the initiation of a reactor scram from 100% power, the insert line of a CRDM breaks off completely at the weld on the pipe connected to the insert port at the CRDM.
WHICH of the following describes the expected response of the rod? a.
. WILL immediately start to insert and WILL fully insert with reactor pressure only 7.*
b.
WILL immediately stad to insert and WILL fully insert with accumulator pressure . only, c.
- WILL NOT insert until reactor and accumulator pressures equalize and WILL insert with reactor pressure only, d.
WILL hydraulically lock and WILL NOT insert e . . L
< SENIOR REACTOR OPERATOR Page 36 of 58 q QUESTION: 058 - (100) - ] . '. Given the following: The plant is operating at 100% power - ' feedwater controlis in automatic.
L - . Then, a transient occurs and RPV level reaches +7" on the narrow range instruments.
WHICH of the following represent the level demand 11 seconds after level passed through +8.9 inches decreasing? . a.
Level demand is 1/4 of the value on the level controller setpoint.
l b.
Level demand is +18" c.
Level demand is +36" - d.
Level demand +40" l l l.' QUESTION: 059- (1.00) Given the following:. -: . The plant is operating at 60 % power.
' Then, a generator trip occurs.
WHICH of the following describes the condition of the Turbine Stop and Control valves in response to this event 7 a.' TSVs remain open, TCVs remain open.
! b.
TSVs remain open, TCVs close.
c.
TSVs close, TCVs remain open.
d.
. TSVs close, TCVs close.
i
,. . . . . _ _.._ ___ _ _ _ __ ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .. _ _ _ SENIOR REACTOR OPERATOR Page 37 of 58 QUESTION: 060 (1.00) Given the following: The plant is in Mode 4 - reactor recirculation pumps are shutdown - RHR is in the Shutdown Cooling Mode of loop "B" - reactor water level is 47 inches on Shutdown Range instruments.
- CHOOSE why the water level is to be maintained at this level.
a.
Ensure proper NPSH for the RHR pump operating in the Shutdown Cooling Mode.
b.
To bring indicated level on scale for Upset Range, allowing work on the other
levelindicators.
c.
Provide sufficient suction head for the Reactor Water Cleanup Pumps.
d.
Assist in natural circulation when RR shutdown to remove decay heat from the fuel QUESTION: 061 (1.00) CHOOSE the method (s) of venting the containment to maintain pressure below the Primary Containment Pressure Limit? a.
Place Continuous Containment Purge in service b.
Vent to the spent fuel pool using the FC retum header.
- c.
Place Containment Atmosphere Monitoring system in service d.
Place Hydrogen Recombiners in service _ - _ - _ _ _ _ _ _
i ! SENIOR REACTOR OPERATOR Page 38 of 58 QUESTION:062 (1.00): ' Given the following conditions: During plant heatup and pressurization reactor pressure is 460 psig and steady - ' all bypass valves are closed. -- reactor water level is.+34" (Narrow Range). - - the feedwater startup controller is in automatic . -- . reactor power is constant.
- Then, the reactor water level control system fails, causing level to INCREASE. ' Assume no . operator action is taken. SELECT the statement below that describes plant response to the - transient. Reactor power WILL... a.
INCREASE due to cold water addition ~ b.' DECREASE due to the shrink from the cold water.
c.
INCREASES as pressure INCREASE due to collapsing voids.
d.
remain constant due to overall power coefficient.
QUES 1lON: 063 (1.00) ' Given the following: The reactoris in power operation.
- WHAT is the potential consequence of prolonged isolation of a control rod (>2 hours)? a.
' Overheating of the associated HCU Accumu'stor Piston Seals reducing the ' ability to insert or SCRAM the control rod b.
Overheating of the associated CRDM Drive Piston Seals reducing the ability to insert or SCRAM the control rod . c.
Erosion of the other'CRDM Drive Piston Seals from excessive Cooling Water - Flow reducing the ability to insert or SCRAM the control rod d.
Erosion of the HCU SCRAM Inlet and Outlet Valve graphitar seats due to l overheating
(- ! l: SENIOR REACTOR OPERATOR Page 39 of 58 I l l QUESTION: 064 (1.00) l Given the following: The reactor is at 30% power during a startup following refueling.
- Reactor pressure is 950 psig - , individual control rod scram time testing surveillance has just started.
! - Then, the Reactor Operator selects the first control rod to be scram time tested. Control rod 28-29 is scrammed, and the test shows the following scram times. The second rod to be checked, control rod 36-09 is scrammed and the test shows following scram times.
Control rod 28-29 Control rod 36-09 0.31 seconds to notch 43 0.29 seconds to notch 43 0.79 seconds to notch 29 0.75 seconds to notch 29 1.44 seconds to notch 13 7.35 seconds to notch 13 WHICH of the following is the required action? a.
Both rods are considered " slow," declare both inoperable and commence a plant shutdown.
b.
Only one rod is considered " slow," one rod is inoperable, a plant shutdown is NOT necessary.
c.
Both rods are considered " slow," neither rod is inoperable, a plant shutdown is NOT necessary.
d.
Only one rod is considered " slow," declare it inoperable, and commence a plant shutdown.
f' L
t i TSENIOR REACTOR OPERATOR. Page 40 of 58 . . QUESTION: 065 (1.00) CHOOSE the reason for securing the RCIC Gland Seal Compressor prior to resotoring the Div 1 Diesel Generator to service when restoring loads following a loss of AC power event, a.
To prevent a compressor shunt trip due to low voltage when the DG is started.
. b.
To ensure sufficient DG field flashing current is available during startup.
c.
To reduce the starting load on the DG which could cause a DG trip on undervoltage.
d.
To prevent the compressor from being load shed after the DG is started by the DG logic.
QUESTION: 066 (1.00) { . A LOCA occurs and all systems function as designed. Five minutes later a loss of off site i power occurs. All automatic actions have taken place.
WHICH of the following components WILL NOT automatically restart? I a.
