IR 05000416/1993301
| ML20059E644 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/21/1993 |
| From: | Bartley J, Lawyer L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20059E628 | List: |
| References | |
| 50-416-93-301, NUDOCS 9311030286 | |
| Download: ML20059E644 (210) | |
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N64g URIVED STATES
[(gtA" Q n- o,, . NUCLEAR REGULAVORY COMR41ss!Ofd REGION 11
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'* $ 101 MARIETTA STHECT, N AT LANTA, GEORGI A 30323
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Report No.: 50-416/93-301 Licensee: Entergy Operations, In P. O. Box 756 Port Gibson, MS 39150 Docket No.: 50-416 License No.: NPF-29 Facility Name: Grand Gulf Inspection Conducted: September 27 - October 1, 1993 Chief Examiner: d,YhI[Ms /0/20/4 3
@nathanH.Bartley Dat'e Signed Accompanying Personnel: G. Buckley, PNL M. Morgan, PNL R. Orton, PNL Approved by:
Lawrence L. Lawyer, Chief i h M Date Signed 3-Operator Licensing Section Operation Branch Division of Reactor Safety SUMMARY l
Scope:
NRC examiners conducted regular, announced operator licensing initial examinations and associated inspection activities during the period .
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September 27-October 1, 1993. Examiners administered examinations under the guidelines of the Examiner Standards (ES), NUREG-1021, Revision 7. Six Senior Reactor Operator (SRO) and four Reactor Operator (RO) candidates received written and operating examination l
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Results: i
Candidate Pass / Fail: j SR0 R0 Total Percent Pass 6 4 10 100 Fail 0 0 0 0%
i Examiners identified a violation with regards to failing to follow procedure for entry into the Radiologically Controlled Area (paragraph 2.d.(2)). j VIO 50-416/93-301-01 j Examiners identified a non-cited violation with regards to an inadequate i procedure (paragraph 2.f.). VIO 50-416/93-301-02 Examiners identified a non-cited violation with regards to failing to control operator aids (paragraph 2.c.(1)). VIO 50-416/93-301-03 ;
l Examiners identified an inspector follow-up item regarding personnel in containment being unable to hear the public address system (paragraph 2.c.(2)). IFI 50-416/93-301-04 Examiners identified a strength regarding simulator support (paragraph 2.c.(3)).
Examiners identified a weakness regarding simulator capability '
(paragraph 2.e.).
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REPORT DETAILS Persons Contacted Licensee Employees
- B. Avara, Operations Training Coordinator
- C. Bell, Simulator Supervisor
- B. Bryant, Operation, Training Supervisor
- C, Cresap, Senior Operations Instructor
- H. Dietrich, Manager, Nuclear Training
- C. Dugger, Manager of Operations
- C. Hayes, Director, Quality
- C, Hicks, Superintendent, Operations
- D. Pace, General Manager
- C. Roberts, Senior Operations Instructor
- R. Ruffin, Licensing Specialist
- W. Shelley, Superintendent, Operations Training Other licensee employees contacted included instructors, engineers, technicians, operators, and office personne NRC Personnel
- R. Bernhard, Senior Resident Inspector C. Hughey, Resident Inspector M. Meeks, Resident Inspector
- Attended exit interview Discussion Results Six SR0 and four R0 candidates passed the initial examinatio No candidates failed the examinatio Examination Development Representatives from the licensee training staff reviewed the written examination in the regional office on September 8-9, 1993. The review was thorough and provided valuable input to improve the accuracy of the information contained in the written examinatio The examination team conducted a preparation visit for the examination on September 13-17, 1993. The examiners utilized this time to walk-down the selected Job Performance Measures (JPM) and simulator scenarios and to resolve the remaining comments on the written examinatio '
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Report Details 2 , Examination Administration The NRC administered the examinations without major problems. The examination team identified an uncontrolled operator aid in the control room and a problem with the public address system in
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containment. A simulator hardware failure required two simulator scenarios to be rescheduled for later in 'he examination wee '
(1) The examiners identified an uncontrolled operator aid in the-control room during the conduct of walkthrough examination Licensee personnel had taped a copy of System Operating Instruction (S0I) 04-1-01-P75-1, " Standby Diesel Generator System," Revision 38, step 3.29 to a moveable podium used by the R0s. This step gave appropriate KVAR settings based on the megawatt loading for the Emergency Diesel Generators . The operator aid was correct but it was not logged into the Operator Aid Log as required by Operations Section Procedure N , " Control and Use of Operations Section Directives,"
Revision 20. Section 6.13.2 required operator aids to be controlled by logging the operator aid into the Operator Aid Log before posting. The NRC considered this as Violation (VIO)
50-416/93-301-03 " Failure to follow procedure which stated the requirements for controlling operator aids." However, based upon the corrective actions, no similar violations identified by the team in the past two years and the lack of willfulness, this NRC identified violation is not being cited because criteria specified in Section VII.B of the NRC Enforcement Policy were satisfie (2) The examination team identified a problem with the facility's public address system in containment while conducting walkthrough ;
examinations on September 28, 1993. The plant was scrammed while an examiner conducted a JPM in containment near the Control Rod Drive system hydraulic control units. The examiner and candidate were unable to hear the announcement telling plant personnel that the plant was going to be scrammed. The examiner and candidate talked to several technicians and maintenance personnel while they were hastily exiting containment and determined they also did not hear any announcements concerning the reactor scram. The examiners talked to control room personnel and determined that an i announcement was made by an operator in the control room prior to ,
the scram. The public address system must be audible in all j areas of containment for response during emergency or abnormal <
conditions. NRC examiners identified this item for follow up as l Inspector Follow-up Item (IFI) 50-416/93-301-04, " Failure of public address system to be audible in all areas of containment."
(3) The simulator failed Monday night of the examination week and was l limited to JPMs only until Thursday morning. The failure caused !
two scenarios scheduled for Tuesday morning to be delayed until )
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Report Details 3 Thursday morning. The simulator support staff worked around the clock to troubleshoot and repair the simulator while keeping it functioning enough to support JPMs on Tuesday and Wednesday. The examiners identified the dedication and technical knowledge of the simulator support personnel as a strengt Candidate performance Candidate performance during the examinations was good. Written examination scores ranged from 82 percent to 95 percent with an average score of 90.4 percent. The examiners identified three '
knowledge area weaknesses from the candidates' performance on the written examinations. The candidates failed three of the ninety JPMs administered. No generic weaknesses were identified during the simulator examinations. The examination team identified a violation ;
during the conduct of one walkthrough examinatio (1) The examiners determined three areas of training weakness during item analysis of the written examination. This conclusion was based on the number of candidates missing specific examination questions. The NRC and the facility training staff reviewed all written examination questions before examination administratio No post-examination review comments were provided by the facility staff on these questions. As a result, it was determined that no overt format or content errors were present that may have led to candidate confusion. Details are provided below. Question numbers are from the SR0 examination. If the question was also asked on the R0 examination, that question number is provided in parenthesi i (a) Question SR0-33 This question dealt with identifying the thermal limit which the Rod Action Control System protects against by limiting rod withdrawal to four notches when reactor power is between the high power and low power setpoints. Five of six ,
candidates answered Maximum Fraction of Limiting Power l Density. One of six answered Maximum Average Planar Linear !
Heat Generation Rate. The correct answer was Minimum l Critical Power Ratio. The examiners changed the question during the facility pre-review because the thermal limit taught in the lesson plan was incorrect. The changes ;
included identifying the correct answer and removing the :
incorrect thermal limit taught in the lesson plan from the I distractors. The examination team and facility agreed this was a valid question with the changes and because the correct thermal limit is identified in the Technical Specification l
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Report Details 4 (b) Question SRO-43 (R0-48)
This question dealt with the minimum Residual Heat Removal (RHR) flow during normal PHR shutdown cooling. Six of ten candidates answered that the minimum flow was 1000 gpm, vice the correct answer that the minimum flow was 4000 gp (c) Question SRO-82 This question dealt with the purpose of the End-of-Cycle Recirculation Pump trip following a Main Turbine trip. Four of six SR0 candidates answered that the trip compensates for lower integral rod worth at the end of cycle, vice the correct answer that the trip recovers the loss of thermal margin at the end of cycl (2) The examiners conducted walkthrough examinations on September 28-30, 1993. During a walkthrough examination on September 30, 1993, a candidate escorted a visitor, followed by the examiner, into the Radiologically Controlled Area (RCA) ;
without being on a Radiation Work Permit (RWP) and without 1 obtaining a self-reading dosimeter. This occurred when, after
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I completing three Control Room JPMs and some administrative I topics, the examiner gave the candidate the initiating cue for an l Auxiliary Building JPM. The candidate stated he understood the cue and led the visitor he was escorting and the examiner out the back door of the Control Room and into the Turbine. building which was part of the RCA. After walking about fifty feet, the examiner recognized where they were and' shortly thereafter the candidate realized that they were not signed onto an RWP and did not have Electronic Alarming Dosimeters (EADs). The candidate immediately took the proper corrective action in accordance with~
plant procedures. Administrative Procedure No. 01-S-08-2,
" Exposure and Contamination Control," Revision 28, section 6. required personnel to have an RWP and the proper dosimetry to enter an RCA. Administrative Procedure No. 01-S-08-3, " Personnel Radiation Exposure Monitoring," Revision 15, section 6. required self-reading pocket dosimeters or EADs for all entries into the RCA. The NRC considered this as VIO 50-416/93-301-01
" Failure to follow procedure which stated the access requirements for entering a Radiologically Controlled Area." Simulator Facility Simulator limitations significantly restricted examiners during examination preparation and affected the content of the operating portion of the examinations. A simulator hardware failure affected the administration of the examination (1) The examiners identified that the malfunction for loss of condenser vacuum did not properly simulate a complete loss of
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Report Details 5 vacuum. When entered at 100 percent severity, vacuum loss stopped at 10 inches of vacuum. This value was greater than the Main Steam Isolation Valves (MSIV) isolation setpoint of 9 inches of vacuum. This impacted the training and testing value of loss of vacuum scenarios in that the operators were able to maintain the reactor feed pumps on lin (2) The simulator did not have the capability to override automatic actuation signals of Emergency Core Cooling Systems (ECCS) while still allowing manual actuation. The simulator also could not override manual initiation pushbuttons for ECCS. This prevented the examiners from evaluating the candidates' response to two plausible events. One was in the event that systems fail to actuate automatically and require manual operation. Another was in the event that systems require initiation via an arm and depress button which failed thereby requiring the system to be manually lined u (3) The simulator model for the Feedwater Control system did not accurately reflect plant response. ~ In a recent plant event, High Pressure Core Spray (HPCS) initiated and the reactor tripped on a high water level signal. In the simulator, the Feedwater Control system in automatic could handle the HPCS initiation and the reactor did not tri (4) The examiners noted during the administration of JPMs for Standby Diesel Generator 11 that on the simulator the output breaker did ,
not shut on several occasions even though all the requirements to I shut the breaker were met. This problem was intermittent in that i the breaker shut during some of the JPM (5) The simulator failed Monday night of the examination week and was !
limited to JPMs only until Thursday morning. The failure was i caused by a bad hard-disk complicated by a failure of the hard-disk controlle Procedures The examiners identified an inadequate procedure during the review of JPMs. During the examination preparation week, a review of one of the 1
facility's JPMs identified an inadequate section in S0I 04-1 01-B21-1 1 Revision 33, Nuclear Boiler System. Section 4.3 of the procedure '
provided two sequences for opening Main Steam Isolation Valves (MSIVs). A note preceding Step 4.3.2c directed the operator to the appropriate sequence. The note directed performance of Steps 4.3.2c(1) and 4.3.2c(2) if opening the MSIVs was not critical or the MSIVs were closed for greater than two hours. The note directed performance of Steps 4.3.2c(3) and 4.3.2c(4) if opening the MSIVs was critical and the MSIVs were closed for less than two hours. The examiners identified the problem in the first sequence. Steps ;
4.3.2c(1) and 4.3.2c(2)(a) thru 4.3.2c(2)(b) isolated the steam loads i off the Main Steam lines, lined up the path to warmup the piping
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downstream of the inboard MSIVs, and directed control of the heatu l Step 4.3.2c(2)(c) stated after heatup and pressurization of downstream piping close.... The problem was that the downstream piping could not i
be pressurized based on the system lineup prior to step 4.3.2c(2)(c) :
and therefore, that step could not be completed. That step confused an instructor during the preparation week and caused a delay in opening the inboard MSIVs while the instructor checked the lineup and procedure and determined what needed to be done to pressurize the downstream piping. This step also confused a candidate who took this path by error during performance of a JPM during the examination wee The NRC considered this as VIO 50-416/93-301-02 " Inadequate procedure 04-1-01-B21-1 to give guidance for opening MSIVs." However, based l upon the corrective actions, no similar violations identified by the i team in the past two years and the lack of willfulness, this NRC identified violation is not being cited because criteria specified in Section VII.B of the NRC Enforcement Policy were satisfie . Action on Previous Inspection Findings (Closed) IFI 50-416/92-07-01, Failure of the facility to reliably screen j candidate i i
This item concerned the 1992 failure of the facility to conduct final l candidate screening using oral boards administered by the Operator 1 Training Evaluation Committee (OTEC). During the September 1993 examination, the examiners reviewed Procedure No. 14-S-02-6, " Licensed '
Operator Training Program Implementation," Revision 16. The examiners also reviewed the final training packages for all candidates initially
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proposed for the September 1993 examination which included the results of '
the OTEC oral boards. The examiners found that the facility had conducted ;
OTEC oral examinations for screening the candidates. The oral examinations were effective and screened out one candidate from the clas l All candidates recommended by the OTEC board passed the NRC examination '
The examiners found the licensee's corrective action to be adequate, and this Inspector Fol. low-up Item is close . Exit Interview i
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At the conclusion of the site visit, the examiners met with !
representatives of the plant staff listed in paragraph 1 to discuss the '
results of the examinations. The licensee did not identify as proprietary any material provided to, or reviewed by the examiners. The examiners further discussed in detail the inspection finding (s) listed belo i Dissenting comments were not received from the license '
Item Number Description and Reference '
VIO 50-416/93-301-01 Violation regarding failure to follow procedure l which stated the access requirements for !
entering a Radiologically Controlled Area (paragraph 2.d.(2)). 1 i
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Report Details 7 VIO 50-416/93-301-02 Violation regarding inadequate procedure 04-1-01-821-1 to give guidance for opening MSIVs (paragraph 2.f.).
VIO 50-416/93-301-03 Violation regarding failure to follow procedure which stated the requirements for controlling operator aids (paragraph 2.c.(1)).
IFI 50-416/93-301-04 Inspector follow-up item regarding failure of public address system to be audible in' all areas of containment (paragraph 2.c.(2)).
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ENCLOSURE 2 SIMULATOR FACILITY REPORT Facility Licensee: Grand Gulf Nuclear Station Facility Docket No(s).: 50-416 Operating Tests Administered On: September 27 - 30, 1993 This form is to be used only to report observations. These observations do not constitute, in and of themselves, audit or inspection findings and are '
not, without further verification and review, indicative of noncorroliance with 10 CFR 55.45(b). These observations do not affect NRC certificat. 1 or approval of the simulation facility other than to provide information that may be used in future evaluations. No licensee action is required solely in response to these observation i i
While conducting the simulator portion of the operating tests, the following i items were observed (if none, so state):
ITEM DESCRIPTION
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Condenser leak When entered at 100 percent severity, the vacuum loss I stopped at 10 inches of vacuum (paragraph 2.e.(1)). l I
Inability to The simulator did not have the capability to override 1 override aut automatic actuation signals of ECCS systems while still initiation allowing manual actuation. The simulator also could not override manual initiation pushbuttons for ECCS systems >
(paragraph 2.e.(2)).
Feedwater Control The simulator model for the Feedwater Control system does not accurately reflect plant response (paragraph 2.e.(3)).
DG 11 Diesel Generator 11 output breaker did not shut on several occasions even though all the requirements to shut the breaker were met (paragraph 2.e.(4)).
Hardware failure The failure was caused by a bad hard-disk complicated by a i failure of the hard-disk controller (paragraph 2.e.(5)). I i
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Nuclear Regulatory Commission Operator Licensing Examination This document is removed from Official Use Only category on date of examinatio NRC Official Use Only
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U. S. NUCLEAR REGULATORY COMMISSION '
SITE SPECIFIC EXAMINATION ,
REACTOR OPERATOR LICENSE
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REGION 2 ,
CANDIDATE'S NAME:
FACILITY: Grand Gulf 1 REACTOR TYPE: BWR-GE6
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DATE ADMINISTERED: 93/10/01
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INSTRUCTIONS TO CANDIDATE:
Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE %
100.00 % TOTALS *
FINAL GRADE All work done on this examination is my ow I have neither given nor ,
received ai Candidate's Signature
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REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blan MULTIPLE CHOICE 023 a b c d 001 a b c d 024 a b c d 002 a b c d 025 a b c d 003 a b c d 026 a b c d 004 a b c d 027 a b c d 005 a b c d 028 a b c d 006 a b c d 029 a b c d 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a b c d 033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c d 016 a b c d 039 a b c d 017 a b c d 040 a b c d 018 a b c d 041 a b c d 019 a b c d 042 a b c d 020 a b c d 043 a b c d 02 a b c d 044 a b c d 022 a b c d 045 a b c d i
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' REAdTOR' OPERATOR Pago, 3 ,
ANSWER S H E.E T -
Multiple Choice (Circle or X your choice)-
If you change your answer, write your selection.in the blan ,.,
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046 a b c d 069 a b c d '
-047 a b c d 070 a b c d 048 a b c d 071 a b c d -!
049 a b c d 072 a b c d 050 a b c d 073 a b c d I
051 a b c d 074 a b c d )
'l 052 a b c d 075 a b c d i
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053 a b c d 076 a b c d 054 a b c d 077 a b c d-
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055 a b c d 078 a b c d 056 a b c d 079 a b c d
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057 a b c d 080 a b c d i
058 a b c d 081- a b c d .!
059 a b c d 082 a b c d l
060 a b c d 083 a b c d 061 a b c d 084 a b c d 062 a b c d 085 a b c d 063 a b c d 086 a b c d 064 a b c d 087 a b c d 065 a b c d 088 a b c d 066 a b c d 089 a b c d 067: a b c d 090 a b c d 068 a b c d 091 a b c d
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0 O REAdTOR. OPERATOR Page ~4 A-N-S W E R SHEET
. Multiple choice (Circle or X your choice)
If you change your_ answer, write your selection in the blan a b c d 093 a b c d 094 a b c d 095 a b c d 096 a b c d 097 a b c d
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098 a b c d 099 a b c d 100 a b c d
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(********** END OF EXAMINATION **********)
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NRC RULES AND GUIDELINES FOR LICENSE' EXAMINATIONS During the administration of this examination the following rules apply: ;
1 Cheating on the examination means an automatic denial of your .i application and could result in more severe penaltie . After the examination has been completed, you must 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. This must be done after you complete the exam-inatio . Restroom trips are to be limited and only one candidate at a time may leav You must avoid all contacts with anyone outside the ;
examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to facilitate legible repro- i duction . Print your name in the blank provided in the upper right-hand corner of the examination cover shee . Fill in the date on the cover sheet of the examination (if )
necessary).
I Print your name in the upper right-hand corner of the first page i of each section of your answer sheet . Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question pag . The point value for each question is indicated in parentheses after the questio . Partial credit will NOT be give . If the intent of a question is unclear, ask questions of the i examiner onl . When you are done and have turned in your examination, leave the examination area as defined by the examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoke .
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OUESTION: 001 (1.00)
The plant is operating at 100% power. A transient occurs that trips the Recirculation Pumps. The following Recirculation Pump breaker indication are observed:
CB 1 A/B OPEN CB 2 A/B OPEN CB 3 A/B CLOSED CB 4 A/B OPEN CB 5 A/B OPEN WHICH ONE (1) of the following is the cause of the Recirculation Pump Trip (RPT)?
a. ATWS RP b. End-of-Cycle RP c. Low feed water flo d. CB 5 A/B tripped on over-curren QUESTION: 002 (1.00)
WHICH ONE (1) of the following is considered a CORE ALTERATION during refueling?
a. Replacement of a control rod blad b. Removal of the steam dryer assembly, c. Withdrawal of a SRM detector, d. Removal of a LPRM detecto l
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- REAOTOR OPERATOR Page 8 QUESTION: 003 (1.00)
WHICH ONE (1) of the following design features of the Control Rod Drive mechanism " Flange" allows the control rod to be scrammed even if its associated scram inlet valve fails to open?
a. Collet piston b. Stop piston c. Cooling water orifice d. Ball check valve QUESTION: 004 (1.00)
WHICH ONE (1) of the following RPS trip signals is allowed to be bypassed?
a. High Scram Discharge Volume Level l
b. High Drywell Pressure c. High Main Steam Line Radiation d. High Reactor Vessel Pressure !
