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) ) FACILITY DOCKET NO: 50-293 FACILITY LICENSE NO: DPR-35 j LICENSEE: Boston Edison Company RFD #1 Rocky Hill Road Plymouth, MA 02360
FACILITY: Pilgrim Nuclear Power Station , EXAMINATION DATES: November 29-December 3,1993 EXAMINERS: C. Sisco, Operations Engineer , C. Tyner, Examiner (EG&G) . fl!Nk7 /b& V CHIEF EXAMINER: C. Sisco, Operations Engineer Date ' BWR Section, Operations Branch Division of Reactor Safety APPROVED BY: c h / Richard J. Conte, Cl[j/f Date ' ' 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 program.
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__ _ _ - _ _ _ _ . . DETAILS 1.0 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 7.
2.0 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 examination.
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 examiners.
The facility staff members who were involved in the preexamination activities signed security agreements to ensure that the examination was not compromised.
3.0 EXAMINATION RESULTS AND ADDITIONAL FINDINGS The results of the examination are summarized below: SRO Pass / Fail Written 3/0 ._ Operating 3/0 Overall 3/0 3.1 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 .. - .
Written Examination i
Strengths J The use of plant Technical SpeciReations to answer application questions.
! Knowledge of Emergency Core Cooling system operations.
't Weaknesses . Knowledge of protective clothing requirements when racking out a draw-out type of electrical breaker.
Knowledge of Emergency Diesel Generator conditions to start the DC drive fuel oil pump.
j Bases of the reactor low water level scram setpoint.
Simulator Examination Strengths Control room command function, crew communications and teamwork was a generic strength during the simulator examinations.
Weaknesses There were no generic weaknesses noted during the simulator examinations.
Walkthrouch Examination . Strengths The use of prints and logic diagrams was a generic strength.
The use of plant procedures and the knowledge and use of the plant technical specifications was a generic strength.
Weaknesses The knowledge of the plant response to the failure of the EPR was a generic weakness.
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Overall Conclusion The NRC examiners concluded that the applicants were well prepared for the examinations.
Also, the facility staff members acting as surrogate operators during the simulator examinations were very helpful to the NRC Examiners.
3.2 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 ' corrective action to revise the procedure.
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 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 aids.
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 examinations.
4.0 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 meeting.
Boston Edison Company E. S. Kraft, Jr.
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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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 , INSTRUCTIONS Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets.
Points for each question are indicated in parentheses after.the question.
The passing grade requires a final grade of at , least 80 percent.
Examination papers will be picked up 4 !' hours after the examination starts.
Al] work done on this examination is my own.
I have neither given nor received aid.
Applicant's signature RESULTS Examination Value 100 Points Applicant's Score Points Applicant's Grade Percent
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. ~SENI'R REACTOR OPERATOR Page. 2 ' ' l ANSWER SHEET l Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
. l MULTIPLE CHOICE 023 a b c d . 001-a b c d 024 a b c d
002 a b c d 025 a b c d { 003 a b c d 026 a b c d I 004 a b c d 027 a b c d i 005 a b c d 028 a b c d
006' a b c d 029 a b c d
007 a b' c d 030 a b c d
008 a b c d 031 a b c d
009 a b c d 032 a b c d 010 a b c d 033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c d 016 a b c d 039 a b c d 017 a b c d 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) ! ' 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
.
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 {
057 a b c d 080 a b c d ! '} 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 ' 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 ! 066 a b c d 089 a b c d.
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067-a b c d 090 a b c d
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- ANSWER SHEET
.:
Multiple Choice-(Circle or X your choice) l ' If you change:your answer, write your selection in the blank.
091 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
i 097 a b c d . 098 a b c-d ,
099 a b c d i
100 a b c d ,
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f i NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination, the following i rules apply: 1.
Cheating on the examination will result in a denial of your
application and could result in'more severe penalties.
2.
After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.
3.
To pass the examination, you must achieve a grade of 80 percent or greater.
. 4.
The point value for each question is indicated in , ' parentheses after the question number.
5.
There is a time limit of 4 hours for completing the examination.
6.
Use only black ink or dark pencil to ensure legible copies.
l 7.
Print your name in the blank provided on the examination ! cover sheet and the answer sheet.
8.
Mark your answers on the answer sheet provided and do not leave any question blank.
-i . 9.
If the intent of a question is unclear, ask questions of the examiner only.
- 30.
Restroom trips are permitted, but only one applicant at a
time will be allowed to leave.
Avoid all contact with ' ' anyone outside the examination room to eliminate even the ' appearance or possibility of cheating.
11.
When you complete the examination, assemble a package . including the examination questions, examination aids, and l ' answer sheets and give it to the examiner or proctor.
Remember to sign the statement on the examination cover , sheet.
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After you have turned in your examination, leave the examination area as defined by the examiner.
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-- . i l . i ! . QUESTION: 001 (1. 00) l When a Tagout is being performed on a centrifugal pump, which of the
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l a.
Control switch l b. Discharge valve l
c.' Auxiliary component power breakers I ! ! d.
Suction valve l ' ! i QUESTION: 002 (1.00) . I Given the following conditions: l
_j -- A 25 year old, male, fully trained radiation worker - -- 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? , i a.
1.25 Rem ' ; '
, b. 5.0 Rem c. 9.0 Rem ., .? d.
12.0 Rem j t t
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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 safety.
Per 10CFR50.54, the Operator at the Controls shall: a. take only those actions that comply with the Technical Specifications.
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 c.
take actions to protect the public health and-safety.
d. obtain verbal NRC concurrence to deviate from Technical Specifications, then take actions to protect the public health and safety.
. 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 error.
IDENTIFY the MINIMUM required approval prior to continuing with the test.
a. ORC Chairman b. Procedure Owner c. Nuclear Watch Engineer d. Nuclear Operations Supervisor
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b ! . . QUESTION: 005 (1.00) i An operator with a current NRC license has maintained his requalification training current but has NOT maintained proficiency.
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 watchstanding.
, b. Complete all licensed operator reactivation requirements and
obtain NRC concurrence.
Complete 40 hours of parallel watchstanding including a complete ! c.
plant tour and shift turnover.
f ! d. Complete seven 8-hour shifts of parallel watchstanding including , a complete plant tour and shift turnover.
, O i QUESTION: 006 (1.00) ! A locked room has a general area dose rate of 1.2 Rem /hr.
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 ! i 't !
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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 p.m.
Monday 7 a.m. to 7 p.m.
-- Tuesday -- 7 a.m.
to 7 p.m.
Wednesday - - 7 a.n. to 5 p.m.
7 a.m. to 3 p.m.
Thursday ' -- Friday 7 a.m. to 3 p.m.
-- 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 p.m.
! b.
7 a.m.
to 11 p.m.
> c. 7 a.m. to 5 p.m.
B d. 7 a.m. to 3 p.m.
, 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 hours.
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 hour.
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 event.
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 hour.
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. . . .. ...- . ..- . '9 . SENIOR REACTOR OPERATOR Page 11 l . , 'l ' QUESTION: 009 (1.00) i What is the MINIMUM emergency classification at which personnel-accountability must be initiated? a. Unusual Event b. Alert q ! 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 Director.
Which of the following MAY be delegated
to the Emergency Plant Manager? !
a. Emergency Classification b. Ordering a Site Evacuation
i c. Authorization of Emergency Exposures , d. Offsite Protective Action Recommendations
! ? ? ! 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 ! .t b. Safety glasses and half-face respirator c. Safety goggles and rubber safety-boots j l d. Safety goggles and half-face respirator ! ! ! ! ., ? - l l i I
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QUESTION: 012 (1.00) j i Which of the following is the MINIMUM required protective equipment when racking out a draw-out type breaker? , a. Full length coat, hard hat, and class "0" gloves i b.
Full length coat, hood, and class "2" gloves c.
Full face shield, class "0" gloves, flame retardant coveralls or
jacket.
d.
