IR 05000293/1993021
| ML20059F529 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/03/1994 |
| From: | Conte R, Sisco C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20059F522 | List: |
| References | |
| RTR-NUREG-1021 50-293-93-21OL, NUDOCS 9401140055 | |
| Download: ML20059F529 (108) | |
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< % U.S. NUCLEAR REGULATORY COMMISSION ;
REGION I j EXAMINATION REPORT NO: 93-21(OL) l
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FACILITY DOCKET NO: 50-293 FACILITY LICENSE NO: DPR-35 j LICENSEE: Boston Edison Company RFD #1 Rocky Hill Road Plymouth, MA 02360
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FACILITY: Pilgrim Nuclear Power Station , EXAMINATION DATES: November 29-December 3,1993 EXAMINERS: C. Sisco, Operations Engineer , C. Tyner, Examiner (EG&G) . CHIEF EXAMINER: /b& V fl!Nk7 C. Sisco, Operations Engineer Date ' BWR Section, Operations Branch Division of Reactor Safety APPROVED BY: c h /
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Richard J. Conte, Cl[j/f Date
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BWR Section, Operations Branch Division of Reactor Safety 9401140055 940105 PDR ADOCK 05000293 V PDR ,
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Summary: Examination Report No. 50-293/93-21 (013: l Initial examinations were administered to one upgrade and two Instant Senior Reactor
- Operator applicants. All applicants passed the examinations and were well prepared for the examinations. Facility generic strengths and weaknesses were listed as feedback to the training progra ,
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__ _ _ - _ _ _ _ . . DETAILS INTRODUCTION The NRC administered initial examinations to one upgrade and two Instant Senior Reactor Operator applicants. The examinations were administered in accordance with NUREG-1021, .
" Operator Licensing Examiner Standards," Revision .0 PREEXAMINATION ACTIVITIES The licensee conducted a review of the written examination in the NRC Regional Office on November 18,1993. The review was productive and as a result only one comment was made by the licensee following the administration of the examinatio The simulator scenarios and Job Performance Measures (JPMs) were validated at the training center during the week of November 29,1993. The facility staff members involved in the examination preparation process were cooperative and helpful to the NRC examiner The facility staff members who were involved in the preexamination activities signed security agreements to ensure that the examination was not compromise .0 EXAMINATION RESULTS AND ADDITIONAL FINDINGS The results of the examination are summarized below:
SRO Pass / Fail Written 3/0
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Operating 3/0 Overall 3/0 Fncility Generic Strengths and Weaknesses The following is a summary of the strengths and weaknesses noted during the initial examination administration. This information is being provided to aid the licensee in upgrading their training progra .- - . - - ..- - - . . - i ..
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Written Examination i
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Strengths J The use of plant Technical SpeciReations to answer application question ! Knowledge of Emergency Core Cooling system operation 't Weaknesses
. Knowledge of protective clothing requirements when racking out a draw-out type of electrical breake Knowledge of Emergency Diesel Generator conditions to start the DC drive fuel oil pum j Bases of the reactor low water level scram setpoin ;
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Simulator Examination Strengths Control room command function, crew communications and teamwork was a generic strength during the simulator examination Weaknesses There were no generic weaknesses noted during the simulator examination ! Walkthrouch Examination . Strengths The use of prints and logic diagrams was a generic strengt The use of plant procedures and the knowledge and use of the plant technical specifications was a generic strengt l Weaknesses The knowledge of the plant response to the failure of the EPR was a generic weaknes , i I _]
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Overall Conclusion The NRC examiners concluded that the applicants were well prepared for the examination Also, the facility staff members acting as surrogate operators during the simulator examinations were very helpful to the NRC Examiner .2 Additional Findings During the onsite preparations for the examination, the NRC Examiners determined that procedure 2.4.38, "LPRM Failure," Rev 11 needed to be enhanced. The enhancement consisted of clarifying the nomenclature in the procedure to be consistent with the control ; room annunciator for an APR.M in an alarm condition. The licensee took prompt and !
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corrective action to revise the procedur The NRC examiners determined that minor inconsistencies existed between the plant control ! room and the plant referenced simulator regarding posted operator aids. The licensee , conducted an audit of the posted operator aids in both the control room and simulator. Based ,i l on this audit, the licensee removed procedure indexes from the simulator. In addition, an Area Temperature Table used during the Emergency Operating Procedures would be made into an operator aid following Plant Status Update Training. The NRC examiners concluded the audit conducted by the licensee was adequate to resolve the inconsistencies between the plant control room and simulator concerning posted operator aid The NRC Examiners noted that the quality of the communications by the on shift control room staff was of a lesser quality than that of the crews observed in the simulator during the q examination l EXIT MEETING An exit meeting was conducted at the training center on December 2,1993. Those in attendance are listed below. The generic strengths and weaknesses as well as the NRC Examiners additional findings were discussed at the exit meetin Boston Edison Company E. S. Kraft, J Vice President, Nuclear Operations L. L. Schmeling Plant Manager J. F. Alexander Nuclear Training Manage T. A. Sullivan Operations Section Manager T. E. Trepanier Chief Operating Engineer T. S. Swan Operations Training Division Manager N. L. Desmond Compliance Division Manager W. J. Green Sr. Training Specialists T. E. Collis Sr. Training Specialists
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U. S. Nuclear Regulatory Commission C. Sisco Chief Examiner C. Tyner Examiner (EG&G)
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EXAMINATION AND ANSWER KEY
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Nuclear Regulatory Commission Operator Licensing Examination
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This document is removed from Official Use Only category on date of examinatio I 8l- / ! ' ' , . , "r/~t.[[ NRC Official Use Only r IN i t.i t u :
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. . U. 8. NUCLEAR REGULATORY COMMISSION BITE-SPECIFIC WRITTEN EXAMINATION APPLICANT INFORMATION Name: R.gion: I Date: 1993/11/29 Facility / Unit: Pilgrim 1 License Level: SRO Reactor Type: GE
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INSTRUCTIONS Use the answer sheets provided to document your answer 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 percent. Examination papers will be picked up 4 !
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hours after the examination start Al] work done on this examination is my ow I have neither given nor received ai Applicant's signature RESULTS Examination Value 100 Points Applicant's Score Points Applicant's Grade Percent
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ANSWER SHEET l l Multiple Choice (Circle or X your choice)
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If you change your answer, write your selection in the blan l
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l MULTIPLE CHOICE 023 a b c d l
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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 I i 005 a b c d 028 a b c d :
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006' a b c d 029 a b c d 1 007 a b' c d 030 a b c d ; 008 a b c d 031 a b c d ;
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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 04 3 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
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ANSWER SHEET i Multiple Choice (Circle or X your choice) !
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If you change your answer, write your selection in the blank.
, 046 a b c d 068 a b c d !i MULTIPLE CHOICE 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 l
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051 a b c d 074 a b c d 052 a b c d 075 a b c d : I 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 {
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057 a b c d 080 a b c d !
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058 a b c d 081 a b c d l 059 a b c d 082 a b c d 060 a b c d 083 a b c d 061 a b c d 084 a b c d i
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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 i
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067- a b c d 090 a b c d ;
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ANSWER SHEET
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Multiple Choice -(Circle or X your choice) l
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If you change:your answer, write your selection in the blan a b c d i MULTIPLE CHOICE , 092 a b c d i
093 a b c d i 094 a b c d 095 a b c d 096 .a b c d 3 i 097 a b c d
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098 a b c- d ,
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Page 5 f i NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination, the following i rules apply: Cheating on the examination will result in a denial of your ; application and could result in'more severe penaltie . After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio : To pass the examination, you must achieve a grade of 80 percent or greate . The point value for each question is indicated in '
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parentheses after the question numbe . There is a time limit of 4 hours for completing the examinatio . Use only black ink or dark pencil to ensure legible copie l Print your name in the blank provided on the examination ! cover sheet and the answer shee . Mark your answers on the answer sheet provided and do not leave any question blan i
. If the intent of a question is unclear, ask questions of the examiner onl . Restroom trips are permitted, but only one applicant at a ;
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time will be allowed to leave. Avoid all contact with ' anyone outside the examination room to eliminate even the ' appearance or possibility of cheatin . When you complete the examination, assemble a package . including the examination questions, examination aids, and l
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answer sheets and give it to the examiner or procto Remember to sign the statement on the examination cover , shee , t 1 After you have turned in your examination, leave the examination area as defined by the examine .
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. QUESTION: 001 (1. 00) l When a Tagout is being performed on a centrifugal pump, which of the :l
:following'should be tagged FIRST? l Control switch l b. Discharge valve l
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c.' Auxiliary component power breakers I
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i QUESTION: 002 (1.00) . I Given the following conditions: l
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-- A 25 year old, male, fully trained radiation worker _j
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-- Lifetime whole body dose is 7.25 Rem ;
-- Date'is June 15, 1993 't
-- No whole body dose has been received during this calendar. year l
-- Current dose history (NRC Form 4) is on file l Per 10CFR20, how much whole body exposure may this individual. receive !
during the remainder of the calendar year?
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i ' .25 Rem '; , 4 b. 5.0 Rem c. 9.0 Rem .,
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QUESTION: 003 (1.00) Conditions develop for which no action consistent with the plant Technical Specifications will provide adequate protection of the public health and safet Per 10CFR50.54, the Operator at the Controls shall: a. take only those actions that comply with the Technical Specification b. take actions to protect the public health and safety and subsequently inform the Nuclear Operations. Supervisor, obtain approval from the Nuclear Operations Supervisor, then take actions to protect the public health and-safet d. obtain verbal NRC concurrence to deviate from Technical Specifications, then take actions to protect the public health and safet . QUESTION: 004 (1.00) The Operator at the Controls is conducting a test in accordance with an approved operating procedure. He identifies what he believes to be an editorial erro IDENTIFY the MINIMUM required approval prior to continuing with the tes a. ORC Chairman b. Procedure Owner c. Nuclear Watch Engineer d. Nuclear Operations Supervisor
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An operator with a current NRC license has maintained his i requalification training current but has NOT maintained proficienc Which of the following is the MINIMUM action required.by the operator l
~(licensee) in order to reactivate his NRC license? l a. Complete 8 hours of parallel watchstandin ,
b. Complete all licensed operator reactivation requirements and
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obtain NRC concurrenc Complete 40 hours of parallel watchstanding including a complete ! plant tour and shift turnove f
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d. Complete seven 8-hour shifts of parallel watchstanding including , a complete plant tour and shift turnove , O i QUESTION: 006 (1.00) ! A locked room has a general area dose rate of 1.2 Rem /h Who controls the keys to such an area? (Assume a routine, non-emergency , access is required.)
a. ALARA Specialist t I b. Nuclear Watch Engineer c. Nuclear Operations Supervisor d. Radiological Operations Division Manager i f s e
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. QUESTION: 007 (1.00) The following hours have been worked by a licensed Reactor Operator: i Sunday -- 7 a.m. to 7 Monday -- 7 a.m. to 7 Tuesday -- 7 to 7 Wednesday - - 7 a.n. to 5 ' Thursday -- 7 a.m. to 3 Friday -- 7 a.m. to 3 Which of the following schedules will allow the operator to work MAXIMUM hours without obtaining special authorization? (Do not include turnover time.)
t a. 7 a.m. to 7 ! a.m. to 11 > c. 7 a.m. to 5 B d. 7 a.m. to 3 , QUESTION: 008 (1.00) . Which of the following describes the REQUIRED emergency notifications in ' response to an emergency declaration? a. Commonwaalth, local authorities, and the NRC must be notified as i soon au possible after declaration of an event not to exceed 15 minutes except for an Unusual Event which may.not exceed 4 hour b. The NRC must be notified within 15 minutes after declaration of an' emergency. Commonwealth and local authorities must be notified as soon as possible thereafter not to exceed one hou c. Commonwealth and local authorities must be notified within 15 minutes and the NRC must be notified within 4 hours after the declaration of the even d. Commonwealth and local authorities must be notified within 15 minutes after declaration of an emergency and the NRC notified ' as soon as possible thereafter not to exceed one hou ; i
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QUESTION: 009 (1.00) i What is the MINIMUM emergency classification at which personnel-accountability must be initiated? a. Unusual Event b. Alert q
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c. Site Area Emergency j i d. General Emergency l l QUESTION: 010 (1.00) , An emergency has occurred and the Nuclear Watch Engineer is acting for ! the On-Call Emergency Directo Which of the following MAY be delegated ; to the Emergency Plant Manager? !