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QUESTION: 005 (1.00) l l
i The indication for the G33-F004 (RWCU Isolation valve) has a flashing red light on the Isolation Valve Status Panel. WHICH ONE (1) of the following is the reason for this indication? ;
I a. The valve has spuriously close b. The valve is in the intermediate positio c. An isolation signal is present and the valve is close d. An isolation signal is present and the valve is ope I
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QUESTION: 006 (1.00) I
The plant is operating at low power. Reactor pressure is being l controlled via the By-pass valves. The Reactor Water Cleanup System is I being used to reject water from the reactor vessel to the Main i Condenser. WHICH ONE (1) of the following is the consequence of opening j the G33-F035, RWCU BLOWDOWN VALVE TO RAD WASTE?-
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a. Over-pressurizing the rad waste pipin :
b. Loss of condenser vacuu l c. RWCU pumps will tri d. Damage to Filter /Deminerlizers
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QUESTION: 007 (1.00)
HPCS automatically initiated on low water level and high drywell ;
pressure. The HPCS injection valve, E22-F004, automatically closes on r high reactor water level. Reactor water level is now -2 inches and decreasing and the high drywell signal is still present. WHICH ONE (1)
of the following . operator actions must be taken to re-inj; t water using HPCS before the water level reaches level 2 (-41.6 inches) ? (Assume no '
other operator actions will be taken.)
a. Reset the automatic initiation logi b. Arm and depress HPCS Manual Initiation pushbutto !
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c. Place the F004 hand switch to the OPEN positio ,
d. Reset the high water level logi i
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- REACTOR OPERATOR Page 10 :
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QUESTION: 008 (1.00) s The 105 second ADS timer has automatically initiated following a loss of coolant accident. WHICH ONE (1) of the following conditions will RESET the 105'second timer?
a. Reactor water level recover b. Drywell pressure decreases below setpoin j c. LPCS pump is shutdow d. The ADS inhibit switches are taken to INHIBI QUESTION: 009 (1.00)
WHICH ONE (1) of the following refueling conditions will result in a rod block with the Mode Switch in the REFUEL position?
a. The refueling platform is positioned over.the reactor core and -
one control rod is withdraw b. The refueling platform is positioned over the reactor core.and the main hoist fuel grapple is loade c. The refueling platform is positioned away from the reactor core ,
and the main hoist fuel grapple is loade !
l d. The refueling platform is positioned away from the reactor core J and one control rod is withdraw j l
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- REACTOR OPERATOR Page 11 QUESTION: 010 (1.00)
You are unable to drive one control ro ONEP-05-1-02-IV-1 states;
" Increase drive water pressure in 25 psi increments and attempt to drive the affected rod."
WHICH ONE (1) of the following describes the action required to increase drive water pressure?
a. Throttle valve F003 is throttled close b. Bypass valve F004 is throttled opene c. Stabilizing valves are opene Flow control valves are close QUESTION: 011 (1.00)
A reactor scram has occurred following a recirculation loop suction line LOC The STA has determined that you are in the unsafe region of the RPV Saturation Temperature curve of EP-2, "RPV Control". WHICH ONE (1)
of the following describes the relationship between indicated and actual reactor vessel level? Assume vessel level is above TA a. Actual reactor vessel water level LESS than indicated due to variable leg flashin b. Actual reactor vessel water level LESS than indicated due to reference leg flashin c. Actual reactor vessel water level GREATER than indicated due to reference leg flashin d. Actual reactor vessel water level GREATER than indicated due to variable leg flashin ~
- REACTOR OPERATOR Page 12 QUESTION: 012 (1.00) ;
The plant is experiencing a small break LOCA. The Feedwater pumps have tripped off and the reactor water level is being maintained at +9 inches with RCIC. Drywell pressure is stable at 1.57 psig. The Recirculation ,
Flow Control Valve "A" HPU oil temperature is 142 deg WHICH ONE (1) of the following has caused a Recirculation Flow Control Valve "A" motion inhibit? ;
a. High drywell pressur b. Loss of both feed pump c. High HPU oil temperatur d. Low reactor water leve QUESTION: 013 (1.00)
The HPCS diesel has started as a result of a LOCA. WHICH ONE (1) of the following will trip the HPCS Diesel Engine with the LOCA signal still present?
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a. High jacket water temperatur b. Low lube oil pressur c. Generator differential curren d. Overcurrent with voltage restrain ,
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' QUESTION: 014 (1.00)
WHlCH ONE (1) of the following is the MAXIMUM number of visitors that can be escorted by one escort in the Vital Area?
a. 2 b. 5 c. 7 d. 10-QUESTION: 015 (1.00)
WHICH ONE (1) of the following functions /isolations is initiated when all four NSSSS manual isolation pushbuttons are armed and depressed?
a. 30 minute timer for suppression pool make-up dump logi b. 105 second timer for ADS activation logi c. RCIC exhaust vacuum breaker isolatio d. LPCS test line isolatio QUESTION: 016 (1.00)
WHICH ONE (1) of the following is-used to calibrate LPRM detectors?-
a. Flux profiles from TIP trace b. Comparison made with PRM detectors in other quadrant c. Neutron to gamma signal ratio _ tes d. Comparison made with associated APRM outpu .- .. -- . ~ . . . -
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-QUESTION: 017 (1.00)
WHICH ONE (1) of the following areas will have its ventilation system [!
isolated on an initiation signal to the Standby Gas Treatment System?
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a. Enclosure building i
b. Radwaste building i i
c. Control Room d. Auxiliary building t
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QUESTION: 018 (1.00)
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-WHICH ONE (1) of the following conditions will result in a Rod Block? IRMs on Range 2 Mode Switch in RUN SRM count is 0.5 cps IRMs on Range 9 :
Mode Switch in RUN i SRM count is 1.5 x 10E5 cps l c. IRMs on Range 2 ;
Mode Switch in STARTUP l SRM count is 0.5 cps ]
1 IRMs on Range 9 i Mode Switch in STARTUP J SRM count is 1.5 x 10E5 cps
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- REACTOR OPERATOR o-O Page 15 QUESTION: 019 (1.00)
During refueling the Fuel Fool Cooling and Cleanup System has failed to maintain the pool temperature below 150 deg F. WHICH ONE (1) of the '
following systems may be used to supplement fuel pool cooling?
a. Residual Heat Removal Plant Chilled Water c. Component Cooling Water
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d. Condensate and Refueling Water Storage and Transfer
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QUESTION: 020 (1.00)
WHICH ONE (1) of the following conditions will cause a half-scram?
a. APRM E is downscale, IRM G is INOP, mode switch is in STARTU IRM H reading 35/125ths on range 4, mode switch is in STARTUP, detector is selected and withdraw c. APRM F is downscale, IRM D is bypassed, mode switch is in RU IRM B reading 115/125ths of scale, mode switch is in STAP. TU J QUESTICN: 021 (1.00)
The Meter Function Switch on a given LPRM/APRM Meter Panel is in the COUNT position. The meter is indicating 75%. WHICH ONE (1) of the following is the number of LPRMs that have been bypassed for that ,
channel? 1 a. 5 ,
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- REACTOR OPERATOR Page 16 QUESTION: 022 (1.00)
WHICH ONE (1) of the following Radwaste Tanks receives high quality, low conductivity wastes?
a. RWCU phase separator Tank b. Equipment Drain Collector Tank c. Floor Drain Collector Tank d. Miscellaneous Waste Receiver Tank QUESTION: 023 (1.00)
The plant is operating at 100% power and LPCS is lined up in the normal standby mode. WHICH ONE (1) of the following conditions will allow the LPCS injection valve (F005) to be opened using its hand switch?
a. Anytime a LOCA signal is presen b. Anytime the LPCS pump is runnin c. Anytime reactor pressure is 400 psi d. Anytime the LPCS Manual Initiation Pushbutton has been armed and depresse QUESTION: 024 (1.00)
A total loss of AC power occurs with Div. I and Div. II ESF buses NOT energize RCIC is manually initiated and maintaining reactor water leve WHICH ONE (1) of the following is the effect that a loss of AC power will have on the RCIC suction transfer due to a high suppression pool level?
a. The CST will drain to the suppression poo b. The suction transfer will occur normall c. The suction path will be isolate d. The suction path will only be available from the CS _ . .
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WHICH ONE (1) of the following is the purpose of the charcoal'adsorber in the standby gas treatment exhaust filter train?
a. Removes iodine, b. Condenses entrained moistur c. Removes particulate matte d. Recombines hydroge QUESTION: 026 (1.00) ;
WHICH ONE (1) of the following is the reason that the mechanical vacuum pumps are NOT operated at reactor powers greater than 5%?
a. Minimize release of radioactive ga Prevent pump runou c. Reduce the pump blade expansio , Prevent erosion of the pump casin QUESTION: 027 (1.00)
The plant is at 100% power, with RFPTs A, and B in service and operatin in three element mode. The M/A Station for RFPT B output signal suddenly fails to zero output. WHICH ONE (1) of the following is the impact on RFPT B speed?
a. Automatically ramps up to the high-speed sto b. Remains at its last called-for valu ;
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c. Automatically ramps down to the minimum called for by the electric automatic positioner (~3000 rpm).
d. Drops to near zero as the governor valve fully close i
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I-QUESTION: 028 (1.00)
l The' fire protection water header pressure has dropped to 128 psi ,
WHICH ONE (1) of the following describes the status of'the fire water protection system? o a. The jockey pump and the electric pump are running; the diesel fire pump is secure b. The electric pump-and the diesel fire pump are running; the -
jockey pump is secured, c. The jockey Inunp is running; the electric fire pump and the
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diesel fire pump are secured . l d. The jockey pump, the electric pump and the diesel fire pump are l runnin QUESTION: 029 (1.00)
A loss of coolant accident (LOCA) has occurred and the appropriate loads ,
have been shed. WHICH ONE (1) of the following components is the FIRST to be re-energized?
a. Standby Service Water Pump A b. Enclosure Building Recirculation Fan A c. Control Room Air Handling Unit A d. Diesel Generator Room Outside Air Fan A
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QUESTION: 030 (1.00) ,
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The following parameters exist on Reactor Recirculation Pump A:
Seal cavity #2 pressure is 400 psi Annunciator RECIRC PMP A OUTR SEAL LEAK HI.is alarme ,
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Annunciator RECIRC PMP A SEAL STG FLO HI/LO is alarmed on low flo WHICH ONE (1) of the following is the indicated failure? Failure of #1 sea b. Failure of #2 sea ,
c. Failure of both seal ,
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d. Plugged restricting orific ;
. QUESTION: 031 (1.00)
Following an Offgas Post-Treatment Radiation High-High-High condition '
the following valve positions are observe ,
F016A/B, Condenser Drain open ,
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F023, Holdup Line Drain closed F034A/B, Cooler Condenser Drain close ,
F045, Adsorber Bypass closed F053A/B, Adsorber Inlet open F054, Prefilter Inlet Drain open F060, Offgas Discharge to Radwaste Vent ' closed WHICH ONE (1) of the following valves is in an abnormal position as a '
result of the High-High-High radiation condition?
a. F016A/B b. F045 .
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' F053A/B d. F054 >
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QUESTION: 032 -(1.00) ,
i WHICH ONE (1) of the following is the minimum allowable operating speed l
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of the RCIC turbine?
a. 1800 rpm
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b. 2000 rpm c. 2200 rpm d. 2400 rpm i
QUESTION: 033 (1.00)
Reactor is operating at 20% power. A loss of condenser vacuum has ,
occurred and is now 14 inches Hg. WHICH ONE (1) of the following ,
automatic actions should have occurred?
a. Reactor feedwater pump tri b. MSIVs isolat c. Offgas train isolates, d. Main steam bypass valves clos ,
QUESTION: 034 (1.00) ,
-g The reactor is operating at 80% power. WHICH ONE (1) of the following trips will be the result of a loss of the flow control reference-signal-to an APRM?
a. INOP b. Downscale c. Upscale neutron flux d. Upscale thermal power
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REACTOR OPERATOR Page'21 QUESTION: 035 (1.00)
WHICH ONE (1) of the following describes the results of installing the SRM shorting links?.
a. Bypasses SRM reactor period tri I b. Allows retraction of an SR c. Bypasses SRM inoperative tri d. Bypasses SRM upscale rod bloc j
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QUESTION: 036 (1.00) 1 WHICH ONE (1) of the following is the MAXIMUM continuous rating for Standby Diesel Generator 11 to maintain safe shutdown conditions? l
.. KW j b. 5740 KW l
c. 7000 KW
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'l l QUESTION: 037 (1. 00 ) d WHICH ONE (1) of the following methods for control rod insertion during an ATWS require the reactor scram.to be reset?
a. Venting CRD Overpiston Volumes i b. Opening Individual Scram Test Switches l c. Manually Venting the Scram Air Header d. Maximizing CRD Cooling Water Differential ;
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l QUESTION: 038 (1.00) H l
A 4.16KV (ITE) electrical circuit breaker is fully ready for remote operation and is currently CLOSED. WHICH ONE (1) of the following describes the breaker response if the control power fuses are blown?
a. The breaker will tri b. The breaker can not be operate c. The breaker can only be operated locall d. The breaker can only be remotely tripped and must be locally shu !
. QUESTION: 039 (1.00)
WHICH ONE (1) of the following is the reason for closing the charging water riser valve, 113, following a single rod scram via the test-switches?
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a. Prevents rod drif '
b. Prevents low charging header pressur c. Prevents loss of coolin .
d. Prevents seal damag QUESTION: 040 (1.00)
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Reactor power is 68% and total core flow is 73% of rated core flo WHICH ONE (1) of the following is the MAXIMUM allowable recirculation ;
flow mismatch?
a. 5%
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b. 10%
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c. 15%
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d. 20%
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-QUESTION: 041 (1.00) :
During a LOCA, the A RHR pump automatically starts in the LPCI mod The pump is then put in Manual Override with the LPCI initiation signals still present. WHICH ONE (1) of the following will restart the pump?
a. Arm and depress the LPCS/RHR A Manual Initiation Pushbutto ,
b. Place the RHR A pump switch to AUT c. Automatic initiation of Containment Spra d..LPCI initiation signal clears and reoccurs before being rese ,
QUESTION: 042 (1.00) .
~1ui ATWS is in progress and HPCS is maintaining reactor water leve The-following is the status of the SLC system:
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SLC OOSVC alarms are presen SLC Out-of-Service switches (A/B) in INO SLC Pump Suction Valve Key Lock Test Switches (A/B) placed to TES Test Tank Outlet Valve (F031) open.
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The Control Room Operator has been directed to initiate SLC regardless of present status. WHICH ONE (1) of the following will prevent the ,
Standby Liquid Control system from injecting sodium pentaborate?
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a. SLC Out-of-Service switch positio b. HPCS injection already in progres c. SLC Pump Suction Valve Key Lock Test Switch positio d. Test Tank Outlet Valve (F031) positio ,
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- REACTOR OPERATOR Page 24 QUESTION: 043 (1.00)
The reactor scrams from 100% power with the master level controller in three-element mode. Reactor water level subsequently decreases below level 3. WHICH ONE (1) of the following is the automatic response of the Feedwater Control System?
a. The level setpoint increases to 54 inches for 10 seconds and then the level controller shifts to manua b. The level setpoint decreases to 10 inches and the m. ster level controller shifts to single element mode, c. The level setpoint increases to 54 inches for 10 seconds and decreases to 18 inche d. One RFPT trips and the other RFPT shifts to Startup Level Control with the setpoint at 54 inche !
QUESTION: 044 (1.00)
WHICH ONE (1) of the following actuates the red light above a Safety Relief Valve Control Switch on panel 1H13-P601 to indicate that the valve is open?
a. Pilot solenoid valv b. Pressure switches on the discharge pipin c. Valve limit switche d. Temperature elements on the discharge pipin REACTOR OPERATOR a
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QUESTION: 045 (1.00) !
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WHICH ONE (1) of the following isolation signals will ONLY close the i Reactor Water Cleanup (RWCU) Pump Suction Outboard Isolation Valve (G33- l F004)? ;
a. Reactor water level b. RWCU high differential flo c. Standby Liquid Control "B" actuatio d. RWCU filter demin inlet high temperature.
QUESTION: 046 (1.00)
WHICH ONE (1) of the following RCIC valves receives a close signal from the RCIC Division I isolation logic?
a. F010, Pump Suction from CST Valve b. F022, Inboard Test Return to CST Valve c. F031, Pump Suction from Suppression Pool Valve d. F045, Turbine Steam Supply Valve QUESTION: 047 (1.00)
WHICH ONE (1) of the following describes the conditions that will open the High Pressure Core Spray Pump Minimum Flow to Suppression Pool Valve (F012)?
a. System flow is 750 gpm and pump discharge pressure is 120 psi b. System flow is 800 gpm and pump discharge pressure is 135 psi c. System flow is 875 gpm and pump discharge pressure is 115 psi d. System flow is 900 gpm and pump discharge pressure is 150 psi . . _ . . ~ . . .. . ._ ,
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'l QUESTION: 048 (1.00)
WHICH ONE (1) of.the following is the MINIMUM'RHR flow during-normal RHR Shutdown Cooling? ,
' gpm b. 2000 gpm ,
c. 3000 gpm '
i d. 4000 gpm
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QUESTION: 049 (1.00)
The reactor'is in Operational Condition 2. WHICH ONE (1) of the 5 following conditions will cause a Main' Steam Line Isolation? 1 a. High Drywell Pressure '!
b. Low Main Steam Line Pressure !
t c. Reactor Water Level 2 ,
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d. High Steam Line Flow l
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. QUESTION: 050 (1.00)
The reactor is operating at 100% power. The LPCS SYS OOSVC annunciator .,
alarms. During the investigation it is determined that.there'is LPCS line break. WHICH ONE (1) of the following is the cause of the alarm? ,
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a. Excessive differential pressure between RHR "A" LPCI ' injection line and the LPCS Spray Sparge ;
b. Excessive differential pressure between the vessel downcomer
. region and the LPCS Spray Sparge l c. Excessive differential pressure between the HPCS Spray Sparger '
and the LPCS Spray Sparger, d. Excessive differential pressure between the LPCS Spray Sparger '
and the above core plate pressure ta ,
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. QUESTION: 051 (1.00)
WHICH ONE.(1) of the following condensate system pumps is powered by ,
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electrical bus 13AD?
a. Condensate pump B i b. Condensate pump C ,
c. Condensate booster pump B i
d. Condensate booster pump C
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QUESTION: 052 (1.00)
WHICH ONE (1) of the following radiation monitoring subsystems use scintillation detectors in a shielded sampler so that background radiation levels will be minimized?
a. Main Steam Line ,
b. Offgas Post Treatment ;
c. Ventilation System !
d. Process Liquid
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QUESTION: 053 (1.00)
WHICH ONE (1) of the following conditions will result in the Control' l Room Ventilation System being isolated?
a. High inlet air freon concentration ;
b. Smoke detected at the fan inle c. Reactor water level of -50 inche ,
d. Loss-of 480 volt AC powe i k
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QUESTION: 054 (1.00)
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WHICH ONE (1) of the following is the power supply to ESF transformer 12?
a. Unit 1 34.5 KV Bus 11 b. Unit 2 34.5 KV Bus 21R c. Port Gibson 115KV line, d. 500 KV West Bu QUESTION: 055 (1.00)
WHICH ONE (1) of the following Main Steam Lines (MSL) supplies steam to the RCIC system?
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a. MSL A b. MSL B c. MSL C d. MSL D
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QUESTION: 056 (1.00)
During a loss of AC power, ESF div. I bus was de-energized for 3 minutes. WHICH ONE (1) of the following is the reason that all ESF di I fluid systems in operation prior to the power loss must be restarted per the applicable SOI?
a. To ensure that fill and vents are completed prior to restar I b. To ensure that pump motors do not exceed their restart. limit .]
c. To ensure that systems are started in order of importanc !
d. To ensure that manual valve line-ups are performe !
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' QUESTION: 057 -(1.00)
, The RCIC system is being flow tested while the reactor is operating at 100% power.' The suppression pool-temperature is increasing. WHICH ONE (1) of the following is the suppression temperature at which the mode switch must immediately be placed in SHUTDOWN? ,
a. 95 deg F b. 105 deg F ;
c. 110 deg F d. 120 deg F ,
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QUESTION: 058 (1.00)
The following conditions exist: 1
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Reactor power is 20%
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High differential temperature condition in the Auxiliary Building due to a fir Main Steam Isolation valves have closed ,
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Rods are being inserted using CRD WHICH ONE (1) of the following systems should be isolated if they are discharging into the Auxiliary building?
a. High Pressure Core Spray b. Reactor Water Cleanup c. Control Rod Drive d. Fire Suppression
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Page 30 l QUESTION: 059 (1.00)
Core reload is in progress. The Continuous Air Monitors (CAM) alarm on the refuel floor. WHICH ONE (1) of the following is the immediate operator action?
a. Evacuate the refuel floo b. Isolate the Fuel Transfer Tub c. Start Standby Gas Treatmen ,
d. Place Containment Cooling Charcoal Filter Trains in Containment Cooling Mod QUESTION: 060 (1.00)
WHICH ONE (1) of the following systems is available for alternate SLC injection?
a. Feedwater b. Reactor Water Cleanup c. Reactor Core Isolation Cooling d. Residual Heat Removal
b)
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- REACTOR OPERATOR Page 31 QUESTION: 061 (1.00)
The reactor is operating at 80% power. All control rods have been withdrawn at least 2 notches. One CRD pump is out of service for maintenanc The second CRD pump trips and cannot be immediately restarted. One scram accumulator is immediately declared inoperabl WHICH ONE (1) of the following conditions requires the operator to scram the reactor?
a. The CRD pumps cannot be restarted within 20 minute b. A second scram accumulator is declared inoperabl c. Three control rods begin to drif d. Control rod drive mechanism high temperature is observe QUESTION: 062 (1.00)
The reactor is operating at rated conditions. A sensing line has broken in the Drywell. Drywell temperature is increasing. WHICH ONE (1) of the following will require that the Drywell Cooling isolation interlocks be defeated in order to establish drywell cooling?
a. Reactor water level is -53 inche b. Drywell pressure is 1.12 psi c. Drywell vent exhaust radiation is 15.2 mR/h d. Div 1 and Div 3 NSSS have been manually actuate !