Full face shield, class "2" gloves, flame retardant coveralls or jacket.
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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? l i a. The reactor water temperature is 165 degrees F with shutdown cooling in service and fuel being removed from the vessel.
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 progress.
j i d. The unit is in cold shutdown with control rod drives being- ] replaced.
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Page 13 l . QUESTION: 014 (1.00) A specific radiation work permit would be required for which of the ~ following activities in a High Radiation Area? j - i a. Chemistry sampling ) b. Operator rounds
c. Radiological Protection surveys d.
Inspection of an area to locate a leak . QUESTION: 015 (1.00) Which of the following changes to a NON-SAFETY RELATED system can be accomplished WITHOUT processing a Temporary Modification?
a. Temporary hose used for system draining b. Temporary pipe supports , c. Plugging floor drains
d. Pulling circuit boards t .
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-.. - - - .~ . . -, -. . _ i . SENIOR REACTOR OPERATOR .Page 14 . 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 Scram.
b. Main Steam Line High Radiation Scram.
, c. Main' Steam Line Isolation Valve Closure Scram.
d. Turbine Stop Valve Closure Scram.
j , QUESTION: 017 ( 1.,00 ) 'While in shutdown cooling, a spurious high drywell pressure signal is i generated.
Select the expected automatic action.
l a. All MO-7 (Torus Suction) valves will open causing the MO-43 (Pump Suction) valves to close. LPCI will then inject.
b. The Shutdown Cooling (SDC) suction valves.(MO-47 & MO-50) will ', close causing the running RHR Pumps in the SDC loop to trip.
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.open.
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) valves.
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) . . . E QUESTION: 018 (1.00) During an initiation of HPCI, turbine exhaust pressure reaches 150 psi.
, ! Select the expected automatic action.
, a. The MO-4 and MO-5 (Inboard and Outboard Isolation Valves) and MO-35 and MO-36 (Torus Suction Valves) will close.
! b. The HPCI turbine stop and control valves both close.
! . c. The HPCI turbine control valve goes closed and the stop valve remains open.
., d. The HPCI turbine stop valve closes and the control valve fully opens.
l QUESTION: 019 -(1.00) > With two pumps operating in an RHR loop, excessive flow through the RHR heat exchangers is procedurally controlled by.
a.
limiting loop flow to 4000 GPM.
- b.
limiting loop flow to 6000 GPM.
c. requiring that the heat exchanger bypass valve-be open.
.; d.' requiring.that the RHR pump manual discharge valve be throttled.
i i QUESTION: 020 (1.00) ! Operating the HPCI turbine at 1800 RPM.will cause: ! a. water hammer in the exhaust line.
! .} b. water hammer in the pump discharge line.
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'! c. Booster Pump cavitation.
l d. erratic control valve operation due to low oil pressure.
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' QUESTION: 021 (1.00) ~. <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 ,' of a valid high torus level signal.
l The MO-22 (CST Suction) will: ! a, close, resulting in a trip of RCIC on low suction pressure.
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b. close when both RCIC suctions from the torus (MO-25 and MO-26) - ; valves are fully opened.
c. close when either of the RCIC suctions from the torus (MO-25 or MO-26) valve are fully opened.
d. close, causing the RCIC suctions from the torus (MO-25 and MO- , 26) to open.
! QUESTION: 022 (1.00) .: RCIC flow rate should be'at least 100 GPM in order to prevent:
a. oscillations in flow indication.
b.
loss of lubrication to pump bearings.
adequate cooling flow to the lube oil cooler.
c.
d. adequate cooling flow to the barometric condenser.
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-QUESTION: 023 (1.00) An ATWS has occurred and the "A" train of Standby Liquid Control (SLC) ~has been initiated.
Which of the following would indicate IMPROPER operation of the SLC.
system? , a. Reactor Cleanup System isolates.
b.
" Loss of Continuity to Squib Valve" alarm occurs.
c.
Squib Valve Ready' light is ON for the "A" trein.
,
d. Squib Valve Ready light is ON for the "B" train.
F k QUESTION: 024 (1.00) ' . 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 running.
i b. High drywell pressure for 2 minutes and one Core Spray Pump a running.
l c. High_drywell pressure and Low-Low reactor water level for 2 minutes with one RHR Pump running.
> d. Low-Low reactor water level for 11 minutes with one RHR Pump running.
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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 level.
SELECT the response of HPCI to' DECREASING level following the trip.
a. The HPCI high level trip must be manually reset for HPCI to automatically start.
, b. HPCI will automatically start at a reactor water level of -49" AND Drywell pressure greater than 2.5 psig.
c. HPCI will start if the HPCI high level trip is reset.
d. HPCI will automatically start at a reactor water level of -49" f I AND the HPCI Aux Oil Pump is manually started.
e QUESTION: 026 (1.00) , Given the following conditions: -- A low reactor water level signal is present.
-- Reactor pressure is 50 psig.
, , 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 open.
b. MO-47 (RHR outboard suction valve) is open.
' c. MO-7A, 7B, 7C, and 7D (torus suction valves) are closed.
' j d. MO-47 and MO-50 (RHR outboard and inboard suction valve) are open.
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. . __ _ . _ . _.. - _ _ i ! . ' SENIOR-REACTOR OPERATOR-Page 19 . . 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 psig.
b.
reactor pressure reaches 400 psig.
i c.
the ADS 2 minute timer times out, d. the ADS 11 minute timer times out.
. t f _ QUESTION: 028 (1.00)
, The Core Spray System is in a normal standby lineup.
The Core Spray Line Break Detection Monitor will detect a break located in the Core Spray injection line: a.
inside the reactor pressure vessel shroud.
b. upstream of the MO-24 (Outboard injection valve).
, i c. in the drywell upstream of the reactor pressure vessel penetration.
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.)
a.
, b.
. c. 4 ' d. 5 QUESTION: 030 (1.00) Alternate Rod Insertion has automatically actuated due to high reactor pressure.
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 signal.
I b. ensure that the Recirculation Pump field breaker trip. timer has timed out.
, i . prevent a loss of instrument air by limiting pressurization of c. the complete Scram Pilot Valve Air Header to once per minute.
d. allow sufficient time for pressure to decrease below the-setpoint prior to reset.
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SELECT the reason differential pressure must be restored to greater.than .1.2 psid.
a. To reduce the volume of water in the downcomers.
r b. Reactor Building to Torus vacuum breakers will 'pora.
Tc incorrect differential pressure during accident.,.. / i t (cur. c. Torus to drywell vacuum breakers will operate improperly.
d. Non-condensibles volume in the drywell will be insufficient for post accident conditions.
+ . 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 scfm.
Which of the following is an expected automatic action.
I a. The "A" fan will trip on a low flow signal.
b. The "B" train inlet damper will open.
c. The "B" fan will start causing.the "A" fan to trip.
d. The cross-tie damper will open to allow suction from the "B" train.
I
. . . . . _. .. ... - - . .- .. -SENIOR'R'EACTOR OPERATOR Page 22 ' .. ! 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? 10 psi a. Fuel oil pressure - Jacket water pressure - 20 psi Diesel speed - 900 rpm J ' b. Fuel oil pressure - 10 psi Jacket water pressure - 20 psi Diesel speed 80 rpm - , 5 psi c. Fuel oil pressure - Jacket water-pressure - 15 psi ' Diesel speed - 80 rpm 25 psi d. Fuel oil pressure - Jacket water pressure - 25 psi Diesel speed - 80 rpm QUESTION: 034 (1.00) ' 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? , ar. Momentarily depress the " Containment Spray' Reset" pushbutton.
, b. Momentarily depress the " Auto Isolation Inboard Valve" - I pushbutton.
, c. Place the "LPCI Override Control" switch for.the desired loop to MANUAL.
d. Place the "RPV Level Override Keylock" switch for the desired
loop to MANUAL OVERRIDE.