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a. Emergency Classification b. Ordering a Site Evacuation 3 i c. Authorization of Emergency Exposures , d. Offsite Protective Action Recommendations ;
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QUESTION: 011 (1.00) l In addition to rubber gloves and aprons, which of the following lists- ! the MINIMUM required protective clothing While handling. sodium- , hypochlorite'in a well ventilated area? j a.. Safety glasses and rubber safety boots !
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d. Safety goggles and half-face respirator !
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QUESTION: 012 (1.00) i Which of the following is the MINIMUM required protective equipment when racking out a draw-out type breaker?
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a. Full length coat, hard hat, and class "0" gloves i Full length coat, hood, and class "2" gloves Full face shield, class "0" gloves, flame retardant coveralls or ; jacke Full face shield, class "2" gloves, flame retardant coveralls or jacke i
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QUESTION: 013 (1.00) i Under which of the following conditions would a licensed Senior Reactor ; Operator be REQUIRED to be in the Control Room? li a. The reactor water temperature is 165 degrees F with shutdown ! cooling in service and fuel being removed from the vesse b. The unit is in hot standby with reactor power in the source ! range.- .; c. The unit.is in cold shutdown with a reactor vessel hydrostatic : test in progres j i d. The unit is in cold shutdown with control rod drives being- ] replace l J
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. QUESTION: 014 (1.00)
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A specific radiation work permit would be required for which of the following activities in a High Radiation Area? j-i a. Chemistry sampling
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b. Operator rounds
c. Radiological Protection surveys Inspection of an area to locate a leak
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QUESTION: 015 (1.00) Which of the following changes to a NON-SAFETY RELATED system can be accomplished WITHOUT processing a Temporary Modification?
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a. Temporary hose used for system draining b. Temporary pipe supports , c. Plugging floor drains ; d. Pulling circuit boards t
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QUESTION: 016 (1.00) Given the following conditions: L
-- A plant startup is in progress . -- The Main Turbine is about to be rolled -- The Mode Switch has been placed in "Run" ,
A tracking LCO would be required if the Technical Specification minimum , number of operable channels is NOT met for the: a. Main Condenser Low Vacuum Scra ; b. Main Steam Line High Radiation Scra , c. Main' Steam Line Isolation Valve Closure Scram.
- d. Turbine Stop Valve Closure Scra j
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QUESTION: 017 ( 1. ,00 )
'While in shutdown cooling, a spurious high drywell pressure signal is i generated. Select the expected automatic actio l a. All MO-7 (Torus Suction) valves will open causing the MO-43 (Pump Suction) valves to close. LPCI will then injec b. The Shutdown Cooling (SDC) suction valves.(MO-47 & MO-50) will ,
close causing the running RHR Pumps in the SDC loop to tri ' c. The MO-29 (LPCI Injection #2) valves in both. loops will be interlocked closed causing the MO-18 (Pump Minimum Flow) valve in the loop aligned for shutdown cooling to.ope d. The RHR pumps will automatically align to inject water to the , selected recirculation loop with suction through the MO-47 and l MO-50 (Shutdown Cooling Outboard and Inboard Isolation) valve ' i i
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. . E QUESTION: 018 (1.00) During an initiation of HPCI, turbine exhaust pressure reaches 150 ps ,
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Select the expected automatic actio , a. The MO-4 and MO-5 (Inboard and Outboard Isolation Valves) and MO-35 and MO-36 (Torus Suction Valves) will clos ! b. The HPCI turbine stop and control valves both clos .
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c. The HPCI turbine control valve goes closed and the stop valve remains ope ., d. The HPCI turbine stop valve closes and the control valve fully open l QUESTION: 019 -(1.00)
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With two pumps operating in an RHR loop, excessive flow through the RHR heat exchangers is procedurally controlled b limiting loop flow to 4000 GP limiting loop flow to 6000 GP c. requiring that the heat exchanger bypass valve-be ope .; d.' requiring.that the RHR pump manual discharge valve be throttle : i i QUESTION: 020 (1.00) ! Operating the HPCI turbine at 1800 RPM.will cause: ! a. water hammer in the exhaust lin !
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b. water hammer in the pump discharge lin l; c. Booster Pump cavitatio '! l d. erratic control valve operation due to low oil pressur !
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' QUESTION: 021
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<RCIC in injecting water from the CST to the RPV due to a valid i initiation signal. Torus level is increasing and results in generation ,
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of a valid high torus level signa l The MO-22 (CST Suction) will:
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a, close, resulting in a trip of RCIC on low suction pressur .
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b. close when both RCIC suctions from the torus (MO-25 and MO-26) -; valves are fully opene c. close when either of the RCIC suctions from the torus (MO-25 or MO-26) valve are fully opene close, causing the RCIC suctions from the torus (MO-25 and MO- , 26) to ope ! QUESTION: 022 (1.00)
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RCIC flow rate should be'at least 100 GPM in order to prevent:
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a. oscillations in flow indicatio loss of lubrication to pump bearing adequate cooling flow to the lube oil coole adequate cooling flow to the barometric condense , i t I
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-QUESTION: 023 (1.00)
An ATWS has occurred and the "A" train of Standby Liquid Control (SLC)
~has been initiate Which of the following would indicate IMPROPER operation of the SL system? ,
a. Reactor Cleanup System isolate " Loss of Continuity to Squib Valve" alarm occur Squib Valve Ready' light is ON for the "A" trei ,
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d. Squib Valve Ready light is ON for the "B" trai F k
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QUESTION: 024
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Which of the following sets of conditions will result in an automatic ADS initiation? a. Low-Low reactor water level for 2 minutes and one Core Spray Pump runnin i b. High drywell pressure for 2 minutes and one Core Spray Pump a runnin l c. High_drywell pressure and Low-Low reactor water level for 2 minutes with one RHR Pump runnin > d. Low-Low reactor water level for 11 minutes with one RHR Pump runnin ,
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QUESTION: 025 (1.00) ! A steam leak occurred in the drywell chusing a re.ctor scram and HPCI initiation.' . Subsequently HPCI tripped on high reactor water leve SELECT the response of HPCI to' DECREASING level following the tri a. The HPCI high level trip must be manually reset for HPCI to automatically star ;
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b. HPCI will automatically start at a reactor water level of -49" AND Drywell pressure greater than 2.5 psi c. HPCI will start if the HPCI high level trip is rese d. HPCI will automatically start at a reactor water level of -49" f I AND the HPCI Aux Oil Pump is manually starte ; e QUESTION: 026 (1.00) , Given the following conditions:
-- A low reactor water level signal is presen Reactor pressure is 50 psi ,
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Which of the following conditions will caus: the MO-29A and MO-29B (RHR i injection valves) to automatically close? a. MO-50 (RHR inboard suction valve) is ope i
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b. MO-47 (RHR outboard suction valve) is ope c. MO-7A, 7B, 7C, and 7D (torus suction valves) are close l
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QUESTION: 027 (1.00) Given the following conditions:
' -- A low low reactor water level signal is present -- Reactor pressure is 850 psig and lowering ' -- Drywell pressure is 2.0 psig and increasing The Core Spray Injection Valves (MO-24 and MO-25) will automatically open when:
a. drywell pressure reaches 2.5 psi b. reactor pressure reaches 400 psi i the ADS 2 minute timer times out, d. the ADS 11 minute timer times ou . t f , _ QUESTION: 028 (1.00) ; The Core Spray System is in a normal standby lineu The Core Spray Line Break Detection Monitor will detect a break located in the Core Spray injection line: inside the reactor pressure vessel shrou b. upstream of the MO-24 (Outboard injection valve). , i c. in the drywell upstream of the reactor pressure vessel penetratio d. between the Mo-24 and Mo-25 (Outboard and inboard injection valves.
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I QUESTION: 029 (1.00) ' What is the-MAXIMUM number of Main Steam Isolation Valves (MSIV) that can be closed at 40% power without resulting in a full reactor scram?
-(Consider only RPS-MSIV relationships.) , .
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d. 5 QUESTION: 030 (1.00) Alternate Rod Insertion has automatically actuated due to high reactor pressur This actuation signal is sealed in for 60 seconds in order to: a. allow the control rods to reach the full in position prior to resetting the signa I b. ensure that the Recirculation Pump field breaker trip. timer has
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timed ou i c. . prevent a loss of instrument air by limiting pressurization of the complete Scram Pilot Valve Air Header to once per minut d. allow sufficient time for pressure to decrease below the-setpoint prior to rese i
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QUESTION: 031 (1.00) Drywell'to torus differential pressure has decreased to 0.75 psid during normal operation SELECT the reason differential pressure must be restored to greater.than
.1.2 psi a. To reduce the volume of water in the downcomer r b. Reactor Building to Torus vacuum breakers will 'por Tc incorrect differential pressure during accident . ,.. / i t (cur .
c. Torus to drywell vacuum breakers will operate improperl d. Non-condensibles volume in the drywell will be insufficient for post accident condition +
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QUESTION: 032 (1.00) Standby Gas Treatment has been running for five minutes due to a valid initiation signal. A piece of plastic is drawn into the "A" fan suction causing flow to drop to 100 scf Which of the following is an expected automatic actio I I a. The "A" fan will trip on a low flow signa b. The "B" train inlet damper will ope c. The "B" fan will start causing.the "A" fan to tri d. The cross-tie damper will open to allow suction from the "B" trai I I
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QUESTION: 033 (1.00) Which of the following will result in an automatic start of the Emergency Diesel Generator DC driven emergency fuel oil pump? a. Fuel oil pressure - 10 psi Jacket water pressure - 20 psi Diesel speed - 900 rpm J
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b. Fuel oil pressure - 10 psi Jacket water pressure - 20 psi Diesel speed - 80 rpm , c. Fuel oil pressure - 5 psi Jacket water-pressure - 15 psi ' Diesel speed - 80 rpm d. Fuel oil pressure - 25 psi Jacket water pressure - 25 psi Diesel speed - 80 rpm QUESTION: 034 (1.00)
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Given the following conditions:
-- Drywell pressure 3.2 psig
-- Reactor water level +30 inches
"i-Which of the following operator actions is required in order to spray the Drywell with the RHR system?