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QUESTION: 063 (1.00)
Following a complete loss of Shutdown Cooling, the bulk reactor coolant water temperature is. increasing at 1 degree F every 5 minutes. The present reactor coolant temperature is 152 degrees F. WHICH ONE (1) of- ,
the following is the MAXIMUM time allowed.to pass before primary '
-containment integrity _must.be established? :
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a. 2 hcurs b. 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> c. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
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d. 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> ,
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QUESTION: 064 (1.00) .i WHICH ONE (1) of the following is the required action following a :
sustained reactor period of +30 seconds during a reactor startup? q a. Scram the reactor, f
b. Manually drive all control rods and shutdown the reacto ;
c. Stop control rod movement until the period increase ,
1 Insert control rods until period increase l
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- REACTOR OPERATOR Page 33 QUESTION: 065 (1.00)
The reactor is operating at 100% powe The following alarms are received:
PSW RADIAL WELL TROUBLE PSW UNIT 1 HDR PRESS LOW PSW UNIT 1 HDR PRESS LOW-LOW The condenser vacuum begins to decrease slowly and the plant chillers are los WHICH ONE (1) of the following is the immediate operator action?
a. Scram the reacto b. Reduce core flow to 50%, drive rods to less than 60% powe c. Close the first stage SJAE suction valve, N62-F003A/ d. Trip the main turbin QUESTION: 066 (1.00)
WHIC'T ONE (1) of the following isolation groups will isolate on high drywell pressure?
a. Main Steam Lines and Drains b. Steam Supply to RHR and RCIC c. Reactor Water Cleanup d. Auxiliary Building Isolation Valves
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Page 34 QUESTION: 067 (1.00)
WHICH ONE (1) of the following is the reason for the reactor vessel high water level scram?
a. Prevent damage to turbine driven system b. Avoid water hammer in the main steam line c. Offset the reactivity effect of cold Seedwater injectio Prevent damage to the steam dryer assembl QUESTION: 068 (1.00)
WHICH ONE (1) of the following plant conditions require entry into EP-3, Containment Control?
a. Suppression Pool Level is 18.95 fee b. Reactor Water Level is -35 inche c. Drywell Temperature is 128 deg d. Hydrogen Concentration 0.35%.
QUESTION: 069 (1.00)
The reactor is operating at 100% power and one recirculation pump trip WHICH ONE (1) of the following observations by the reactor operator defines THERMAL HYDRAULIC INSTABILITY? Peak-to-peak APRM swings of 5% rated powe Peak-to-peak LPRM swings of 5 watts /sq c c. Unexplained sustained increase in APRM leve d. Oscillations of 5 inches in reactor water leve . __ _ _ _ . _ . . _
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i QUESTION: -070' (1.00) '!
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lWHICH ONE (1) of the following isolate on loss of 125 VDC? 1
'a. RCIC :!
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b. MSIVs c. MSL Drain Valves d. Auxiliary Building ;l
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QUESTION: 071 (1.00) ,
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The reactor is experiencing high reactor water level during startu :
WHICH ONE (1) of the following parameters limits the amount of flow that l can be rejected to the condenser? 'i a. Condenser vacuum q b. Non-regenerative heat exchanger outlet 1temperatur c. Regenerative heat exchanger inlet temperatur >
d. Filter /demineralizer flo !
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QUESTION: 072 (1.00)
The reactor is operating at 20% power. WHICH ONE (1) of-the following i should be immediately verified following a Turbine / Generator trip?.
a. Turbine Aux Oil Pump start '
b. Recirculation Pumps shift to slow spee c. Reactor Scram, d. Generator Output Breakers ope ;
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QUESTION: 073 (1.00)
WHICH ONE (1) of the=following Main Steam Line Drain Line Isolation Valves has remote shutdown capability at Panel 295?
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a. B21-F016 b. B21-F019 c. B21-F021
d. B21-F067A
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QUESTION: 074 (1.00)
The ADS air system header begins to leal,and header pressure begins to ;
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decrease. WHICH ONE (1) of the following is the MINIMUM header pressure before the ADS system is considered inoperaisle?
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a. 170 psig
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b. 160 psig c. 150 psig d. 140 psig ,
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' REACTOR OPERATOR Page 37 QUESTION: 075 (1.00)
A complete loss of instrument air has occurred. No control rods are drifting. WHICH ONE (1) of the following is required in accordance with procedure 05-1-02-V-9, " Loss of Instrument Air"?
a. Attempt to restore service air to the affected drywell coolers air-operated damper b. Determine the cause for the loss of instrument air and attempt to restor c. Reduce power to less than 50% and check for thetmal hydraulic instabilitie d. Manually open the service water makeup valves to the fire water storage tan QUESTION: 076 (1.00)
The following plant conditions exist: ,
Power level, 90%
Drywell temperature 125 deg. F Drywell Pressure 1.1 psi RCIC Room between the maximum normal and the maximum safe operating area water leve ;
WHICH ONE (1) of the following Emergency Procedures should be entered? j a. EP-2, RPV Control !
b. EP-2A, RPV Control-ATWS c. EP-3, Containment Control d. EP-4, Auxiliary Building and Radioactive Release Control
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' REACTOR OPERATOR Page 38 QUESTION: 077 (1.00)
RCIC was initiated prior to the Control Room being evacuated due to a fir WHICH ONE (1) of the following RCIC turbine isolation / trips can be reset from the DIV I Remote Shutdown Panel?
a. Turbine overspee Low steam supply pressur c. High PCIC equipment area temperatur Low pump suction pressur QUESTION: 078 (1.00)
While operating the plant at 95% power, a valve in the extraction steam line to a feedwater heater inadvertently closes and cannot be reopene WHICH ONE (1) of the following actions should immediately be taken?
a. Manually insert control rods per STA recommendation :
b. Gradually open the bypass valve on the operating feedwater strin c. Reduce recirculation flow until thermal power is 75% of rate Initiate a manual scra l
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' REACTOR OPERATOR
'" O Page 39 QUESTION: 079 (1.00)
The plant is operating at 100% power when a complete loss of Component Cooling Water is experienced. As the Reactor operator you inserted a manual scra WHICH ONE (1) of the following immediate operator actions should be taken?
a. Shift recirculation pumps from fast to slow spee b. Manually trip the recirculation pumps upon any increase in motor winding temperatur c. Close the FCVs to minimum upon any increase in recirculation pump temperatur Start the second CRD pump and maximize purge flow to recirculation pump seal QUESTION: 080 (1.00)
The plant was operating at 100% power when the "A" recirculation pump tripped resulting in the following condition:
Reactor Power 60%
Core Flow 44%
Recirculation loop "B" flow 44,500 gp WHICH ONE (1) of the following is the appropriate action to be taken in this condition? (See attachment)
a. Insert a manual scram within one minut Increase recirculation loop "B" flow until core flow is greater i
'
than 45%. Immediately reduce thermal power to within region III or IV l
' CLOSE the "A" FC l
- _ _ _ _ _ _ _ _ _ _ _ _ _ .
- REACTOR OPERhTOR Page 40 QUESTION: 081 (1.00)
The plant has scrammed due to high reactor pressur The reactor pressure peaked at 1145 psig. WHICH ONE (1) of the following shows the expected status of the scram pilot solenoids, backup scram solenoids and ARI solenoids?
SCRAM PILOT BACKUP SCRAM ARI SOLENOIDS SOLENOIDS SOLENOIDS a. de-energized de-energized energized b, de-energized energized energized c. energized de-energized de-energized d. energized energized de-energized QUESTION: 082 (1.00)
WHICH ONE (1) of the following is the reason that ADS is inhibited before SLC initiation?
a. To prevent a decrease in natural circulation resulting in inadequate boron mixin b. To prevent rapid cooldown during depressurization resulting in a reactivity excursio c. To prevent a rapid injection of cold, unborated water resulting in a rapid increase in powe d. To prevent an increase in natural circulation resulting in decreased voiding and an increase in powe . .
' REACTOR OPERATOR Page 41 l
.I QUESTION: 083 (1.00) -
,
i The Suppression Pool Makeup System (SPMU) has actuated on a~ low-low -)
suppression pool level and high drywell pressure. . WHICH ONE (1) of.the i following will allow manual closure of E30-F001A/B and E30-F002A/B, l Suppression Pool Dump Valves?
.
a. The SPMU mode switch is placed in OF ;
b. The low-low suppression pool level has cleare j c. The 29 minute timer has timed ou d. The SPMU Dump test switches are placed in TES ;
QUESTION: 084 (1.00)
WHICH ONE (1) of the following is the significance of 62 feet in the Containment during Containment Flooding?
a. Places maximum stress on the Containment Structur b. Level at which the RPV should be vente ;
c. Top of the weir wal d. Top of active fue ;
!
QUESTION: 085 (1.00) ]
WHICH ONE (1) of the following is the consequence of NOT venting the RP !
during Containment Flooding? j a. Suppresses RPV water level increase, b. Over-pressurizes the RP c. Allows buildup of explosive gases in the RP I d. Violates Maximum Containment Water Level Limi .
.
.- , .
' REACTOR OPERATOR Page 42 QUESTION: 086 (1.00)
The reactor is being refueled and a leak develops that begins to drain the reactor cavity. WHICH ONE (1) of the following vessel instrument ranges will be the first one to detect the level decrease?
a. Shutdown Range b. Upset Range c. Fuel Zone Range d. Wide Range QUESTION: 087 (1.00)
WHICH ONE (1) of the following is the purpose of the End-of-Cycle Recirculation Pump Trip (EOC-RPT) following a Main Turbine trip?
a. Protects the Recirculation Pumps during the transien b. Recovers the loss of thernal margin at the end-of-cycl c. Reduces the consequences of a failure to scram at the end-of-cycl d. Compensates for the lower integral rod worth at the end-of-cycl QUESTION: 088 (1.00)
A station blackout has occurred. WHICH ONE (1) of the following would prevent Div 1 ESF bus from being energized by the Div 3 diesel generator?
a. Drywell pressure is 1.54 psi )
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b. Reactor water level is -52 inche c. An ATWS is in progres d. Containment temperature is 94 deg l I
i l
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- b,.
~ REACTOR OPERATOR Page 43 QUESTION: 089 (1.00)
The fuel has been removed from the reactor. WHICH ONE (1) of the following is the MINIMUM number that make up the site fire brigade?
(Assume that there are no absences.)
a. 3 b. 4 i
c. 5 l d. 6 i
QUESTION: 090 (1.00) l i
WHICH ONE (1) of the following is the one individual that may be allowed j
'
to make changes to the Reactor Recirculation Flow Control Valve position at 85% power WITHOUT direct supervisio a. A system engineer, holding an inactive SRO license, to check valve response, provided the consent of the
" Operator-at-the-Controls" is obtaine )
b. Shift Engineer, provided the knowledge and consent of the ,
" Operator-at-the-Controls" is obtaine l
c. An unlicensed individual in the SRO Training Program, provided the consent of the licensed Shift Superintendent is obtaine d. The Control Room Operator assisting the
" Operator-at-the-Controls".
. . . _ _ . _ _ .
- REACTOR ' OPERATOR O- O Page 44'
"
I
. QUESTION: 091 (1.00)
As a licensed RO returning from two weeks vacation, WHICH ONE (1) of the-following is required prior to your relief of.the on-shift: Reactor Operator? ,
a. Review CRO log back to last watc <
b. Review plant-status check sheets back to last watc c. Review CRO log for last seven day d. Review plant status check sheets for last seven day QUESTION: 092 (1.00)
Given the following conditions: ,
--
Today's date is January 1, 199 A fully tralued, male radiation worker is 25 years old toda He has a current NRC Form His' previous lifatime whole body exposure is 26 Re WHICH ONE (1) of the following is the MAXIMUM radiation exposure for 1993 allowed by Federa?. Regulations for this individual?
a. 1.25 rem / quarter for the entire year -I b. 3 rem / quarter for 3 quarters in the year c. 1.25 rem / quarter for 2 quarters in the year d. 3 rem / quarter not to exceed 5 rem total for the year 3
+
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- REACTOR OP.SRATOR (/ 'U Page 45 QUESTION: 093 (1.00) .
-,
WHICH ONE (1) of the following is the MAXIMUM number.of people allowed in the horseshoe area without special authorization?. b. 4 c. 5 t t
QUESTION: 094 (1.00) -
You have been directed to lock the mode switch in SHUTDOWN. WHICH ONE
.(1) of the following is the location of the key after the mode switch-has been locked?
a. Mode switch, b. Operations key cabine ,
c. Control Room Operator's desk drawe d. Shift Superintendent's key cabine QUESTION: 095 (1.00)
According to Grand Gulf Technical Specifications, WHICH ONE (1) of the following is the minimum shift staffing on-site when the reactor is at 5% power?
_ _
- REACTOR OPERATOR e
"
4~
Page 46 QUESTION: 096 (1.00)
Work histories for FOUR (4) operators are given below. WHICH ONE (1) of the following operators meets the administrative requirements?
Work History- Excluding Shift Turnover Time Fr Sa Su Mo Tu We T F Sa OFF OFF 1500 1500 1500 0000 0800 0800 0800 2400 2400 2400 0800 OFF OFF OFF 0700 0700 0700 0700 0000 0800 1700 1700 1700 1700 0800 d ., 1200 1200 1200 1200 0700 0700 0700 0700 0000 2200 2200 2200 2200 1500 1500 1500 1500 0800 l
QUESTION: 097 (1.00)
You are working on a weekend and have been given a controlled procedure and been asked to verify that it is the latest revision. The TSO computer is down. WHICH ONE (1) of the following is the approved method?
a. Verify that the procedure is the same revision as the one in the control roo Check the Master Log in Document Contro Check the Master Index in the control roo d. Have an SRO verify all of the steps are correc l l
}
' REACTOR OPERATOR Page 47 QUESTION: 098 (1.00)
WHICH ONE (1) of the following non-safety related systems does NOT require independent verification following each refueling outage?
a. N19 Condensate b. N22 Condensate Cleanup c. N34 Lube Oil d. C84 Met Monitoring QUESTION: 099 (1.00)
You are performing the quarterly physical verification of outstanding clearances when you find an unattached tag. WHICH ONE (1) of the following actions should be taken FIRST?
a. Document the location of the tag on Attachment VII of procedure 01-S-06-1, " Protective Tagging System".
b. Notify the Plant Supervisor, c. Place the component in the position required by the ta d. Stop the work in the area of the equipmen QUESTION: 100 (1.00)
You are on a plant walk-through with an NRC license examiner when the examiner brushes against some equipment. His shirt is checked for contamination. No Alpha contamination is foun WHICH ONE (1) of the following is the MINIMUM limit for determining if the shirt is contaminated with Beta-Gamma?
a. Deteccable (above background)
b. 50 cpm / scan above background cpm / scan above background d. 250 cpm / scan above background (********** END OF EXAMINATION **********)
. - . . . , - - - - -
REACTOR OPERATOR Page 48
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I-AN'SWER: 001 -(1.00)
e REFERENCE:
!
' Grand Gulf: OP-LO-SYS-LP-B33-1-08, LO 31b, p 8 .- Grand Gulf: System Test #2,-Q-11 KA: 202001K127 [4 .1/ 4 . 3 ]
202001K127 ..(KA's)
'
ANSWER: 002 (1.00)
- REFERENCE: Grand Gulf: OP-LO-SYS-LP-F11/17, LO 8, p. 5 . Grand Gulf: System Test 2, Q-2 KA: 201003G005 [3.3/3.9]
,
201003G005 ..(KA's)
-l
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ANSWER: 003 (1.00) j l
REFERENCE: :
-!
i Grand Gulf: OP-LO-SYS-LP-C11-1B-02, LO 3, p. 12 Grand Gulf: System Test 2, Q-46 .
I KA: 201003G007 [3. 6/3. 6] l
'l 201003G007 ..(KA's)
!
!
l i
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_ . _ . _ _ -
-
- 1
-
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" REACTOR OPERATOR Page 49 ANSWER: 004 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-C71-03, LO Sa, p. 34 KA: 212000K412 [ 3 . 9 /4 .1)
212000K412 .(KA's)
ANSWER: 005 (1.00) REFERENCE: Grand Gulf: OP-M72-501, LO 3d, p. 7 KA: 223002A101 [3.5/3.5)
223002A101 ..(KA's)
ANSWER: 006 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-G33/36-05, LO 10a Grand Gulf: 04-1-01-G33-1, Section KA: 204000A205 [2.7/2.8)
204000A205 ..(KA's)
ANSWER: 007 (1.00) ' '
REACTOR OPERATOR Page 50 REFERENCE: Grand Gulf: OP-LP-SYS-E22-1-04, LO 6a, p. 3 '
KA: 209002A415 [3.6/3.6]
209002A415 ..(KA's)
- ANSWER: 008 (1.00) REFERElsCE:
'
' Grand Gulf: OP-LO-SYS-LP-E22-2-03, LO 8, p. 14, 15, and 2 (NOTE:
when the inhibit switch is in INHIBIT, the timer does not advance and is not reset - see page 21 of the reference)
KA: 218000A402 [4.2/4.2]
218000A402 ..(KA's)
l
i ANSWER: 009 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-HN-C11-2, LO 5, Table 1 KA: 234000A302 [ 3 .1/ 3 . 7 ) l l
234000A302 ..(KA's)
!
ANSWER: 010 (1.00) .
n -
,/N,
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' REACTOR OPERATOR Page 51 REFERENCE: Grand Gulf: OP-LO-SYS-LP-C11-1A, LO 3a, p. KA: 201001K408 [ 3 .1/ 3 . 0 ]
201001K408 ..(KA's)
ANSWER: 011 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-B21-05, LO Sd, p. 26-27 KA: 216000K513 [3.5/3.6]
216000K513 ..(KA's)
ANSWER: 012 (1.00) REFERENCE: Grand Gulf: OP-LP-SYS-LP-B33-2-04, p. 25, LO 6 KA: 202002A108 [3.4/3.4]
202002A108 ..(KA's) !
ANSWER: 013 (1.00) l l
l
(\
q)
- REACTOR OPEggyog Page 52 l
l
REFERENCE: Grand Gulf: OP-LO-SYS-LP-P81-03, p. 40, LO Sc KA: 264000K402 [4.0/4.2)
264000K402 ..(KA's)
ANSWER: 014 (1.00) REFERENCE: - Grand Gulf: AP-01-S-11-10, p. 46 KA: 294001K105 [3.2/3.7)
294001K105 . . ( KA' s ) i
,
'
ANSWER: 015 (1.00)
i I REFERENCE: Grand Gulf: OP-LO-SYS-LP-E30-03, p. 15, LO S . Grand Gulf: OP-LO-SYS-LP-M71-04, p. 76, LO 4 KA: 223002K403 [3.5/3.6]
223002K403 ..(KA's)
ANSWER: 016 (1.00)
a.
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-.- - -.- . . . - - . . . - .. ... . .. . . . _ . . . .- .
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- REACTOR OPERATOR' '
Page:53 i
REFERENCE: . Grand Gulf: OP-LO-SYS-LP-C51-3-03, p. 24, LO 6 KA: 215005K113 [2.6/3.0)
,
215005K113 ..(KA's) ;
i
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ANSWER: 017 (1.00) 4
- ,
REFERENCE: Grand Gulf: OP-LO-SYS-LP-T48-05, p. 48, LO 9f i
KA: 288000K103 [3.7/3.7]
288000K103 ..(KA's)
. ANSWER: 018 ' (1. 0 0 )
, t REFERENCE: ; Grand-Gulf: OP-LO-SYS-LP-C51-1-03, p. 27, LO.6a: ,
KA: 215004K401 [3.7/3.7] ,
.i I
215C04K401 ..(KA's)
.-!
l ANSWER: 019 (1.00) !
,
t
,
., _ _ . , , = . - . -
- - , . . - . - , - , - , - . - . - -
..
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- REACTOR OPERATOR Page 54 REFERENCE: . Grand Gulf: OP-LO-SYS-LP-G41/46-05, p. 16 Grand Gulf: OP-LO-SYS-LP-E12-06,-LO le KA: 233000K102 [2.9/3.0]
f 233000K102 ..(KA's)
ANSWER: 020 (1.00) j [+1.0]
REFERENCE: Grand Gulf: OP-LO-SYS-LP-C51-4-03, p. 23, LO 3 KA: 212000K602 [3.7/3.9]
212000K602 ..(KA's)
ANSWER: 021 (1.00) ,
!
REFERENCE- Grand Gulf: OP-LO-SYS-LP-C51-3-03, p. 22, LO 2b ;
.1 KA: 215005K501 [2.8/2.9]
i 215005K501 ..(KA's) .]
~!
l ANSWER: 022 (1.00) -!
l i l l
\
I l
l-l
Or'
- REACTOR OPERATOR Page 55 REFERENCE: Grand Gulf: OP-LP-SYS-LO-G17, p. 6, LO 3 KA: 268000G007 [2. 8 / 3.1]
268000G007 ..(KA's)
ANSWER: 023 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-E21-05, p. 26, LO KA: 209001A403 [3.7/3.6]
209001A403 ..(KA's)
ANSWER: 024 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-E51-03, p. 75, LO 9a. (NOTE: The suction valves are DC powered.)