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. _ _ . - . ... _. _ _ __. _ _. _ _ _. _ . _ .. SEMIOR' REACTOR OPERATOR Page 23' . QUESTION: 035 (1.00) Due to a loss of ventilation, Reactor Building temperature has INCREASED by 20 degrees F.
i Assuming a constant reactor water level,' INDICATED narrow range reactor water level would: , a.
increase due to an decrease in variable leg density.
b.
increase due to an decrease in reference. leg density.
I c. decreasa due to an decrease in variable leg density.
d. decr:ase due to an decrease in reference leg density.
. QUESTION: 036 (1.00) I Select the condition that will bypass ALL SRM rod blocks.
a. Reactor modo switch in RUN.
b. All IRM range switches are on Range 4.
" c. All SRM detectors are fully withdrawn.
d. All SRM channels read 200 cps.
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SENIOR REACTOR OPERATOR Page 24 ' , . ! QUESTION: 037 (1.00)
While operating at 80% power, a failure occurs which causes the > ' electrical pressure regulator to send.a full "Open" signal to the ! turbine control valves.
Select the expected automatic response.
(Assume no operator actions ' taken.)
i a. The MSIVs will close when reactor pressure reaches 880 psig.
b. The reactor will scram when reactor pressure reaches 600 psig.
- c. The mechanical pressure regulator will take control and maintain pressure slightly lower than pressure prior to the transient.
> d. The mechanical pressure regulator will take control and maintain , pressure slightly higher than pressure prior to the transient.
, 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
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 valve.
b. excessive scram time of that rod during a reactor scram.
j c. damage to the accumulator in the event of a reactor scram.
d. damage to the drive mechanism in the event of a reactor scram.
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,. - - . . . .. .... ~...... .. - - . SENIOR REACTOR OPERATOR Page 25 , . b . QUESTION: 039 (1.00) ' .Which of the following will result in a control rod drift alarm? . A control rod that is: a. not selected for movement passes an " odd" reed switch.. 't ! b. not selected for movement passes an "even" reed switch.
- c. selected for motion but not receiving a motion signal passes an
" "even" limit switch.
d.
selected for motion and is being moved passes 2 " odd" limit j switches within 2 seconds.
j
QUESTION: 040 (1.00) , ' The individual Reactor Recirculation Pump speed controllers are in " Manual".
> Pump speed is limited to a MINIMUM of 26% in order to prevent: l . a. unstable recirc MG set operation.
b. cavitation of the jet pumps due to low. flow.
l c. temperature stratification in the reactor vessel.
d. excessive current from the generator due to low voltage from the volts /hz regulator.
i 'f
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_ _ _. _ _ , , - .. SENIOR REACTOR OPERATOR Page 26 o 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
Total Reactor Water Cleanup system flow must NOT exceed 150 gpm in order to prevent: a.
excessive vibration in the regenerative heat exchanger due to flow.
b. excessive temperature load on the non-regenerative heat exchanger.
! c. inadequate net positive suction head on the Reactor Cleanup + Pumps.
d.
excessive differential pressure across the filter demineralizers.
, 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 hours.
, In this condition, how is RHR "A" pump minimum flow protection provided? a. Administratively limiting minimum RHR pump flow to 500 GPM.
b. Administratively limiting minimum RHR pump flod to 1800 GPM.
I c. Automatic opening of-the MO-18A (minimum flow) valve when flow . ' drops to less than 2500 GPM for 10 seconds.
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 seconds.
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_- -_ - . .... -. . - . - - .- - _ __... I . SENIOR REACTOR OPERATOR Page 27 l . QUESTION: 043 (1.00) Which thermal limit does the Rod Block Monitor (RBM) prevent exceeding? I Average Planar Linear Heat Generation Rate (APLHGR) a.
b. Linear Heat Generation Rate (LHGR)
, 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 shutdown.
-- IRM "G" has failed " Upscale" and has NOT been bypassed.
-- 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%.
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, , , . - _ _ _ _ .... . - . ... . -SENIOR REACTOR OPERATOR Page 28 - , i , QUESTION: 045 (1.00) During plant operation at 100% power, APRM Channels "A" & "C" upscale trip units become inoperable.
, Which of the following actions will comply with Technical ' Specifications? , , a. Insert all operable control rods within 24 hours.
i b.
Insert a half scram on RPS "A" within 12 hours.
i i c.
Insert a half scram on RPS "B" within 12 hours.
d. Reduce power to IRM range and place the Mode Switch in "Startup/ Hot Standby" within 24 hours.
! ! QUESTION: 046 (1.00) , Torus cooling is in service using the "A" RHR Pump and.is in the process
of being secured.
l 1' The RHR pump should be secured when the system flow drops to 2000 GPM in order to- '! a. prevent air from becoming entrained in the system.
b. prevent opening the MO-18A (pump minimum flow) valve.
c. prevent excessive differential pressure across the MO-36A-(Torus Cooling) Valve.
.! d. prevent excessive differential pressure across the MO-34A-(Torus Cooling / Spray Block) Valve.
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_ .. ... _ _. . .- .-. .. _ _ _ - _ - -. . .. SENIOR REACTOR OPERATOR Page 29 i I 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 failed monitor
- 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 QUESTION: 048 (1.00) ! After operating for seven days, the "A" Reactor Feedwater Pump was manually tripped.
.j Which of the following is true concerning restarting of the "A" Reactor Feedwater Pump? j ! a. Pump starts are limited to three (3) starts in any one hour.
b. Pump starts are limited to five (5) starts in any one hour.
-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 start.
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? 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".
? ! } QUESTION: 050 (1.00) While operating at 90% power, Stator Cooling Water inlet pressure l suddenly drops to 10 psig.
l ! Which of the following conditions would result in a reactor scram? 'l i Time after pressure decrease - 1 minute -; a.
' -- Generator stator amps - 15,000 amps l -- Reactor power - 70%
-- Time after pressure decrease - 2 minutes ~ b.
-- Generator stator amps - 13500 amps i -- Roactor power - 60% i -- c. -- Time after pressure decrease - 4 minutes
Generator stator amps - 5250 amps
- - - Reactor power - 35% l -- _ d.
-- Time after pressure decrease - 5 minutes ) Generator stator amps - 4450 amps
-- Reactor power - 29% ! -- l . , ._- - ., _, _ __ . - . . _ _ _, - i
. . . ~ _ - __ . . . SENIOR REACTOR OPERATOR Page~31 ,. -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 rod.
j a.
200 psig
! b.
400 psig .l c.
600 psig d.'800 psig i QUESTION: 052 (1.00) . , > A TIP trace is being performed when a high drywell pressure signal 'f occurs.
l Select the expected automatic action.
a. The shear valve fires with the detector still in the core.
[ b. The ball valve closes with the detector still in the core.
i c. The detector withdraws into its shield and the ball valve closes.
! d. The detector withdraws into its shield and the shear valve
fires.
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QUESTION: 053 (1.00) Select the Refuel Floor Ventilation Exhaust Radiation Monitor trip ~ combination below that will result in a Reactor Building ventilation isolation.
! a.
"A" and "C" are " upscale".
' b.
"A" and "D" are " upscale".
c.
"A" and "D" are "downscale".
d.
"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 , ' condition -- All Condensate Pumps are running ' ,
SELECT the expected automatic system response.
a. The "A" RFP will trip immediately.
] i b. The "B" RFP will trip immediately.
c. The Condensate Pumps min-flow valve will immediately close.
d. The condensate Pumps min-flow valve closes after a one (1) minute time delay.
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SENIOR REACTOR. OPERATOR Page 33 s ) -QUESTION: 055 (1.00)~ Power for RPS MG Set "A" is supplied from: a.
B-8.
b. B-10.
! c. B-22.
d. B-23.
, 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 , d.
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 occurred.
l Which of the following conditions would require entry into EOP-2, , " Failure to Scram."
