IR 05000293/1992022
| ML20127E898 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/06/1993 |
| From: | Conte R, Walker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20127E754 | List: |
| References | |
| 50-293-92-22OL, NUDOCS 9301200083 | |
| Download: ML20127E898 (258) | |
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lixamination Iteport No.: 92-22 (01.)
Facility Docket No.: 50-293 1:acility License No.: DPit 35
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Licensee: lioston tidison Company * itFD #1 Itocky 11i11 lload Plymouth, hiassachusetts 02360
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Facility: Pilgrim Nucicar Power Station 11xamination Dates: November 16 - 20, 1992 IIxaminers: T. Walker, Senior Operations lingineer , S Ilansell, Operations lingineer . C. Tyner, lixaminer (!!G&G) J. llanck, lixaminer (EG&G) Chief lixaminer: e W //G/ T. Wder, Senior Operations !!ngineer Date
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EXECUTIVE SUMMARY Initial examinations were administerej to three Senior Reactor Operator (SitO) instant and seven Reactor Operator (110) applicants. Six of seven RO and all of the SitO applicants passed bothl ortions of the examinations. One RO did not pass the written examinatio Requalification retake examinations were administered to one licensed SRO (walk through examination only) and one licensed RO (written examination only). Ikith licensed operators passed the requalification retake examinations. The examinees were well prepared for the ' examinations. The applicants' communications during the simulator portion of the initial examinations were a noted strength. The facility staff was very cooperative during the examination preparation and administratio During the examination process, discrepancies were identified between the limergency Plan training materials and procedures and management expectations which indicate a weakness in , the interface between Operations and limergency Planning departments in limergency Plan training for licensed operators.
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DETAllii n
- INTRODUCTION The NRC administered initial examinations to three Senior Reactor Operator (SRO) instant and seven Reactor Operator (RO) applicants. The examinations were administered in accordance with NURiiG 1021, Ilxaminer Standards, proposed Revision 7.
The NRC administered requalification retake examinations to one licensed SRO (walk-through portion only) and to one licensed RO (written portion only) who did not pass the NRC administered requalincation examinations in hiay 1992. The examinations were administered in accordance with NURl!G 1021, Examiner Standards, Revision ' PREEXAAllNATION ACTIVITIES Several problems related to learning objectives were identified during preparation of the , initial examinations. The reference materials that were initially submitted for examination preparation did not contain any learning objectives related to the use of procedures or administrative topics. At the NRC's request, the facility provided the on-shift training materials, which included student handouts and task lists. The student handouts contained objectives related to procedurest however, these objectives were of a generic nature and did * not contain specific conditions and standards of performance. An administrative task list was provided, but no learning objectives related to administrative toples were available. The ' materials that were initially submitted also did not contain learning objectives related to Technical Specifications for ROs Technical Specification training materials, including learning objectives, for ROs were also provided at the NRC's request. Problems were also identified with the timergency Plan training materials and learning objectives. These problems are discussed in Section 3.4 of this repor The facility reviewed the written examinations in the facility training center during the week of November 2,1992. A large number of changes had to be made to the initial examinations as a result of this review, including 15 questions (out of 130) that had to be replaced or revised significantly. Problems with learning objectives or lack of specine learning objectives were the cause of the majority of the significant revisions and replacements. The licensee agreed to review and revise the faulty learning objective The simulator scenarios and Job Performance hicasures (JPhis) were validated during the week of November 2,1992, on the facility's simulator and in the plant. Several changes had to be made due to the limitations of the simulator as noted in Attachment 5. One JPhi had to ; be replaced because of inconsistencies between the Emergency Plan training materials and ' management expectations as discussed in Section The facility staff who were involved with these reviews signed security agreements to ensure that the initial and requalifica'. ion examinations were not compromised. The personnel involved in the preexamination reviews are identified in Attachment _ _ _ . _ _- _ _ _ _ _ __ __ _ _ _ _ ._ _