KA: 217000K601 [3.4/3.5]
217000K601 ..(KA's)
ANSWER: 025 (1.00) rx J
REACTOR OPERATOR Page 56 )
REFERENCE: Grand Gulf: OP-LO-SYS-LP-T48-05, p. 12, LO 3 i KA: 261000G007 [3. 5 /3. 7)
261000G007 ..(KA's)
ANSWER: 026 (1.00)
a.
REFERENCE: Grand Gulf: OP-LO-SYS-LP-N62-04, p. 7-8, LO 8 KA: 256000G010 [ 3 .1/2 . 9 )
256000G010 ..(KA's)
ANSWER: 027 (1.00)
b.
REFERENCE: Grand Gulf: OP-LO-SYS-LP-C34-04, p. 24, LO 4 KA: 259002K302 [3.7/3.7)
259002K302 ..(KA's)
ANSWER: 028 (1.00) .. .. ~. . . - . - . - -. .. - .-
. .-
REACTOR OPERATO O Page 57-
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i REFERENCE:
l 1. . Grand Gulf: OP-LO-SYS-LP-P64-05, p. 32, LO S KA: 286000K401 [3.4/3.6)
286000K401 ..(KA's)
!
ANSWER: 029 (1.00)
i >
,
REFERENCE:
' Grand Gulf: OP-LO-SYS-LP-R21-05, TABLE 1, LO S KA: 262001A304 [3.4/3.6]
i 262001A304 ..(KA's)
.
'
ANSWER: 030 (1.00)- REFERENCE: , Grand Gulf: OP-LO-SYS-LP-B33-1-08, p. 39, LO 13 KA: 202001A210 [3.5/3.9)
i 202001A210 ..(KA's) l
ANSWER: 031 (1.00)
) l
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l
!
!
l
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I
,
l J
i
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-* REACTOR OPERATOR o O Page:58
!
REFERENCE: Grand Gulf: OP-LO-SYS-LP-N64/65-03, p. 33, LO 6 ;
KA: 271000K408 -[3.1/3.3]
271000K408 ..(KA's)
s
,i ANSWER: 032 (1.00) REFERENCE: Grand Gulf: Procedure 04-101-E51-1 Grand Gulf: OP-LO-SYS-LP-E51-03, LO 1 KA: 217000G010 [3.4/3.5]
217000G010 ..(KA's)
>
'
ANSWER: 033 (1.00) , >
REFERENCE:
KA: 295002K205 [2.7/2.7]
295002K205 ..(KA's)
ANSWER: 034 (1.00) , .
' ' " ' ' -<v ,- _ . - - -- ---- _.--
- .- -- .
' REACTOR OPERATOR O O Page~59 REFERENCE: Grand Gulf: OP-LO-SYS-LP-C51-4-03, p. 13, LO 2 KA: 215005K604 [3.1/3.2) ,
215005K604 ..(KA's) (
,
ANSWER: 035 (1.00)
i .
REFERENCE:
- Grand Gulf: OP-LO-SYS-LP-C51-1-03, p. 28, LO 6 KA: 215004K101 [3.6/3.7) ,
215004K101 ..(KA's)
ANSWER: 036 (1.00) '
REFERENCE: Grand Gulf: 04-1-01-P75-1, rev. 38, 3.8 and 3.9, p. 4 Grand Gulf: OP-LO-SYS-LP-P75-05, LO B KA: 264000A203 [3.4/3.4)
264000A203 ..(KA's) 'l ANSWER: 037 (1.00) l l
l l
l l
l
4 S
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,r x, LJ
' REACTOR OPERATop Page 60 REFERENCE: Grand Gulf: 05-S-01-EP-2, Attachment 22 Grand Gulf: OP-LP-EP-LP-004-02, LO KA: 295015K204 [4.0/4.13 295015K204 ..(KA's)
ANSWER: 038 (1.00) REFERENCE: Grand Gulf: OP-SSV-AE-LP-005-02, p. 18 KA: 262001K502 [2.6/2.93 262001K502 ..(KA's)
ANSWER: 039 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-C11-1A, p. 21, LO 8 KA: 201003K101 [3.2/3.3]
!
201003K101 ..(KA's)
,
ANSWER: 040 (1.00) I !
l l
(,_
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- REACTOR OPERATOR Page 61 RE"ERENCE: Grand Gulf: 04-1-01-B33-1, 3.11 Grand Gulf: OP-LO-SYS-LP-B33-2-04, LO 9a KA: 202002A204 [3.0/3.2]
202002A204 ..(KA's)
ANSWER: 041 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-E12-06, p. 59-60, LO 9 KA: 226001K409 [3.2/3.4]
226001K409 ..(KA's)
ANSWER: 042 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-C41-08, p. 25, LO B KA: 211000A308 [4.2/4.2]
211000A308 ..(KA's)
ANSWER: 043 (1.00) __- - . - _ _ _ - _ _ _ _ - _ _ _ - - _ - - _ _ .
, .- . .- -- . . . - . . .. =
' REACTOR OPERATOR Page 62 ,
~ REFERENCE: ,
- L . Grand Gulf: OP-LO-SYS-LP-C34-04, p. 21, LO 4 KA: 295009K202 [3.9/3.9]
295009K202 ..(KA's) ;
ANSWER: 044 (1.00)
i
- '
'i REFERENCE:
, Grand Gulf: OP-LO-SYS-LP-E22-2-03, p. 26, LO KA: 239002A308 (3.6/3.6]
'
239002A308 ..(KA's)
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ANSWER: 045 (1.00) i REFERENCE: Grand Gulf: OP-LO-SYS-LP-M71-04, p. 39, LO S KA: 204000K111 [3.5/3.7]
204000K111 ..(KA's)
ANSWER: 046 (1.00) I
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' REACTOR OPERATOR Page 63-REFERENCE: Grand Gulf: OP-LO-SYS-LP-E51-03, p. 48, LO 6 KA: 217000A301 [3.5/3.5]
217000A301 ..(KA's)
.'. ANSWER: 047 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-E22-1-04, p. 35, LO 6 KA: 209002A301 [3.3/3.3]
209002A301 ..(KA's)
ANSWER: 048 (1.00) REFERENCE: Grand Gulf: 04-1-01-E12-1, rev. 49, p. 5, step 3.6.3 . Grand Gulf: OP-LO-SYS-LP-E12-06, LO 14 KA: 205000G010 [3.2/3.3]
205000G010 ..(KA's)
ANSWER: 049 (1.00) '"
/
) )
' REACTOR OPERATOR Page 64 REFERENCE: l
' Grand Gulf: OP-LO-SYS-LP-M71-04, p. 59, LO S KA: 223002K404 [3.2/3.6]
223002K404 ..(KA's)
ANSWER: 050 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-E21-05, p. 31, LO 8 KA: 209001K404 [3.0/3.2]
209001K404 ..(KA's)
ANSWER: 051 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-N19-04, p. 15-16, LO 4 KA: 256000K201 [2. 7/2. 8]
256000K201 ..(KA's)
ANSWER: 052 (1.00) ;
i i
l l
' REACTOR OPERATOR Page 65 REFERENCE: Grand Gulf: OP-LO-SYS-LP-D17-05, p. 14, LO 5 KA: 272000G007 [3.5/3.5]
272000G007 ..(KA's)
ANSWER: 053 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-Z51-02, p. 27, LO S KA: 290003K401 [ 3 .1/ 3 . 2 )
29000:4K403 ..(KA's)
ANSWER: 054 (1.00) REFERENCE: Grand Gulf: OP-LP-SYS-LP-R21-05, p. 13, LO 3 KA: 262001K201 [3.3/3.6]
262001K201 ..(KA's)
ANSWER: 055 (1.00) ) ]
!
)
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REACTOR OPERATOR Page 66-
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REFERENCE: i i Grand Gulf: OP-LO-SYS-LP-E51-03, p. 13, LO 9 KA: 239001K119 [3.1/3.2]
239001K119 ..(KA's) !
ANSWER: 056 (1.00) .
REFERENCE: Grand Gulf: '05-1-02-I-4, p. 3 !
2 Grand Gulf: OP-LO-DT-LP-029-01, LO ,
'
KA: 295003G007 [3.2/3.6]
,
295003G007 ..(KA's)
,
ANSWER: 057 (1.00) ,
!
. !
REFERENCE: l
- Grand Gulf
- Technical Specification 3/4.6.3, p. 6-2 KA: 295013G003 [3.3/4.2]
295013G003 ..(KA's)
!
ANSWER: 058 (1.00) , ;
.:
l
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i
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. . . - , - ..- - - . _ . - .. - . _ _ _ .
_ - - _- _ ___ __-________-_--___ _ ___-____- ____-___ . . _ _ _ _ _ _ _ _ _ _ _ _ - - .
' REACTOR OPERATOR A" O Page 67 REFERENCE: Grand Gulf: OP-LO-EP-LP-006-01, p. 11, LO KA: 295032K303 [3. 8 /3. 9)
295032K303 ..(KA'o)
ANSWER: 059 (1.00) REFERENCE: Grand Gulf: 05-1-02-II-8, .
KA: 295023G010 [3.8/3.9]
295023G010 ..(KA's)
ANSWER: 060 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-004-02, p. 10, LO KA: 295037K't . 3 [ 3 . 4 /4 .1] l l
295037K213 ..(KA's)
ANSWER: 061 (1.00) l )
i
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..
. _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
A
- O,-
- REACTOR OPERATOR Page 68 REFERENCE: Grand Gulf: 05-1-02-IV-1, p. KA: 295022A201 [3.5/3.6]
295022 A201 ..(KA's)
ANSWER: 062 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-005-02, p. 10, LO KA: 295012K202 [3.6/3.7]
295012K202 ..(KA's) ;
i
ANSWER: 063 (1.00) I REFERENCE: Grand Gulf: Technical Specification, Table 1.2, 3/4.6. j l
KA: 295021A201 [3.5/3.6] l 295021A201 ..(KA's) ;
ANSWER: 064 (1.00) _____ _ _ _ _ _ - - _ _ _
_ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - -
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' REACTOR OPERATOR Page 69 REFERENCE: Grand Gulf: 03-1-01-1, step 2. KA: 295014A202 [3.9/3.9]
295014A202 ..(KA's)
ANSWER: 065 (1.00) REFERENCE: Grand Gulf: 05-1-02-V-11, p. 1 KA: 295018K302 [3.3/3.4]
295018K302 ..(KA's)
ANSWER: 066 (1.00) REFERENCE: Grand Gulf: 05-1-02-III-5, p. 1 KA: 295024K217 [3.0/3.3]
295024K217 ..(KA's)
ANSWER: 067 (1.00) {
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- REACTOR OPERATOR Page 70
. REFERENCE: Grand Gulf: OP-LO-SYS-LP-C71-03, p. 13, LO 3 KA: 295008K302 [3.6/3.9]
295008K302 ..(KA's)
ANSWER: 068 (1.00) REFERENCE: Grand Gulf: EP-3 Grand Gulf: OP-LO-EP-LP-005-02, LO KA: 295029G011 [4.2/4.5)
295029G011 ..(KA's)
ANSWER: 069 (1.00) REFERENCE: Grand Gulf: 05-01-02-III-3, step 4. KA: 295001A106 [3.3/3.4) I 295001A106 ..(KA's)
ANSWER: 070 (1.00) I l
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Page 71 REFERENCE: Grand Gulf: OP-LO-SYS-LP-M71-04, p. 79, LO 9 KA: 295004A204 [3.2/3.3)
295004 A204 ..(KA's)
ANSWER: 071 (1.00) REFERENCE: Grand Gulf: SD G33/G36, p. 45-4 . Grand Gulf: OP-LO-SYS-LP-G33/36-05, LO 10 KA: 295008K209 [3 .1/ 3 .1]
295008K209 ..(KA'e)
ANSWER: 072 (1.00) REFERENCE: Grand Gulf: 05-1-02-I-2, KA: 295005G010 [3. 8/3. 6)
295005G010 ..(K\'s)
ANSWER: 073 (1.00) I i
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REACTOR OPERATOR o O Page 72
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q REFERENCE: Grand Gulf: OP-LO-SYS-LP-C61-00, Table 3, LO 3 KA: 295016G006 [4.1/4.1] j l
295016G006 ..(KA's)
' ANSWER: 074 (1.00) REFERENCE: Grand Gulf: 04-1-01-P53-1, p. 10 - Grand Gulf: OP-LO-SYS-LP-P53-04, LO 9a KA: 295019K218 [3.5/3.5]
295019K218 ..(KA's)
t ANSWER: 075 (1.00) REFERENCE: Grand Gulf: Loss of Instrument Air, 05-1-02-V-9, page 1 e KA: 295019G010 [3.7/3.4] '
.295019G010 ..(KA's)
ANSWER: 076 (1.00) ; .
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Page 73
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- REACTOR OPERATOR REFERENCE: Grand Gulf: OP-LO-EP-LP-006-01, L.O. 4, page 6 KA: 295036G011 [3. 8 /4 .1)
295036G011 ..(KA's)
ANSWER: 077 (1.00) REFERENCE: Grand Gulf: 05-1-02-II-1, p. 10 Grand Gulf: OP-LO-SYS-LP-C61-00, LO 4 KA: 295016K201 [4.4/4.5]
295016K201 ..(KA's)
ANSWER: 078 (1.00) REFERENCE: Grand Gulf: 05-1-02-V-5, Section KA: 295014G010 [4.0/3.9)
295014G010 ..(KA's)
ANSWER: 079 (1.00) ._ - - _-. .
- REACTOR OPERATOR a"
O Page 74 REFERENCE: Grand Gulf: 05-1-02-V-1, " Loss of Component Cooling Water", page 1 KA: 295018K303 [ 3 .1/ 3 . 3 ]
295018K303 ..(KA's)
ANSWER: 080 (1.00) REFERENCE: Grand Gulf: Decrease in Recirculation System Flow Rate, 05-1-02-III-3, page KA: 295001A201 [3. 5/3. 8]
295001A201 ..(KA's)
ANSWER: 081 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-C11-1A, L.O. 3c and 3f KA: 295025A107 [4 .1/4 .1]
295025A107 ..(KA's)
ANSWER: 082 (1.00) l i
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- REACTOR OPERATOR Page 73 REFERENCE: Grand Gulf: OP-LO-EP-LP-004-02, , LO KA: 295014K106 [3.8/3.9]
295014K106 ..(KA's)
ANSWER: 083 (1.00) REFERENCE: Grand Gulf: 04-1-01-E30-1, p . 1, step KA: 295030A104 [4. 0/4. 0]
295030A104 ..(KA's)
ANSWER: 084 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-003-02, p. 28, LO KA: 295031K101 [4. 6/4. 7]
295031K101 ..(KA's)
ANSWER: 085 (1.00) I l
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i REFERENCE:
- Grand Gulf: OP-LO-EP-LP-003-02, p. 29, LO KA: 295031K201 [4.4/4.4]
?
295031K201 ..(KA's)
i ANSWER: 086 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-B21-05, p. 17, LO 3 KA: 295009K201 [3.9/4.0) .,
295009K201 ..(KA's)
ANSWER: 087 (1.00) , ,
>
REFERENCE: Grand Gulf: OP-LO-SYS-LP-B33-1-08, p. 80, LO 30 KA: 295006K306 [3.2/3.3] )
295006K306 ..(KA's)
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ANSWER: 088 (1.00) !
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REACTOR OPERATOR Page 77 REFERENCE: Grand Gulf: 05-1-02-I-4, p. . Grand Gulf: Facility Question ONEP-1 KA: 295003K202 [4 .1/ 4 . 2 ]
295003K202 ..(KA's)
ANSWER: 089 (1.00)
c.
REFERENCE: Grand Gulf: Conduct of operations procedure 01-S-06-2 , 6.5. . Grand Gulf: OP-LO-AD-LP-001-05, LO D.21 KA: 294001K116 [3.5/3.8]
294001K116 ..(KA's)
ANSWER: 090 (1.00)
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REFERENCE: l Grand Gu]t: 01-S-06-2, 6.4.5 and 6. ' Grand Gulf: OP-LO-AD-LP-001-05, " Conduct of Operations," L.O. D.1 KA: 294001A109 [3.3/4.2]
294001A109 ..(KA's)
ANSWER: 091 (1.00) [+1.0)
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- REACTOR OPERATOR Page 78 REFERENCE: Grand Gulf: 02-S-10-4, . Grand Gulf: OP-LO-AD-LP-001-05, " Shift Relief and Turnover," LO KA: 294001A106 [3.4/3.6)
294001A10C ..(KA's)
ANSWER: 092 (1.00) REFERENCE: Grand Gulf: 01-S-08-2, 6.3. . 10CFR20.101, " Radiation Dose Standards for Individuals in Restricted Areas" KA: 294001K103 [3.3/3.8)
294001K103 ..(KA's)
ANSWER: 093 (1.00) REFERENCE: I Grand Gulf: 01-S-06-4, p. 3 2 Grand Gulf: OP-LO-AD-LP-001-05, LO C7 KA: 294001A111 [3.3/4.3]
294001A111 ..(KA's)
ANSWER: 094 (1.00) !
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- REACTOR OPERATOR Page 79
REFERENCE: ; Grand' Gulf: 02-S-01-9, 6. KA: 294001K102 [3.9/4.5]
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294001K102 ..(KA's)
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ANSWER: 095 (1.00)
b REFERENCE: Grand Gulf: Technical Specification, Table 6. .
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K/A: 294001A103 [2. 7/3. 7]
294001A103 ..(KA's) '
ANSWER: 096 (1.00) ;
d J REFERENCE: Grand Gulf: Conduct of operations 01-S-06-2, Attachment I K/A: 294001A111 [3.3/4.3]
294001A111 ..(KA's)
ANSWER: 097 (1.00) ; . . _ - - . ... .- - -. - -
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- REACTOR OPERATOR Page 80 REFERENCE: Grand Gulf: Procedure 01-S-02-1, page 19 Grand Gulf: OP-LO-AD-LP-001-05, L.O. A.15
[2.9/3.4]
KA: 294001A101 [2.9/3.4]
,
294001A101 ..(KA's) !
ANSWER: 098 (1.00) REFERENCE: Grand Gulf: Independent Verification Program, 01-S-06-29, 6. . Grand Gulf: OP-LO-AD-LP-001, L.O. E3 KA: 294001K101 [3.7/3.7]
294001K101 ..(KA's)
ANSWER: 099 (1.00) REFERENCE: Grand Gulf: Protective Tagging System, 01-S-06-1, page 18 ; Grand Gulf: OP-LO-AD-LP-001-05, L.O. J.22 KA: 294001K102 [3.9/4.5]
294001K102 ..(KA's)
ANSWER: 100 (1.00) i i
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O s-Page 81 REFERENCE: Grand Gulf: Exposure and Contamination control, 01-S-08-2, page 10 KA: 294001K103 [3.3/3.8]
294001K103 ..(KA's)
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TEST CROSS REFERENCE Page 1 RO Exam BWR Reactor Orqanized by Quest ion Number I QUESTION VALUE REFERENCE 001 1.00 9000001 002 1.00 9000002 003 1.00 9000003 004 1.00 9000005 005 1.00 9000006 006 1.00 9000007 007 1.00 9000008 008 1.00 9000009 ,
009 1.00 9000010 010 1.00 9000011 011 1.00 9000012 012 1.00 9000013 013 1.00 9000014 014 1.00 9000015 015 1.00 9000017 016 1.00 9000018 017 1.00 9000019 018 1.00 9000020 019 1.00 9000022 020 1.00 9000023 021 1.00 9000024 022 1.00 9000025 023 1.00 9000026 024 1.00 9000028 025 1.00 9000029 026 1.00 9000030 027 1.00 9000031 028 1.00 9000032 029 1.00 9000033 030 1.00 9000035 031 1.00 9000037 032 1.00 9000038 033 1.00 9000042 034 1.00 9000044 035 1.00 9000045 036 1.00 9000046 037 1.00 9000047 038 1.00 9000048 039 1.00 9000049 040 1.00 9000050 041 1.00 9000052 042 1.00 9000053 043 1.00 9000054 044 1.00 9000055 045 1.00 9000056 046 1.00 9000057 047 1.00 9000058 048 1.00 9000059 049 1.00 9000060
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TEST CROSS REFERENCE Page 2 RO Exam BWR Reactor Organized by Quest ion Number QUESTION VALUE REFERENCE 050 1.00 9000061 051 1.00 9000062 052 1.00 9000063 053 1.00 9000064 054 1.00 9000065 055 1.00 9000066 056 1.00 9000067 057 1.00 9000068 058 1.00 9000070 059 1.00 9000071 060 1.00 9000073 061 1.00 9000074 062 1.00 9000075 063 1.00 9000076 064 1.00 9000077 065 1.00 9000079 066 1.00 9000081 067 1.00 9000082 068 1.00 9000083 069 1.00 9000087 070 1.00 9000088 071 1.00 9000089 072 1.00 9000091 073 1.00 9000092 074 1.00 9000093 075 1.00 9000094 076 1.00 9000095 077 1.00 9000096 078 1.00 9000097 079 1.00 9000098 080 1.00 9000099 081 1.00 9000101 082 1.00 9000102 083 1.00 9000103 084 1.00 9000104 085 1.00 9000105 086 1.00 9000107 087 1.00 9000108 088 1.00 9000109 089 1.00 9000111 090 1.00 20940 091 1.00 9000114 092 1.00 22789 093 1.00 9000118 094 1.00 9000120 095 1.00 9000122 096 1.00 9000123 097 1.00 9000125 098 1.00 9000126
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TEST CROSS REFERENCE e*
Page 3 RO Exam BWR Reactar Organized by Quest ion Number QUESTION VALUE REFERENCE 099 1.00 9000127 100 1.00 9000130
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100.00
@ m & W e e 100.00 l
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Organized by KA Group l
PLANT WIDE GENERICS 4 QUESTION VALUE KA 097 1.00 294001A101 095 1.00 294001A103 i 091 1.00 294001A106 090 1.00 294001A109 093 1.00 294001A111'
096 1.00 294001A111 098 1.00 294001K101 099 1.00 294001K102 094 1.00 294001K102 100 1.00 294001K103 092 1.00 294001K103 014 1.00 294001K105 089 1.00 294001K116
_.____
PWG Total 13.00 PLANT SYSTEMS Group I
,
QUESTION VALUE KA 010 1.00 201001K408 012 1.00 202002A108 040 1.00 202002A204 023 1.00 209001A403 050 1.00 209001K404 047 1.00 209002A301 ,
007 1.00 209002A415 042 1.00 211000A308 004 1.00 212000K412 020 1.00 212000K602 035 1.00 215004K101 018 1.00 215004K401 016 1.00 215005K113 021 1.00 215005K501 034 1.00 215005K604 4 011 216000K513
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1.00 046 1.00 217000A301 032 1.00 217000G010 024 1.00 217000K601 008 1.00 218000A402 005 1.00 223002A101 015* 1.00 223002K403 049 1.00 223002K404 044 1.00 239002A308
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l RO Exam BWR Reactor Organized by KA Group
' PLANT SYSTEMS Group I QUESTION VALUE KA 027 1.00 259002K302 025 1.00 261000G007 I 036 1.00 264000A203 i 013 1.00 264000K402
______
PS-I Total 28.00 Group II QUESTION VALUE KA
,
002 1.00 201003G005 003 1.00 201003G007 039 1.00 201003K101 030 1.00 202001A210 001 1.00 202001K127 006 1.00 204000A205 045 1.00 204000K111 048 1.00 205000G010 ,
041 1.00 226001K409 055 1.00 239001K119 026 1.00 256000G010 051 1.00 256000K201 '
029 1.00 262001A304 054 1.00 262001K201 038 1.00 262001K502 031 1.00 271000K408 052 1.00 272000G007 028 1.00 286000K401 053 1.00 290003K401
......