. a. Entered _directly based on the failure to' scram.
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 rods.
l c.
remain at position 04.
d. Entered.upon verification that the automatic scram signal did .not result in insertion of control rods to or beyond 02.
,
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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 psig.
Securing sprays at 2.5 psig is done to ensure that: primary containment pressure does.NOT drop below atmospheric.
a.
b. primary containment pressure does NOT drop below the scram setpoint.
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 F.
b. the torus water temperature will not exceed the heat capacity temperature limit.
c.
the torus water will not be heated to the point that the RHR and Core Spray pumps lose adequate net positive suction head.
d. the torus water will not be heated to the point that Drywell , pressure increases solely due to high torus temperature.
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a'
" QUESTION: 060 (1.00).
Which condition would require entry into EOP-03, " Primary Containment
Control"?
a. Torus water temperature is 82 degrees F.
] b. Drywell temperature is 145 degrees F.
c. Torus pressure is 1.7 psig.
! d. Drywell hydrogen concentration is 2.5% t i ! ' QUESTION: 061 (1.00) . The reactor low water level scram setpoint is selected to: . } prevent violating the Minimum Critical Power Ratio Operating a.
Limit.
i b. prevent violating the Minimum Critical Power Ratio Safety Limit.
c. ensure a reactor scram has been completed prior to a Group I t ! isolation, d. ensure that fuel uncovery does not occur during a level , transient at power.
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' QUESTION: 062 (1.00) ! , 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 setpoint.
l ! b. decrease to the point that the reactor scrams on low water
- !
i.evel.
l increase to compensate for the feed flow / steam-flow mismatch.but
c.
will remain below the turbine trip setpoint.
d. increase to the point that a turbine trip occurs which will result in a reactor scram.
! ! 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 - 90 inches.
Which of the following will become uncovered at 90 inches?
a. Drywell downcomers ' b. SRV spargers c. HPCI steam exhaust line , d. RCIC steam exhaust line
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. . _. . _ . _. - . ., SENIOR REACTOR OPERATOR _Page 37= ] '\\ . 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? ! a. Unusual Event .. ! b. Alert l c. Site Area Emergency d. General Emergency
QUESTION: 065 (1.00) . Reactor power begins to increase at 5%.per minute with no apparent i
cause.
l At what' point is the operator is REQUIRED to manually scram the reactor? < a. Any APRM has two LPRMs inputs of 110% or greater.
b. Any APRM has three LPRM inputs of 110% or greater.- ! i c. Two LPRMs within a string reach 110% or greater.
d. Three LPRMs within a string reach 110% or greater.
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. - ! . SENIOR REACTOR OPERATOR Page 38 Y ! ! QUESTION: 066 (1.00)
With reactor power at 50% on the 50% load line, a rapid increase in off- ,
- gas radiation level' occurs.
SELECT the REQUIRED operator actions for the Reactor Recirculation , Pumps.
, I The Reactor Recirculation Pumps shall: a. be reduced in speed to lower activity but maintained above 31.5.
j Mlb/ hour.
' b. be reduced to minimum speed, then tripped if necessary to terminate the increase in activity.
i be reduced to minimum speed if necessary to terminate the ! c.
' increase in activity.
d. be tripped immediately to lower power to 40% as rapidly as
possible.
, QUESTION: 067-(1.00)
! -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? .
' This sequence is designed to: a. avoid high local pool temperatures.
s b. utilize the highest capacity valve first.
c.
ensure that relief valves nearest torus temperature detectors are operated last.
d. delay discharging a relief valve into the same area as the HPCI-steam exhaust.
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! , QUESTION: 068 (1.00)
Boron injection using the Standby Liquid Control (SLC) System has been , ' initiated in accordance with EOP-02, "RPV Control, Failure-To-Scram.
l The SLC Pumps should: k a. automatically trip on low suction pressure.
] b.
automatically trip on low SLC tank level at 50 gallons.
}
be manually tripped if SLC tank level drops-to 0 gallons, c.
d. be manually tripped if SLC tank level drops by 1600 gallons.
! !
? E-QUESTION: 069 (1.00) . A plant transient is in progress that results in a rapidly rising. torus j pressure.
. 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 spargers.
b. fatigue failure of the torus-drywell' vacuum breakers.- operation of the reactor building-torus vacuum breakers which i c.
' would cause deinertion.
Ed. downcomer failure and the resultant pressurization of the torus airspace.
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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 running.
a.
b. Both loops of RHR with one RHR pump running in each loop.
Both loops of RHR with both RHR pumps running in both loops.
c.
d. A total of two RHR pumps in either or both RHR loops.
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 , SELECT the MSIV isolation signal EOP-02 AUTHORIZES bypassing for these conditions, a. Low RPV water level b.
High main steam line radiation level c. Low main condenser vacuum d. High main steam line flow
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, SENIOR REACTOR OPERATOR Page 41
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, b. a turbine trip from 100% WITHOUT turbine bypass capability.
c. MSIV closure with the reactor scramming as a direct result of valve closure.
d. MSIV closure with the reactor scramming as a result of the high flux transient.
QUESTION: 073 (1.00) A Given the following conditions: -- The plant is shutdown with core alterations in progress.
-- All Source Range Monitoring (SRM) channels were reading a minimum of 3 cps prior to the start of core alterations.
SELECT the conditions allowing SRM counts to drop BELOW 3 cps.
a. A portable external neutron source is in continuous use to verify SRM operability.
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 method.
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SENIOR REACTOR OPERATOR Page 42.
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? - l a. LFRM oscillations of 15 watts /cm occur.
' b.
SRM short period alarm periodically alarms then clears.
c. APRM oscillations of 15 percent peak to peak are observed.
d. An LPRM upscale alarm and downscale alarm exists in two separate detectors within an LPRM string.
l QUESTION: 075 (1.00)
Given the following conditions: -- The Station Blackout.(SBO) diesel generator is carrying Bus A6 -- The "A" Emergency Diesel Generator (EDG) is supplying Bus A5- ' The SBO diesel generator load is limited to 1700 KW in order to: a. allow for starting the "B" or "D" RHR Pumps.
i b. allow for starting the "B" Core Spray Pump.
c.
prevent overloading the SBO DG if the "A" EDG failed.
! d. prevent an overspeed trip in the event the output breaker is opened.
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, -.~. _.. - . . -- - - .\\ .. SEMIOR REACTOR OPERATOR Page 43' Di ) i - ~ QUESTION: 076 (1.00) , Observing-the Drywell spray Initiation Limit curve when initiating drywell sprays will prevent: , a. operation of the drywell-torus vacuum breakers.
, , b. exceeding the negative design pressure of the drywell j c. the occurrence of an evaporative cooling pressure drop.
d. drawing non-condensible gasses from the torus into the drywell.
t
I s-QUESTION: 077 (1.00) ' While at rated pressure the Mode Switch is taken to SHUTDOWN.to scram the reactor.
All blue scram lights on C905 remain EXTINGUISHED.
[ l This is indicative of: i failure of the backup scram valves to de-energize.
j a.
b. hydraulic locking of the scram discharge volume.
[ t c. electrical failure of the Reactor Protection System.
d. failure of individual hydraulic control accumulators to supply f pressure to scram.
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- QUESTION: 078 (1.00)
I During refueling activities with a bundle passing through the transfer canal, the refuel floor area radiation monitor alarms.
i Select the REQUIRED action.
a.
Return the' bundle to its previous location in the spent fuel l pool.
b. Locate the nearest open area and lower the fuel bundle into that area.
c. Leave fuel bundle in its current location and' ensure all-personnel evacuate the refuel floor.
d. Direct Radiological Protection personnel to conduct a survey of the area to verify _the alarm.
! ' -QUESTION: 079 (1.00) - While operating at 100% power, reactor pressure starts to increase as main generator output starts to decrease.