High Pressure Core Spray Pump l b.
Drywell Chillers c.
Shutdown Service Water Pumps d.
Residual Heat Removal Pumps i
' SENIOR REACTOR OPERATOR Page 41 of 58 QUESTION: 067 (1.00) Smoke has filled the control room requiring a main control room evacuation. WHICH of the following is an immediate Operator Action to be performed prior to leaving the main control room? :- ' a.
Place the Reactor Mode Switch in SHUTDOWN.
' b.
Trip the Reactor Recirculation pumps.
c.
Initiate the Standby Liquid Control system.
d.
Sound the Containment Evacuation Alarm.
i ! QUESTION: 068 (1.00) Given the following: A refueling accident in the Containrnent Building Fuel Transfer Pool resulted in a fuel bundle being dropped and is damaged.
Then, the control room annunciator HI RAD INITIATION SGTS is received.
SELECT the statement that descibes the expected plant response that prevents a High Off-Site Release Rate.
~ a.
VF (Fuel Building) supply and exhaust dampers remain open, the supply and exhaust fans trip on the high radiation signal.
b.
' VF (Fuel Building) supply and exhaust dampers isolate and the supply and exhaust fans trip.
~ c.
VF (Fuel Building) exhaust dampers isolate and the exhaust fans trip when the ')' damprs_ isolate, the supply fans continue to operate.
d.
VF (Fuel Building) supply dampers isolate and the supply fans trip on the high radiation signal, the exhaust fans continue to operate.
l
SENIOR REACTOR OPERATOR Page 42 of 58 - QUESTION:069 (1.00) .4 ; CHOOSE the immediate Operator Action for CPS. No. 4011.02, Spent Fuel Pool Abnormal l - Water Level DECREASE.
- . a.
Suspend any fuellcore component lift operations after placing loads in a safe condition.
-i b.
- Sound the containment evacuation alarm.
c.
Verify all appropriate automatic actions occur and manually perform any that I have NOT.
d.
Start standby gas treatment and establish secondary containment integrity, l ~' QUESTION: 070 (100) . Given the following: - .The plant is at full power one loop of RHR is in service in the Suppression Pool cooling mode for routine -- cooldown.
Then, an SRV has inadvertently openet The other loop of RHR was immediately started in
' Suppression Pool cooling.
WHICH of the following WILL result if the SRV CAN NOT be reclosed? a.
the suppression pool WILL heat up a bit, but when the second loop is maximized the heat-up WILL stop.
, . continued cooldown of the pool wdh both loops in operation, but at a reduced ,. b.
rate due to the open SRV.
. c.
cooldown of the pool. Each heat exchanger is sized to equal the heat rate of , one open SRV.
d; heat-up of the suppression pool as the two heat exchangers CAN NOT equal heat rate of an open SR ! i SENIOR REACTOR OPERATOR Page 43 of 58 L ,. 07/ ' QUESTION:q 00) ' A large LOCA has occurred, the following conditions exists:
suppression pool water level: 50 feet - drywell temperature: 250 deg F - containment pressure: 50 psig - reactor pressure: 50 psig - RPV levelis: 50" wide range j - ' WHICH of the following actions is directed by the EOPs given the above conditions? ' 's.
Stop adding water from outside primary containment.
b.
Dump upper pools, c.
Inject with all ECCS systems.
d.
Start pumps in all available alternate injection systems.
s i f
_ l V . ~ SENIOR REACTOR OPERATOR Page 44 of 58 - QUESTION: 072 (1.00) iGiven thE following: , A plant transient has occurred, and EOP-1 RPV Control and EOP-6 Primary Containment - Control have been entered.
Current conditions are: , Drywell pressure is 6.5 psig - Containment pressure is 3.7 psig and slowly increasing - . Suppression Pool Level is 17 ft.1 in. and steady - Containmer.t Sprays have been initiated but only the A RHR pump is being used - , in Containment Spray. The B RHR pump is needed to maintain RPV level above TAF.
WHICH of the following is the appropriate action? a.
Depressurize at Normal Cooldown Rates b.
Switch the B RHR Pump to Containment Spray c.
Blow Down (and Scram) d.
Dump the Upper Pools
p
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i . - SEN!OR REACTOR OPERATOR Page 45 of 58 . QUESTION: 073 (1.00) . You are in EOP-1 RPV Control and you determine that Blow Down WILL most likely be - required; Following directions in EOP-1, you anticipate Blow Down and open the Main Turbine Bypass Valves to the Main Condenser. As a result of this action, the conditions which were leading you to Blow Down are NOT as severe, and it is obvious that the EOP WILL no longer require Blow Down. Y2u-chould; a.
Enter EOP-3 RPV Depressurization and Blow Down since the determination to anticipate Blow Down was made b.
Open some SRVs to INCREASE the cooldown rate i c.
Recognize that your action mitigated the severity of the condition, and determine j that Blow Down is no longer required i d.
Defer Blow Down until the depressurization into the Main Condenser is complete
, QUESTION:074 (1.00) Given a small break LOCA has occurred which has deprea.surized the RPV to 300 psig.
Select the condition under which the core is being adequately cooled: i-a.
RPV level is at -205 inches, with a RHR pump injecting.
i b.
RPV level is being maintained at -162 inches, with CRD pumps injecting.
I i ! c.
RPV level is -245 inches, and 'one SRV is open.
d.- .RPV level is -170 inches with Low Pressure Core Spray injecting.
[
.
r: . , SENIOR REACTOR OPERATOR Page 46 of 58 , .. , ' QUESTION:075 (1.00) The plant is in an ATWS condition and is operating in EOP-1 A, ATWS RPV Control, Level Leg idirects the operator to hold RPV water level between -162" and +52" using ONLY the following. l-systems: Condensate /Feedwater - L .CRD- - ! RCIC ' - ! -- RHR (through shutdown cooling) - h ' ll Why are these systems specifically designated for use in these conditions? ! s.