PS-II Total 19.00 Group III
QUESTION VALUE KA 019 1.00 233000K102 009 1.00 234000A302 022 1.00 268000G007 017 1.00 288000K103
._____
PS-III Total 4.00
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O~' TEST CROSS REFERENCE LO Page 6 RO Exam BWR Reactor Organized by KA Group i
PLANT SYSTEMS QUESTION VALUE KA PS Total 51.00
EMERGENCY PLANT EVOLUTIONS Group I VALUE KA
'
QUESTION 072 1.00 295005G010 087 1.00 295006K306 086 1.00 295009K201 ;
043 1.00 295009K202 064 1.00 295014A202 078 1.00 295014G010 082 1.00 295014K106 037 1.00 295015K204 066 1.00 295024K217 081 1.00 295025A107 084 1.00 295031K101 085 1.00 295031K201 060 1.00 295037K213
..__ EPE-I Total 13.00 ;
Group II !
QUESTION VALUE KA 069 1.00 295001A106 j 080 1.00 295001A201 :
033 1.00 295002K205 056 1.00 295003G007 088 1.00 295003K202 070 1.00 295004A204 ,
071 1.00 29500BK209 !
067 1.00 295008K302 ,
062 1.00 295012K202 ;
057 1.00 295013G003 !
073 1.00 295016G006 l 077 1.00 295016K201 '
065 1.00 295018K302 079 1.00 295018K303 075 1.00 295019G010 074 1.00 295019K218 061 1.00 295022A201 068 1.00 295029G011 l l
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TEST CROSS REFERENCE Page 7 RO Exam BWR Reactor Orqanized by KA Group EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA 083 1.00 295030A104
______
EPE-II Total 19.00 Group III QUESTION VALUE KA 063 1.00 295021A201 059 1.00 295023G010 058 1.00 295032K303 076 1.00 295036G011
______
EPE-III Total 4.00
______
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EPE Total 36.00
______
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Test Total 100.00
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M b REAdTOR OPERATOR Pags 1 ANSWER KEY MULTIPLE CHOICE 023 c 001 a 024 b 002 a 025 a 003 d 026 a 004 a 027 b 005 d 028 a 006 b 029 b 007 d 030 b 008 a 031 d 009 b 032 b 010 a 033 a 011 b 034 d 012 a 035 c 013 c 036 c 014 a 037 b 015 a 038 c 016 a 039 d 017 d 040 a 018 c 041 c 019 a 042 d 020 a 043 c 021 b 044 b 022 b 045 d
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a 9 REAdTOR' OPERATOR 9 ANSWER XEY 046 c 069 c 047 b 070 d 049 d 071 b 049 d 072 d 050 a 073 b 051 c 074 c 052 d 075 b 053 c 076 d 054 c 077 d 055 a 078 c 056 a 079 b 057 c 080 c 058 b 081 b 059 a 082 c 060 c 083 a 061 b 084 d 062 a 085 a 063 c 086 a 064 d 087 b 065 b 088 b 066 d 089 c 067 c 090 d 068 a 091 c
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.REAOTOR OPERATOR Pago 3 i
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ANSWER KEY
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l 092 b ,
i 093 a 094 d ;
095 b I 096 d .
097 b 098 b
.099 b 100 c
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NRC pfficial Use Only Sb 4 n h 0u /0 fd f Of c
/ L{ a.r N M 000 Nuclear Regulatory Commission Operator Licensing Examination This document is removed from Official Use Only category on date of examinatio NRC Official Use Only l
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' f U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE i l
REGION 2 CANDIDATE'S NAME:
FACILITY: Grand Gulf 1 REACTOR TYPE: BWR-GE6 DATE ADMINISTERED: 93/10/01 INSTRUCTIONS TO CANDIDATE:
Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheet Points for each question are indicated in parentheses after the questio The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE %
100.00 % TOTALS FINAL GRADE All work done on this examination is my ow I have neither given nor received ai Candidate's Signature t
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SENIOR REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blan MULTIPLE CHOICE 023 a b c d 001 a b c d 024 a b c d 002 a b c d 025 a b c d 003 a b c d 026 a b c d 004 a b c d 027 a b c d 005 a b c d 028 a b c d 5 ~
006 a b c d 029 a b c d 007 a b c d 030 a b c d , , _ _
008 a b c (' __
031 a b c d 009 a b c d 032 a b c d 010 a b c d 033 a b c d
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011 a b c d 034 a b c d _ , ,
012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c d 016 a b c d 039 a b c d 017 a b c d 040 a b c d 018 a b c d 041 a b c d 019 a b c d 042 a b c d 020 a b c d __
043 a b c d 021 a b c d 044 a b c d 022 a b c d 045 a b c d i
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. . . SENIOR REACTOR OPERATOR-O O Paga 3 A N S-W E R SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in'the blan a b c d 069 a b c -d 047 a b c d 070 a b c d 048 a b c d 071 a b c d 049 a b c d 072 a b c d 050 a b c d 073 a b c d 051 a b c d 074 a b c d ,
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l 052 a b c d 075 a b c d 1 053 a b c d 076 a b c d 054 a b c d 077 a b c d 055 a b c d 078 a b c d 056 a b c d 079 a b c d ,
057 a b c d - - _ . .
080 a b c d l 058 a b c d 081 a b c d l
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059 a b c d 082 a b c d 060 a b c d 083 a b c d l
061 a b c d 084 a b c d 062 a b c d 085 a b c d 063 a b c d 086 a b c d 064 a b c d 087 a b, c d
065 a b c d 088 a b c d ,
066 a b c d 089 a b c d l 061 a b c d 090 a b c d 068 a b c d 091 a b c d
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SEMIOR REACTOR OPERATOR PCgs 4 ANSWER SHEET Multiple choice (Circle or X your choice)
If you change your answer, write your selection in the blan a b c d 093 a b c d 094 a b c d 095 a b c d 096 a b c d 097 a b c d
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098 a b c d 099 a b c d 100 a b c d
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(********** END OF EXAMINATION **********)
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e O Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . After the examination has been completed, you must sign the statetent on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be done after you complete the exam-inatio . Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating, i Use black ink or dark pencil only to facilitate legible repro- l duction I Print your name in the blank provided in the upper right-hand corner of the examination cover shee . Fill in the date on the cover sheet of the examination (if ;
necessary). Print your name in the upper right-hand corner of the first page of each section of your answer sheet . Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question pag . The point value for each questi n is indicated in parentheses after the questio . Partial credit will NOT be give . If the intent of a question is unclear, ask questions of the examiner onl . When you are done and have turned in your examination, leave the examination area as defined by the examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoke ,
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., SEN,IOR'. REACTOR OPERATOR Page 7.-
. QUESTION: 001 '(l' 00)-
.
The plant is operating at 100% power. A transient occurs that trips the Recirculation Pumps. The following Recirculation Pump breaker indication are observed:
CB 1 A/B OPEN CB 2 A/B OPEN CB 3 A/B CLOSED CB 4 A/B OPEN CB 5 A/B OPEN WHICH ONE (1) of the following is the cause of the Recirculation Pump >
Trip (RPT)? ,
,
c. Low feed water flo , CB 5 A/B tripped on over-curren .
QUESTION: 002 (1.00)
WHICH ONE (1) of the following design features of the Control Rod Drive mechanism " Flange" allows the control rod to be scrammed even if its associated scram inlet valve fails to open?
I a. Collet. piston ~
b. Stop piston c. Cooling water orifice d. Ball check valve q
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Page 0 QUESTION: 003 (1.00)
WHICH ONE (1) of the following is the bases for maintaining the fuel storage pool 22 feet 8 inches above the top of the reactor pressure vessel flange during refueling?
a. To provide adequate net positive suction head to the Fuel Pool Cooling Cleanup Pump b. To maintain a reservoir of water for suppression pool makeup, c. To provide spent fuel decay heat removal for 7 days without makeu d. To remove the iodine gap activity released from a fuel ruptur QUESTION: 004 (1.00)
The indication for the G33-F004 (RWCU Isolation valve) has a flashing red light on the Isolation Valve Status Panel. WHICH CNE (1) of the following is the reason for this indication?
a. The valve has spuriously closed, b. The valve is in the intermediate positio '
c. An isolation signal is present and the valve is close d. An isolation signal is present and the valve is ope lf /'}
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SENIOR REACTOR OPERATOR Page 9 QUESTION: 005 (1.00)
The plant is operating at low power. Reactor pressure is being controlled via the By-pass valves. The Reactor Water Cleanup System is being used to reject water from the reactor vessel to the Main Condenser. WHICH ONE (1) of the following is the consequence of opening the G33-F035, RWCU BLOWDOWN VALVE TO RAD WASTE?
a. Over-pressurizing the rad waste pipin Loss of condenser vacuu c. RWCU pumps will tri d. Damage to Filter /Deminerlizers QUESTION: 006 (1.00)
HPCS automatically initiated on low water level and high drywell pressure. The HPCS injection valve, E22-F004, automatically closes on high reactor water level. Reactor water level is now -2 inches and decreasing and the high drywell signal is still present. WHICH ONE (1)
of the following operator actions must be taken to re-inject water using HPCS before the water level reaches level 2 (-41. 6 inches) ? (Assume no other operator actions will be taken.)
, a. Reset the automatic initiation logi b. Arm and depress HPCS Manual Initiation pushbutto c. Place the F004 hand switch to the OPEN positio d. Reset the high water level logi _ _ . _ _ . . . _ .. . . _ _ _ . _ _ . .
, SEN,IOR~ REACTOR OPERATOR O O Page,10
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QUESTION: 007 (1.00) :
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WHICH ONE'(1) of the.following refueling conditions will-result in a rod block with the Mode Switch in the REFUEL position?-
a. The refueling platform is positioned over the reactor core and one control rod is withdraw b. The refueling platform is positioned over the reactor core and the main hoist fuel grapple is loade t c. The refueling platform is positioned away from the reactor core and the main hoist fuel grapple is loade d. The refueling platform is positioned away from the reactor core-and one control rod is withdraw !
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QUESTION: 008 (1.00) {
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You are unable to drive one control ro ONEP-05-1-02-IV-1. states;
" Increase drive water pressure in 25 psi increments and attempt to drive ,
the affected rod."
WHICH ONE (1)~of the following describes the action required to-increase 5 drive water pressure? ,
a. Throttle valve F003 is throttled close b. Bypass valve F004 is throttled opene c. Stabilizing valves are opene d. Flow control valves are close I i
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- l QUESTION: 009 (1.00) ;
A reactor scram has occurred following a recirculation loop suction line LOC The STA has determined that you are in the unsafe region of the i RPV Saturation Temperature curve of EP-2, "RPV Control". .WHICH ONE (1) ,
of the following describes the relationship between indicated and actual reactor vessel level? Assume vessel level is above TA '
a. Actual reactor vessel water level LESS than indicated due to variable leg flashin '
b. Actual reactor vessel water level LESS than indicated due to reference leg flashin c. Actual reactor vessel water level GREATER than indicated due to reference leg flashin >
d. Actual reactor vessel water level GREATER than indicated due to variable leg flashin QUESTION: 010 (1.00)
The plant is experiencing a small break LOCA. The Feedwater pumps have tripped off and_the reactor water level is being maintained at +9 inches '
with RCIC. Drywell pressure is stable at 1.57 psig. The Recirculation
' Flow Control Valve "A" HPU oil temperature is 142 deg WHICH ONE (1) of the following has caused a Recirculation Flow Control i Valve "A" motion inhibit? !
a. High drywell pressure, Loss of both feed pump ,
c. High HPU oil temperatur d. Low reactor water leve !
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, SEN,IOR REACTOR OPERATOR Page 12 QUESTION: 011 (1.00)
The HPCS diesel has starten as a result of a LOCA. WHICH ONE (1) of the following will trip the HPCS Diesel Engine with the LOCA signal still present?
a. High jacket water temperatur b. Low lube oil pressur c. Generator differential curren d. Overcurrent with voltage restrain QUESTION: 012 (1.00)
WHICH ONE (1) of the following is the MAXIMUM number of visitors that can be escorted by one escort in the Vital Area?
a. 2 b. 5 c. 7 d. 10 QUESTION: 013 (1.00)
WHICH ONE (1) of the following is the condition that exists when the Suppression Pool level is at 14 feet?
a. Suppression Pool level cannot be measure b. The upper most horizontal vents are uncovere c. Suppression Pool temperature cannot be determine d. The RCIC turbine exhaust is uncovere O
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, SEN,IOR REACTOR OPERATOR Page 13 QUESTION: 014 (1.00)
WHICH ONE (1) of the following functions /isolations is initiated when all four NSSSS manual isolation pushbuttons are armed and depressed?
a. 30 minute timer for suppression pool make-up dump logi b. 105 second timer for ADS activation logic, c. RCIC exhaust vacuum breaker isolatio d. LPCS test line isolatio QUESTION: 015 (1.00)
WHICH ONE (1) of the following is used to calibrate LPRM detectors?
a. Flux profiles from TIP trace b. Comparison made with LPRM detectors in other quadrant c. Neutron to gamma signal ratio tes d. Comparison made with associated APRM outpu QUESTION: 016 (1.00)
WHICH ONE (1) of the following areas will have its ventilation system isolated on an initiation signal to the Standby Gas Treatment System?
a. Enclosure building l l
b. Radwaste building c. Control Room d. Auxiliary building I
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. SEN,IOR REACTOR: OPERATOR Page 14~
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-QUESTION: 017 (1.00)
WHICH ONE (1) of the following conditions will result in a Rod Block?
a.-IRMs on Range 2 Mode Switch in RUN-SRM count is 0.5 cps b. IRMs on Range 9 Mode Switch in RUN '
SRM count is 1.5 x 10E5 cps c. IRMs on Range 2 Mode Switch in STARTUP SRM count is 0.5 cps
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d. IRMs on Range 9 Mode Switch in STARTUP SRM count is 1.5 x 10E5 cps
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QUESTION: 018 (1.00)
WHICH ONE (1) of the following is the basis for maintaining the containment average temperature below 90 deg F?
a. Ensures that mechanical equipment in the containment does not degrade due to high temperature during the design life of the plan b. Ensures that electrical components in the containment do not degrade due to high temperature during the design life of the .;
plan c. Ensures personnel access to safety related equipment in the containment without risk of heat stres d. Ensures that the containment peak air temperature during a LOCA does not exceed the design temperature of the containmen . - . - . . - . .- _- ~ ..
, SEN,IOR REACTOR OPERATOR Page 15 QUESTION: 019 (1.00)
WHICH ONE (1) of the following conditions will cause a half-scram?
a. APRM E is downscale, IRM G is INOP, mode switch is in STARTU b. IRM H reading 35/125ths on range 4, mode switch is in STARTUP, detector is selected and withdraw c. APRM F is downscale, IRM D is bypassed, mode switch is in RU d. IRM B reading 115/125ths of scale, mode switch is in STARTU QUESTION: 020 (1.00)
WHICH ONE (1) of the following Radwaste Tanks receives high quality, low conductivity wastes?
a. RWCU phase separator Tank b. Equipment Drain Collector Tank c. Floor Drain Collector Tank d. Miscellaneous Waste Receiver Tank l
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QUESTION: 021 (1.00)
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During a reactor plant heatup, MSIVs were stroked at 800 psig in-accordance with a plant surveillance procedure. Both B21-F022A and B21-F028A close in 2.3 and 2.5 seconds respectively, failing the surveillanc During the following 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B21-F028A is repaired and returned to operable status. Estimated repair time for B21-F022A is 2 ,
hour WHICH ONE (1) of the following is the required action'for the current conditions? j t
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a. Continue operation at the present power level until B21-F022A can be repaire '
b. Isolate the affected main steam line by use of a deactivated MSIV in the closed position, c. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN t within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ; .
d. Be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, HOT SHUTDOWN l within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and, COLD SHUTDOWN within the j subsequent 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l i
'i QUESTION: 022 (1.00) :
A total loss of AC power occurs with Div. I and Div. II ESF buses _NOT energize RCIC is manually initiated and maintaining reactor water .
leve WHICH ONE (1) of the following is the effect that a loss of AC i
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power will have on the RCIC suction transfer due to a high suppression pool level? !
a. The CST will drain to the suppression poo b. The suction transfer will occur normall c. The suction path will be isolate ,
d. The suction path will only be available from the CS l
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QUESTION: 023 (1.00) .
WHICH ONE (1) of the following is the purpose of the charcoal adsorber in the standby gas treatment exhaust filter train?
a. Removes iodin Condenses entrained moistur [
i c. Removes particulate matte ,
d. Recombines hydroge QUESTION: 024 (1.00)
The plant is at 100% power, with RFPTs A, and B in service and operating in three element mode. The M/A Station for RFPT B output signal suddenly fails to zero outpu WHICH ONE (1) of the following is-the impact on RFPT B speed?
a. Automatically ramps up to the high-speed sto b. Remains at its last called-for valu c. Automatically ramps down to the minimum called for by the electric automatic positioner (~3000 rpm).
d. Drops to near zero as the governor valve fully close '
l QUESTION: 025 (1.00)
I A loss of coolant accident (LOCA) has occurred and the appropriate loads l have been shed. WHICH ONE (1) of the following components is the FIRST i to be re-energized? j l
a. Standby Service Water Pump A ,
b. Enclosure Building Recirculation Fan A c. Control Room Air Handling Unit A d. Diesel Generator Room Outside Air Fan A
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SENJOR REACTOR OPERATOR Page 18 QUESTION: 026 (1.00)
The plant is operating at rated conditions. An electrician reports that the average specific gravity of all the connected cells for Battery 1A3 is 1.2 Five cells are reported to have specific gravities 1.18 WHICH ONE (1) of the following is the MAXIMUM time limit to restore the battery to operable limits before reactor shutdown must begin? hours hours days days QUESTION: 027 (1.00)
The following parameters exist on Reactor Recirculation Pump A:
Seal cavity #2 pressure is 400 psi Annunciator RECIRC PMP A OUTR SEAL LEAK HI is alarme Annunciator RECIRC PMP A SEAL STG FLO HI/LO is alarmed on low flo WHICH ONE (1) of the following is the indicated failure? Failure of #1 sea Failure of #2 sea Failure of both seal Plugged restricting orific I l
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, SEN,IOR REACTOR OPERATOR 'Page 19 QUESTION: 028 (1.00)
A loss of coolant accident has occurred while the reactor was operating at 100% power. Reactor pressure is now 27 psig. Drywell pressure is 2.52 psig. The inboard MSIVs have been closed. Main steam line pressure is 15 psig. The Standby Gas Treatment' system has been secure The Main Steam Isolation Valves Leakage Control System (MSIV-LCS) is required to be put into servic '
WHICH ONE (1) of the following will prevent activation of the MSIV-LCS?
a. High reactor pressur b. No Standby Gas Treatment trains are runnin c. Outboard MSIVs are ope d. High drywell pressur QUESTION: 029 (1.00)
Following an Offgas Post-Treatment Radiation High-High-High condition the following valve positions are observe F016A/B, Condenser Drain open F023, Holdup Line Drain closed F034A/B, Cooler Condenser Drain closed j F045, Adsorber Bypass closed F053A/B, Adsorber Inlet open F054, Prefilter Inlet Drain open F060, Offgas Discharge to Radwaste Vent closed WHICH ONE (1) of the following valves is in an abnormal position as a result of the High-High-High radiation condition?
a. F016A/B F045 c. F053A/B F054 ;
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' QUESTION: 030 (1.00)
WHICH ONE (1) of the following is the reason ~that the RCIC is'not ,
operated below 2000 rpm? l l'
a. To prevent chattering of turbine exhaust check valv b. To avoid operation at turbine critical: spee c. To assure that net positive suction :bs maintaine <
d. To avoid " steam cutting" the governor valv l QUESTION: 031 (1.00)
The Standby Liquid Control (SLC) system is initiated following an ATW WHICH ONE (1) of the following conditions have to be met before the SLC system can be secured?