, Select the FIRST action the operator is directed to take to gain control of pressure.
a. Take the EPR POWER control switch to the OFF position.
b. Hold the EPR SET PT control switch in the LOWER position.
c. Hold the MPR SET PT control switch in the RAISE position.
- d. Hold the MPR SET PT control switch in the LOWER position.
i i l
. - . -... - -.. . . -. -. - . . I SENIOR REACTOR OPERATOR Page 45 ] w 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 failure.
i . b. prevent high radiation levels in the area of the Mechanical Vacuum Pump.
! c.
limit release of activity from the main condenser.
l d.
limit release of activity past the seals on the mechanical' vacuum pump.
QUESTION: 081 (1.00)
Given the following conditions: -- The plant is in cold sautdown -- Both Reactor Recirculation Pumps are "off" -- RHR "A" Loop is tagged out j -- 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 i a. +40 inches.
b. +50 inches.
c. +55 inches.
d. +60 inches.
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_. .~ ... - -... . _ - - ~.. ... . ,. - . -.. . . i . SENIOR REACTOR OPERATOR Page 46 l '. -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.)
, a.
13 minutes b.
18 minutes c.
22 minutes i d.
32 minutes i i i QUESTION: 083 (1.00) The "A" Control Rod Drive (CRD) Pump has tripped.
What are the MINIMUM , conditions requiring a reactor scram? a.
It is determined that the standby CRD Pump cannot be started.
b. It is determined that the CRD Flow Control Valve cannot be l closed.
j \\ c. Two accumulator low pressure alarms occur within a nine rod l square array.
j d. Three accumulator low pressure alarms occur within a nine rod I square array.
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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 vacuum.
Taking generator load below 30% will cause: fatigue damage to the high pressura turbine.
a.
, b. fatigue damage to the low pressure turbine last stages.
i c. excessive thrust on the low pressure turbines.
, d.
flashing in the piping between the high pressure and low pressure turbines.
i ! , QUESTION: 085 (1.00) With the "B" TBCCW Pump out of service, the "A" TBCCW Pump trips and
cannot be restarted.
Which of the following is a required 4IMMEDIATE action? , a. Close the MSIVs after the turbine trips.
b.
Place the HPCI turbine in the injection mode.
c. Place the RCIC turbine in the pressure control mode.
l
- '
d. Reduce the number of running feedwater pumps to one.
r i s , i ! , D - - ~ -- ,,.,
. .. _ . -. .. .- _ _ .-. .... I '. I SENIOR REACTOR OPERATOR Page 48 ... . . QUESTION: 086-(1.00) . 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 cooling.
a.
b. closed to minimize voiding on the RHR side of the heat exchanger.
c. opened to minimize voiding on the-RBCCW side of the heat exchanger.
. i d. opened to minimize voiding on the RHR side of the heat ' exchanger.
, . . t ) ) , l
I i - -., _ - .l b
. _. - _ - _ ,. . SENIOR REACTOR OPERATOR Page 49.
-4 QUESTION: 087 (1.00) , l . Given the'following conditions: -- There has been a nitrogen line leak in the drywell -- 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 conditions.
a. The ADS SRVs will not be operable until another source of air or
nitrogen is made available.
b.
Each ADS SRV will be able to be operated through one complete open and close cycle using the air / nitrogen remaining in the , lines, c. ADS SRV operation is assured on an extended basis from an installed backup pneumatic source.
. d. The ADS SRVs will not be operable for the plant conditions as stated above.
i i QUESTION: 088 (1.00) 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 properly.
b.
failure of the Reactor Building to torus vacuum breakers to operate properly.
c.
insufficient air space in the torus to relieve negative drywell pressure as drywell pressure is lowered.
d. overpressure in the torus due to non-condensibles being transferred from the drywell.
'
, . _.... . -, -
.-- -.. , i l ~ -SENIOR REACTOR OPERATOR Page 50. l . i a 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 the: ' a. containment pressure indicators.
b. containment wall.
c. reactor pressure indicators.
d. reactor pressure vessel water level indicators.
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-valve.
] . b. close the pump discharge valve for 5 minutes, then reopen the I valve.
-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 , --- ,, _, _,,
. _.
. . . _, . _.. - _.
. .. - _ _
. zSENIOR REACTOR OPERATOR Page 51 l . QUESTION: 091 (1.00)' While operating at 100% power, a spurious signal causes a Group II and i Group VI isolation.
. , The spurious signal was from: , a. low reactor pressure.
. b. RWCU high flow.
c. high drywell pressure.
d.
low reactor water level.
l QUESTION: 092 (1.00) l , 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 Values.
Alternate depressurization is required in order to: . a. allow the RWCU system to stay in service.
b. protect the secondary containment from failure.
i c. allow personnel access to the area.
j d. reduce the leak rate thereby allowing reactor building ventilation to be placed back in service.
'
! ., - -. - -. . - ~ .-.
__.
_... - = ... _ _. _ _ . _ __ . __ -.. - _ _ _ __ i . SENIOR REACTOR OPERATOR Page 52 . , I QUESTION: 093 (1.00) The following conditions exist: l ' -- 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 Flood.
, b. perform Coitainment Flood.
' c. perform Occam Cooling.
! d.
perform Alternate Depressurization.
, 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 MSIVs.
, b. Loss of indicating lights on the inboard MSIVs.
c. Closure of Reactor Building "A" isolation dampers.
l d.
Closure of Reactor Building "B" isolation dampers.
! > I h .
l .--,,.,y.
,,.-,.,-.m ....,... ,,,-., _ .. - . m., , _ - - .,, ....~m.
. -. -..-4 - - - -
J . l SENIOR REACTOR OPERATOR Page 53
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 dryers.
Which of the following automatic actions should have PRECEDED opening of AO-4310? a.
Feedwater control valve lockup.
b.
Isolation of the service air header.
c. Opening of the scram pilot air header trip valve.
d.
Isolation of the non-essential instrument air header.
QUESTION: 096 (1.00) Which of the following conditions would violate secondary containment integrity? a.
Both Drywell personnel access doors are open.
b. Reactor Water Cleanup MO-2 (RWCU Suction) valve is failed open.
c. Reactor Building ventilation is secured due to dampers failing closed.
d.
One Refuel Floor Exhaust Fan damper is failed open with the other Refuel Floor Exhaust Fan damper open and fully operable.
i . . . - - - -
_ .. . .. _ _. _, . _ . _ _. _... . .- _ . . . SENIOR REACTOR OPERATOR Page 54 e
' QUESTION: 097 (1.00) l-A main turbine trip occurs from 100% power.
Which of the following j i would be an expected automatic response? i ' a. Turbine combined intermcdiate valves go full open.
i b. Extraction steam to feedwater heaters will auto-isolate.
c. Reactor pressure will be maintained at 930-940 psig by the Mechanical Pressure Regulator.
d. Reactor pressure will be maintained at 960-970 psig by the Mechanical Pressure Regulator.
> . QUESTION: 098 (1.00) , Given the following conditions: ! -- Power is 100%. -- Shift manning is at the Technical Specification MINIMUM.
-- The STA is filling a licensed SRO position.
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 , b.
1 unlicensed operator, 1 licensed SRO, and the STA ! c.
1 unlicensed operator and 1 licensed RO d.
2 unlicensed operators - >
i i t , , m ._ . ,_ - __ .. _ n . . . - -,:--
. . , _..-. . . ..-- .- -_ _ -.. ... SENIOR REACTOR OPERATOR- 'Page'55 o I QUESTION: 099 (1.00) .j 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 running.
c. minimize pulsations in system discharge pressure.
{ d.
ensure sodium pentaborate remains suspended in solution.
, ! ! . 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
d.