These systems all have a reactor grade water source to inject into the reactor.- b.
At this point in the ATWS, reactor pressure precludes use of other systems.
c.
They provide the operator with much more precise level control during the - ATWS.
d.
Their point of injection outside the shroud provides 'some " pre-heating" of the - water.
,. ! ~ L QUESTION:076 (1.00)
- The plant is ope, rating at 93% power. The A CRO then observes:
Reactor Power increasing slightly to 95%. - ROD DF'JFT alarm.
- WHICH of the following desenbes the IMMEDIATE operator response? a.
Reduce reactor power by 50 MWe by reducing reactor recirculation flow, b.
Reduce reactor power to less than 80 % by reducing reactor recirculation flow.. c.: Lockup Rod Control and Information System (RC & IS) by depressing both the white clock frequency " Scan Test" and " Master Test" pushbuttons until the " inoperative" light comes on, d.
Manually SCRAM the reactor.
-
SENIOR REACTOR OPERATOR Page 47 of 58 QUESTION: 077 (1.00) While operating in EOP 6, " Primary Containment Control", the operator is directed to perform an emergency RPV depressurization when plant conditions "CAN NOT be held below the Heat Capacity Limit".
IDENTIFY plant conditions that MUST be evaluated to make the decision to blowdown.
a.
Primary Containment water level and containment pressure b.
RPV pressure and suppression pool temperature c.
Drywell pressure and drywell temperature i d.
RPV pressure and drywell temperature l QUESTION:078 (1.00) Given the following: An unisolable pipe break in the RWCU system in secondary containment has occurred. Then, j the secondary conaninment fails and off site release rate INCREASES to the General Emergency Level (2 x 10E6 micro ct' ries per second).
j 'Nf " , WHICH of the following is the correct action to perform? a.
Isolate all systems except those needed for reactor shutdown, core cooling or fire fighting.
i !- b.
Isolate all systems except those needed for reactor shutdown and core cooling.
Restart the secondary containment ventilation, i c.
Scram the reactor and enter EOP-1.
l d.
Scram the reactor, enter EOP-1, and enter EOP-3 to blowdow. SENIOR REACTOR OPERATOR ~ Page 48 of 58 QUESTION: 079 (1.00) The Standby Liquid Control System has been initiated in accordance with CPS No. 4411.10, EOP SLC OPERATIONS. While verifying automatic actions, the operator finds the Reactor Water Cleanup System isolation Valves have failed to close.
From the following, CHOOSE how continued operation of the Reactor Water Cleanup System. _would affect the ability of the SLC System to shut down the reactor.
a.
Shut down ability would DECREASE because the boron is removed from the . reactor coolant by the RWCU filter domineralizers, b.
More boron would be required because the RWCU piping and components MUST also be filled, reducing the amount in the vessel.
c.
Shut down ability would DECREASE due to lowering reactor water level via the RWCU blowdown line to the Main Condenser.
d.
More water volume because of RWCU still in service would INCREASE the time for SLC to achieve reactor shutdown.
l ' QUESTION: 080 (1.00) l-Scram signals are present on Division I and Division 11 of RPS. The plant is stable with the turbine on line at 50% load. The C Area operator reports that the scram air header has NOT
- depressurized, and the scram pilot solenoids on the HCU's checked all feel warm.
< t WHICH of the following would depressurize the scram air header? a.
De-energize the backup scram valve pilot solenoids by opening their breakers.
l; . b.
Reset the scram to energize the scram valve pilot solenoids and allow the SDV to drain.
- . De-energize the ARI solenoids by opening their breakers.
c.
d.
De-energize the scram pilot solenoids by opening their breakers.
i
n.
' / I SENIOR REACTOR OPERATOR Page 49 of 58 , , f l-l - QUEbTION: 081 (1.00) l . Given the following:, . A large LOCA has occurred - the reactor has successfully scrammed - EOP-6 has been entered -- ' all RHR pumps are maintaining RPV water level.
- Currently conditions are: - containment temperature has INCREASED to 190 degrees F - containment pressure has INCREASED to 1.9 psig.
- WHICH of the following is the required action? a.
Vent the Containment in accordance with 4411.06.
b.
Enter EOP-1, then blowdown in accordance with EOP-3.
c.
Gtart Containment Sprays.
i . d.
Dump upper pool.
l i l QUESTION: 082 (1.00) l When starting the A RHR pump from the Remote Shutdown Panel for Suppression Pool , Cooling, what is the response of the pump MINIMUM Flow Valve? '- , a.
The operator MUST open the valve but it WILL auto close when flow is >1100 lf gpm.
L
- b.
- All automatic functions of the MINIMUM flow valve are bypassed in this
' condition.
. . c.
Valve WILL open when the pump is started but requires operator action to close.
L d Valve operates as it does when starting the pump from the Main Control Room.
i F
SENIOR REACTOR OPERATOR Page 50 of 58 QUESTION:083 (1.00) WHAT is the MAXIMUM flow rate, listed below, that can be obtained without exceeding the RCIC NPSHNortex Limit (Detail Z)? a.
600 gpm . b.
610 gpm c.
625 gpm d.
750 gpm i QUESTION: 084 (1.00) The Plant is operating at 90% Rated Power when suddenly RR Pump "B" trips to off. From the - following choose the appropriate operator response.
a.
Scram the Reactor if the Restricted Zone is entered.
] b.
Shut the associated 1833-F0238 "B" Loop Suction Valve.
c.
If the Restricted Region is entered, wait until core instabilities are observed, then Scram the Reactor.
d.
Open the FCV on the "A" RR pump as required to keep flow from causing a entry into the Restricted Zone.
&
- SENIOR REACTOR OPGRATOR. Page 51 of 58 . QUESTION: 085 (1.00) . WHICH of the following would be the most likely consequence of failing to re-open the discharge valve of a tripped Reactor Recirculation (RR) Pump 5 minutes after shutting it? a.