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a. All APRMs are downscal i l
b. All control rods are inserted to'or beyond position 0 H c. Suppression pool temperature is reduced to less than 110 deg d. The SLC tank level is reduced to below 1000 gallon QUESTION: 032 (1.00)
The ADS valves have automatically opened and the initiation signals are still present. WHICH ONE (1) of the following ADS controls, when actuated, will close the ADS valves?
a. Manual ADS Inhibit Switch b. Lo-Lo Set Logic Reset Pushbutton c. ADS Hi Drywell Pressure Reset Pushbutton d. ADS Logic Reset Pushbutton
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SENIOR REACTOR OPERATOR Page 21 QUESTION: 033 (1.00)
The Rod Action Control system limits rod withdrawal to four notches when reactor power is between low power set point and the high power set poin The Rod Action Control system is designed to prevent WHICH ONE (1) of the following thermal limits from being exceeded?
a. Average Planar Linear Heat Generation Rate (APLHGR)
b. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
c. Minimum Critical Power Ratio (MCPR)
d. Maximum Fraction of Limiting Power Density (MFLPD)
QUESTION: 034 (1.00)
The reactor is operating at 80% powe WHICH ONE (1) of the following trips will be the result of a loss of the flow control reference signal to an APRM? INOP b. Downscale c. Upscale neutron flux d. Upscale thernal power QUESTION: 035 (1.00)
WHICH ONE (1) of the following is the MAXIMUM continuous rating for Standby Diesel Generator 11 to maintain safe shutdown conditions?
a. 5000 KW KW KW KW l
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, SEN,IOR REACTOR OPERATOR Page 22 QUESTION: 036 (1.00)
WHICH ONE (1) of the following methods for control rod insertion during an ATWS require the reactor scram to be reset?
a. Venting CRD Overpiston Volumes b. Opening Individual Scram Test Switches c. Manually Venting the Scram Air Header d. Maximizing CRD Cooling Water Differential QUESTION: 037 (1.00)
WHICH ONE (1) of the following is the reason that the operator is directed to shutdown the reactor following an indicated jet pump failure?
a. There may be loose parts in the vesse b. There may be a reduced capability to reflood the core following a LOC c. There may be a violation of core thermal limit d. There may be a unobserved entry into the restricted regions of the power to flow ma QUESTION: 038 (1.00)
During a LOCA, the A RHR pump automatically starts in the LPCI mod The pump is then put in Manual Override with the LPCI initiation signals still present. WHICH ONE (1) of the following will restart the pump?
a. Arm and depress the LPCS/RHR A Manual Initiation Pushbutto b. Place the RHR A pump switch to AUT c. Automatic initiation of Containment Spra d. LPCI initiation signal clears and reoccurs before being rese _ _ _ _
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-QUESTION: 039 (1.00)
An ATWS is in progress and HPCS is maintaining reactor water leve The following is the status of the SLC system: ;
SLC OOSVC alarms are presen ,
SLC Out-of-Service switches (A/B) in INO SLC Pump Suction Valve Key Lock Test Switches (A/B) placed to TES Test Tank Outlet Valve (F031) ope The Control Room Operator has been directed to initiate SLC regardless of present status. WHICH ONE (1) of the following will-prevent the !
Standby _ Liquid Control system from injecting sodium pentaborate? ;
a. SLC Out-of-Service switch positio b. HPCS injection already in progres c. SLC Pump Suction Valve Key Lock Test Switch positio .i l
d. Test Tank Outlet Valve (F031) positio l-i
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QUESTION: 040 (1.00)
The reactor scrams from 100% power with the master level controller in three-element mode. Reactor water level subsequently decreases below level 3. WHICH ONE (1) of the following is the automatic response of the Feedwater Control System?
a. The level setpoint increases to 54 inches for 10 seconds and then the level controller shifts to manual, b. The level setpoint decreases to 18 inches and the master level controller shifts to single element mod c. The level setpoint increases to 54 inches for 10 seconds and decreases to 18 inche ,
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d. One RFPT trips and the other RFPT shifts to Startup Level Control with the setpoint at 54 inche _
. SENIOR REACTOR OPERATOR Page 24 QUESTION: 041 (1.00)
WHICH ONE (1) of the following actuttes the red light above a Safety Relief Valve Control Switch on panel 1H13-P601 to indicate that the valve is open?
a. Pilot solenoid valv b. Pressure switches on the discharge pipin c. Valve limit switche d. Temperature elements on the discharge pipin QUESTION: 042 (1.00)
WHICH ONE (1) of the following isolation signals will ONLY close the Reactor Water Cleanup (RWCU) Pump Suction Outboard Isolation Va3ve (G33-F004)?
a. Reactor water level b. RWCU high differential flo c. Standby Liquid Control "B" actuatio d. RWCU filter demin inlet high temperatur QUESTION: 043 (1.00)
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WHICH ONE (1) of the following is the MINIMUM RHR flow during normal RHR l Shutdown Cooling? I a. 1000 gpm b. 2000 gpm c. 3000 gpm d. 4000 gpm
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SEMIOR REACTOR OPERATOR Page 25 QUESTION: 044 (1.00)
WHICH ONE (1) of the following condensate system pumps is powered by electrical bus 13AD?
a. Condensate pump B b. Condensate pump C c. Condensate booster pump B d. Condensate booster pump C QUESTION: 045 (1.00)
WHICH ONE (1) of the following radiation monitoring subsystems use scintillation detectors in a shielded sampler so that background radiation levels will be minimized?
a. Main Steam Line b. Offgas Post Treatment c. Ventilation System Process Liquid QUESTION: 046 (1.00)
WHICH ONE (1) of the following conditions will result in the Control Room Ventilation System being isolated?
a. High inlet air freon concentration b. Smoke detected at the fan inle c. Reactor water level of -50 inche Loss of 480 volt AC powe h
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, SENIOR REACTOR OPERATOR Page 26 QUESTION: 047 (1.00)
During a loss of AC power, ESF div. I bus was de-energized for 3 minutes. WHICH ONE (1) of the following is the reason that all ESF di I fluid systems in operation prior to the power loss must be restarted per the applicable SOI?
a. To ensure that fill and vents are completed prior to restar b. To ensure that pump motors do not exceed their restart limit c. To ensure that systems are started in order of importanc d. To ensure that manual valve line-ups are performe i QUESTION: 048 (1.00)
The RCIC system is being flow tested while the reactor is operating at 100% power. The suppression pool temperature is increasing. WHICH ONE (1) of the following is the suppression temperature at which the mode switch must immediately be placed in SHUTDOWN? deg F deg F deg F deg F
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l QUESTION: 049 (1.00) ,
EP-3 " Con'taf ament Temperature Control" requires termination of ;
containment sprays if containment pressure is below.1.23.psi WHICH
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ONE (1) of the following describes the basis for this step? l l'
a. Rapidly cooling containment will drive the RPV saturation temperature into the unsafe region and make level indication t inaccurat Increase differential pressure between the Containment and the Auxiliary building threatens the integrity of the Auxiliary buildin c. Subsequent Emergency Depressurization may be required to stay in ,
the safe region of the HCL i d. Terminating containment spray at this pressure avoids containment failure due to negative pressur ;
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QUESTION: 050 (1.00)
i The tcilowing conditions exist:
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Reactor power is 20% ;
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High differential temperature condition in the Auxiliary Building due to a fir ;
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Main Steam Isolation valves have closed
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HPCS is maintaining RPV level '
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Rods are being inserted using CRD WHICH ONE (1) of the following systems should be isolated if they are I
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discharging into the Auxiliary building?
a. High Pressure Core Spray l l
b. Reactor Water rieanup ]
c. Control Rod Drive Fire Suppression i
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- QUESTION: 051 (1.00)
Core reload is in progress. The Continuous Air Monitors (CAM) alarm on-
'the refuel floor. WHICH ONE (1) of the following is the immediate operator action?- -!
a. Evacuate the refuel floo b. Isolate the Fuel Transfer Tub c. Start Standby _ Gas Treatment'.
l d. Place Containment Cooling Charcoal Filter Trains in Containment Cooling Mod QUESTION: 052 (1.00)
'WHICH ONE (1) of the following describes'why lowering RPV level during
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an ATWS increases the void fraction and reduces reactor power? Inhibits natural circulatio b. Prevents localized power peak ,
c. Increases moderator temperatur Increases steam removal rat ,
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.. SEN,IOR REACTOR OPERATOR Page 29
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QUESTION: 053 (1.00) :
The reactor is operating at 80% power. All control rods have been '
withdrawn at least 2 notches. One CRD pump is out of service for maintenance. The second CRD pump trips and cannot be immediately ,
i restarted. One scram accumulator is immediately declared inoperabl WHICH ONE (1) of the following conditions requires the operator to scram the reactor?
a. The CRD pumps cannot be restarted within 20 minute b. A second scram accumulator is declared inoperable, c. Three control rods begin to drif d. Control rod drive mechanism high temperature is observe i QUESTION: 054 (1.00) :
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i The reactor is operating at rated conditions. A sensing line has broken in the Drywel Drywell temperature is increasing. WHICH ONE (1) of ;
the following will require that the Drywell Cooling isolation interlocks j be defeated in order to establish drywell cooling? _,
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a. Reactor water level is -53 inche l
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b. Drywell pressure le 1.12 psi ,
t c. Drywell vent exhaust radiation is 15.2 mR/h ;
I d. Div 1 and Div 3 NSSS have been manually actuate !
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Following a complete loss of Shutdown Cooling, the bulk reactor coolant water temperature is increasing at 1 degree F every 5 minutes. The present reactor coolant temperature is 152 degrees F. WHICH ONE (1) of the following is the MAXIMUM time allowed to pass before primary containment integrity must be established?
a. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> hours c. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hours QUESTION: 056 (1.00)
WHICH ONE (1) of the following is the required action following a sustained reactor period of +30 seconds during a reactor startup?
a. Scram the reacto b. Manually drive all control rods and shutdown the reacto c. Stop control rod movement until the period increase Insert control rods until period increase QUESTION: 057 (1.00)
WHICH ONE (1) of the following is the most severely challenged from an inadvertent MSIV isolation during a high powered ATWS?
a. Fuel integrit j
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b. RPV integrit c. Containment integrit d. Core cooling integrit I i
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The reactor is operating at 100% powe The following alarms are received:
PSW RADIAL WELL TROUBLE PSW UNIT 1 HDR PRESS LOW PSW UNIT 1 HDR PRESS LOW-LOW The condenser vacuum begins to decrease slowly and the plant chillers are los WHICH ONE (1) of the following is the immediate operator action?
a. Scram the reactor, b. Reduce core flow to 50%, drive rods to less than 60% powe c. Close the first stage SJAE suction valve, N62-F003A/ d. Trip the main turbin QUESTION: 059 (1.00)
WHICH ONE (1) of the following conditions describe entry into steam cooling?
a. No systems are available for injection, reactor water level decreases below -210 inche b. No systems are available for injection, reactor pressure is above 176 psig and reactor water level decreases below -167 inche c. Automatic initiation of the Automatic Depressurization Syste d. Reactor water level cannot be determined and all SRVs are ope .. -- .- -. .- .- -- .. . . .. . . .
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QUESTION: 060 (1.00) i
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WHICH ONE (1) of'the following is the reason for the reactor! vessel high'
water level scram? i a. Prevent damage to turbine driven system :
b. Avoid water hammer in the main steam line !
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c. Offset the reactivity effect of cold feedwater injection'
d. Prevent damage to the steam dryer assembl >
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QUESTION: 061 (1.00)
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The reactor is in an ATWS condition. The reactor-operator is preparing-to initiate the SLC system. WHICH ONE (1) of the following is the ,
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reason for down shifting Recirculation Pumps prior to tripping the pumps? ;
a. Provide adequate mixing for boro ;
' Prevent main turbine tri c. Reduce mechanical stress on recirculation syste d. Avoid localized power spike ;
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' QUESTION: 062 (1.00)
The.following conditions are observed after a Loss of Coolant Accident:
Reactor Pressure 90 psig 166' elevation temp in the drywell 252 deg. F 139' elevation temp in containment 192 deg. F Upset. Range Level Indication 146_ inches Shutdevn Range Level Indication 121 inches ,
Wide Range Level Indication 10 inches *
Fuel Zone Range Indication -117_ inches WHICH ONE (1) of the following indicates actual level? (See' attachments)
a. Upset Range b. Shutdown Range c. Wide Range '
d. Fuel Zone Range !
QUESTION: 063 (1.00)
Following a small break LOCA, indicated wide range reactor level is -20" and slowly increasing due to RCIC injection from the CS Other plant parameters are as follows:
Suppression pool level 22 feet Suppression pool temp 140 deg. F Containment pressure 4.0 psig WHICH ONE (1) of the following identifies the MAXIMUM allowable reactor pressure. (See attachments)
a. 500 psig b. 600 psig c. 700 psig psig L
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QUESTION: 064 ( 1.00)
The reactor is operating at 100% power and one recirculation pump trips' .
WHICH ONE (1) of the following observations by the reactor operator defines THERMAL HYDRAULIC INSTABILITY?
a. Peak-to-peak APRM swings of 5% rated powe Peak-to-peak LPRM swings of 5 watts /sq c c. Unexplained sustained increase in APRM leve d. Oscillations of 5 inches in reactor water leve QUESTION: 065 (1.00)
WHICH ONE (1) of the following isolate on-loss of 125 VDC?
a. RCIC b. MSIVs c. MSL Drain Valves d. Auxiliary Building QUESTION: 066 (1.00)
The offsite radiation release rate is above the. Alert limit and the Turbine Building Ventilation has shutdown. .WHICH ONE (1) of the following is the reason EP-4 directs that the Turbine Building Ventilation be restarted?
a. To naintain positive pressure in the Turbine. Buildin b.-To preserve Turbine Building accessibilit c. To prevent another entry condition into EP-' d. To assure max safe temperature limits are not reache . SENIOR REACTOR OPERATOR Page 35 QUESTION: 067 (1.00)
The reactor is operating at 20% power. WHICH ONE (1) of the following should be immediately verified following a Turbine / Generator trip?
a. Turbine Aux Oil Pump starts, b. Recirculation Pumps shift to slow spee c. Reactor Scra d. Generator Output Breakers ope QUESTION: 068 (1.00)
WHICH ONE (1) of the following Main Steam Line Drain Line Isolation Valves has remote shutdown capability at Panel 295?
a. B21-F016 b. B21-F019 c. B21-F021 d. B21-F067A QUESTION: 069 (1.00)
The ADS air system header begins to leak and header pressure begins to decreas WHICH ONE (1) of the following is the MINIMUM header pressure before the ADS system is considered inoperable?
a. 170 psig b. 160 psig c. 150 psig d. 140 psig i
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.A complete loss of instrument air has occurred. No control rods are '
drifting. WHICH ONE (1) of the following is required in accordance with ,
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procedure 05-1-02-V-9,-" Loss of Instrument Air"?
a. Attempt to restore service air to the affected drywell coolers air-operated damper b. Determine-the cause for the loss of instrument air and attempt to restor ,
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c. Reduce power to less than 50% and check ~for thermal hydraulic instabilitie ,
d. Manually open the service water makeup valves to the fire water storage tan i
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QUESTION: 071 (1.00) .i RCIC was initiated prior to the Control Room being evacuated due to a fir WHICH ONE (1) of the following RCIC turbine isolation / trips can ,
be reset from the DIV I Remote Shutdown Panel?' l a. Turbine overspee I b. Low steam supply pressur c. High RCIC equipment area temperatur [
d. Low pump suct:.on pressur !
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. SENIOR REACTOR OPERATOR Page 37 QUESTION: 072 (1.00)
While operating the plant at 95% power, a valve in the extraction steam line to a feedwater heater inadvertently closes and cannot be reopene WHICH ONE (1) of the following actions should immediately be taken?
a. Manually insert control rods per STA recommendation b. Gradually open the bypass valve on the operating feedwater strin c. Reduce recirculation flow until thermal power is 75% of rate d. Initiate a manual scra QUESTION: 073 (1.00)
The plant is operating at 100% power when a complete loss of Component Cooling Water is experienced. As the Reactor operator you inserted a manual scram. WHICH ONE (1) of the following immediate operator actions should be taken?
a. Shift recirculation pumps from fast to slow spee b. Manually trip the recirculation pumps upon any increase in motor winding temperatur c. Close the FCVs to minimum upon any increase in recirculation pump temperature, d. Start the second CRD pump and maximize purge flow to recirculation pump seal _ .. -- _ . _ . , , . , , . . - . - . _ - - - - . . . - . . _. _ _ _ . _ _ . _ _ _ _ _ _ . -.__..____i
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. SENJOR REACTOR OPERATOR QUESTION: 074 (1.00)
The plant was operating at 100% power when the "A" recirculation pump tripped resulting in the following condition:
Reactor Power 60%
Core Flow 44%
Recirculation loop "B" flow 44,500 gp WHICH ONE (1) of the following is the appropriate action to be taken in this condition? (See attachment)
a. Insert a manual scram within one minut b. Increase recirculation loop "B" flow until core flow is greater than 45%.
c. Immediately reduce thermal power to within region III or IV d. CLOSE the "A" FC QUESTION: 075 (1.00)
You have entered EP-2, "RPV Control", and have executed the steps down to the decision diamond that asks, "IS SP WATER LEVEL ABOVE 10.5 FT?".
WHICH ONE (1) of the following is the reason for requiring the suppression pool level to be greater than 10.5 feet before depressurizing?
a. To ensure no ECCS pump exceeds the vortex limits during the reflood after the blowdow b. To ensure that the HCLL is not a limiting factor during the addition of heat to the suppression poo c. To ensure that the suppression pool level can be monitored during heat addition to the srppression poo d. To ensure the SRVs will not discharge steam directly into the containment airspac l
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. SENJOR REACTOR OPERATOR QUESTION: 076 (1.00)
The plant has scrammed due to high reactor pressure. The reactor pressure peaked at 1145 psig. WHICH ONE (1) of the following shows the expected status of the scram pilot solenoids, backup scram sole acids and ARI solenoids?
SCRAM PILOT BACKUP SCRAM ARI SOLENOIDS SOLENOIDS SOLENOIDS a. de-energized de-energized energized b. de-energized energized energized c. energized de-energized de-energized d. energized energized de-energized QUESTION: 077 (1.00)
WHICH ONE (1) of the following is the reason that ADS is inhibited before SLC initiation?
a. To prevent a decrease in natural circulatica resulting in inadequate boron mixin b. To prevent rapid cooldown during depressurization resulting in a reactivity excursio c. To prevent a rapid injection of cold, unborated water resulting in a rapid increase in powe d. To prevent an increase in natural circulation resulting in decreased voiding and an increase in powe . .. . . . .- -- -. .. . - .
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' O Page 40 QUESTION: 078 (1.00) *
The Suppression Pool Makeup System (SPMU) has actuated on a low-lo f suppression pool level and high drywell pressure. WHICH ONE (1) of the ,
following will allow manual closure of E30-F001A/B and E30-F002A/B, j Suppression Pool Dump Valves? >
a. The SPMU mode switch is placed in OF b. The low-low suppression pool level has cleare c. The 29 minute timer has timed ou ,
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d. The SPMU Dump test switches are placed in TES .
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QUESTION: 079 (1.00) ,
WHICH ONE (1) of the following is the significance of 62 feet in the Containment during Containment Flooding? > Places maximum stress on the Containment Structur b. Level at which the RPV should be vente c. Top of the weir wal '
d. Top of active fue QUESTION: 080 (1.00) i WHICH ONE (1) of the following is the consequence of NOT venting the RPV during Containment Flooding?
a. Suppresses RPV water level increas b. Over-pressurizes the RP c. Allows buildup of explosive gases in the RP d. Violates Maximum Containment Water Level Limit.
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-During execution of EP-4, " Auxiliary Building.and Radioactivity Release Control", entry into EP-2, "RPV Control", is directe WHICH ONE (1) of '
the'following is the reason for entering.RPV Control?
a. Reduce the energy of the primary syste y
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b. Maintain primary containment integrit .t c. Assure adequate core coolin .