51' elev Turbine Building ! , ,
.
i ! l . s (********** END OF EXAMINATION **********) ,
- , -, m _
. _ _ __ _ .. SENIOR REACTOR OPERATOR Page
, , ANSWER KEY
MULTIPLE CHOICE 023 c 001 a 024 c 002 c 025 c 003 c 026 d ,
004 d 027 b , 005 c 028 c 006 d 029 c 007 c 030 a 008 d 031 a 009 c 032 b l l ' 010 c 033 a ' 011 c 034 c 012 b 035 b j ! ') 013 b 036 a 014 a 037 a 015 a 038 d 016 d 039 a 017 b 040 c 018-b 041 c 019 c 042 b 020 a 043 c 021 b 044 a 022 a 045 b i !
... / . Page. 2 SENIOR-REACTOR OPERATOR , ANSWER KEY , i , l , ' d 046 a A ND C.
068 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 ! 054 c 077 c { 055 d 078 b , t ' 056 a 079 d , I 057 c 080 c 058 a 081 b k f 059 b 082 c- ! 060 a 083 c
061 b 084 b { 062 b 085 a
063 a 086 c ' i ' 064 b 087 c 065 c 088 a ! l 066 c 089 b ..
067 a 090 b
! ! I f
! t , _ , -- ..
. .. - _. .. -..... - .. .~ - - _.- h .; . ' ~ SENIOR REACTOR OPERATOR-' Page
.a.
ANSWER K.E Y i ! I ti
? ! f 091 d . , MULTIPLE CHOICE-l -1
092 b l 093 b
094 b
095 b k 096 d
i 097 b , , 098 a
099 c 100 c , ! , > i-
i , l
i , i t i I i f , . l , (********** END OF EXAMINATION **********) .. . .. . . _ _.. _. _ _. ,
. i . SENIOR REACTOR OPERATOR Page 56-i 'a .! -ANSWER: 001 (1.00) a.
' REFERENCE: Tagging Procedure No.
1.4.5, page 14, section 4.2.5[2] I [3.9/4.5]
! 294001K102 ..(KA's) i ANSWER: 002 (1.00) . C.
, i REFERENCE:
. 10CFR20.101(b) [3.3/3.8] ] 294001K103 ..'(KA's) ANSWER: 003 (1.00) i c.
REFERENCE: i
10CFR50.54.x and 10CFR50.54.y Procedure 1.3.6, page 5, section 5.1[2] '; I [3.3/4.2] l 294001A109 ..(KA's) ANSWER: 004 (1.00) d
.... ._ . _ -. _ _. _ , _. . _. _. . . - __ .. T . 1 SENIOR REACTOR OPERATOR Page 57' b .! REFERENCE: Temporary Changes to Approved Procedures, No.
1.3.4-3, page 8, section 6.1[1] [4.2/4.2] > 294001A102 ..(KA's) , I ANSWER: 005 (1.00) . C.
, l' REFERENCE: , Procedure 1.3.34, page 33, section 6.4[4] ~ 10CFR55.53(f) [2.7/3.7] l' 294001A103 ..(KA's) > ANSWER: 006 (1.00) d'
REFERENCE: Procedure 6.1-012 sections 8.2[1] and 3.0 [3.3/3.8] 294001K103 ..(KA's) AusWER: 007 (1.00) . C.
l l ! t k l i l [ ! .- - . _... _. _ _. _... _..... - -.. _ ~., _ _. . _ _, _. _., _. _, , _ _,,, , i .
- l SENIOR REACTOR OPERATOR Page 58 j 1- . REFERENCE:
CAF-probably 1.3.67 ) [2.7/3.7] 294001A103 ..(KA's)
, ANSWER: 008 (1.00)
d.
, ' REFERENCE: 10CFR50 Appendix E, Section IV.D.3 10CFR50. 72 (a) (3) ' EP-IP-110, pages 5-6, section 5.2.3-5.2.4 I' [2.9/4.7] 294001A116 ..(KA's) ' . ANSWER: 009 (1.00) < ' c.
' REFERENCE: EP-IP-130, page 7, section 5.5 , ! [2.9/4.7) , 294001A116 ..(KA's) ! .! . I ANSWER: 010 (1.00) l c.
, f . $ ,
_ _ _.
.. -... .... _.. _.
. . . -_ .. . -. .
. SENIOR REACTOR OPERATOR Page 59
, %. ' REFERENCE: . I EP-IP-200, page 1 of attachment 1 ' i [2.9/4.7] { ! ! 294001A116 ..(KA's) ANSWER: 011 (1.00) ' C.
. REFERENCE: ' Procedure 1.4.10, page 4, section 5.2 , [3.1/3.4] 294001K110 ..(KA's)
ANSWER: 012 (1.00) b.
REFERENCE: a Procedure 1.4.33, page 10, section 5.3.2[2] [3.3/3.6]
294001K107 ..(KA's)
ANSWER: 013 (1.00) b.
..- - - - - - -. -..
.- _, -- . -. =... -.- - - - ..- . ,, _ i . ' . ' ' SENIOR REACTOR OPERATOR Page 60 i REFERENCE: Procedure 1.3.34, page 32, section 6.3 i ! -[3.3/4.2] - i !
294001A109 ..(KA's) l t h . ANSWER: 014 (1.00) a.
, REFERENCE: Procedure 6.1-022, page 8, section 6.0[4] g 'l {3.3/3.8] .'.
294001K103 ..(KA's) ,
ANSWER: 015 (1.00) l a.
' REFERENCE: , Procedure 1.5.9, page 8, section 4.2.1 'f -
! [4.2/4.2] 294001A102 ..(KA's) l t ANSWER: 016 (1.00) , d.
! t i l' , ' . . - _ ,- .
, . . _.. . - ._ .._ . _. _. - -...~_ - ,. . -
.
SENIOR REACTOR OPERATOR Page 61 . REFERENCE: , i Procedure 1.3.34.2, page 4, Definitions l T.S.
Table 3.1.1
[3.4/3.6] 294001A106 ..(KA's) . .
ANSWER: 017 (1.00) t , b.
, REFERENCE: , l Procedure 2.2.19, page 15, section 4.2.4 Procedure 2.2.19, page 148, section 5.3(3)(b)
' [3.6/3.7]
! 203000K114 ..(KA's) .
ANSWER: 018 (1.00)
b.
'
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) -{ . ANSWER: 019 (1.00)
) c.
, '
i
.. . ..
. --. - . . - ., . 4 ' SENIOR' REACTOR OPERATOR Page 62 . '
. ? - REFERENCE: Procedure 2.2.19, page 19, section 5.3[7] l [3.7/3.9) i e 203000G010 (KA's) 'I .. -,
--
' ANSWER: 020 (1.00) , ! a.
, - 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 b.
REFERENCE: , Instructor Guide 0-RO-02-09-04, section III.3, ELO 12 l [3.5/3.5) i l 217000K603 (KA's)
.. . ) ANSWER: 022 (1.00) ) a, P
Tv--t ek-- " W-rW t'=*-W ev71-r m e.w-
4 re-w
. . .. _ __ .. . _. _ _ _.. _ _ _ _. -. .. .. _, _ . ,_ y . SENIOR REACTOR OPERATOR Page-63 i , ! REFERENCE: .; Instr'2ctor Guide O-RO-02-09-04, section V.4, ELO 8 i
[3.1/3.2]
! , 2170C0A211 ..(KA's) i ' ANSWER: 023 (1.00)
c.
h REFERENCE: 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) ? c.
REFERENCE:
ADS Reference Text, Figure 7 (no learning objectives listed in this document)
- a i
l [3.8/3.8]
+ -218000K501 ..(KA's) -! ANSWER: 025 (1.00)
c.
. d ., -,, -- c.
6.
% ,.,..-i- ---
, I . SENIOR REACTOR OPERATOR Page 64 -. REFERENCE: High Pressure Coolant Injection System (HPCI) Reference Text, page HPCI 55.
[4.3/4.4] l l 206000A101 ..(KA's) l ! .i ANSWER: 026 (1.00) d.