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3,0 EXAMINATION RESULTS AND RELATED FINDINGS, OllSERVATIONS AND CONCl,USIONS Exainination Results The results of the examinations are summarized below: Initial Examinations SRO RO ; Pass / Fail Pass / Fall l Written 3/0 6/1 Operating 3/0 7/0 Overall 3/0 6/1 Requalification Retake SRO RO Examinations Pass / Fail Pass / Fail Written N/A 1/0 Operating 1/0 N/A Overall 1/0 1/0 Generic Strengths and Weaknesses The following is a summary of the strengths and weaknesses noted during initial examination administration. This information is being provided to aid the licensee in upgrading their training progra Written IIxainination strengths:
- Knowledges and abilities associated with emergency and abnormal events (RO only)
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- Ability to interpret Technical Specifications (SRO only)
Weaknesses: The following subjects were missed by at least forty percent of the applicants that were evaluated on the subject indicating a generic weakness in the subject: _ - . _ . _ _ , _ . . _:. . _ . _ . . . _ _ . _ _ . . - - - - . _ - - . - _ . _ _ , _ . . - - ---
i - Knowledge of minimum shift crew composition in accordance with Technical i Specifications (SRO only) - Ability to determine composition of a radioactive release (SRO only) l - Knowledge of restrictions on HPCI operation during implementation of the EOPs ! - Ability to identify and predict the plant response for an APRh! flow reference mismatch condition . - Ability to predict the system response to a failed control rod reed switch - Ability to determine the cause and predict the plant response to a reactor feed pump ! low net positive suction head condition
- Ability to predict the system response to a Standby Gas Treatment train heater trip -
Ability to predict the HPCI system response to an increase in torus water level (RO only)
. - Ability to predict the response of the htSIVs te a loss of instrument air and DC power -
Knowledge of the requirement for tripping the turbine on high vibration during a startup (RO only)
> - Ability to determine the correct method for reactor pressure control following a failure to scram and h1SIV closure (SRO only) - Ability to determine the appropriate actions for primary containment hydrogen control (SRO only)
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- Ability to determine the correct actions for venting primary containment with radioactive release rates above Technical Specification limits (SRO only)
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Ability to identify secondary containment conditions which would require Alternate RPV Depressurization (SRO only) Operating Tests (Simulator and Walk-through) Strengths:
- SRO briefings to the crew during abnormal and emergency events
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- Crew communications and team skills (also observed in the Control Room by actual l operating crews)
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- Response to JPMs which required use of an alternate procedure or procedure path (alternate path JPMs)
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- Applicant familiarity with the examination process Weaknesses:
- Ability to perform immediate actions for abnormal and emergency events without the use of the procedure; specifically, the immediate actions for a Control Room-cvacuation
- Ability to assume the responsibilities of a Control Room communicator during an {
event which requires implementation of the Emergency Plan; specifically, ability to (
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use the Plant Data Phone (RO only)
- Lack of familiarity with the procedure for a loss of Shutdown Cooling (SRO only) - Knowledge of the interlock that causes a Recirculation Pump trip if t'ie pump discharge valve is taken to the close position (RO only) Procedures During the preparation of the examinations, the NRC identified several problems with facility procedures. The following are examples of procedure discrepancies that were identified during examination preparatio PNPS Proc. 5.3.26, "RPV injection During Emergencies," provides direction for cross-ticing the Firewater system to the Feedwater system for use when implementing the EOPs. PNPS 5.3.26 states that the methods for RPV injection may only be used when specifically directed by the EOPs None of the EOP Dowcharts provide-direction to cross tie Firewater to Feedwater for RPV injection. It is not clear whether Firewater cross-tied to Feedwater can be used when implementing the EOP Sectiori 4.3 of PNPS Proc. 2.4.54, " Loss of All Fire Suppression Pumps or Loss of Redundancy in the Fire Water Supply System," which provides direction for isolating a leak on the Firewater ring header contains several technical errors, Portions of the procedure do not accomplish the intended function and/or fail to restore the Firewater system to an operable status following isolatio The licensee was very responsive when the problems were identined and agreed to review and correct the procedures accordingl l