Erroneous core flow data would be indicated on the core flow instruments due to the instruments automatically subtracting the reverse core data based on the position of the RR Pump breaker.
b.
Stratification and cooldown of the RR loop would occur due to failing to re-open the discharge valve.
c.
The tripped RR Pump would NOT receive any recirculation flow into it's seal area and the pump seal could ultimately fail.
d.
The differential pressure resulting from the loop cooldowa could hydraulically lock the valve closed.
I QUESTION: 086 (1.00) j Which of the following is an acceptable method to exit the Controlled Entry Region that is : i indicated on the " Stability Control & Power / Flow Operating Map? I a.
Promptly exiting the region by using the Cram Array is an acceptable method.- b.
Reducing core flow via the use of throttling closed the Reactor Recirculation flow Control Valves to reduce core flow is an acceptable method.
t c.
Inserting control rods in reverse sequence is an acceptable method.
d.
Closing extrachon steam valves to DECREASE the final Feedwater temperature is an acceptable method, t , ! l
- 1-- . , , . 1 SENIOR REACTOR OPERATOR Page 52 of 58 QOESTION: 087- (1.00) WHICH of the following would require an emergency depressurization? < - - a.
HPCS pump room temperature and Area Radiation Limits above Max Safe j . b.
Fuel Bldg Gen Area El. 712' and Fuel Bldg Pipe Valve Room above Max Safe Area Radiation Limits , c.
LPCS area above' Max Safe Water Level and Max Safe Area Radiation Limits . d.
. Aux Bldg Steam Tunnel above Max Normal Temperature and Aux Bldg Gas ' Cont Bourdary (EL 762' West) sbove Max Safe Area Temperature Limits , QUESTION: 088 (1.00) Given the following: The plant is in a transient condition. The following indications are noted: The Mode Switch is in Shutdown.
- 18 rods arc between notch 04 and 36 ' - - Reactor power is approximately 40%. l - ' The MSIVs are open; the main turbine is on line.
- L Reactor pressure is being maintained at 920 psig.
- (- Reactor waterlevelis 46 inches.
- Which of the following WILL initially cause reactor level to DECREASE? a.
Emergency depressurization.
" b.
Boron injection.
c.
Turbine Trip.
- d.
Recirc Pump Trip.
) l
' SENIOR REACTOR OPERATOR Page 53 of 58 QUESTION: 08g (1.00)' Given the following: , l - TURB TRIP EHC SYS has alarmed.
WHICH of the following describes the response of the MSR 1 A/1B DRN VLVs (1TD-MSR(1-4))? a.
Do not Auto close, but can be opened by the control switches.
b.
Auto open and are interlocked to prevent closing by the control switches.
c.
Auto close and are interlocked to prevent opening by the control switches.
?- . ' .d.
Do not Auto oper., but can be closed by the control switches.
QUESTION: 090 (1.00)
- Given the following:
' The plant is operating at 100% power - feedwater level control is in automistic.
- . Then, the B Steam flow signal to the Reactor Water Level Control System fails upscale.
. . Assume NO operator action. WHAT is the effect on the ACTUAL Reactor Vessel water level? Reactor water level WILL... (
- a.
continuously DECREASE until low level scram setpoint is reached.
b.- INCREASE until terminated by a high level feed pump trip.
, c.
DECREASE to a lower level and stabilize.
d.
- INCREASE to a higher level and stabilize.
<
i e SENIOR REACTOR OPERATOR Page 54 of 58 QUESTION: 091 (1.00) WHICH of the following requires entry into EOP-6, PRIMARY. CONTAINMENT CONTROL?. a.
Suppression PoolTemperature i rees F.
blecosrecfnaswy q- ,r-~~ 19 origeW3 ca4ueJ. o.n . b.
Suppression Pool Levelis 9.3 eet.
_;59 $F s clt a q u} W ( O e d. . f Drywell Temperature ishdegrees F.
- g,f gg
,y, c.
d.
Drywell Hydrogen Concentration is QUESTION: 092 (1.00) WHAT isolations WILL 4410.00 C006 (DRYWELL COOLING ISOLATIONS) defeat? a.
Level 1, Level 2, and high drywell pressure b.
Level 1, Level 3 and high drywell pressure c.
Level 1, Level 2, and Containment Exhaust High Radiation d.
Level 1, Level 3, and Containment Exhaust High Radiation . l.
.
SENIOR REACTOR OPERATOR Page 55 of 58 ' QUESTION: 093 (1.00) While operating at 75 percent reactor power, the following alarms are received: CCW EXPANSION TANK LVL Hl/LO CCW PMP AUTO TRIP in addition to the CCW System, WHICH of the following should be monitored / tracked during this - event? a.
Reactor Recirculation Pumps b.
Fuel Building HVAC cooling coils l I c.
Drywell Chillers d.
Auxiliary Building HVAC cooling coils ) QUESTION: 094 (1.00) . Given the following: The Containment Compressed Gas Supply isolation Valves, ilA012B and 11A013B have i automatically isolated.
- WHICH of the following signals can cause this condition? a.
Lcw IA/SA header pressure of 70 psig.
b.
High Drywell pressure and/or Level 1.
i c.
An ADS automatic actuation signal.
d.' High Drywell pressure and/or Level 2.
! l
-
' , SENIOR REACTOR OPERATOR Page 56 of 58 2 QUESTION: 095 (1.00)' Given the following: A Main Control Room evacuation has occured - plant control has been established at the Remote Shutdown Panel.
- WHICH Safety Relief Valve (SRV) has the MOST air available for operation following a loss of instrument air. (Assume no operator actions have been taken for the loss of instrument air.) ' a.
F051G j i - b.
F051D c.
F051C d.
F051B QUESTION: 96 (1.00) Given the following conditions: The plant has just experienced a complete loss of Shutdown Cooling - Temperature readings indicate a 1 degree F INCREASE in bulk water - temperature every 10 minutes Assume the reactor vessel head is installed - ' No other parameters change - Current temperature is 158 degrees F.