REFERENCE: PCIS Reference Text, page PCIS-15 section 3 [3.4/3.5] , 223002K108 ..(KA's) . ANSWER: 027 (1.00) ! b.
' REFERENCE: j Procedure 2.2.20, page 6, section 4.2 i
[3.1/3.1] 209001A208 ..(KA's) . ANSWER: 028 (1.00)
'! C.
I ! ! ! ! -- - - -.,,,, -. - ,. - - - -. - -
. . _. _.. m . . _ _ _ _ _ _ ... _ _ _. - . _ _ _ _ _ _, - , .
. SENIOR REACTOR OPERATOR Page 65 t ! g
- REFERENCE:
CS Reference Text, page CS-10, section 7 , [3.3/3.6] , 209001A205 ..(KA's) , ' ANSWER: 029 (1.00) c.
REFERENCE:
RPS Reference Text, MSIV Closure Scram Coincidence Logic Diagram ! ! [3.6/3.7] i 212000K114 ..(KA's) , ANSWER: 030 (1.00) ' .i a.
REFERENCE: RPS Reference Text, page RPS-16, section 13 [4.1/4.1] 212000A417 ..(KA's) ANSWER: 031 (1.00) a.
! -, ,, - - - -, . ..- ... .. - . . _.. _.. -, - - ... _ -- .
i - 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) ' . ANSWER: 032 (1.00) b.
REFERENCE: SGTS Reference Text, page SGTS-13, section D.1 [3.2/3.3] , 261000A301 .. ( l' A 's) . . ANSWER: 033 (1.00) , a.
. REFERENCE: Procedure 2.2.8, page 23, section 7.7.1 j j [3.6/3.6] 264000K602 ..(KA's) -j ANSWER: 034 (1.00) C.
i l l j __ _,. -.. - ,, _ _ _.. _. _ _., _ _, _,, . ,, . _. _. -- . .. .. .
_- _ -. .. - - -.. - - - .... ._
.. SENIOR REACTOR OPERATOR .Page 67
REFERENCE: RHR Reference Text, page RHR-28 [3.7/3.5] 226001G009 ..(KA's) i ANSWER: 035 (1.00)
b REFERENCE: A k Nuclear Boiler Instrumentation Reference Text, page NBI-4 [3.6/3.8) 216000K507 ..(KA's)
ANSWER: 036 (1.00) ! a.
, REFERENCE: SRM Reference Text, page SRM-11, section 4-Interlocks and Trips [3.7/3.7] 215004K401 ..(KA's) ! . . ANSWER: 037 (1.00)- a.
'l 'l _ . . - . -
. SENIOR REACTOR OPERATOR Page 68 i REFERENCE: Procedure 2.4.37, page 5, section 5.0 [4.2/4.3] 241000K302 ..(KA's) ANSWER: 038 (1.00) , d.
> REFEREMCE: Procedure 2.2.87, page 18, precaution 5.
[3.2/3.3] , , 201001G010 ..(KA's) ,. ANSWER: 039 (1.00) a.
I REFERENCE: ! RMC/RPIS Reference Text, page RMC-15, section D.3.c [3.6/3.6] . 201002K403 ..(KA's) ANSWER: 040 (1.00) C.
] ) , . .
,. . .... - - . .- -. .- . 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) ? C.
, REFERENCE: s Procedure 2.2.83, page 10, section 5.1[5] [3.2/3.2) 204000G010 ..(KA's) !
ANSWER: 042 (1.00) , b.
, REFERENCE: Procedure 2.2.19, Attachment 7, sheet 1, section 2.0[1] [3.2/3.3) ! 205000G010 ..(KA's) - i ~i ANSWER: 043 (1.00) ' i C.
l
> -i , !
._,,. -... .. ,-. _.
. . . -. .. .
.-- _ . - _. .. _. . -. _. _. _ . . - _
- 4'
< SENIOR REACTOR OPERATOR Page170 t.
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) , a.
REFERENCE: IRM Reference Text, page IRM-12, section 4
[4.0/4.0] , 215003K402 ..(KA's) , l ANSWER: 045 (1.00) b.
, REFERENCE: j Technical Specifications Table 3.1.1 J IRM Reference Text, page IRM-12
[3.4/4.1] 215003G011 ..(KA's) ANSWER: 046 (1.00)
$ggD C,. CGS
a.
i i .:,. ~. .- _ .i
- ! SENIOR REACTOR OPERATOR Page 71 k ! REFERENCE: Procedure 2.2.19, page 25, section 7.1.2 [3.4/3.5] 219000G010 ..(KA's) ANSWER: 047 (1.00) c.
REFERENCE: PF't Reference Text, page PRM-6-5/89 and Figure 1 [3.6/3.8] 272000K109 ..(KA's) ANSWER: 048 (1.00) a.
REFERENCE: Procedure 2.2.96, page 11, section 5.2.1[1] [3.2/3.3] 259001G010 ..(KA's) ANSWER: 049 (1.00) d.
_
. . . .. -.... _ _ = . - . ,
I _ SENIOR REACTOR OPERATOR Page 72
- \\.
'
- REFERENCE:
PRM Reference Text, page PRM-22-5/89, section 4 (3.1/3.5] 271000A112 ..(KA's)
ANSWER: 050 (1.00) c.
REFERENCE: SCW Reference Text, page SCW-12, section E.3 (3.6/3.6] 245000A301 ..(KA's) $ ANSWER: 051 (1.00) d.
REFERENCE: CRDM Reference Text, page CRDM-10 [3.6/3.7] 201003K404 ..(KA's) ANSWER: 052 (1.00) c.
l l
4
. , - . . - -.. - . . ,
__ _ _ - _ _. - . - - _ . . _ - SENIOR REACTOR' OPERATOR Page 73 . 4.
REFERENCE: . TIP Reference Text, page TIP-22-5/89 [3.4/3.5] 215001K401 ..(KA's) ' i ANSWER: 053 (1.00) ! b.
REFERENCE: PRM Reference Text, page PRM-12-5/89
[3.4/3.4]
288000K102 ..(KA's)
. ANSWER: 054 (1.00) c.
REFERENCE: i $ Condensate and Feedwater System Reference Text, Page FCS-16
[3.3/3.3] .
256000K102 ..(KA's)
i ' ANSWER: 055 (1.00) l d.
- - , - . . . , . --
_. _ _ - _ _-_ _ _ _ _ _ _ _ _ _ ___ _ __ ______. -. _._ ___.._._.._-,
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) a.
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) , C.
REFERENCE: l Procedure 2.1.6, Revision 34, sections 4.0 and 6.2[2] s- [4.5/4.6] 295006A201 (KA's) .. , L , ANSWER: 058 (1.00) ! , a.
\\ i ,
.--.-...-..,...,+n .,i----. _,.. - -~-.r., ..-----.--.--*-----.----..------------.------.,w-__ E -
-.. _ -. - -. - .. - -. -- i
. -SENIOR' REACTOR OPERATOR.
Page 75 .i
REFERENCE * ~ l Training Module 0-RO-03-04-05, ELO 7 and page IG-6, section 5.b , [3.6/3.9] 295024G007 ..(KA's) ANSWER: 059 (1.00) b.
REFERENCE: Training Module O-RO-03-04-04, ELO 20, page IG-53, section 1.a (3.4/3.8] \\ 295026G007 ..(KA's) ANSWER: 060 (1.00) a.
. . ' REFERENCE: .i EOP-03 Entry Conditions i [4.4/4.6] c 295026G011 ..(KA's) i t ANSWER: 061 (1.00) { . b.
i , 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) b.
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) a.
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) c.
REFERENCE:
Procedure 2.4.13, Revision 9, page 3 of 5, Immediate Operator.
Actions [4.0/3.9] i 295014G010 ..(KA's) . ANSWER: 066 (1.00) ! c.
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) a.
! , ..-r - .. . _... - _ _, _ _....-,. _ _. -,. . ._ _ _ -..