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ar. Momentarily depress the " Containment Spray' Reset" pushbutto , I b. Momentarily depress the " Auto Isolation Inboard Valve" - pushbutto , c. Place the "LPCI Override Control" switch for.the desired loop to MANUA d. Place the "RPV Level Override Keylock" switch for the desired
loop to MANUAL OVERRID ,
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QUESTION: 035 (1.00) Due to a loss of ventilation, Reactor Building temperature has INCREASED
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by 20 degrees i Assuming a constant reactor water level,' INDICATED narrow range reactor water level would: , increase due to an decrease in variable leg densit increase due to an decrease in reference. leg densit I c. decreasa due to an decrease in variable leg densit d. decr:ase due to an decrease in reference leg densit . QUESTION: 036 (1.00) I
Select the condition that will bypass ALL SRM rod block a. Reactor modo switch in RU b. All IRM range switches are on Range " c. All SRM detectors are fully withdraw d. All SRM channels read 200 cp P P-i
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QUESTION: 037 (1.00) : While operating at 80% power, a failure occurs which causes the >
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electrical pressure regulator to send.a full "Open" signal to the
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turbine control valve ' Select the expected automatic respons (Assume no operator actions taken.) i a. The MSIVs will close when reactor pressure reaches 880 psi b. The reactor will scram when reactor pressure reaches 600 psi c. The mechanical pressure regulator will take control and maintain > pressure slightly lower than pressure prior to the transien d. The mechanical pressure regulator will take control and maintain , pressure slightly higher than pressure prior to the transien , i QUESTION: 038 (1.00) When valving in a CRD hydraulic control accumulator, the 305-102 (Withdraw Riser Isolation Valve) and the 305-112 (Scram Discharge Riser ;
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Isolation Valve) are required to be open prior to opening the 305-101-(Insert Riser Isolation Valve). I This will directly prevent: j a. a single rod scram when opening the 305-101 valv I b. excessive scram time of that rod during a reactor scra j c. damage to the accumulator in the event of a reactor scra l d. damage to the drive mechanism in the event of a reactor scra ')
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QUESTION: 039 (1.00)
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.Which of the following will result in a control rod drift alarm?
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A control rod that is:
't a. not selected for movement passes an " odd" reed switc !
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b. not selected for movement passes an "even" reed switc c. selected for motion but not receiving a motion signal passes an
" "even" limit switc d. selected for motion and is being moved passes 2 " odd" limit j switches within 2 second j
QUESTION: 040 (1.00) ,
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The individual Reactor Recirculation Pump speed controllers are in
" Manual".
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Pump speed is limited to a MINIMUM of 26% in order to prevent: a. unstable recirc MG set operatio ; b. cavitation of the jet pumps due to low. flo l c. temperature stratification in the reactor vesse d. excessive current from the generator due to low voltage from the volts /hz regulato i
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i QUESTION: 041 (1.00) ) Given the following conditions:
-- A reactor startup is in progress
-- The reactor is not yet critical
-- Reactor temperature is 175 degrees F
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Total Reactor Water Cleanup system flow must NOT exceed 150 gpm in order to prevent: excessive vibration in the regenerative heat exchanger due to flo ; b. excessive temperature load on the non-regenerative heat exchange !
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c. inadequate net positive suction head on the Reactor Cleanup Pump : excessive differential pressure across the filter demineralizer , f-QUESTION: 042 (1.00) l, The RHR "A" loop /RHR "A" pump is being placed in the shutdown cooling mode and is expected to remain in shutdown cooling for five hour , In this condition, how is RHR "A" pump minimum flow protection provided? a. Administratively limiting minimum RHR pump flow to 500 GP b. Administratively limiting minimum RHR pump flod to 1800 GP I c. Automatic opening of-the MO-18A (minimum flow) valve when flow .
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drops to less than 2500 GPM for 10 second d. Automatic opening of the MO-34A and MO-36A (suppression chamber block valve and test line block valve) valve when flow drops to less than 2500 GPM for/ 10 second ' s
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. QUESTION: 043 (1.00) Which thermal limit does the Rod Block Monitor (RBM) prevent exceeding? Average Planar Linear Heat Generation Rate (APLHGR) I b. Linear Heat Generation Rate (LHGR) ;
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c. Minimum Critical Power Ratio (MCPR) d. Maximum Fraction of Limiting Power Density (MFLPD) QUESTION: 044 (1.00) * Given the following conditions:
-- Reactor power is being reduced during a plant shutdow l -- IRM "G" has failed " Upscale" and has NOT been bypasse ! -- The Mode Switch is in "RUN". i At what point will the IRM "G" upscale failure cause a half scram?
a. APRM channel "E" drops to 3%. b. APRM channel "B" drops to 3%. c. APRM channel "C" drops to 3%. d. APRM channel "D" drops to 3%. l l
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QUESTION: 045 (1.00) During plant operation at 100% power, APRM Channels "A" & "C" upscale trip units become inoperabl , Which of the following actions will comply with Technical
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a. Insert all operable control rods within 24 hour i i Insert a half scram on RPS "A" within 12 hour i Insert a half scram on RPS "B" within 12 hour d. Reduce power to IRM range and place the Mode Switch in
"Startup/ Hot Standby" within 24 hour !
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QUESTION: 046 (1.00) , Torus cooling is in service using the "A" RHR Pump and.is in the process ; of being secure l
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The RHR pump should be secured when the system flow drops to 2000 GPM in order to-
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a. prevent air from becoming entrained in the syste b. prevent opening the MO-18A (pump minimum flow) valv c. prevent excessive differential pressure across the MO-36A-(Torus Cooling) Valv .! d. prevent excessive differential pressure across the MO-34A-(Torus Cooling / Spray Block) Valv ;
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SENIOR REACTOR OPERATOR Page 29 i I I l l QUESTION: 047 (1.00) ! l Given the following conditions:
-- Reactor power is 100%
-- The "A" Main Steam Line Radiation Monitor has failed "Downscale" !
-- No Technical Specification actions have been taken for the l failed monitor ;
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Which of the following will cause a Group I Isolation? a. The "B" Main Steam Line Radiation Monitor fails " upscale".
' b. The "C" Main Steam Line Radiation Monitor fails "downscale".
l c. The "C" and "D" Main Steam Line Radiation Monitors fail {
" upscale".
d. The "B" and "D" Main Steam Line Radiation Monitors fail ;
" upscale". l t
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QUESTION: 048 (1.00)
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After operating for seven days, the "A" Reactor Feedwater Pump was manually trippe .j Which of the following is true concerning restarting of the "A" Reactor Feedwater Pump? j
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a. Pump starts are limited to three (3) starts in any one hou b. Pump starts are limited to five (5) starts in any one hou l-i c. A cooldown period of 45 minutes is required between pump starts.' ; i d. The pump discharge valve must be' closed prior to pump star I
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^ QUESTION: 049 (1.00) The Adsorber system operation keylock control switch is in the AUTO , position.
., Which of the following conditions will result in closure of the adsorber ;
-bypass valve and' opening of the adsorber inlet valve?
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a. Both Main Stack Off-Gas radiation monitors go "downscale".
b. The "B" Main Stack Off-Gas radiation monitor goes " upscale".
c. The "A" and "D" Main Steam Line Radiation monitors go ,
"downscale". q d. The "A" Augmented Off-Gas post treatment radiation monitor goes * " upscale".
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QUESTION: 050 (1.00) } While operating at 90% power, Stator Cooling Water inlet pressure l suddenly drops to 10 psi l
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Which of the following conditions would result in a reactor scram? 'l i ' Time after pressure decrease - 1 minute -;
-- Generator stator amps - 15,000 amps l --
Reactor power - 70% ;
~ Time after pressure decrease - 2 minutes --
Generator stator amps - 13500 amps i
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Roactor power - 60% i c. -- Time after pressure decrease - 4 minutes ;
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Generator stator amps - 5250 amps ;
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Reactor power_
- 35% l Time after pressure decrease - 5 minutes
- 4450 amps
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Generator stator amps :
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Reactor power - 29% ! l
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,. -QUESTION: 051 (1.00)
Concerning alternate control rod insertion during an emergency condition- , Select the MINIMUM reactor pressure that will assist the CRD accumulator ! in scramming its associated control ro j psig :
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b. 400 psig .l psig d.'800 psig i QUESTION: 052 (1.00) >
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A TIP trace is being performed when a high drywell pressure signal 'f occur l Select the expected automatic actio a. The shear valve fires with the detector still in the cor [ b. The ball valve closes with the detector still in the cor i c. The detector withdraws into its shield and the ball valve close ! d. The detector withdraws into its shield and the shear valve ; fire ,
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QUESTION: 053 (1.00) Select the Refuel Floor Ventilation Exhaust Radiation Monitor trip
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combination below that will result in a Reactor Building ventilation isolatio ! "A" and "C" are " upscale".
' b. "A" and "D" are " upscale". "A" and "D" are "downscale". "A" is " upscale" and "B" is "downscale".
. h QUESTION: 054 (1.00) Given the following conditions:
-- The plant is operating at 100% power
-- All Reactor Feedwater Pumps (RFP) are running
-- The RFP Tripping Sequence Selector is selected to sequence "BCA"
-- The "A" RFP has just experienced a low net positive suction head '
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condition '
-- All Condensate Pumps are running ,
SELECT the expected automatic system respons ; a. The "A" RFP will trip immediatel ] i b. The "B" RFP will trip immediatel c. The Condensate Pumps min-flow valve will immediately clos d. The condensate Pumps min-flow valve closes after a one (1) minute time dela ~ -. -
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) -QUESTION: 055 (1.00)~
Power for RPS MG Set "A" is supplied from: B- b. B-1 ! c. B-2 d. B-2 , QUESTION: 056 (1.00) , In the event of a loss of vital AC (Y2), which Control Room instrument ; should be used to monitor reactor water level? a. Narrow Range indicators on panel C905 b. FWLC Range indicators on panel C905 c. Wide Range / Fuel' Zone indicators on panel C903 , Shutdown Level indicator on panel C904 QUESTION: 057 (1.00) The plant was operating at 100% power when the APRM chart recorders , indicate that a high APRM scram should have occurre : l Which of the following conditions would require entry into EOP-2, ,
" Failure to Scram."