- Assuming no operator actions, how long before's mode change occurs? -1 a.
540 minutes b. - 420 minutes c.
360 minutes d.
200 minutes 'I ' ,.
_
I SENIOR REACTOR OPERATOR Page 57 of 58 . QUESTION: 097.(1.00) = Given the following: l Reactor pressure is 855 psig 'i - control rod 22-11 is at position 00 with the it's nitrogen accumulator isolated for l - ' repair.
Then, the operating Control Rod Drive (CRD) pump trips and the ACCUM TROUBLE . annunicator is received for the rods 18-27 (at position 00) and 38-23 (at position 48). WHAT action required? a.
Immediately place the reactor mode switch to " SHUTDOWN".
b.
If any other accumulator becomes inoperable for a withdrawn rod, immediately - place the reactor mode switch in " SHUTDOWN."
c.
If any other accumulator becomes inoperable for a wdhdrawn rod, start a CRD q pump within 20 minutes or place the reactor mode switch in " SHUTDOWN."
' d.
Start a CRD pump within 20 minutes or place the reactor mode switch in " SHUTDOWN."
QUESTION:098 (1.00) Given the following: ATWS actions are in progress.
-- drywell temperature has INCREASED to the unsafe region of Detail B, RPV - Saturation Temperature.
A loss of all level instruments has been declared - RPV flooding has been entered.
- WHY are the RPV Water Level Instruments taken to be unusable? The combination of... l a.
drywell pressure and temperature makes the reference legs unreliable ' b.' drywell pressure and RPV temperature makes the reference legs unreliable c.
drywell pressure and temperature makes the variable legs unreliable.
- d.
drywell temperature and RPV pressure make the reference legs unreliable , f )
i SENIOR REACTOR OPERATOR Page 58 of 58 , . QUESTION: 099 (1.00) The initiating signals for a valid suppression pool dump actuation have occurred and are still present.
~ WHAT actions are necessary to close the Division i Suppression Pool Dump Valves (1SM001 A and 1SM002A) and have them remain closed.
The Division i Suppression Pool Dump Valves may be closed... a.
after the " Suppression Pool Dump Valve Mnde Selector Div I" is placed in " Disable", b.
after the 25 minute delay timer from the initiation signal has timed out.
, c.
after the "SM System Div I in Test" switch is placed in the " Test" position.
~ d.
after the "LCPS/LPCI FM RHR A Seal in Reset" pushbutton on P601 is depressed.
QUESTION:100 (1.00) A valid Fuel Building Exhaust High Radiation signal comes in. WHAT is the response of the < Standby Gas Treatment System (VG).
a.
BOTH trains of VG start. The following ventilation systems are running: VRNQ, VF, CCP b.
BOTH trains of VG start. VF is secured. The following ventilation systems are running: VRNQ, CCP.
c.
ONE train of VG starts. The following ventilation systems are secured: VRNQ, VF, CCP.
d. ' ONE train of VG starts. VF is secured. The following ventilation systems are ' running: VRNQ, CCP. Only i (..... END OF EXAMINATION """**") ' .
SENIOR REACTOR OPERATOR Page 1 of 11 REFERENCE MATERIAL . ANSWER: 006 (1.00)' ANSWER: - 001 (1.00) a.
c.
REFERENCE: - REFERENCE: CPS 1406.01, AnnunciatorTracking 10CFR55.53, Conditions of licenses.
Program, Rev.10, page 4, LP 87460, CPS No.1401.01, Objective 1.4 - Section 8.1.3 & 8.1.4.b 2.1.2 ..(KA's) LP87468, Objective 1.5 2.1.11 ..(KA's) ANSWER: 002 (1.00) b.
ANSWER: 007 (1.00) REFERENCE: b Technical Specifications 5.2.2.c REFERENCE: LP87592. Objective 1.4.4 CPS No.1005.01, CPS PROCEDURES 2.1.4 - ..(KA's) AND DOCUMENTS, Rev 38, page 7.- LP85132, Objective.1.1.1 ANSWER: 003 (1.00) 2.2.12 ..(KA's) c-REFERENCE: ' CPS No. 4005.01, Rev.13, Loss Of ANSWER: 008 (1.00) Feedwater Heating note and c.
Immediate actions on page 2: REFERENCE: LP87505, Objective.1.2 CPS 1014.01 SAFETY TAGGING, Rev. 22, 2.1.20- ..(KA's) page 14, NOTE above etep 8.3.1.4 ' LP85140, Objective.1.2.8 . 2.2.13 ..(KA's) - ANSWER:. 004 (1.00) c REFERENCE: ANSWER: 009 (1.00) CPS No.1052.01, CONDUCT OF b . SYSTEM LINEUPS, Rev 7, page 14, REFERENCE: - Section 8.5.2.1.2/.3 LP85147.1 LP85110,l Objective.1.1 CPS 1014.05, Preparation of Post-2.1.29 ..(KA's) Maintenance Testing, Section 3.3.
2.2.19 ..(KA's)
- ANSWER: 005 (1.00)' d.
ANCWER: 010 (1.00) REFERENCE: c.
- w , . LP87633 ~.1.3 REFERENCE: 10CFR55.53, Conditions of licenses.
Technical Specifications B3.9.1, Refueling ORM 6.8.3 Equipment interlocks 2.2.6 ..(KA's) LP85234, Objective ~.1 j 2.2.25 ..(KA's) ! l
. ' SENIOR REACTOR OPERATOR Page 2 of 11 ANSWER: 011 (1.00) ANSWER: 016 (1.00) c.
d.
REFERENCE: - REFERENCE: Technical Specifications SR 3.9.2.2: EPIP, EC-01, CPS No. Emergency Perform CHANNEL FUNCTIONAL Response Organization and Staffing, form TEST,7 days 28, Title Emergency Manager, item 3, LP87630, Objectvie.1.2.2 Receive command authority.