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d. Monitor RPV parameter f i
QUESTION: 082 (1.00)
WHICH ONE (1) of the following is the purpose of the End-of-Cycle -'
Recirculation Pump Trip (EOC-RPT) following a Main Turbine trip?- i a. Protects the Recirculation Pumps during the transien b. Recovers the loss of therral margin at the end-of-cycl '
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c. Reduces the consequences of a failure to scram'at.the end-of-cycl ]
d. Compensates for the lower integral rod worth at the ;
end-of-cycl .;
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QUESTION: 083 (1.00)
A station blackout has occurred. WHICH ONE (1) of the following would prevent Div 1 ESF bus from being energized by the Div 3 diesel generator? ,
a. Drywell pressure is 1.54 psi b. Reactor water level is -52 inche c. An ATWS is in progres ,
d. Containment temperature is 94 deg f
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Page 42 QUESTION: 084 (1.00)
The Reactor Feedwater Pumps trip on high reactor water level. WHICH ONE (1) of the following describes when the pump discharge valves can be reopened?
a. When the reactor high level trip clears, b. When the two minute interlock times ou c. When the RFPT Trip Reset pushbutton is depresse d. When the Feedwater Turbine coasts dow QUESTION: 085 (1.00)
The fuel has been removed from the reactor. WHICH ONE (1) of the following is the MINIMUM number that make up the site fire brigade?
(Assume that there are no absences.)
a. 3 b. 4 c. 5 d. 6
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While operating at rated conditions, a reactor coolant chemistry sample indicates an activity level of 4.4 uCi/ gram Dose Equivalent I-13 WHICH ONE (1) of the following is required action?
a. Reduce reactor power to 15% of rated and sample every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> until coolant activity is within limit Increase sampling frequency to every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce reactor power until coolant activity is within limit c. Be in at least HOT SHUTDOWN with the MSIVs closed within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> QUESTION: 087 (1.00)
The plant is operating at 100% power. Control Room Air Conditioning Unit B002A has been declared out of service due to a faulty controller on the Condenser Pressure Flow Control Valve (Z51-F073B). Control Room temperature is 85 deg F and increasing. WHICH ONE (1) of the following is the MINIMUM priority that should be assigned to the Work Order to repair Z51-F073B? Priority 1 Priority 2 Priority 4 Priority 5
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SENJOR REACTOR' OPERATOR -Page-44 QUESTION: 088 (1.' 0 0 )
WHICH ONE (1) of the'following is NOT designated as an Item Control Area?
a. Fuel Transfer Canal b. Cask Storage Pool c. TIPS Storage Area d'. Upper Containment Pool QUESTION: 089 (1.00)
WHICH ONE (1) of the following is the MAXIMUM number of-people allowed in the horseshoe area without special authorization?
a. 3 b. 4 c. 5 , . . .. _ _ _ . . _ . . _ . _ _
7 SENIOR REACTOR. OPERATOR O O .Page 45
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QUESTION: 090 (1.00)
In. order to protect the health and safety of the public, an action which departs from Technical Specifications is required to be immediatel performed by a licensed reactor operato WHICH ONE (1) of the following describes the course of action this operator is permitted to take?
a. Immediately take whatever action is required without.further directio b. Notify the Control Room Operator then perform the requiredL actio c. Perform the required action.then notify the Control ~ Room Operator or Shift Supervisor of his actio ,
d. Obtain approval from the Shift Supervisor:and then perform the actio :
QUESTION: 091 (1.00) {
You have entered a Technical Specification Action Statement requiringL that the plant be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ;
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WHICH ONE (1) of the following is the administrative requirement for the time for notification of the NRC?
a. When the Tech Spec Action Statement is entere i b. After it is determined how long it will take to get to HOT *
SHUTDOW c. Within one hour of commencing plant shutdown d. Within one hour of reaching HOT SHUTDOW >
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, SENIOR REACTOR OPERATOR Page 46 QUESTION: 092 (1.00)
Ac0ording to Grand Gulf Technical Specifications, WHICH ONE (1) of the following is the minimum shift staffing on-site when the reactor is at 5% power?
SS SRO RO AO QUESTION: 093 (1.00)
Work histories for FOUR (4) operators are given below. WHICH ONE (1) of the following operators meets the administrative requirements?
Work History- Excluding Shift Turnover Time Fr Sa Su Mon. Tu We T F Sa OFF OFF 1500 1500 1500 0000 0800 0800 0800 2400 2400 2400 0800 OFF OFF OFF 0700 0700 0700 0700 0000 0800 1700 1700 1700 1700 0800 O
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. SENIOR REACTOR OPERATOR Page 47 QUESTION: 094 (1.00)
During refueling operation a seismic event occurs at the GGNS. The seismic event was verified by the in-plant as greater than the OBE leve The refueling operators report that the spent fuel does not appear to be damaged and the radiation levels are the same as before the seismic event. WHICH ONE (1) of the following Emergency Plan Classifications should be declared? Unusual Event b. Alert c. Site Area Emergency d. General Emergency QUESTION: 095 (1.00)
You are working on a weekend and have been given a controlled procedure and been asked to verify that it is the latest revision. The TSO computer is down. WHICH ONE (1) of the following is the approved method?
a. Verify that the procedure is the same revision as the one in the control roo Check the Master Log in Document Control, c. Check the Master Index in the control roo d. Have an SRO verify all of the steps are correc i
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. SENTOR REACTOR OPERATOR Page 48 l i
QUESTION: 096 (1.00) ,
l WHICH ONE (1) of the following non-safety related systems does NOT require independent verification following each refueling outage?
a. N19 Condensate b. N22 Condensate Cleanup c. N34 Lube Oil d. C84 Met Monitoring QUESTION: 097 (1.00)
You are performing the quarterly physical verification of outstanding clearances when you find an unattached tag. WHICH ONE (1) of the following actions should be taken FIRST?
a. Document the location of the tag on Attachment VII of procedure 01-S-06-1, " Protective Tagging System".
b. Notify the Plant Superviso c. Place the component in the position required by the ta d. Stop the work in the area of the equipmen . . . .-. . - - - -
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SENIOR REACTOR OPERATOR Page 49 QUESTION: 098 (1.00) q An Unusual Event has been declared. WHICH ONE (1) of_the following states the Shift Superintendent's responsibilities for notification?
a. He shall notify the NRC immediately after notification of the appropriate state and local agencies and not later than one hour-after the time the licensee declares the Unusual Even b. He shall notify the appropriate state and local agencies immediately after notification of the NRC and not later than one hour after the time the licensee declares the Unusual Even c. He shall notify the NRC within one hour after the notification-of the appropriate state and local agencies and not later than two hours after the time the licensee declares the Unusual Even d. He shall notify the appropriate state and local agencies immediately after notification of the NRC and not later than two hours after the time the licensee declares the Unusual Event.
QUESTION: 099 (1.00)
The reactor is critical and an entry into the drywell is necessar WHICH ONE (1) of the following selections states the MAXIMUM reactor power level and the MINIMUM people required to make the entry?
a. 10% power and 2 people % power and 1 person c. 10% power and 3 people % power and 2 people
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, SENIOR' REACTOR OPERATOR Page 50 QUESTION:.100- (1.00)
You are on a plant walk-through with an NRC license examiner when the-examiner brushes against some equipment. His shirt is checked for contamination. No Alpha contamination is foun WHICH ONE (1) of the following is the MINIMUM limit for determining if the shirt is contaminated with Beta-Gamma?
a. Detectable (above background) cpm / scan above background c. 100 cpm / scan above background d. 250 cpm / scan above background (********** END OF EXAMINATION **********)
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ANSWER: 001 (1.00) {
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REFERENCE: -
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- Grand Gulf: OP-LO-SYS-LP-B33-1-08, LO 31b, p 8 . Grand Gulf: System Test #2,. Q-11 ,
KA: 202001K127 [4 .1/4 . 3 ] .
I 202001K127 ..(KA's)
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ANSWER: 002 (1.00) , :
REFERENCE: ,
i Grand Gulf: OP-LO-SYS-LP-C11-1B-02, LO 3, p. 12 : Grand Gulf: System Test 2, Q-46 KA: 201003G007 [3.6/3.6]
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201003G007 ..(KA's) '!
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ANSWER: 003 (1.00)
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REFERENCE: .
. Grand Gulf: Technical Specification Bases 3/4. ,
' Grand Gulf: OP-LO-SYS-LP-F11/17, LO KA: 295023A202 [3.4/3.7] ,
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295023A202 ..(KA's)
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. SENIOR REACTOR OPERATOR Page 52 RNSWER: 004 (1.00) REFERENCE: Grand Gulf: OP-M72-501, LO 3d, p. 7 KA: 223002A101 [3.5/3.5)
223002A101 ..(KA's)
ANSWER: 005 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-G33/36-05, LO 10a Grand Gulf: 04-1-01-G33-1, Section KA: 204000A205 [2. 7/2. 8)
204000A205 ..(KA's)
ANSWER: 006 (1.00) REFERENCE: Grand Gulf: OP-LP-SYS-E22-1-04, LO 6a, p. 3 KA: 209002A415 [3.6/3.6)
209002A415 ..(KA's)
ANSWER: 007 (1.00) _ - - _ . _ _ - _ -
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. SENJOR REACTOR OPERATOR Page 53
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i Grand Gulf: OP-LO-SYS-HN-C11-2, LO 5, Table 1 )
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KA: 234000A302 [3.1/3. 7) ;
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234000A302 ..(KA's) ;
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j ANSWER: 008 (1.00) :
i REFERENCE: 'l l Grand Gulf: OP-LO-SYS-LP-C11-1A, LO 3a, p. 9; ;
KA: 201001K408 [3 .1/3. 0)
201001K408 ..(KA's)
ANSWER: 009 (1.00) i
^ REFERENCE: Grand Gulf: OP-LO-SYS-LP-B21-05, LO 5d, p. 26-27 KA: 216000K513 [3.5/3.6]
i 216000K513 ..(KA's)
ANSWER: 010 (1.00) t
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. Page 5 REFERENCE: Grand Gulf: OP-LP-SYS-LP-B33-2-04, p. 25, LO 6 KA: 202002A108 [3.4/3.4]
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202002A108 ..(KA's)
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ANSWER: 011 (1.00) ,
REFERENCE: 1 l.
' Grand Gulf: OP-LO-SYS-LP-P81-03, p. 40, LO Sc l
KA: 264000K402 [4.0/4.2] ,
264000K402 ..(KA's)
ANSWER: 012 (1.00) ! .
REFERENCE: i Grand Gulf: AP-01-S-11-10, p. 46 KA: 294001K105 [3.2/3.7] '
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294001K105 ..(KA's)
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ANSWER: 013 (1.00) -r
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. SENIOR REACTOR OPERATOR Page 55 REFERENCE: Grand Gulf: EP-3, Caution 2
, KA: 295030A202 [3.9/3.9]
295030A202 ..(KA's)
ANSWER: 014 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-E30-03, p. 15, LO S . Grand Gulf: OP-LO-SYS-LP-M71-04, p. 76, LO 4 KA: 223002K403 [3.5/3.6]
223002K403 ..(KA's)
ANSWER: 015 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-C51-3-03, p. 24, LO 6 KA: 215005K113 [2. 6/3. 0]
215005K113 ..(KA's)
AFSWER: 016 (1.00) i
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REFERENCE: Grand Gulf: OP-LO-SYS-LP-T48-05, p. 48, LO 9f KA: 288000K103 [3.7/3.7)
288000K103 ..(KA's)
ANSWER: 017 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-C51-1-03, p. 27, LO 6a KA: 215004K401 [3.7/3.7)
215004K401 ..(KA's)
ANSWER: 018 (1.00) REFERENCE: Grand Gulf: Technical Specification Bases, 3/4.6.1.8, p. B 3/4 6- . Grand Gulf: OP-LP-SYS-LP-M41-05, LO 1 KA: 223001G006 [3. 0/4. 0)
223001G006 ..(KA's)
ANSWER: 019 (1.00) [+ 1. 0 )
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~ REFERENCE: ;
! Grand Gulf: OP-LO-SYS-LP-C51-4-03, p. 23, LO 3 KA: 212000K602 [3.7/3.9] ,
n 212000K602 ..(KA's) b a
ANSWER: 020 (1.00) . REFERENCE: -
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, Grand Gulf: OP-LP-SYS-LO-G17, p. 6, LO 3 !
KA: 268000G007 [2.8/3.1]
i 268000G007 ..(KA's)
ANSWER: 021 (1.00) ,
' REFERENCE: Grand Gulf: Technical Specification, 3. ,
i KA: 223002G005 [3.1/4.1]
I 223002G005 ..(KA's) I l
ANSWER: 022 (1.00) i <
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a REFERENCE:
' Grand Gulf: OP-LO-SYS-LP-E51-03, p. 75, LO 9a. (NOTE: The suction valves are DC powered.)
KA: 217000K601 [3.4/3.5] .
217000K601 ..(KA's) .
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ANSWER: 023 (1.00) I
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-REFERENCE: - Grand Gulf: OP-LO-SYS-LP-T48-05, p. 12, LO 3 KA: 261000G007 [3.5/3.7) ,
261000G007 ..(KA's)
ANSWER: 024 (1.00) i
- REFERENCE: ! Grand Gulf: OP-LO-SYS-LP-C34-04, p. 24, LO 4 KA: 259002K302 [3. 7/3. 7]
259002K302 ..(KA's) ,
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ANSWER: 025 (1.00)
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REFERENCE: Grand Gulf: OP-LO-SYS-LP-R21-05, TABLE 1, LO S KA: 262001A304 [3.4/3.6]
262001A304 ..(KA's)
ANSWER: 026 (1.00) REFERENCE: Grand Gulf: Technical Specifications, Table 4.8.2.1-1, Grand Gulf: OP-LO-SYS-LP-L11-02, LO KA: 263000G005 [ 3 .1/ 3 . 8 )
263000G005 ..(KA's)
ANSWER: 027 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-B33-1-08, p. 39, LO 13 KA: 202001A210 [3.5/3.9)
202001A210 ..(KA's)
ANSWER: 028 (1.00) j l
, SENIOR REACTOR OPERATOR Page 50 REFERENCE: Grand Gulf: OP-LO-SYS-LP-E32/38-02, p. 14 and 18, LO KA: 239003A301 [3. 0/2. 8]
239003A301 ..(KA's)
ANSWER: 029 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-N64/65-03, p. 33, LO 6 KA: 271000K408 [ 3 .1/ 3 . 3 )
271000K408 ..(KA's)
ANSWER: 030 (1.00) REFERENCE: Grand Gulf: Procedure 04-101-E51-1 Grand Gulf: OP-LO-SYS-LP-E51-03, LO 1 KA: 217000G010 [3.4/3.5]
217000G010 ..(KA's)
ANSWER: 031 (1.00)
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t
._ , - SENIOR REACTOR. OPERATOR Page . 61 --
,
REFERENCE: Grand Gulf: EP-2A KA: 255037A104 [4. 5/4. 5] '
295037A104 ..(KA's) .!
!
ANSWER: 032 (1.00) ,
I :
REFERENCE: ;
- Grand Gulf: 04-1-01-B21-1, p. 10 Grand Gulf: OP-LO-SYS-E22-2-03, Figure 4,.LO 3 '! Grand Gulf: OP-LO-SYS-RN-E22-2-00, p. 13
KA: 218000K501 [3.8/3.8) ,
218000K501- ..(KA's)
.
ANSWER: 033 (1.00)
i ;
REFERENCE: Grand Gulf: OP-LO-SYS-LP-C11-2-04, p. 21 ~
'
KA: 201005K510 [3.2/3.3]
t 201005K510 ..(KA's) j
ANSWER: 034 (1.00) t
'
. ,
t
'
i i
-_ . . . . - - - , . , , . - -,- -
n., . ,-,
, -. . . - . _-. . - . .
_,fSENIOR REACTOR OPERATOR O LO Page 62-i
!
REFERENCE: !
1 Grand Gulf: OP-LO-SYS-LP-C51-4-03, p. 13, LO 2 j KA: 215005K604 {3.1/3.2)
215005K604 ..(KA's) ,
l ANSWER: 035 (1.00) : ,
REFERENCE: ,
t Grand Gulf: 04-1-01-P75-1, rev. 38, 3.8 and 3.9, p. 4 Grand Gulf: OP-LO-SYS-LP-P75-05, LO 8 KA: 264000A203 [3.4/3.4]
264000A203 ..(KA's)
ANSWER: 036 (1.00) REFERENCE: ; Grand Gulf: 05-S-01-EP-2, Attachment 22 Grand Gulf: OP-LP-EP-LP-004-02, LO KA: 295015K204 [4 . 0 / 4 .1)
295015K204 ..(KA's)
ANSWER: 037 (1.00) ,
,
--.
,. . . - - . . .
...; SENIOR-REACTOR OPERATOR ' Page 63-
')
' REFERENCE: j q Grand Gulf: Technical Specification Bases 3/4. . Grand Gulf: OP-LO-SYS-LP-B33-2-04, LO 1 KA: 202001K601 [3.5/3.7) !
'l 202001K601 ..(KA's) i
~!
-
.!
ANSWER: 038 (1.00) .]
REFERENCE: Grand Gulf: OP-LO-SYS-LP-E12-06, p. 59-60, LO 9 KA: 226001K409 [3.2/3.4]
i 226001K409 ..(KA's)
ANSWER: 039 (1.00) i l REFERENCE: Grand Gulf: OP-LO-SYS-LP-C41-08, p. 25, LO 8 KA: 211000A308 [4.2/4.2] -
211000A308 ..(KA's)
ANSWER: 040 (1.00) t-
.-----,r t- --
< m - + r --- , ~ - r - - - - - - - - - - - - - - - - - - - -
SENIOR REAOTOR OPERATOR e
'"
Page 64
.
REFERENCE: Grand Gulf: OP-LO-SYS-LP-C34-04, p. 21, LO 4 KA: 295009K202 [3.9/3.9]
295009K202 ..(KA's)
ANSWER: 041 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-E22-2-03, p. 26, LO KA: 239002A308 [3.6/3.6]
239002A308 ..(KA's)
ANSWER: 042 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-M71-04, p. 39, LO S KA: 204000K111 [3.5/3.7]
l l
204000K111 I
..(KA's)
I (1.00)
'
ANSWE )
!
l
. SENIOR REACTOR OPERATOR Page 65 REFERENCE: Grand Gulf: 04-1-01-E12-1, rev. 49, p. 5, step 3.6.3 . Grand Gulf: OP-LO-SYS-LP-E12-06, LO 14 KA: 205000G010 [3.2/3.3]
205000G010 ..(KA's)
ANSWER: 044 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-N19-04, p. 15-16, LO 4 KA: 256000K201 [2. 7/2. 8)
256000K201 ..(KA's)
ANSWER: 045 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-D17-05, p. 14, LO 5 KA: 272000G007 [3.5/3.5)
, 272000G007 ..(KA's)
ANSWER: 046 (1.00) I i
. - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _
- -- . . . .. . . . . . . ~ . - - - - . . - ~ . . . . . _ ~ . .
.
. SENIOR REACTOR OPERATOR. O LO Pagel66
.
'
_ . .
REFERENCE:
- Grand Gulf: OP-LO-SYS-LP-Z51-02, p. 27, Lo Sa.-
KA: 290003K401 [3.1/3. 2 ]
290003K401 ..(KA's)
'
ANSWER: 047- (1.00)
. REFERENCE: Grand Gulf: 05-1-02-I-4, p. 3 2 Grand Gulf: OP-LO-DT-LP-029-01, LO ;
, KA: 295003G007 [3.2/3.6]
295003G007 ..(KA's)
ANSWER: 048 (1.00)
. '
. REFERENCE: ' Grand Gulf: Technical Specification-3/4.6.3, p. 6-2 >
KA: 295013G003 [3.3/4.2]
295013G003 ..(KA's)
ANSWER: 049 (1.00) . ._ . _ . . _ _ _ _ .. .
. SENIOR REACTOR OPERATOR Pagel 67--
REFERENCE: d Grand Gulf: OP-LO-EP-LP-005-02, p. 12, LO KA: 295011K301 [3.6/3.93 295011K301 ..(KA's)
ANSWER: 050 (1.00) ,
- REFERENCE: ; Grand Gulf: OP-LO-EP-LP-006-01, p. 11, LO t KA: 295032K303 [3.8/3.9) i l
295032K303 ..(KA's)
,
ANSWER: 051 (1.00) ; '
REFERENCE: Grand Gulf: 05-1-02-II-8, .
I KA: 295023G010 [3.8/3.9)
.
295023G010 ..(KA's)
!
I 052 (1.00)
'
ANSWER: , __
. . .- . . -
'
.- ' SENIOR REACTOR OPERATOR - Page:68--
,
-
REFERENCE: ,
4 Grand Gulf: OP-LO-3P-LP-004-02, p. 21, LO ,
.
KA: 295037K303 [4.1/4. 5] .
.
'295037K303 . . (KA's) ,
ANSWER: 053 (1.00) *
! .
REFERENCE: ,
i Grand Gulf: 05-1-02-IV-1, . -
!
KA: 295022A201 [3. 5/3. 6] ,
!
295022A201 . . (KA's)
i
'
ANSWER: -054 (1.00) ,
,
REFERENCE: Grand Gulf: OP-LO-EP-LP-005-02, p. 10, LO KA: 295012K202 [3.6/3.7] I 295012K202 . . (KA's) 1
,
ANSWER: 055 (1.00) ,
!