. a. Entered _directly based on the failure to' scra l b. . Entered.following insnrtion of a manual scram only if power l remains above 3%, ; Entered following insertion of a manual scram if 2 or more rod l remain at position 0 Entered.upon verification that the automatic scram signal did
.not result in insertion of control rods to or beyond 0 ,
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. QUESTION: 058 (1.00)
LWhen~ attempting to lower drywell temperature with drywell sprays, the operator is directed to secure sprays if drywell pressure drops below 2.5 psi Securing sprays at 2.5 psig is done to ensure that: a. primary containment pressure does.NOT drop below atmospheri b. primary containment pressure does NOT drop below the scram setpoin c. reactor building to torus vacuum breakers do NOT open.
l d. torus to dr.>311 vacuum breakers do NOT open.
l QUESTION: 059 (1.00) Injecting boron , prior to exceeding the Boron Injection Initiation Temperature (BIIT) will ensure that: a. the torus water temperature will not exceed 160 degrees b. the torus water temperature will not exceed the heat capacity temperature limi c. the torus water will not be heated to the point that the RHR and Core Spray pumps lose adequate net positive suction hea d. the torus water will not be heated to the point that Drywell , pressure increases solely due to high torus temperatur l _ _ - _ _ -
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QUESTION: 060 (1.00).
Which condition would require entry into EOP-03, " Primary Containment ; Control"? ;
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a. Torus water temperature is 82 degrees ] b. Drywell temperature is 145 degrees c. Torus pressure is 1.7 psi Drywell hydrogen concentration is 2.5% ! t i
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QUESTION: 061 (1.00) . The reactor low water level scram setpoint is selected to: .
} prevent violating the Minimum Critical Power Ratio Operating Limi i b. prevent violating the Minimum Critical Power Ratio Safety Limi c. ensure a reactor scram has been completed prior to a Group I t
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isolation, d. ensure that fuel uncovery does not occur during a level , transient at powe l
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' QUESTION: 062 (1.00) !
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While operating at 100% power, the Feedwater Level Control steam flow l summer output fails to "zero". l-With no operator action, reactor water level will: a. decrease to compensate for the feed flow / steam flow mismatch but will remain above the low water level scram setpoin l
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b. decrease to the point that the reactor scrams on low water :! i.eve l increase to compensate for the feed flow / steam-flow mismatch.but ; will remain below the turbine trip setpoin ; d. increase to the point that a turbine trip occurs which will result in a reactor scra !
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QUESTION: 063 (1.00) i When executing EOP-03, " Primary Containment Control", Alternate r Depressurization is required if torus level cannot be maintained above 'l
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90 inche Which of the following will become uncovered at 90 inches?
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a. Drywell downcomers
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b. SRV spargers c. HPCI steam exhaust line , d. RCIC steam exhaust line
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QUESTION: 064 (1.00) ! Conditions have developed which require entry into EOP-05, ;
' " Radioactivity Release Control".
I What is the MINIMUM 9mergency classification that must be made?
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a. Unusual Event ..
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b. Alert l c. Site Area Emergency d. General Emergency
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QUESTION: 065 (1.00)
. Reactor power begins to increase at 5%.per minute with no apparent i caus :
At what' point is the operator is REQUIRED to manually scram the reactor? l
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a. Any APRM has two LPRMs inputs of 110% or greate b. Any APRM has three LPRM inputs of 110% or greater.- ! i c. Two LPRMs within a string reach 110% or greate d. Three LPRMs within a string reach 110% or greate . D
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QUESTION: 066 (1.00) ; With reactor power at 50% on the 50% load line, a rapid increase in off- ,
; gas radiation level' occur SELECT the REQUIRED operator actions for the Reactor Recirculation ,
Pump , I The Reactor Recirculation Pumps shall: a. be reduced in speed to lower activity but maintained above 3 j Mlb/ hou ' b. be reduced to minimum speed, then tripped if necessary to terminate the increase in activit i be reduced to minimum speed if necessary to terminate the !
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increase in activit d. be tripped immediately to lower power to 40% as rapidly as ; possibl ,
QUESTION: 067- (1.00)
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-When using SRVs for pressure control in EOP-01, "RPV Control", the operator is directed to use the sequence "B-C-D-A". l Why is this the preferred sequence?
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This sequence is designed to: a. avoid high local pool temperature s b. utilize the highest capacity valve firs ; c. ensure that relief valves nearest torus temperature detectors are operated las d. delay discharging a relief valve into the same area as the HPCI-steam exhaus ,
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QUESTION: 068 (1.00)
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Boron injection using the Standby Liquid Control (SLC) System has been ,
' initiated in accordance with EOP-02, "RPV Control, Failure-To-Scra l The SLC Pumps should:
k a. automatically trip on low suction pressur ] automatically trip on low SLC tank level at 50 gallon }
; be manually tripped if SLC tank level drops-to 0 gallons, d. be manually tripped if SLC tank level drops by 1600 gallon !
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E-QUESTION: 069 (1.00) . A plant transient is in progress that results in a rapidly rising. torus j pressur . Torus spray initiation shall be directed prior to torus bottom pressure ; reaching 16 psig in order to prevent: -;
a. fatigue failure of the Safety Relief Valve sparger b. fatigue failure of the torus-drywell' vacuum breakers.- operation of the reactor building-torus vacuum breakers which i
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would cause deinertio Ed. downcomer failure and the resultant pressurization of the torus airspac ,
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SENIOR REACTOR OPERATOR Page 40 4-QUESTION: 070 (1.00) During execution of EOP-03, " Primary Containment Control", the operator is directed to " Maximize torus cooling". Assuming no RHR pumps are needed to provide adequate core cooling, which of the following RHR system lineups is the MINIMUM that will comply with this direction? One RHR loop with both RHR pumps runnin b. Both loops of RHR with one RHR pump running in each loo Both loops of RHR with both RHR pumps running in both loop d. A total of two RHR pumps in either or both RHR loop QUESTION: 071 (1.00) Given the following conditions:
-- A plant transient has occurred which required entry into EOP-02,
"RPV Control, Failure-To-Scram"
-- The MSIVs are closed
-- The operators are attempting to establish pressure control using the main turbine bypass valves
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SELECT the MSIV isolation signal EOP-02 AUTHORIZES bypassing for these conditions, a. Low RPV water level High main steam line radiation level c. Low main condenser vacuum d. High main steam line flow l
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QUESTION: 072 (1.00) The size of the Safety Valves and Safety Relief Valves is based upon: a. a turbine trip from 100% WITH turbine bypass capability, a turbine trip from 100% WITHOUT turbine bypass capabilit c. MSIV closure with the reactor scramming as a direct result of valve closur d. MSIV closure with the reactor scramming as a result of the high flux transien QUESTION: 073 (1.00) A Given the following conditions:
-- The plant is shutdown with core alterations in progres All Source Range Monitoring (SRM) channels were reading a minimum of 3 cps prior to the start of core alteration SELECT the conditions allowing SRM counts to drop BELOW 3 cp a. A portable external neutron source is in continuous use to verify SRM operabilit b. All control rods are able to be verified fully inserted every 4 hours, c. The count rate on the affected channels had doubled prior to the decrease, d. The core is being unloaded via the spiral unloading fuel movement metho .
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SENIOR REACTOR OPERATOR Page 4 i QUESTION: 074 (1.00) Given the following conditions:
-- The plant is operating on the 100% load line -- The "A" Reactor Recirculation Pump trips -- Calculations indicate that total core flow is 23 Mlb/hr Which of the following conditions REQUIRE a manual reactor scram? -
a. LFRM oscillations of 15 watts /cm occu l
' SRM short period alarm periodically alarms then clear c. APRM oscillations of 15 percent peak to peak are observe d. An LPRM upscale alarm and downscale alarm exists in two separate detectors within an LPRM strin l QUESTION: 075 (1.00)
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Given the following conditions:
-- The Station Blackout.(SBO) diesel generator is carrying Bus A6
-- The "A" Emergency Diesel Generator (EDG) is supplying Bus A5-
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The SBO diesel generator load is limited to 1700 KW in order to: a. allow for starting the "B" or "D" RHR Pump ; i b. allow for starting the "B" Core Spray Pum prevent overloading the SBO DG if the "A" EDG faile ! d. prevent an overspeed trip in the event the output breaker is opene I l
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~ QUESTION: 076 (1.00) ,
Observing-the Drywell spray Initiation Limit curve when initiating drywell sprays will prevent:
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a. operation of the drywell-torus vacuum breaker ,
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b. exceeding the negative design pressure of the drywell j c. the occurrence of an evaporative cooling pressure dro d. drawing non-condensible gasses from the torus into the drywel t
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I s-QUESTION: 077 (1.00)
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While at rated pressure the Mode Switch is taken to SHUTDOWN.to scram the reactor. All blue scram lights on C905 remain EXTINGUISHE [ l This is indicative of: i failure of the backup scram valves to de-energiz j b. hydraulic locking of the scram discharge volum [ t c. electrical failure of the Reactor Protection Syste d. failure of individual hydraulic control accumulators to supply f
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pressure to scram.
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' SENIOR REACTOR OPERATOR' Page 44 ; . : QUESTION: 078 (1.00) ;
I During refueling activities with a bundle passing through the transfer canal, the refuel floor area radiation monitor alarm i Select the REQUIRED actio Return the' bundle to its previous location in the spent fuel l poo b. Locate the nearest open area and lower the fuel bundle into that are c. Leave fuel bundle in its current location and' ensure all-personnel evacuate the refuel floo d. Direct Radiological Protection personnel to conduct a survey of the area to verify _the alarm.
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-QUESTION: 079 (1.00) !
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While operating at 100% power, reactor pressure starts to increase as main generator output starts to decreas , Select the FIRST action the operator is directed to take to gain control of pressur a. Take the EPR POWER control switch to the OFF positio b. Hold the EPR SET PT control switch in the LOWER positio c. Hold the MPR SET PT control switch in the RAISE positio * d. Hold the MPR SET PT control switch in the LOWER positio I
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QUESTION: 080 (1.00) ,
'The mechanical vacuum pump will automatically isolate on high main steam line radiation in order to:
a. give the-operator the first indication of possible fuel element failur i
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b. prevent high radiation levels in the area of the Mechanical Vacuum Pum :
! limit release of activity from the main condense l limit release of activity past the seals on the mechanical' l vacuum pum QUESTION: 081 (1.00) :
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Given the following conditions:
-- The plant is in cold sautdown l -- Both Reactor Recirculation Pumps are "off" ;
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-- RHR "A" Loop is tagged out -- A pipe rupture occurs between the MO-28B and MO-29B (LPCI i Injection #1 and LPCI Injection #2) valves j Actual reactor water level should be raised to a MINIMUM of: j l
i a. +40 inche b. +50 inche c. +55 inche d. +60 inche ! l
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'. -QUESTION: 082 (1.00) i i
Given the following conditions: I
-- The plant is shutdown for a non-retueling outage -- A loss of shutdown cooling has just occurred -- Reactor water temperature is 145 degrees F and increasing at 3 .
degrees F per minute f How many minutes can elapse before primary containment integrity MUST be intact? (Round answer to nearest minute.) , minutes minutes minutes i minutes i i i QUESTION: 083 (1.00) The "A" Control Rod Drive (CRD) Pump has trippe What are the MINIMUM , conditions requiring a reactor scram? ; It is determined that the standby CRD Pump cannot be starte It is determined that the CRD Flow Control Valve cannot be l close j
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c. Two accumulator low pressure alarms occur within a nine rod l square arra j d. Three accumulator low pressure alarms occur within a nine rod l I square arra \ l
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W t SENIOR REACTOR OPERATOR Page 4 i
- QUESTION: 084 (1.00)
Due'to an incorrect off-gas lineup, condenser vacuum has reached 24 ' inches Hg. Generatur load is being reduced in an attempt to restore vacuu Taking generator load below 30% will cause: fatigue damage to the high pressura turbin , fatigue damage to the low pressure turbine last stage i c. excessive thrust on the low pressure turbine , flashing in the piping between the high pressure and low i pressure turbine !