2.2.24 ..(KA's) LP87536, Objective.1.4.2 2.4.37 ..(KA's) ' ANSWER: 012 (1.00) a ANSWER: 017 (1.00) REFERENCE: d LP88602, Objective 1.13 . REFERENCE: 10CFR20 Appendix B, table 1, column 3 for - EPIP FE-02, OSC Operations, page 5,- DAC.
Section 4.6 Relocation of the OSC 2.3.1 ..(KA's) . LP11016 Terminal Objective 2, Enabling l Objective 1 ' 2.4.42 ..(KA's) ANSWER: 013 (1.00) d L REFERENCE: ANSWER: 018 (1.00)
- EPIP RA-03, Radiological Exposure b
Guidelines, Section 4.1.2 REFERENCE: l . LP 11016 . LP 85201, Control Rod Drive Hydraulics l LP85234, Objective.1.1.
LP85201, Objective.1.1.7 2.3.4 ..(KA's) 201001K408 ..(KA's) ,
- ANSWER: 014 (1.00)
ANSWER: 019 (1.00) c a . REFERENCE: REFERENCE: LP85110, Objective.1 LP87629, Objective.1.2.1 CPS No.1052.01, Section 6.1 CPS No. 9082.01, 2.3.10' ..(KA's) 262001K202 ..(KA's) LANSWER: 015 (1.00) c . ANSWER: 020 (1.00) REFERENCE: d.. ~ LP88602, Objective.1.18 REFERENCE: CPS No.1893.00 CPS OPERATIONAL LP 85401.1, Rod Control and Information FIRE PROTECTION PROGRAM, Rev. 6, System
- Section 3.0 201005
..(KA's) 2.4.25.
..(KA's) -
. SENIOR REACTOR OPERATOR Page 3 of 11 ANSWER: 021 (1.00) ANSWER: ~ 026 (1.00) c.
b REFERENCE: REFERENCE: CPS No. 3312.01, Residual Heat Removal, LP 85209, Objective 1.7 Rev. 28, pg 23, note after step 8.1.11.8 209001K408 ..(KA's) LP87205 .1.7.1 226001K612 ..(KA's) ANSWER: 027 (1.00) c - ANSWER: 022 (1.00) REFERENCE: d LP85380, Objective.1.2.7 REFERENCE: 209002A301 ..(KA's) . See CPS No. 3314.02, Section 8.1.3 and . I Figure i to find tank level less than low level alarm.
ANSWER: 028 (1.00) LP87201, Objective.1.6.12 c.
211000A208 '..(KA's) REFERENCE: LP 85211, Objective.1.4.2 211000A308 ..(KA's) ANSWER: 023 (1.00) .d REFERENCE: ANSWER: 029 (1.00) LP85202, Objective.1.18.2 c 202002K103 ..(KA's) REFERENCE: LP87411, Objective.1.4.3 ORM:.66(w-8)+48% RTP Table 1 Notes ANSWER: 024 (1.00) Attachment 2-3 page 115 d 212000K307 ..(KA's) REFERENCE: CPS No. 4008.01 Abnormal Reactor Coolant Flow, Rev.15, page 5.
ANSWER: 030 (1.00) i LP 85402, Reactor Recirculation Flow d Control REFERENCE: 202002A207 ..(KA's) LP85482, Objective 1.4.2 215003K102 ..(KA's) ANSWER: 025 (1.00) c ANSWER: 031 (1.00) REFERENCE: d EOP-4 REFERENCE: . LP87556, Objective.1.2.3 LP 85215 SRMs, Objective.1.1.10 203000K502 ..(KA's) 215004K101 ..(KA's) . k
l
- i SENIOR REACTOR OPERATOR Page 4 of 11 ANSWER: 032 (1.00) ANSWER: 037 (1.00) - b b REFERENCE: REFERENCE LP 87411, Objective.1.1.2 LP87218, Objective.1.2.1 &.2 215005A404 ..(KA's) 218000K501 ANSWER: 038 (1.00) ANSWER: 033 (1.00) d c REFERENCE: REFERENCE: LP85205, Objective 1.4.6 - LP 87411.1.1.2 Annunciator procedures CPS No. 3312.01, Residual Heat Rem. oval, 5006-2H and 5005-3B Rev. 28, pg 17, Note at .215005A104 ..(KA's) top of page - 223001A304 ..(KA's) ANSWER: 034 (1.00) d ANSWER: 039 (1.00)- REFERENCE: a { LP 85423, Objective.1.6.5 REFERENCE: 216000K501 ..(KA's) LP 85205, RHR, Obj 1.4.5 223002K108 ..(KA's) ANSWER: 035 (1.00) a ANSWER: 040 (1.00) REFERENCE: a LP 87241, section 1.4.1 and LP 87248 REFERENCE: 241000K306 ..(KA's) LP 85433, index item 1.2.2 239002K405-..(KA's) ANSWER: 036 (1.00) d ANSWER: 041 (1.00) . REFERENCE: a 5063-2D: RCIC PUMP SUCTION REFERENCE: PRESSURE LOW: AUTO ACTIONS:RCIC LP87630.1.1 Turbine trip 234000K403 .'.(KA's) LP 85217.1.5.6, Reactor Core Isolation Cooling - CPS No. 3310.01, Rev.15, pg 30, Appendix ANSWER: 042 (1.00) ] 'B c i 217000A301 ..(KA's) REFERENCE-
CPS No. 4201.01 LOSS OF DC POWER, Rev. 3, pg 2, Step 3.1 and Site Safety Standards j LP 85234, Objective.1 263000K303 ..(KA's) i i
SENIOR REACTOR OPERATOR Page 5 of 11 ANSWER: 043 (1.00) ANSWER: 049 (1.00) 'b c REFERENCE: REFERENCE: LP85202, Objective.1.6.1 85205,.1.11 - RHR CPS No. 3302.01, Rev. 21, pg. 40, step 230000K606 ..(KA's) 8.2.4.2 202001A218 ..(KA's) ANSWER: 050 (1.00) d ANSWER: 044 (1.00) REFERENCE: ] d LP8701, Objective.1.1.5 ) REFERENCE: CPS No. 3304.02, ROD CONTROL AND LP85204, Objective.1.4.1 INFORMATION SYSTEM (RC&lS), Sections 5000-1C, Annunciator Response procedure 8.1.7.1 & 8.1.7.2.