-.. _.
. -. -. .. =. ... _ . ~.. . . - _.. -... _ - . .. ' SENIOR REACTOR OPERATOR Page 78 O .' REFERENCE: 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) C.
REFERENCE: j EOP-02, Q-11 i [4.3/4.4] 295037A203 ..(KA's)
! ] ANSWER: 069 (1.00) d.
REFERENCE: Instructor Guide O-RO-03-04-05, ELO-6, page IG-26 [3.5/3.8] 295024K302 ..(KA's) ' ANSWER:- 070 (1.00) '] . b.
1 > i 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) a.
REFERENCE: EOP-02, step P-4 [3.7/4.4] - 295015G012 ..(KA's) l t i ANSWER: 072 (1.00) d.
.! 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 d.
, ! .!
,
, , ... , -., _. - -.. - ,, .. - - -. - .- -.
_. --. _ . _ . _. _ _._ _ _ _.. -. . _.... _ _ _ ... _ _ _ SENIOR REACTOR OPERATOR Page 80 .- REFERENCE: l Technical Specifications, page 203, section 3.10.B i [2.9/3.6] l , .295023G007 ..(KA's) i ANSWER: 074 (1.00) c.
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 c.
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) b.
' -i
,
.,. - ~. ... _. _ - _ .. _ _
- - - . - . . - . - I I .. ENIOR REACTOR OPERATOR Page 81 S e REFERENCE: i Instructor Guide O-RO-03-04-02, Page IG-71, section P.1 ' [3.6/3.8] (KA's) 295010G007 ..
-ANSWER: 077 (1.00) c.
REFERENCE: Procedure 5.3.23, Revision 10, pages 4 and 6 of 16, section 2.1[1]. and 2.2 , [4.0/4.1] 295015K204 (KA's) .. ANSWER: 078 (. 00) b.
REFERENCE: Procedure 5.4.3, Revision 10, page 2 of 4, section 3.0 .[3.8/3.9] 295023G010 (KA's) .. , ANSWER: 079 (1.00) , d.
, j i i
, -,, -. _.
_ __.
,, _, , .,,,.
... . .._..-. -... - -.. _. .. .. ~ -. . - -. .. 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) i ANSWER: 080 (1.00) C.
, i REFERENCE: Technical Specifications, page 193f, section 3/4.8.G Bases , [2.8/3.8] , 295038G004 ..(KA's) , , ' ANSWER: 081 (1.00) b.
! 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) C.
, i
I e . , b I .,..,,..... _, -,, _ -. . . .,... . _ . . -. , _
. _ _... .. ... . . _. _. _ -- -. . _. - _ _ _ _ _ . _ _ _ .! -.
SENIOR REACTOR OPERATOR Page 83 i i a
, . REFERENCE: ' Technical Specifications, page 155, section 3.7.A 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 4.0 , i ! [3.7/3.5]
i ! 295002G010 ..(KA's)
I ! ANSWER: 084 (1.00) , , b.
., ! ! ' REFERENCE:
Procedure 2.4.36, Revision 13, page 5 of 6, section 5.0[6]
i [3.2/3.2] '
295002G007 ..(KA's) ' . I ANSWER: 085 (1.00) ,
d.
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 3.0 [3.4/3.3] 295018G010 ..(KA's) .; ANSWER: 086 (1.00) c.
' REFERENCE: Procedure 2.4.42, Revision 14, page 6 of 7, section 5.0[4] [3.1/3.3] 295018K303 ..(KA's) ! < ANSWER: 087 (1.00) c.
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) , 8.
, f
, i ' f n , f [ -- - -,, -, . ,.
,..... .. l . SENIOR REACTOR OPERATOR Page 85
REFERENCE: Instructor Guide 0-RO-03-04-05, ELO 11, page IG-7, section 7.a [3.4/3.8] 295028G007 ..(KA's) l ANSWER: 089 (1.00) b.
l l REFERENCE: l Instructor Guide O-RO-03-04-05, ELO 15, page IG-7, section 8.a [3.6/3.9] 295028K301 ..(KA's) ANSWER: 090 (1.00) b.
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) d.
_ - _ _ _ - _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
- .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) b.
REFERENCE: Instructor Guide O-RO-03-04-06, ELO 5, page IG-9, section L.2 , [3.3/3.5) 295033K301 ..(KA's) ANSWER: 093 (1.00) b.
l , REFERENCE: Instructor Guide O-RO-03-04-05, ELO 12, page IG-15, section B.1.a ) i [3.1/3.2] 295029K207 ..(KA's) ANSWER: 094 (1.00) ~b.
i
i . _
. _ _ __ _ _ T* . SENIOR REACTOR OPERATOR Page 87 g_ REFERENCE: , Procedure 5.3.13, Revision 12, page 2 of 7, section 1.2 [3.3/3.3] 295004K203 ..(KA's) i ANSWER: 095 (1.00) b.
, . . REFERENCE: Procedure 5.3.8, Revision 16, page 2 of 9, section 2.0 [3.2/3.2] , 295019K214 ..(KA's) ' .: ! ANSWER: 096 (1.00) i d.
REFERENCE: Technical Specification Definitions, page 4 [2.8/3.9] ! ! 295035G003 ..(KA's) J ANSWER: 097 (1.00) b.
, , t b t i
- ! 1. , < , -.. , _ _ _ _ _ - --
_ _. _ .. . . _ _ _. _. _. _ - $ . SENIOR REACTOR OPERATOR.
Page 88
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- sf
- , 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 ] a.
REFERENCE: . i
PNPS Technical Specifications, Table 6.2-1.. IG: 0-RO-06-01-04, Technical Specification Design Factors and
" ' Admin Controls," ELO 5 i [3.5/3.8]
. l
i J 294001K116 ..(KA's) ANSWER * 099 (1.00) C.
, ' 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) c.
.i
l
vm .
- r 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'.4.
i OJT Guide O-RO-04-04, " Emergency Tasks," Task 62.
' ) '[4.1/4.1] , . 295016G006 ..(KA's) . . ' .
! , , -. ~ ! 'I
i i i . , i , h a . I h
i ! , . . (********** END OF EXAMINATION **********) ,
.-. s _,. . ., ~, ,- . - - ,.4 ,
l i . I TEST CROSS REFERENCE Page
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
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 C.8 1.00 9000038 l 039 1.00 9000039 ' 040 1.00 9000040 041 1.00 9000041
042 1.00 9000042 i 043 1.00 9000043 044 1.00 9000044 045 1.00 9000045 046 1.00 9000046 047 1.00 9000047 i 048 1.00 9000048 049 1.00 9000049 i . . -..- ... .. - .,.. . . -... -,.. -. ..
.. 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 052 1.00 9000052 053 1.00 .9000053 054 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
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 096 1.00 9000096 j , 097 1.00 9000097 098 1.00 22581 ! P 's.
-,. - . - ,, . . . .. ~ - - _ . . -i--- -
.. - . _. .. . . .,
. -, t TEST CROSS REFERENCE Page
,
SRO.
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 . ' a - , .. -. . .... _
_ _ __ _ _ _ - .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
003.
1.00 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 f i 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 1.00.
223002K108 034 1.00 226001G009 i , , __ - . -_ _ . . -. - - .. j
,,.. . .. - . - - . .- -. ~. -.- . TEST CROSS REFERENCE .Page-5 ,O twd S R'O Exam BWR Reactor-q.
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
- & ' 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 - 058 1.00 295024G007
2 ' 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
, 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 i
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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 answer.
NRC RESOLUTION OF FACILITY COMMENT Question 046: Accepted answers a and c as correct answers.
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A1TACIIMENT 3 SIMULATION FACILITY REPORT Facility Licensee: Pilgrint Facility Docket No. 5Da91
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 observations.
While conducting the simulator portion of the operating tests, the following items were observed.
[Qhi DESCRilrr1ON-NONE- , 5 }}