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QUESTION: 085 (1.00) With the "B" TBCCW Pump out of service, the "A" TBCCW Pump trips and ; cannot be restarte Which of the following is a required 4IMMEDIATE action? , a. Close the MSIVs after the turbine trip Place the HPCI turbine in the injection mod ; c. Place the RCIC turbine in the pressure control mod l
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d. Reduce the number of running feedwater pumps to on r i s
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SENIOR REACTOR OPERATOR Page 48
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. QUESTION: 086- (1.00)
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Given the following conditions: :
-- The "A" loop of RHR is in shutdown cooling -- Reactor pressure at 50 psig * -- A complete loss of RBCCW has just occurred If Shutdown Cooling flow is to be maintained, the Heat Exchanger Bypass Valve (MO-16A) should be fully: closed to provide the maximum available coolin b. closed to minimize voiding on the RHR side of the heat exchange c. opened to minimize voiding on the-RBCCW side of the heat exchange .
i d. opened to minimize voiding on the RHR side of the heat
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exchange ,
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SENIOR REACTOR OPERATOR Page 4 QUESTION: 087 (1.00) , l
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Given the'following conditions:
-- There has been a nitrogen line leak in the drywell
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-- The Nitrogen / Air Isolation Valve to the Drywell (AO-4356) has been closed IDENTIFY the operational concern for the ADS safety relief valves (SRV)
for the above condition ; a. The ADS SRVs will not be operable until another source of air or 1 nitrogen is made availabl Each ADS SRV will be able to be operated through one complete open and close cycle using the air / nitrogen remaining in the , lines, ADS SRV operation is assured on an extended basis from an installed backup pneumatic sourc . d. The ADS SRVs will not be operable for the plant conditions as stated abov i QUESTION: 088 (1.00) i If drywell spray was initiated with torus level at 190 inches, containment failure could occur due to: ! l a. failure of the torus to.drywell vacuum breakers to operate ; properl b. failure of the Reactor Building to torus vacuum breakers to operate properl c. insufficient air space in the torus to relieve negative drywell pressure as drywell pressure is lowere d. overpressure in the torus due to non-condensibles being transferred from the drywell.
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l-SENIOR REACTOR OPERATOR Page 50 . i a l
!i QUESTION: -089 (1.00)
If drywell temperatures cannot be maintained below.280 degrees F, alternate depressurization is required in order to prevent failure of j
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the: a. containment pressure indicator b. containment wal c. reactor pressure indicator d. reactor pressure vessel water level indicator QUESTION: 090 (1.00) ; a In order to prevent reverse pump rotation and idle loop temperature stratification following a single Reactor Recirculation Pump trip, it is necessary.to .i l a, close the pump suction valve for 5 minutes, then reopen the- ! valv ]
.
b. close the pump discharge valve for 5 minutes, then reopen the I valv j
c. close the pump discharge valve, then immediately reopen the valve approximately 10%. d. close the pump suctico and discharge valve for 5 minutes,.then reopen both valves.
P
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zSENIOR REACTOR OPERATOR Page 51 l
. QUESTION: 091 (1.00)'
While operating at 100% power, a spurious signal causes a Group II and i
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Group VI isolatio , The spurious signal was from:
,
a. low reactor pressur . b. RWCU high flo c. high drywell pressur low reactor water leve l QUESTION: 092 (1.00) l
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Unisolable Reactor Water Cleanup (RWCU) system leaks have developed such
'that radiation levels in the "A" Pump area and "B" Pump area are ABOVE Maximum Safe Operating Value Alternate depressurization is required in order to:
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a. allow the RWCU system to stay in servic b. protect the secondary containment from failur i c. allow personnel access to the are j d. reduce the leak rate thereby allowing reactor building ventilation to be placed back in servic !
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SENIOR REACTOR OPERATOR Page 52 !
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I QUESTION: 093 (1.00) The following conditions exist: l
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-- Torus water level 120 inches
-- Reactor water level -150 inches and decreasing
-- Reactor pressure 10 psig
-- All available injection systems, injection subsystems, and alternate injection subsystems are injecting
;
The operator is required to: , a. perform RPV Floo , b. perform Coitainment Flood.
' c. perform Occam Coolin ! perform Alternate Depressurizatio , QUESTION: 094 (1.00) Which of the following would occur as a result of losing Essential DC f Bus D-6? a. Loss of indicating lights on the outboard MSIV , b. Loss of indicating lights on the inboard MSIV c. Closure of Reactor Building "A" isolation damper l Closure of Reactor Building "B" isolation damper !
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I h
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. J l SENIOR REACTOR OPERATOR Page 53 ) 1 l l l l QUESTION: 095 (1.00) Given the following conditions:
-- Instrument air header pressure is decreasing with no apparent cause -- AO-4310 (INST AIR DRYR BYP VLV) has automatically bypassed the instrument air dryer Which of the following automatic actions should have PRECEDED opening of AO-4310? Feedwater control valve locku Isolation of the service air heade c. Opening of the scram pilot air header trip valv Isolation of the non-essential instrument air heade QUESTION: 096 (1.00)
Which of the following conditions would violate secondary containment integrity? Both Drywell personnel access doors are ope b. Reactor Water Cleanup MO-2 (RWCU Suction) valve is failed ope c. Reactor Building ventilation is secured due to dampers failing close One Refuel Floor Exhaust Fan damper is failed open with the other Refuel Floor Exhaust Fan damper open and fully operabl l i l
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SENIOR REACTOR OPERATOR Page 54 e
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QUESTION: 097 (1.00) l j-A main turbine trip occurs from 100% powe Which of the following would be an expected automatic response? i i
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a. Turbine combined intermcdiate valves go full ope i Extraction steam to feedwater heaters will auto-isolat c. Reactor pressure will be maintained at 930-940 psig by the Mechanical Pressure Regulato d. Reactor pressure will be maintained at 960-970 psig by the Mechanical Pressure Regulato >
. QUESTION: 098 (1.00) ,
Given the following conditions:
!
-- Power is 100%.
-- Shift manning is at the Technical Specification MINIMU The STA is filling a licensed SRO positio i
-- A fire has been reported in the Turbine Building, WHICH of the following lists the MAXIMUM number and composition of operations personnel that can respond to the fire?
-
(All operations personnel on duty are qualified to be Fire Brigade _ members.)
a. 2 unlicensed operators, 1 licensed RO, and the STA , unlicensed operator, 1 licensed SRO, and the STA
! unlicensed operator and 1 licensed RO unlicensed operators ->
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... SENIOR REACTOR OPERATOR- 'Page'55 o I QUESTION: 099 (1.00) .j
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The purpose of the Standby Liquid Control (SLC) Pump discharge accumulator is to: a. provide a means of injection during a failure of both SLC pumps.- b. prevent spurious operation of the discharge relief valves when-both pumps are runnin c. minimize pulsations in system discharge pressur { d. ensure sodium pentaborate remains suspended in solutio ,
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QUESTION: 100 (1.00) , The Control Room has been evacuated due to a fire in the Cable Spreading Room. At WHICH of the following locations would you find an Alternate : Shutdown toolbox? ! a. 3' elev Aux Bay i b. 23' elev RPS MG Room , c. 37' elev Switchgear Room i
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d. 51' elev Turbine Building
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s (********** END OF EXAMINATION **********)
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.. ! SENIOR REACTOR OPERATOR Page 1 ,
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ANSWER KEY
MULTIPLE CHOICE 023 c 001 a 024 c 002 c 025 c 003 c 026 d ,
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004 d 027 b
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005 c 028 c 006 d 029 c 007 c 030 a 008 d 031 a
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009 c 032 b l l
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010 c 033 a
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011 c 034 c 012 b 035 b j
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013 b 036 a ') 014 a 037 a l 015 a 038 d 016 d 039 a I 017 b 040 c 018- b 041 c 019 c 042 b 020 a 043 c 021 b 044 a
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022 a 045 b i
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SENIOR-REACTOR OPERATOR , ANSWER KEY
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d 046 a A ND c MULTIPLE CHOICE 069 d i e 047 c 070 b > 048 a 071 a !
'l 049 d 072 d ! !
050- c 073 d j 051 d 074 c m 052 c 075 c l 053 b- 076 b i
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054 c 077 c { 055 d 078 b , t
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056 a 079 d , I 057 c 080 c 058 a 081 b k 059 b 082 c- f
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060 a 083 c ; 061 b 084 b 062 b 085 a {
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063 a 086 c ' i
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064 b 087 c 065 c 088 a ! l 066 c 089 b ..
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067 a 090 b :
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~ SENIOR REACTOR OPERATOR-' Page 3
. ANSWER K.E Y i! I t i
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091 d f
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MULTIPLE CHOICE -l-1
092 b l 093 b ; 094 b ; 095 b k 096 d : i 097 b ,
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098 a ; 099 c 100 c ,
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l (********** END OF EXAMINATION **********) _ . . _ . _ _ . ,
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SENIOR REACTOR OPERATOR Page 56- i 'a
.! -ANSWER: 001 (1.00) '
REFERENCE: Tagging Procedure No. 1.4.5, page 14, section 4.2.5[2] I
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[3.9/4.5]
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294001K102 ..(KA's) i ANSWER: 002 (1.00)
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i REFERENCE: ;
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10CFR20.101(b)
[3.3/3.8] ]
294001K103 ..'(KA's) ANSWER: 003 (1.00) i REFERENCE: i
:
10CFR50.54.x and 10CFR50.5 l Procedure 1.3.6, page 5, section 5.1[2] '; I
[3.3/4.2]
l l 294001A109 ..(KA's) ANSWER: 004 (1.00) d
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1 SENIOR REACTOR OPERATOR Page 57' b .! REFERENCE: Temporary Changes to Approved Procedures, N .3.4-3, page 8, section 6.1[1]
[4.2/4.2] >
294001A102 ..(KA's) , I ANSWER: 005 (1.00) . , l' REFERENCE: , Procedure 1.3.34, page 33, section 6.4[4] ~ 10CFR55.53(f)
[2.7/3.7] l'
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294001A103 ..(KA's) ANSWER: 006 (1.00) d' ; REFERENCE: Procedure 6.1-012 sections 8.2[1] and [3.3/3.8] 294001K103 ..(KA's) . AusWER: 007 (1.00) C.