, ~ , 204000A213...(KA's) LER, Scram April 9,1996 214000K105 ..(KA's) ANSWER: 045 (1.00) d ANSWER: 051 (1.00) REFERENCE: d LP85433, Objective.1.4 REFERENCE: CPS No. 4003.01, Remote Shutdown, Rev.
. LP87570, Objective 1.8.2 11, pg 37, Caution to begin Section G.4 259001A207 ..(KA's) 205000K403 ..(KA's) ANSWER: 052 (1.00) ANSWER:. 046 (1.00) a a . _ REFERENCE: REFERENCE: LP86576, Objective.1.8 j 87205,.1.7.1 (page 36) CPS No. 3500.01,4.6: ,219000K101 ..(KA's) 262002K401 ..(KA's) ANSWER: 047 (1.00) ANSWER: 053 (1.00) c d REFERENCE: REFERENCE: LP86201, CRD Hydraulics, Section 1.8.9 LP85261, Objective.1.2.3 .201001K408 ..(KA's) CPS No.1893.04, M310,719' Control HVAC Equipment Area, Rev. 4, ANSWER: 048 (1.00)- Pre-fire plan d 286000A304 ..(KA's) i REFERENCE: LP 85447, Objective.1.4.1 - Control Room HVAC System, CPS No.
5052-7M. 290003K101 ..(KA's) ,
SENIOR REACTOR OPERATOR' Page 6 of 11 ANSWER: 054 (1.00) . ANSWER: 060 (1.00) c d REFERENCE: REFERENCE:
' LP85205, Objective.1.2.5 CPS No. 3303.01, REACTOR WATER CPS No. 3312.01 RESIDUAL HEAT CLEANUP, Rev. 20, pg 8, Step 6.7 REMOVAL LP87442, Objective.1.4 268000K502 ..(KA's) 295009K105 ..(KA's) ) ANSWER: 055 (1.00) ANSWER: 061 (1.00) c b REFERENCE: REFERENCE:- LP87448, Section 10.1.1.6 4411.06 Emergency Containment Venting, 290002K403 ..(KA's) Purging, And Vacuum Relief Section 2.5: Vent to Spent fuel Pool Using FC Return Header ANSWER: 056 (1.00) 4402.01, EOP-6, 87558 b LP87558, Objective.1.2.3 REFERENCE: 295010A101 ..(KA's) LP87621, Objective.1.1 Technical Specifications 2.1.1.2 I 290002K507_..(KA's) ANSWER: 062 (1.00) a REFERENCE: ANSWER: 57 (t ')0) LP 87498,.1.9.2 Transient and Event a worksheet REFERENCE: 295014K206 ..(KA's) LP87400, Section 1.4. 2 201004K404 ..(KA's) ANSWER: 063 (1.00) b ANSWER: 058 (1.00) REFERENCE: b CPS No. 3304.01, Control Rod Drive REFERENCE: Hydraulics, Rev. 24, pg 36, . LP87570,.1.7.1 ' Note to step 8.2.3.5 i-295006A102 ..(KA's) LP86201, Objective.1.8.4 295005K201 ..(KA's) ANSWER: 059 (1.00) d ANSWER: 064 (1.00) REFERENCE: b.
~ LP85245, Main Turbine / Generator, REFERENCE: II.A.1.1.3 LP87622, Objective.1.2.4 295007A105-..(KA's) Technical Specification Table 3.1.4-1 and note 2 to the Table 295006A107 ..(KA's) !
> SENIOR REACTOR OPERATOR Page 7 of ii - ANSWER: - 065 (1.00) ANSWER: 070 (1.00) b d REFERENCE: REFERENCE: LP 87513, Sections 1.7 and 1.9 LP87509, Suppression Pool, Step 1.7 295003K106 ..(KA's) LP85234, Objective.1 295013K201 ..(KA's) ANSWER: 066 (1.00)- b ANSWER: 071.(1.00) . . REFERENCE: a .' CPS No. 4001.02, Automatic isolation, Rev.
REFERENCE: 12, Group 11 checklist EOP-6 I . LP87513, Objective.1.4.1 LP87526, Objective.1
- 295003K204'..(KA's)
295024K101 ..(KA's) ANSWER:. 067 (1.00) ANSWER: 072 (1.00) d c REFERENCE: REFERENCE: LP87503, Objective. s.1 Pressure Control Leg of.EOP-6
4003.01, Remote Shutdown, Rev.11, Step LP87558, Objective.1.8.6 ' ' 3.1.3, page 3.
295024A209 ..(KA's) 295016K201 ..(KA's) . ANSWER: 073 (1.00)- . ANSWER: LO68 (1.00) c-b REFERENCE: REFERENCE: LP87552, Objective.1.9.5 LP. 85449, Section 1.3.4 RPV Control (EOP-1) 295017K204 ..(KA's) 295025A201 ..(KA's) ANSWER: 074 (1.00) . ANSWER: 069 (1.00)- b a: REFERENCE: REFERENCE: . LP87552, Objective.1.8.5.2 - CPS No. 4011.02, Spent Fuel Pool RPV Control (EOP-1) Abnormal Water Level Decrease, Rev. 3, 295031A204 ..(KA's) pg 2, step 3.1 , LP87297, Objective.1.2 - - 295023K101 ..(KA's) ANSWER: 075-(1.00) .d.
REFERENCE: AT'NS RPV Control (EOP 1A) LP87553, Objective '.1.3.2 295037A202 ..(KA's) -
l 1 SENIOR REACTOR OPERATOR . Page 8 of 11' l ' ANSWER: 076 (1.00).