,
E
- * , --e..---~q-r ,,y 4. .
4 e-._ -
. SENJOR REACTOR OPERATOR Page 69 REFERENCE: Grand Gulf: Technical Specification, Table 1.2, 3/4.6. KA: 295021A201 [3.5/3.6]
295021A201 ..(KA's)
ANSWER: 056 (1.00) REFERENCE: Grand Gulf: 03-1-01-1, step 2. KA: 295014A202 [3.9/3.9]
295014A202 ..(KA's)
ANSWER: 057 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-004-02, p. 9, LO KA: 295020K101 [3.7/3.9]
295020K101 ..(KA's)
ANSWER: 058 (1.00) S
~
. SENIOR REACTOR OPERATOR () '"
Page 70 REFERENCE: Grand Gulf: 05-1-02-V-11, p. 1 KA: 295018K302 [3.3/3.4]
295018K302 ..(KA's)
ANSWER: 059 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-003-02, p. 20, LO KA: 295031K304 [4. 0/4. 3 ]
295031K304 ..(KA's)
ANSWER: 060 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-C71-03, p. 13, LO 3 KA: 295008K302 [3.6/3.9]
295008K302 ..(KA's)
ANSWER: 061 (1.00) . . .. . . .- . .. . . . . . . =. ,:.SENIORTREACTOR' OPERATOR 'O 'O
.
~ Page.7 *
,
"
t
. REFERENCE: '
Grand Gulf: OP-LP-EP-LP-004-02, p. 8,,LO ' KA: 295037K301 [4.1/4.2]
295037K301 ..(KA's)
- ANSWER: 062 (1.00) ,
. REFERENCE: Grand Gulf: EP-2, Caution KA: 295027K102 [3.0/3.2]
,
295027K102 ..(KA's)
ANSWER: 063 (1. 0 0 ) . >
, REFERENCE: Grand Gulf: EP-3, figure (For the exam attachments use all EP figures) Grand Gulf: OP-LO-EP-LP-005-02, LO KA: 295026A203 [3.9/4.0] !
>
'l 295026A203 ..(KA's)
!
ANSWER: 064 (1.00) -
l ,
.
l
..
I i
i p -3w .- --,.~y ,-,,v, , , , w
O
'"'
M
'"'
. SENIOR REACTOR OPERATOR Page 72 l REFERENCE: Grand Gulf: 05-01-02-III-3, step 4. ,
i KA: 295001A106 [3.3/3.4]
295001A106 ..(KA's)
ANSWER: 065 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-M71-04, p. 79, LO 9 KA: 295004 A204 [3.2/3.3]
295004A204 ..(KA's)
ANSWER: 066 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-006-01, p. 8, LO KA: 295034K204 [3.9/3.9]
295034K204 ..(KA's)
ANSWER: 067 (1.00) .
.""
.
'"
Page 73 l
..SENJOR REACTOR OPERATOR
,
REFERENCE: Grand Gulf: 05-1-02-I-2, p. 1 KA: 295005G010 [3.8/3.6)
295005G010 ..(KA's)
ANSWER: 068 (1.00) REFERENCE: Grand Gulf: OP-LO-SYS-LP-C61-00, Table 3, LO 3 KA: 295016G006 [4.1/4.1)
295016G006 ..(KA's)
ANSWER: 069 (1.00) REFERENCE: l Grand Gulf: 04-1-01-P53-1, p. 10 Grand Gulf: OP-LO-SYS-LP-P53-04, LO 9a KA: 295019K218 [3.5/3.5)
i 295019h218 ..(KA's) )
.
'
ANSWER: 070 (1.00)
b.
_
.; SENIOR REACTOR: OPERATOR Page 74 REFERENCE: Grand Gulf:' Loss of Instrument Air, 05-1-02-V-9,-page 1 KA: 295019G010 - [3. 7/3.4]- i 295019G010 ..(KA's)
ANSWER: 071 (1.00)
. REFERENCE: Grand Gulf: 05-1-02-II-1, p. 10 Grand Gulf: OP-LO-SYS-LP-C61-00, LO 4 KA: 295016K201 [4.4/4.5]
295016K201 ..(KA's)
ANSWER: 072 (1.00) REFERENCE: Grand Gulf: 05-1-02-V-5, Section KA: 295014G010 [4.0/3.9]
295014G010 ..(KA's)
ANSWER: 073 (1.00) a ,.- w ,
M
'"
\
'
J
, SENIOR REACTOR OPERATOR Page 75 REFERENCE: Grand Gulf: 05-1-02-V-1, " Loss of Component Cooling Water", page 1 KA: 295018K303 [3.1/3.3]
295018K303 ..(KA's)
ANSWER: 074 (1.00) REFERENCE: Grand Gulf: Decrease in Recirculation System Flow Rate, 05-1-02-III-3, page KA: 295001A201 [3.5/3.8]
295001A201 ..(KA's)
l ANSWER: 075 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-003-02, step 15, KA: 295030K208 [3.5/3.8)
295030K208 ..(KA's)
ANSWER: 076 (1.00) o SENIOR REACTOR OPERATOR Page 76
. REFERENCE: Grand Gulf: OP-LO-SYS-LP-C11-1A, L.O. 3c and 3f KA: 295025A107 [4.1/4.1]
295025A107 ..(KA's)
ANSWER: 077 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-004-02, , LO KA: 295014K106 [3.8/3.9]
295014K106 ..(KA's)
ANSWER: 078 (1.00) REFERENCE: Grand Gulf: 04-1-01-E30-1, , step KA: 295030A104 [4. 0/4. 0]
295030A104 ..(KA's)
ANSWER: 079 (1.00) .
a""
Page 77
. SENJOR REACTOR OPERATOR REFERENCE: Grand Gulf: OP-LO-EP-LP-003-02, p. 28, LO KA: 295031K101 [4. 6/4. 7]
295031K101 ..(KA's)
ANSWER: 080 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-003-02, p. 29, LO KA: 295031K201 [4. 4 /4. 4 ]
295031K201 ..(KA's)
ANSWER: 081 (1.00) REFERENCE: Grand Gulf: OP-LO-EP-LP-006-01, p. 9, LO KA: 295017K304 [3.6/3.8)
295017K304 ..(KA's)
ANSWER: 082 (1.00) .-.- .. ~. - . . . . - . - . . .~. ~ . ~ , .. .
t
.
- SENTORLREACTOR OPERATOR O 0 cPage178
.l
-
';
REFERENCE: >
i l '. Grand Gulf: OP-LO-SYS-LP-B33-1-08, p. 80,-LO 30 .j KA: 295006K306 [3.2/3.3]
!
!
295006K306 ..(KA's) 4 ANSWER: 083 (1.00) . !
-REFERENCE:
.
i Grand Gulf: 05-1-02-I-4, p. ; Grand Gulf: Facility-Question ONEP-1 ~l
KA: 295003K202 [4.1/4 . 2]
.i 295003K202 ..(KA's)
ANSNER: 084 (1.00) 3 REFERENCE: ;
- Grand Gulf
- OP-LO-SYS-LP-N21-02, p. 42, LO 6 ,
-;
KA: 295008A108 [3.5/3.5]
295008A108 ..(KA's)
ANSWER: 085 (1.00) ___ , _ . - . , _
I
) h""
. SENIOR REACTOR OPEb 70R Page 79 REFERENCE: Grand Gulf: Conduct of operations procedure 01-S-06-2 , 6.5. . Grand Gulf: OP-LO-AD-LP-001-05, LO D.21 KA: 294001K116 [3.5/3.8]
294001K116 ..(KA's)
ANSWER: 086 (1.00) [+1.0)
REFERENCE: Grand Gulf: T.S. 3. . Grand Gulf: OP-LO-PB-LP-001-03, License / Tech Specs /10 CFR, L. KA: 294001A114 [2.9/3.4)
294001A114 ..(KA's)
ANSWER: 087 (1.00) REFERENCE: Grand Gulf: 01-S-07-1, p. 23 Grand Gulf: OP-LO-AD-LP-001-05, p. 11, LO I.11 KA: 294001K102 [3.9/4.5)
294001K102 ..(KA's)
ANSWER: 088 (1.00) , - . , . . ... -. - - ~ .-. . . _ . . . . . .. .-
.
'
.iSENIOR REACTOR OPERATOR-O 10 .Page 80 .
.
REFERENCE:
1.- Grand Gulf: 01-S-06-15, p. 17 Grand Gulf: .OP-LO-PB-LP-002-06, LO 11 , Grand Gulf: Facility Question, No. 29, Procedure Test :
!
' KA: 294001K105 [3.2/3.7) j
'
294001K105 ..(KA's)
l
!
ANSWER: 089 (1.00) , !
, ,
REFERENCE: [ Grand Gulf: 01-S-06-4, p. 3 2 Grand Gulf: OP-LO-AD-LP-001-05, LO C7 ;
KA: 294001A111 [3.3/4.3] .
294001A111 ..(KA's)
ANSWER: 090 (1.00) REFERENCE: Grand Gulf: 01-S-06-2 6.2.1. (4 )
'
2 .. Grand Gulf: 10CFR50.54y ,
,
KA: 294001A102 [4.2/4.2]
.
294001A102 ..(KA's)
.
ANSWER: 091 (1.00) , ,
l
.
-- +,w a ,mr -n-r 2- --a,r.-- . . . e
M
. SCQOR REACTOR OPERATOR Page 81 REFERENCE: Grand Gulf: Conduct of operations 01-S-06-2, 6. . Grand Gulf: OP-LO-AD-LP-001-05, LO D.26 KA: 294001A105 [3. 4 /3. 8)
294001A105 ..(KA's)
ANSWER: 092 (1.00)
b REFERENCE: Grand Gulf: Technical Specification, Table 6. K/A: 294001A103 [2.7/3.7)
294001A103 ..(KA's)
ANSWER: 093 (1.00)
d REFERENCE: Grand Gulf: Conduct of operations 01-S-06-2, Attachment I K/A: 294001A111 [3.3/4.3)
294001A111 ..(KA's)
ANSWER: 094 (1.00)
b
. SENIOR REACTOR OPERATOR e'"
e'"
Page 82 REFERENCE: Grand Gulf: Activation of the Emergency Plan 10-S-01-1, Attachment I, page 25 ;
K/A: 294001A116 [2.9/4.7]
294001A116 ..(KA's)
ANSWER: 095 (1.00) REFERENCE: Grand Gulf: Procedure 01-S-02-1, page 19 Grand Gulf: OP-LO-AD-LP-001-05, L.O. A.15
[2.9/3.4]
KA: 294001A101 [2.9/3.4]
294001A101 ..(KA's)
ANSWER: 096 (1.00) REFERENCE: Grand Gulf: Independent Verification Program, 01-S-06-29, 6. . Grand Gulf: OP-LO-AD-LP-001, L.O. E3 KA: 294001K101 [3.7/3.7]
294001K101 ..(KA's)
l ANSWER: 097 (1.00) _ _ - - _ - _ _ _ _ _ _ _ _ -
c
, . . SENJOR-REACTOR OPERATOR Page 83--
,
REFERENCE: Grand Gulf: Protective Tagging System, 01-S-06-1, page 18 Grand Gulf: OP-LO-AD-LP-001-05, L.O. J.22 KA: 294001K102 [3.9/4.5]
294001K102 ..(KA's)
, ANSWER: 098 (1.00)
l
- ' REFERENCE: Grand Gulf: Incident Reports and Reportable Events, 01-S-06-5' Grand Gulf: OP-LO-AD-LP-001-05, L.O. KA: 294001A116 [2.9/4.7]
294001A116 ..(KA's)
ANSWER: 099 (1.00) REFERENCE: Grand Gulf: Containment and Drywell Access Control, 01-S-06-7, page 5 and 6 Grand Gulf: OP-LO-AD-LP-001-05, L.O. K5 KA: '294001K114 [3.2/3.4]
294001K114 ..(KA's)
ANSWER: 100 (1.00)
C '.
- _ - _ - _ _ _ _ _ _ - _ - _ _ - _ - _ - _ _
. SENTOR REACTOR OPERATOR
. Page 84 REFERENCE: Grand Gulf: Exposure and Contamination control,.01-S-08-2, page 10 KA: 294001K103 [3.3/3.8]
294001K103 ..(KA's)
(********** END OF EXAMINATION **********)
..
__ _ _ _ - - - _ _ _ _ -
- . _ .._ _ _ _ . _ ..- _ -
. . . .
. O TEST CROSS. REFERENCE O Page_ 1
,
SRO Exam B W R1 Reactor !
Orga'nized by Quest ion Numb'er ,
I QUESTION VALUE REFERENCE
'
001 1.00 9000C01 002 1.00 9000003' *
003 1.00 9000004 004 1.00 9000006 :
005 1.00 9000007 l 006 1.00 9000008 t 007 1.00 9000010 l 008 1.00 9000011 '
009 1.00 9000012 010 1.00 9000013 '
011 1.00 9000014 012 1.00 9000015 "
.
013 1.00 9000016 014 1.00 9000017 .i 015 1.00 9000018 016 1.00 9000019 ;
017 1.00 9000020 018 1.00 9000021 :
019 1.00 9000023 7 020 1.00 9000025 021 1.00 9000027 022 1.00 9000028 023 1.00 9000029 'l 024 1.00 9000031 ;
025 1.00 9000033 ;
026 1.00- 9000034 027 1.00 9000035 028 1.00 9000036 i
"
029 1.00 900003" 030 1.00 90000';9 031 1.00 9000';40 032 1.00 9000041 033 1.00 9000043 ,
034 1.00 9000044 035 1.00 9000046 I 036 1.00 9000047 1 037 1.00 9000051 038 1.00 9000052 -
039 1.00 9000053 t
^
040 1.00 9000054 041 1.00 9000055 042 1.00 9000056-043 1.00 9000059 044 1.00 9000062 045 1.00 9000063 046 1.00 9000064 047 1.00 9000067 048 1.00 9000068 049- 1.00 9000069
,
.[
h
V
, , TEST CROSS REFERENCE Page 2 SRO Exam BWR Reactor Organized by 0ueat ion Number QUESTION VALUE REFERENCE 050 1.00 9000070 051 1.00 9000071 052 1.00 9000072 053 1.00 9000074 054 1.00 9000075 055 1.00 9000076 056 1.00 9000077 057 1.00 9000078 058 1.00 9000079 059 1.00 9000080 060 1.00 9000082 061 1.00 9000084 062 1.00 9000085 063 1.00 9000086 064 1.00 9000087 065 1.00 9000088 066 1.00 9000090 067 1.00 9000091 068 1.00 9000092 069 1.00 9000093 070 1.00 9000094 071 1.00 9000096 072 1.00 9000097 073 1.00 9000098 074 1.00 9000099 075 1.00 9000100 076 1.00 9000101 077 1.00 9000102 078 1.00 9000103 079 1.00 9000104 080 1.00 9000105 081 1.00 9000106 082 1.00 9000108 083 1.00 9000109 084 1.00 9000110 085 1.00 9000111 086 1.00 20936 087 1.00 9000115 088 1.00 9000117 089 1.00 9000118 090 1.00 22805 091 1.00 9000121 092 2.00 9000122 093 1.00 9000123 094 1.00 9000124 095 1.00 9000125 096 1.00 9000126 097 1.00 9000127 098 1.00 9000128
_ - ., - .=. .- .. -. - - - . . .
- . .
O TEST CROSS REFERENCE 0 Page 3'
l
'
S-R O Exam BWR Reactor -
Organized by Quest ~ ion Number ,
!
QUESTION VALUE REFERENCE i 099 1.00 9000129 100 1.00 9000130
......
100.00
......
... .. ;
100.00 i
%
$
,
- I,
P
,
l
-1
1
?
., . , , . . - - , , . . ~ . . r , - . - _ . - . . _ - - . . - - - , - , e
. . . . , _ - . . - - . _ . . .- - - . ~ . - . -
"
.
0 TEST CROSS REFERENCE O .Page. 4-
. ;
~
Reactor Organized by KA Group ,
,
PLANT WIDE GENERICS !
!
'
QUESTION VALUE KA 095 1.00 294001A101 ~I 090 1.00 294001A102 .;
'
092 1.00 294001A103 091 1.00 294001A105 .
089 1.00 294001A111 !
093 1.00 294001A111 086 1.00 294001A114 l 094 1.00 294001A116 j 098 1.00 294001A116 .
096 1.00 294001K101 !
087 1.00 294001K102 ,
097 1.00 294001K102 '
100 1.00 294001K103 088 1.00 294001K105 .
012 1.00 294001K105 I 099 1.00 294001K114 i 085 1.00 294001K116
......
PWG Total 17.00 PLANT SYSTEMS Group I QUESTION VALUE la '
033 1.00 201005K510 010 1.00 202002A108 l 006 1.00 209002A415 -!
039 1.00 211000A308 019 1.00 212000K602 .)
017 1.00 215004K401 !
015 1.00 215005K113 034 1.00 215005K604 009 1.00 216000K513 030 1.00 217000G010 022 1.00 217000K601 032 1.00 218000K501 018 1.00 223001G006 004 1.00 223002A101 023 1.00 223002G005 014 1.00 223002K403 038 1.00 226001K409 041 1.00 239002A308 024 1.00 259002K302 023 1.00 261000G007 i
. , , , , . , , -- , ,- - . . , -
- . . . - - - ,. . . - .
,
a
'"'
TEST CROSS REFERENCE O Page 5
,
SRO Exam BWR Reactor Orqanized by KA Group PLANT SYSTEMS l Group I QUESTION VALUE KA 025 1.00 262001A304 035 1.00 264000A203 011 1.00 264000K402
......
PS-I Total 23.00 Group II QUESTION VALUE KA 008 1.00 201001K408 027 1.00 202001A210 001 1.00 202001K127 j 037 1.00 202001K601 005 1.00 204000A205 j 042 1.00 204000K111 043 1.00 205000G010 007 1.00 234000A302 i 028 239003A301
'
1.00 026 1.00 263000G005 029 1.00 271000K408 045 1.00 272000G007 l 046 1.00 290003K401 ;
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PS-II Total 13.00 l l
Group III QUESTION VALUE KA 002 1.00 201003G007 l 044 1.00 256000K201 l 020 1 00 268000G007 1 016 1.00 288000K103 i
.... PS-III Total 4.00
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PS Total 40.00 EMERGENCY PLANT EVOLUTIONS j Group I
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TEST CROSS REFERENCE O Page 6
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SRO Exam BWR Reactor Organized by KA Group EMERGENCY PLANT EVOLUTIONS Group I QUESTION VALUE KA 047 1.00 295003G007 083 1.00 295003K202 082 1.00 295006K306 040 1.00 295009K202 048 1.00 295013G003 056 1.00 295014A202 072 1.00 295014G010 077 1.00 295014K106 036 1.00 295015K204 068 1.00 295016G006 071 1.00 295016K201 081 1.00 295017K304 i 003 1.00 295023A202 )
051
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1.00 295023G010 076 1.00 295025A107 063 1.00 295026A203 062 1.00 295027K102 078 1.00 295030A104 013 1.00 295030A202 075 1.00 295030K208 079 1.00 295031K101 080 1.00 295031K201 059 1.00 295031K304 031 1.00 295037A104 061 1.00 295037K301 052 1.00 295037K303
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EPE-I Total 26.00 Group II QUESTION VALUE KA 064 1.00 295001A106 074 1.00 295001A201 065 1.00 295004A204 067 1.00 295005G010 084 1.00 295008A108 060 1.00 295008K302 049 1.00 295011K301 054 1.00 295012K202 058 1.00 295018K302 073 1.00 295018K303 070 1.00 295019G010 069 1.00 295019K218
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. .. TEST CROSS REFERENCE Page 7'
i SRO Exam BWR Reactor j Organized by KA Group
EMERGENCY PLANT EVOLUTIONS .
Group II ,
QUESTION VALUE KA
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057 1.00 295020K101 055 1.00 295021A201 053 1.00 295022A201 050 1.00 295032K303 '
066 1.00 295034K204
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EPE-II Total 17.00
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EPE Total 43.00 l
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Test Total 100.00 i
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g (g)
. SENJOR REA:.: TOR OPERATOR Pago 1 ANSWER KEY MULTIPLE CHOICE 023 a 001 a 024 b 002 d 025 b 003 d 026 a 004 d 027 b 005 b 028 a 006 d 029 d 007 b 030 a 008 a 031 b 009 b 032 d 010 a 033 c 011 c 034 d 012 a 035 c 013 c 036 b 014 a 037 b '
015 a 038 c 016 d 039 d 017 c 040 c 018 d 041 b 019 a 042 d
020 b 043 d 021 b 044 c 022 b 045 d
. . - - ,
- - * . SENIOR REACTOR OPERATOR O O Pago- 2'
ANSWER KEY 046 c 069 c 047 a 070 b 048 c 071 d 049 d 072 c 050 b 073 b 051- a 074 c 052 a 075 d 053 b 076 b 054 a 077 c 055 c 078 a 056 d 079 d 057 c 080 a 058 b 081 b 082 b 060 c 083 b 061 b 084 b 062 c 085 c 063 b 086 c-064 c 087 b 065 d 088 a 066 b 089 a 067 d 090 d 068 b 091 c
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e SEMIOR REACTOR OPERATOR Pega 3 ANSWER KEY 092 b 093 d 094 b 095 b 096 b 097 b 098 a 099 d 100 c
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(********** END OF EXAMINATION **********) j
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