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l l SENIOR REACTOR OPERATOR Page 58 1- j
. REFERENCE: l
CAF-probably 1.3.67 ) l
[2.7/3.7]
294001A103 ..(KA's) ;
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ANSWER: 008 (1.00) ; ,
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REFERENCE: 10CFR50 Appendix E, Section IV. CFR50. 72 (a) (3) ' EP-IP-110, pages 5-6, section 5.2.3-5. I
' [2.9/4.7]
294001A116 ..(KA's)
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ANSWER: 009 (1.00)
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REFERENCE: EP-IP-130, page 7, section ,
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294001A116 ..(KA's) !
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I ANSWER: 010 (1.00) l , f
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, SENIOR REACTOR OPERATOR Page 59 ;
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REFERENCE:
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EP-IP-200, page 1 of attachment 1 i
[2.9/4.7] {
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294001A116 ..(KA's) : ANSWER: 011 (1.00)
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C.
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REFERENCE: Procedure 1.4.10, page 4, section ,
[3.1/3.4]
294001K110 ..(KA's)
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ANSWER: 012 (1.00) REFERENCE: a Procedure 1.4.33, page 10, section 5.3.2[2]
[3.3/3.6] ;
294001K107 ..(KA's) : ANSWER: 013 (1.00) . - . .
.- _, -- = . . . - .- - . - . - - ..- . , , _
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' SENIOR REACTOR OPERATOR Page 60 i
REFERENCE: Procedure 1.3.34, page 32, section i
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294001A109 ..(KA's) l t h
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ANSWER: 014 (1.00) , REFERENCE: Procedure 6.1-022, page 8, section 6.0[4] 'l g
{3.3/3.8]
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294001K103 ..(KA's) ,
:
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ANSWER: 015 (1.00) l ' REFERENCE: , Procedure 1.5.9, page 8, section 4. 'f
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[4.2/4.2]
294001A102 ..(KA's) l t ANSWER: 016 (1.00) ,
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SENIOR REACTOR OPERATOR Page 61
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REFERENCE: , i l Procedure 1.3.34.2, page 4, Definitions T.S. Table 3. :
:
[3.4/3.6]
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294001A106 ..(KA's) .
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ANSWER: 017 (1.00) t
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REFERENCE: , l Procedure 2.2.19, page 15, section 4. Procedure 2.2.19, page 148, section 5.3(3)(b) ;
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[3.6/3.7]
:
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203000K114 ..(KA's)
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:
ANSWER: 018 (1.00) ;
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I REFERENCE: Instructor Guide O-RO-02-09-03, section VII.C.2, ELO 6 i i
(3.8/3.9] j l
206000K401 ..(KA's) -{
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ANSWER: 019 (1.00) 3
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l SENIOR' REACTOR OPERATOR Page 62 .I
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- REFERENCE:
Procedure 2.2.19, page 19, section 5.3[7] l
[3.7/3.9) i e
203000G010 'I
. . (KA's)
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' ANSWER: 020 (1.00) !
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- REFERENCE:
-
:
Procedure 2.2.21.5, page 10, section 1.0 CAUTION ,;
[3.3/3.4] .i
206000K501 . . (KA's) . l t ANSWER: 021 (1.00) : i REFERENCE: ,
:
Instructor Guide 0-RO-02-09-04, section III.3, ELO 12 l l
[3.5/3.5)
i l 217000K603 . . (KA's) 1
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)
ANSWER: 022 (1.00) ) a, P * "
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y
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SENIOR REACTOR OPERATOR Page-63 i
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REFERENCE:
.;
Instr'2ctor Guide O-RO-02-09-04, section V.4, ELO 8 i
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[3.1/3.2]
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2170C0A211 ..(KA's) i 023 (1.00)
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ANSWER:
* REFERENCE: h SLC Reference Text, page SLC-13, section D.1.b (no learning ,
objectives in this document)
[3.8/3.9] !
:
t I 211000A102 ..(KA's)
!
ANSWER: 024 (1.00) ? : REFERENCE: 3 ADS Reference Text, Figure 7 (no learning objectives listed in this document) : a i
[3.8/3.8] l
:
+
-218000K501 ..(KA's)
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ANSWER: 025 (1.00)
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d
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SENIOR REACTOR OPERATOR Page 64
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REFERENCE: High Pressure Coolant Injection System (HPCI) Reference Text, page HPCI 5 [4.3/4.4] l l 206000A101 ..(KA's) l
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.i ANSWER: 026 (1.00) REFERENCE:
PCIS Reference Text, page PCIS-15 section 3
[3.4/3.5] ,
223002K108 ..(KA's)
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l ANSWER: 027 (1.00) ! '
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REFERENCE: l j Procedure 2.2.20, page 6, section I i
[3.1/3.1]
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209001A208 ..(KA's) . ANSWER: 028 (1.00)
:
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:
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m . . _ _ _ _ _ _ _ _ _ . - . _ _ _ _ _ _,
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SENIOR REACTOR OPERATOR Page 65 t g !
: REFERENCE: ;
CS Reference Text, page CS-10, section 7 ,
[3.3/3.6] ,
209001A205 ..(KA's)
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ANSWER: 029 (1.00) REFERENCE:
:
RPS Reference Text, MSIV Closure Scram Coincidence Logic Diagram !
! [3.6/3.7]
i 212000K114 ..(KA's)
, '
ANSWER: 030 (1.00)
.i REFERENCE:
RPS Reference Text, page RPS-16, section 13
[4.1/4.1]
212000A417 ..(KA's) ANSWER: 031 (1.00) l l ! -, , , - - - - ,
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- SENIOR REACTOR OPERATOR Page 66
.s REFERENCE: i Primary Containment Atmosphere Control Reference Text., page PCI.C-2, , section 4 )
[3.0*/4.O]
!
223001G006 ..(KA's)
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ANSWER: 032 (1.00) REFERENCE: SGTS Reference Text, page SGTS-13, section [3.2/3.3] , 261000A301 . . ( l' A 's) .
.
ANSWER: 033 (1.00)
, .
REFERENCE:
Procedure 2.2.8, page 23, section 7. j j
[3.6/3.6]
264000K602 ..(KA's) -j ANSWER: 034 (1.00) i l I l j __ _ , . - . . - , , _ _ _ . . _ . _ _ . , _ _ , _ , , . , , . _ . _ . -- __ . _ . . _ . . _ .
_- _ - . .. - - - . . - - - ... . ._
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.. SENIOR REACTOR OPERATOR .Page 67
REFERENCE: RHR Reference Text, page RHR-28
[3.7/3.5]
226001G009 ..(KA's) i ANSWER: 035 (1.00)
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b REFERENCE: A k Nuclear Boiler Instrumentation Reference Text, page NBI-4
[3.6/3.8)
216000K507 ..(KA's)
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ANSWER: 036 (1.00) !
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REFERENCE: SRM Reference Text, page SRM-11, section 4-Interlocks and Trips
[3.7/3.7]
215004K401 ..(KA's)
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. ANSWER: 037 (1.00)- ; l
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'l
'l
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. SENIOR REACTOR OPERATOR Page 68 i REFERENCE: Procedure 2.4.37, page 5, section [4.2/4.3] 241000K302 ..(KA's) ANSWER: 038 (1.00) , > REFEREMCE: Procedure 2.2.87, page 18, precaution [3.2/3.3] ,
,
201001G010 ..(KA's)
,.
ANSWER: 039 (1.00) I REFERENCE:
!
RMC/RPIS Reference Text, page RMC-15, section D. [3.6/3.6] . I 201002K403 ..(KA's) ANSWER: 040 (1.00) ]
)
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,. . . . . . - - . .- -. .- .
SENIOR REACTOR OPERATOR' Page 69
' REFERENCE:
!
ER Rs erence Text, page RR-27, section D.I.b .
[3.2/3.3)
202001K103 ..(KA's)
.
. ANSWER: 041 -(1.00)
? ,
REFERENCE: s Procedure 2.2.83, page 10, section 5.1[5]
[3.2/3.2)
204000G010 ..(KA's)
!
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ANSWER: 042 (1.00)
, ,
REFERENCE: Procedure 2.2.19, Attachment 7, sheet 1, section 2.0[1]
[3.2/3.3)
!
205000G010 ..(KA's) - i
~i 043 (1.00)
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ANSWER: i l
:
>
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._, , . - . . . .. ,-. . . . - . _ . . . _
.-- _ . - _ . . . _ . . - . _ . _ . _ . . - _ : 4' < SENIOR REACTOR OPERATOR Page170 l . REFERENCE: ,
Rod Block Monitor Reference Text, page RBM-1, section 4 ;
-[3.3/3.4]
'!
215002G004 ..(KA's) , l i ANSWER: 044 (1.00) ,
; REFERENCE:
IRM Reference Text, page IRM-12, section 4 *
[4.0/4.0] ,
215003K402 ..(KA's)
,
l ANSWER: 045 (1.00) , REFERENCE: 1 j Technical Specifications Table 3. J IRM Reference Text, page IRM-12 1
[3.4/4.1]
215003G011 ..(KA's) ANSWER: 046 (1.00) CGS 1 a. $ggD C,. 1 i i l
.:,. ~. .- _
.i
- !
SENIOR REACTOR OPERATOR Page 71 ! k REFERENCE: Procedure 2.2.19, page 25, section 7. [3.4/3.5] 219000G010 ..(KA's) ANSWER: 047 (1.00) REFERENCE: PF't Reference Text, page PRM-6-5/89 and Figure 1
[3.6/3.8]
272000K109 ..(KA's) ANSWER: 048 (1.00) REFERENCE: Procedure 2.2.96, page 11, section 5.2.1[1]
[3.2/3.3]
259001G010 ..(KA's) ANSWER: 049 (1.00) d.
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I _ SENIOR REACTOR OPERATOR Page 72
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: REFERENCE:
PRM Reference Text, page PRM-22-5/89, section 4 (3.1/3.5] 271000A112 ..(KA's)
(1.00)
*
ANSWER: 050 REFERENCE: SCW Reference Text, page SCW-12, section (3.6/3.6] 245000A301 ..(KA's)
$
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ANSWER: 051 (1.00) l REFERENCE: CRDM Reference Text, page CRDM-10
[3.6/3.7]
201003K404 ..(KA's) ANSWER: 052 (1.00) l l
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SENIOR REACTOR' OPERATOR Page 73 .i REFERENCE: . TIP Reference Text, page TIP-22-5/89 I
[3.4/3.5]
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215001K401 ..(KA's) i ANSWER: 053 (1.00) ! REFERENCE: PRM Reference Text, page PRM-12-5/89
[3.4/3.4]
;
288000K102 ..(KA's)
;
.
ANSWER: 054 (1.00) REFERENCE: i
$
Condensate and Feedwater System Reference Text, Page FCS-16 :
[3.3/3.3] .
:
:
256000K102 ..(KA's) ;
:
i
' ANSWER: 055 (1.00) l , - . ._ . _ _ __ _ , . --
_ . _ _ - _ _-_ _ _ _ _ _ _ _ _ _ ___ _ __ ______. - . _._ ___.._._.._-,
*
SENIOR' REACTOR OPERATOR Page.74 *
,
k' REFERENCE: .