. ANSWERi 081.(1.00) .h a- .. _ b REFERENCE: REFERENCE: < CPS 4007.02, inadvertent Rod Movement, EOP-6, IF/THEN Statement under Rev. 9, step 3.3 containment temperature LP87506, Objective.1.2 LP 87558, Objective.1
- 295014A105_..(KA's)
'295027K103 ..(KA's)
- ANSWER: 077 (1.00)
ANSWER: ' 082 (1.00) b b ' < . REFERENCE: REFERENCE: -! - LP87558-CPS No. 4003.01, Remote Shutdown, Rev.
LP87558, Objective.1, 11, pg 34, Appendix F, step F.4.16 & 17 '295026K301 ..(KA's) LP85433, Objective 1.2.1 '295018K303 -..(KA's) ANSWER: 078 (1.00) 'd ! REFERENCE: ANSWER: 083 (1.00) EOP-9, Radioactivity Release Control, c directs the scram of the reactor, entry into REFERENCE: p ' EOP 1 and EOP 3 to for the purpose of ' LP85217, Objective.1.5, RCIC blowing down.
295030K102 -..(KA's) LP87560, Objective.1.7 295038K304 ..(KA's) - ANSWER: 084 (1.00)
a l.
- ANSWER: : 079 (1.00)
REFERENCE: j a CPS No. 4008.01, Abnormal Reactor [ REFERENCE:
- Coolant Flow, Rev.15, pg 3, step 3.1 l-
_ LP85211, Summary Section LP87508, Objective.1.2 LP85211, Objective.1.5 295001K102 ..(KA's) 295037K103'..(KA's) ANSWER: 080 (1.00) ANSWER: 085 (1.00) dL b REFERENCE: REFERENCE: j . LP 87212, Objective.1.8 CPS No. 4008.01, Abnormal Reactor 295037K201 ..(KA's) Coolant Flow, Rev.15, pg 5, step 6.1 LP87508, Objective.1.3.2 , 295001A101 ..(KA's) r t NIOR REACTOR OPERATOR Page 9 of 11 ANSWER: 086 (1.00) ANSWER: 092 (1.00) c a = REFERENCE: REFERENCE: CPS No. 4008.01, Abnormal Reactor LP87558, Objective.1.10
Coolant Flow, Rev.15, pg 6, step 6.3.d Primary Containment Control (EOP-6) LP87508, Objective.1.6.2 295012A101 ..(KA's) 295001A201 ..(KA's) - ANSWER: 093 (1.00) ANSWER: 087 (1.00) a b REFERENCE: REFERENCE: LP 85208, Objective.1.9
LP87559, Objective.1.1 5040.02, Annunciator Response Procedure, EOP-8, Secondary Containment Control Rev. 25, step 2.d and 3203.01, ' 295033K301 ..(KA's) 295018A102 ..(KA's) e ' ANSWER: 088 (1.00) ANSWER: 094 (1.00) e d REFERENCE: REFERENCE: LP874981.9.7.1.10.1.11, Step 4 CPS No. 4001.02, Automatic isolation, Rev.
295005K103 ..(KA's) 13, checklist C001 group 8 _ LP85239, Objective.1.6.9 295019K303 ..(KA's) ANSWER: 089 (1.00) b REFERENCE: ANSWER: 095 (1.00) 5007-1B TURB TRIP EHC SYS, Rev. 24, a.
pg 3, Auto Action Step 1 REFERENCE: LP85245, Objective.1.5.2 LP85239, Section 1.1.15 ' 295005A203 ..(KA's) 295019K214 ..(KA's) _ ANSWER: 090 (1.00) ANSWER: 096 (1.00) d b REFERENCE: REFERENCE: LP87570, Feedwater Control, Section Technical Specifications pg 1.0-7, Table IV.1.8.2 1.1-1 295008K203 ..(KA's) 295021K201 ..(KA's) - ANSWER: 091 (1.00) ANSWER: 097 (1.00) c d REFERENCE: REFERENCE: LP87558, Section 1.1 LP85201,.1.8.17 295012K101 ..(KA's) Technical Specifications 3.1.5.b.
295022K207 ..(KA's) ,
SENIOR REACTOR OPERATOR Page 10 of 11 ! ANSWER: 098 (1.00) d l REFERENCE: l LP87553,.1.1, EOP-1A, ATWS RPV Control, Detail B.
295028K303 ..(KA's) ANSWER: 099 (1.00) a REFERENCE: J 87408 page 8 { ' LP87408, Objective.1 j 295029A201 ..(KA's) ANSWER: 100 (1.00) b REFERENCE: LP85261 STANDBY GAS TREATMENT SYSTEM, item 1.4.1 295033K204 ..(KA's) (.........* END OF EXAMINATION "**""") l-
l SENIOR REACTOR OPERATOR Page 11 of 11 ANSWER KEY MULTIPLE CHOICE 001 c 021 c 041 a 061 b 081 b 002 b 022 d 042 c 062 a 082 b 003 c 023 d 043 b 063 b 083 c 004 c 024 d 044 d 064 b 084 a 005 d 025 c 045 d 065 b 085 b 006 a 026 b 046 a 066 b 086 c 007 b 027 c 047 c 067 d 087 b 008 c 028 c 048 d 068 b 088 c 009 b 029 c 049 c 069 a 089 b 010 c 030 d 050 d 070 d 090 d 011 c 031 d 051 d 071 a 091 c . 012 a 032 b 052 a 072 c 092 a J 013 d 033 c 053 d 073 c 093 a l 014 c 034 d 054 c 074 b 094 d ' 015 c 035 a 055 c 075 d 095 a 016 d 036 d 056 b 076 a 096 b 017 d 037 b 057 a 077 b 097 d 013 b 038 d 058 b 078 d 098 d 019 a 039 a 059 d 079 a 099 a , 020 d 040 a 060 d 080 d 100 b l ("*"*"" END OF EXAMINATION "*"""*) }}