,
RPS Reference Text Figure 3
, [3.2/3.3]
212000K201 . . (KA's) l i ANSWER: 056 (1. 00) REFERENCE: Procedure 5.3.6, Revision 12<, page 4, section 3.0[1]
[3.9/4.1]
295003G010 . . (KA's) ,
-
t
!
ANSWER: 057 (1.00)
, REFERENCE:
Procedure 2.1.6, Revision 34, sections 4.0 and 6.2[2] l s-
[4.5/4.6]
, 295006A201 . . (KA's) L , ANSWER: 058 (1.00) ! , a.
\ i ,
.--.-...-..,...,+n .,i---- . _ , . . - -~-.r., ..-----.--.--*-----.----..------------.------.,w-__ __ ___E -
-.. _ - . - - . - . . - - . --
i
. 1-SENIOR' REACTOR OPERATO Page 75
.i 1
~
REFERENCE * l Training Module 0-RO-03-04-05, ELO 7 and page IG-6, section I
, [3.6/3.9]
295024G007 ..(KA's) ANSWER: 059 (1.00) REFERENCE: Training Module O-RO-03-04-04, ELO 20, page IG-53, section (3.4/3.8]
\
295026G007 ..(KA's) ANSWER: 060 (1.00) .
.
'
REFERENCE:
.i EOP-03 Entry Conditions i [4.4/4.6]
c 295026G011 ..(KA's) i t ANSWER: 061 (1.00) { . i
,
i
! '
i s
, - - - - - , , . .
-
, , - - . . , -- ,_____m
,
SENIOR REACTOR OPERATOR Page 76 ;
- .
.
REFERENCE: !
;
Technical Specifications, page 37, section 3.1 bases !
[3.3/4.2) !
;
,
295009G004 ..(KA's)
$
ANSWER: 062 (1.00) I REFERENCE: Instructor Guide O-RO-02-04-02, ELO 82, page IG-41
[3.9/3.9]
295009K202 ..(KA's) ANSWER: 063 (1.00) REFERENCE: Instructor Guide O-RO-03-04-05, page IG-23 (no facility specific learning objective identified)
[3.8/4.1]
295030K101 ..(KA's) ANSWER: 064 (1.00) . . . . __ _ _ . . _ _ _ . _ . . . . . . . ._ ____m- . . .
<!
..
'
SENIOR REACTOR OPERATOR Page 77
= i ' REFERENCE:
Instructor Guide 0-RO-03-04-07, ELO 3, page IG-2 !
. . [4.2/4.5] l i
l 295017G011 ..(KA's) i i
'
ANSWER: 065 (1.00) ;
;
REFERENCE:
;
Procedure 2.4.13, Revision 9, page 3 of 5, Immediate Operato ; Actions
[4.0/3.9]
i 295014G010 ..(KA's)
.
ANSWER: 066 (1.00)
! REFERENCE:
Procedure 2.4.40, Revision 11, page 2 of 4, section 3.0, Immediate Operator Action '
[3.8/3.6]
295038G010 ..(KA's) ANSWER: 067 (1.00) ! ,
..-r - .. . _ . . . - _ _ , _ _ . . . .-, . _ _ . - , . . ._ _ _ - . . _
- . . . - . - . . . . = . . . . _ . ~ . . . . - _ . . - . . . _ -
..
' SENIOR REACTOR OPERATOR Page 78 O
.' REFERENCE: l Instructor Guide O-RO-03-04-03, ELO 1.d, page IG-27
[3.9/4.1]
.i
'295007A104 ..(KA's) i ANSWER: 068 (1.00) REFERENCE: j EOP-02, Q-11 i
[4.3/4.4]
295037A203 ..(KA's) ;
!
]
ANSWER: 069 (1.00) l REFERENCE: Instructor Guide O-RO-03-04-05, ELO-6, page IG-26
[3.5/3.8]
295024K302 ..(KA's) ! l I l
' ANSWER:- 070 (1.00) '] .
1 > i l l l l l . - - . . . ~ . . . _ . . _ _ _
., . , . _ , . _ _ _ _ _ _ _ _ .
... -. .. . . ~ . . . ... - . . .. . -. . _ . ~
SENIOR REACTOR OPERATOR Page-79
.
,
REFERENCE:
+
Instructor Guide O-RO-03-04-05, ELO 14, page IG-11
[3.9/3.9]
295013A101 ..(KA's) ! t r ANSWER: 071 (1.00) REFERENCE: EOP-02, step P-4
[3.7/4.4]
-
295015G012 ..(KA's) l t i ANSWER: 072 (1.00) .! r
!
REFERENCE: I Technical Specifications, page 145, Safety and Relief Valve Bases ,
[3.1/4.2]
;
295025G004 ..(KA's) .j
-'1 ANSWER: 073 (1.00)
-1 ,
!
.!3
,
;
,
,
. . . , - . , _ . - - . . - ,, . . - - - . - .- -.
_ . --. _ . _ . _ . _ _ ._ _ _ _ . . - . . _ . . . . _ _ _ . . . _ _ _ l
SENIOR REACTOR OPERATOR Page 80
.-
l REFERENCE: l l Technical Specifications, page 203, section 3.1 i
[2.9/3.6] l
,
.295023G007 ..(KA's)
i ANSWER: 074 (1.00) ! REFERENCE: Procedure 2.4.17, Revision 14, page 5 of 9, section 4.1[4] ;
[4.1/4.2] ,
295014A201 ..(KA's) l ANSWER: 075 (1.00) , , i ; REFERENCE: ! Procedure 2.2.146, Revision 7, page 9 of 37, section 6.3[6] l
[3.8/4.0] ;
.
!
295003K106 ..(KA's) . L ANSWER: 076 (1.00) ' i
,
__ . , . - _ _ ~ . . . . _ . _ - _
. . _ _
- - - . - . . - . -
I l
.. I ENIOR S REACTOR OPERATOR Page 81 e
REFERENCE: i l Instructor Guide O-RO-03-04-02, Page IG-71, section !
' [3.6/3.8]
295010G007 . . (KA's)
: -ANSWER: 077 (1.00) REFERENCE:
Procedure 5.3.23, Revision 10, pages 4 and 6 of 16, section 2.1[1]. , and [4.0/4.1] 295015K204 . . (KA's) ANSWER: 078 (. 00) REFERENCE: Procedure 5.4.3, Revision 10, page 2 of 4, section .[3.8/3.9] 295023G010 . . (KA's)
,
ANSWER: 079 (1.00)
, ,
j i i
:
1 , - ,, -. _. _ _ , , _ , , . , , , .
. . . . .._. .-. - . . . - - . . _ . . . .. ~ - . . - - . ..
i
.
SENIOR REACTOR OPERATOR Page 82-
.,
REFERENCE: !
Procedure 2.4.37, Revision 10, page 2 of 6, section 3.0[1](b)
[3.5/3.7]
295007K201 ..(KA's) l i ANSWER: 080 (1.00) , i REFERENCE: Technical Specifications, page 193f, section 3/4.8.G Bases ,
[2.8/3.8] ,
295038G004 ..(KA's) ,
,
'
ANSWER: 081 (1.00)
! REFERENCE:
Procedure 2.4.25, Revision 15, page 3 of 7, section 4.0[4]
; [3.3/3.4]
295021K301 ..(KA's)
- i A
ANSWER: 082 (1.00) , i
I e . , b I
. , . . , , . . . . . _ , - , , _ - . . . . , . . . . _ . _ _ . - . _ , _
. _ _ . . . . . .. . . . _ . _ . _ -- - . . _ . - _ _ _ _ _ . _ _ _
.!
-.
;
SENIOR REACTOR OPERATOR Page 83 i i a *
,
'
. REFERENCE:
Technical Specifications, page 155, section 3. Procedure 2.4.25, Revision 15, page 3 of 7, section 4.0[5] , i
.[3.1/3.6] !
295021G003 ..(KA's) !
,
'
ANSWER: 083 (1.00) !
!
C.
-,
'
REFERENCE: Procedure 2.4.4, Revision 9, page 3 of 5, section , i
! [3.7/3.5] ;
i
!
295002G010 ..(KA's) ; I
!
ANSWER: 084 (1.00) ,
, .,
!
!
' REFERENCE:
;
Procedure 2.4.36, Revision 13, page 5 of 6, section 5.0[6] ; i '
[3.2/3.2] 2 295002G007 ..(KA's) '
.
I ANSWER: 085 (1.00)
,
1 t
--. " ~ - - - ' w te - a m--ur , ww-,y ,.,
- .. . . .
. -!
,
SENIOR REACTOR OPERATOR Page 84 o REFERENCE: Procedure 2.4.41, Revision 14, page 3 of 7, section [3.4/3.3] 295018G010 ..(KA's)
.;
ANSWER: 086 (1.00) ' REFERENCE: Procedure 2.4.42, Revision 14, page 6 of 7, section 5.0[4]
[3.1/3.3]
295018K303 ..(KA's)
!
(1.00)
<
ANSWER: 087 REFERENCE: ,
:
Procedure 2.4.21, Revision 5, page 3 of 4, section 5.0[2] ,
> [3.5/3.5]
295019K218 ..(KA's) ,
.
'
ANSWER: 088 (1.00)
, ,
f
,
i
'
n f
,
f
[ - , , - , . ,.
.. , . . . . .
l
.
SENIOR REACTOR OPERATOR Page 85
;
REFERENCE: Instructor Guide 0-RO-03-04-05, ELO 11, page IG-7, section [3.4/3.8] 295028G007 ..(KA's) l ANSWER: 089 (1.00) ; l l REFERENCE: l Instructor Guide O-RO-03-04-05, ELO 15, page IG-7, section [3.6/3.9] 295028K301 ..(KA's) ANSWER: 090 (1.00) REFERENCE: Procedure 2.4.17, Revision 14, page 8 of 9, section 5.0[5]
[3.2/3.6]
295001K305 ..(KA's) ANSWER:. 091 (1.00) _ - _ _ _ - _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
-
.l
.. SENIOR REACTOR OPERATOR Page 86
.
REFERENCE: PCIS Reference, Text, pages PCIS 75 and PCIS 76 (3.4/3.8) 295020A206 ..(KA's)
!
,
b ANSWER: 092 (1.00) REFERENCE: Instructor Guide O-RO-03-04-06, ELO 5, page IG-9, section ,
[3.3/3.5)
l 295033K301 ..(KA's)
ANSWER: 093 (1.00) l, REFERENCE: ; Instructor Guide O-RO-03-04-05, ELO 12, page IG-15, section B. ) i
[3.1/3.2]
295029K207 ..(KA's) ANSWER: 094 (1.00)
~ i
i
!
;
. _ _ __ _ _
T* . SENIOR REACTOR OPERATOR Page 87 g_ REFERENCE:
,
Procedure 5.3.13, Revision 12, page 2 of 7, section [3.3/3.3] 295004K203 ..(KA's) i ANSWER: 095 (1.00) ,
. . REFERENCE:
Procedure 5.3.8, Revision 16, page 2 of 9, section [3.2/3.2] , 295019K214 ..(KA's) '
.:
!
ANSWER: 096 (1.00) i REFERENCE: Technical Specification Definitions, page 4
[2.8/3.9] !
!
295035G003 ..(KA's) J ANSWER: 097 (1.00) ,
,
t b t i
-
l
!
1 .
, < , - . .
_ _ .
-
_ .. . . _ _ _ . _ . _ . _
$ .
SENIOR REACTOR OPERATO Page 88 3
:sf i
-
,
REFERENCE: Instructor Guide O-RO-02-04-01, page IG-16-7/90
[2.9/3.0] !
>
,
295005K202 ..(KA's) '
!
ANSWER: 098 (1.00) l t ] REFERENCE: .i
:
PNPS Technical Specifications, Table 6.2- l IG: 0-RO-06-01-04, " Technical Specification Design Factors and ;
'
Admin Controls," ELO 5 i
[3.5/3.8] 1
.l l
;
i l J 294001K116 ..(KA's)
!
ANSWER * 099 (1.00) ,
'l REFERENCE:
, IG: 0-RO-02-06-06, " Standby Liquid Control System", pg IG-5
[4.1/4.1]
211000G004 ..(KA's) ,
,
.1 ANSWER: 100 (1.00) l
.i
;
!
*
l
vm . - 1'8'- Y '- - "-v "W
- .- .. . . - . - . . - - . . . . . . . ..
I SENIOR REACTOR OPERATOR Page 89
../ . ;
REFERENCE: R PNPS Proc. No. 2.4.143,'Rev. 13, " Shutdown from Outside Control ; Room," page'. l OJT Guide O-RO-04-04, " Emergency Tasks," Task 6 i
')
'[4.1/4.1]
,
. 295016G006 ..(KA's) .
.
'
.
:
!
,
,
-.
~
!
'I
;
i i i
.
,
i
,
h a
.
I h
; !
,
.
.
(********** END OF EXAMINATION **********)
,
s ., ~, ,- . - - ,.4 , *
l i
. I TEST CROSS REFERENCE Page 1 4,
SRO Exam BWR Reactor organ 1 zed by Question Number ; J QUESTION VALUE REFERENCE * 001 1.00 9000001 , 002 1.00 9000002 003 1.00 9000003 004 1.00 9000004 - 005 1.00 9000005 !
'
006 1.00 9000006 007 1.00 22134 008 1.00 9000008 009 1.00 9000009 010 1.00 9000010 011 1.00 9000011 , 012 1.00 9000012 ' 013 1.00 9000013 014 1.00 9000014 , 015 1.00 9000015 016 1.00 9000016 017 1.00 9000017 018 1.00 9000018 ; 019 1.00 9000019 l 020 1.00 9000020 021 1.00 9000021 022 1.00 9000022 ! 023 1.00 9000023 , 024 1.00 9000024 025 1.00 9000025 026 1.00 9000026 027 1.00 9000027
-;
028 1.00 9000028 029 1.00 9000029 030 1.00 9000030 s 1 031 1.00 9000031 032 1.00 9000032 033 1.00 9000033 034 1.00 9000034 035 1.00 9000035 ; 036 1.00 9000036 * 037 1.00 9000037 .00 9000038 l l 039 1.00 9000039 ' 040 1.00 9000040 041 1.00 9000041 1 042 1.00 9000042 i 043 1.00 9000043 044 1.00 9000044 045 1.00 9000045 046 1.00 9000046 i 047 1.00 9000047 l 048 1.00 9000048 049 1.00 9000049 i l
. - . . . - , . . - . ..
..
TEST CROSS REFERENCE Page. 2 y SRO Exam BWR Reactor organized by Question Number QUESTION VALUE REFERENCE 050 1.00 9000050 : 051 1.00 9000051 I 052 1.00 9000052 053 054 1.00 .9000053 l 1.00 9000054 l 055 1.00 9000055 - 056 1.00 9000056 057 1.00 9000057 058 1.00 9000058 059 1.00 9000059 , 060 1.00 9000060 061 1.00 9000061 062 1.00 9000062 063 1.00 9000063 064 1.00 9000064 ' 065 1.00 9000065 066 1.00 9000066 . 067 1.00 9000067 068 1.00 9000068 069 1.00 9000069 070 1.00 9000070 071 1.00 9000071 ; 072 1.00 9000072 073 1.00 9000073 074 1.00 9000074 1 075 1.00 9000075 ;
'
076 1.00 9000076 077 1.00 9000077 078 1.00 9000078 079 1.00 9000079 080 1.00 9000080 081 1.00 9000081 082 1.00 9000082 , 083 1.00 9000083 ! 084 1.00 9000084 > 085 1.00 9000085 086 1.00 9000086 , 087 1.00 9000087 ' 088 1.00 9000088 ,
'
089 1.00 9000089 090 1.00 9000090 091 1.00 9000091 ' 092 1.00 9000092 093 1.00 9000093 094 1.00 9000094 $ 095 1.00 9000095 I , 096 1.00 9000096 j 097 1.00 9000097 098 1.00 22581
!
P
' , . - _ . - ,, .
.. -
. _ . .. . .
. ,
l
.-, l t
TEST CROSS REFERENCE Page 3 i
, * 1 SR Exam BWR Reactor f l
0rganized by Quention Number .
!
QUESTION VALUE REFERENCE l 099 1.00 18566 ! 100 1.00 22665 ______
,
100.00 l
'
______ ______ ; 100.00 i t i
i l l l l l l
'l .
' a I
- , . . _ - . . . . . . _
_ _ __ _ _ _ - .m._... -
. . _ . ._ . - _ .
t
. i TEST CROSS REFERENCE- Page.14
S. R O Exam B W R- Reactor Organized b y~ KA Group
:
PLANT WIDE GENERICS QUESTION VALUE KA
:
015 1.00 294001A102 { 004 1.00 294001A102 ! 007 1.00 294001A103 .; 005 1.00 294001A103 j 016 1.00 294001A106 1 00 .00 294001A109 l 013 1.00 294001A109 010 1.00 294001A116 009 1.00 294001A116 008 1.00 294001A116 , 001 1.00 294001K102 002 1.00 294001K103 006 1.00 294001K103 ! 014 1.00 294001K103 i f 012 1.00 294001K107 -l 011 1.00 294001K110
*
098 1.00 294001K116
------ PWG Total 17.00
'l PLANT SYSTEMS Group I ;
;
QUESTION VALUE KA 019 1.00 203000G010 , 017 1.00 203000K114 025 1.00 206000A101 , 018 1.00 206000K401 020 1.00 206000K501 028 1.00 209001A205 . 027 1.00 209001A208 023 1.00 211000A102 099 1.00 211000G004 030 1.00 212000A417 l 029 1.00 212000K114 055 1.00 212000K201 036' 1.00- 215004K401 ; 035 1.00 216000K507 022 1.00 217000A211 021 1.00 217000K603
024 1.00 218000K501 - 031 1.00 223001G006 026 1.0 K108 034 1.00 226001G009 i
, , __
j
-
. -_ _
. _ _ . - . - -
__ ..
, , . . . . . - . - - . .- - . ~ . -.- . TEST CROSS REFERENCE .Page- 5
,O tw d - S R'O Exam BWR Reactor Organized by KA Group !
,
PLANT SYSTEMS ,
'!
Group I
{
QUESTION VALUE KA 037 1.00 241000K302 '! 032 1.00 261000A301 033
'
1.00 264000K602 ______ PS-I Total 23.00
.
Group II * j
:
QUESTION VALUE KA q
;
'
038 1.00 201001G010 039 1.00 201002K403 - 040 1.00 -202001K103 041 1.00 204000G010 042 1.00 205000G010 043 1.00 215002G004 045 1.00 215003G011 I 044 1.00 215003K402 046 1.00 219000G010 050 1.00 245000A301 048 1.00 259001G010 049 1.00 271000A112 047 1.00 272000K109 ,
-
______ PS-II Total 13.00 Group III i QUESTION VALUE KA 051 1.00 201003K404 'I 052 1.00 215001K401 054 1.00 256000K102 053 1.00 288000K102 , ______ PS-III Total 4.00 ,
-
______ ______
,
PS Total 40.00 l
.;',
EMERGENCY PLANT EVOLUTIONS Group I j
!
'
. . _ . . . . . _ _
_ .
..
'
TEST CROSS REFERENCE Page 6
-&
'
SRO Exam BWR Reactor Organized by KA Group
.
EMERGENCY-PLANT EVOLUTIONS
,
Group I QUESTION VALUE KA ; 056 1.00 295003G010 075 1.00 295003K106 057 1.00 295006A201 , 067 1.00 295007A104 079 1.00 295007K201 , 061 1.00 295009G004 - 062 1.00 295009K202 076 1.00 295010G007 070 1.00 295013A101 074 1.00 295014A201 065 1.00 295014G010 071 1.00 295015G012 077 1.00 295015K204 100 1.00 295016G006 064 1.00 295017G011 073 1.00 295023G007 078 1.00 295023G010 - 2 058 1.00 295024G007 4
'
069 1.00 295024K302 , 072 1.00 295025G004 059 1.00 295026G007 060 1.00 295026G011 063 1.00 295030K101 068 1.00 295037A203 080 1.00 295038G004 066 1.00 295038G010 , ______ EPE-I Total 26.00 Group II QUESTION VALUE KA 090 1.00 295001K305 084 1.00 295002G007 083 1.00 295002G010 094 1.00 295004K203 097 1.00 295005K202 085 1.00 295018G010 086 1.00 295018K303 095 1.00 295019K214 087 1.00 295019K223 091 1.00 295020A206 082 1.00 295021G003 081 1.00 295021K301 _ _ _ __ .. , _ :.._- :
.
. '
TEST CROSS REFERENCE , Page 7 V SRO Exam BWR Reactor organized by KA Group EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA 088 1.00 295028G007 089 1.00 295028K301 093 1.00 295029K207 092 1.00 295033K301 096 1.00 295035G003 ______ EPE-II Total 17.00 ______ ______ EPE Total 43.00 ______ ______ ______ Test Total 100.00
,
!
;
,
l l l I i l
~ U ATTACIIMENT 2 NRC RESOLUTION OF NRC RESOLUTION OF FACILITY COMMENT FACILITY COMMENT Question 046: Accept answer c as a correct answer in addition to answer a. PNPS procedure 2.2.19, " Residual Heat Removal System," Section 5.3, " Precautions an i Limitations," Number 12 provides the justification to accept answer e as an addit J correct answe NRC RESOLUTION OF FACILITY COMMENT Question 046: Accepted answers a and c as correct answer i l
I
A1TACIIMENT 3 SIMULATION FACILITY REPORT Facility Licensee: Pilgrint Facility Docket No. 5Da91 9 Operating Test Administered on: November 30.1993 This form is to le used only to report observations. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future evaluations. No licensee action is required in response to these observation While conducting the simulator portion of the operating tests, the following items were observe [Qhi DESCRilrr1ON-NONE-
,
5 }}