3,4 Emergency Plan During the preparation of the written examinations, the NRC identified that the instructions for desigr.ation of an Assembly Area in the procedures responding to a Site Area Emergency and a General Emergency were not consistent with Emergency Planning department expectations. EP-IP-130, " Site Area Emergency," and EP-IP-140, " General Emergency," provide guidance on designating Assembly Areas for site evacuation. If a release is in progress, the Chiltonville Training Center or the Kingston Warehouse should be designated based on wind direction, in accordance with EP-IP-130 and EP-IP-140, if the event is being upgraded from an Alert with no rele_se in progress, the I&S Building should be the designated Assemb;y Area. According to a PNPS emergency planning manager, I&S Building is the primary Assembly Aiea and should be designated regardless of the timing of the event classification. Chiltonville or Kingston would only be used as the Assembly Area in the case of a ground level release or an elevated release with a release path that would preclude evacuation to the I&S Building. Procedures that are inconsistent with management expectations and training could cause confusian during an emergency that requires evacuation of onsite personnel. The emergency planni..g manager indicated that the procedures would be revised as necessary to coincide with management expectations,
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The training materials, specifically, the learning objectives related to Emergency Plan training for SRO licensing provided for examination preparation, were not consistent with Emergency Planning and Operations management expectations. The PNPS expectations for SRO knowledge and abilities were not clearly defined as indicated by the following discrepancies in the Emergency Plan training material for SRO '
- Unit Guide O-RO-06-03, " Emergency Plan Training for Senior Reactor Operators,"
references an abbreviated version of the Emergency Public Information Organization and Position Training module (T-ER-01-01-19), but the detailed module (T-ER-01-01-13) was provided with the examination materials. It was not clear which module is applicable to SRO license trainin Unit Guide 0-RO-06-03 references module T-ER-01-01-80 for dose assessment and Protective Action Recommendation (PAR) training. This module was provided with the examination materials; however, it is not the module that is used for SRO training. According to an Emergency Planning manager, module T-ER-01-0181, the overview version of dose assessment and PAR training, is used for SRO trainin This discrepancy was identified during the on-site JPM validation and resulted in replacement of a JPM Based on the learning objectives in the training' materials that , were provided, the NRC preparid a JPM to perform an off-site dose calculation. The Operations and Training department personnel participating in the examination . preparation indicated that this was not an appropriate task for an SRO. Task 91 of-Unit Guide 0-RO-04-10, "SRO On-Shift Tasks," specifies that an SRO must estimate offsite release rates. The NRC agreed to replace the JPM for this examination; however, the NRC expects an SRO to understand the dose assessment process _
-. - . - . . thoroughly in order to perform the responsibilities of the Emergency Director in approving offsite notifications and determining PAR The discrepancies in the Emergency Plan training materials ano o 'har management expectations indicate a weakness in Operations involvement in the Emergency Plan portion of the licensed operator training program. The identified weakness in the RO applicants' ability to assume the responsibilities of a Control Room communicator during an emergency event also indicated a weakness in Emergency Plan trainin .5 In-Plant Observations The examiners noted that the material condition of the plant was good. Access into the plant and through the radiological contv1 areas was smooth. The control room atmosphere was -
conducive to the conduct of the exan.inations. The examiners also observed excellent ' communication by the control room .rew .0 EXIT MEETING An exit meeting was conducted November 20,1992, following the administration of the examinations. Attendees are listed in Attachment 1. The facility presented their comments on the written examinations (Attachment 4). The NRC discussed generic findings regarding the applicants performance on the operating tests, simulator fidelity problems (Attachment 6), and the problems encountered during preparation for the examinations (Sections 2.0,3.3,'and - 3.4).
ATTACllMENT 1 PERSONS CONTACTED 1.icensee Personnel I E. T. Boulette Vice President Nuclear Operations / Station Director (3) E. S. Kraft, J Plant Manager (2), (3) W. C. Rothert Director Nuclear Engineering (3) J. F. Alexander Nuclear Training Manager (2), (3) ' H. R. Balfour Operations Training Section Manager (2), (3)
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T. .A. Sullivan Operations Section Manager (1), (2), (3) A. R. Shiever Operator Training Division Manager - Initial (1), (2), (3) T. S. Swan Operator Training Division Manager - Requalification(l) N. L. Desmond Compliance Division Manager (2) R. L. Cannon Senior Compliance Engineer (2), (3) W. J. Green Senior Operator Training Specialist (3) M. Santiago Senior Operator Training Specialist (1) E. Olsen Senior Operator Training Specialist (1) P. V. Gallant Operator Training Specialist (1) R. F. Balduc Operator Training Instructor (Protech) (1) D. A. Whiting Operator Training Instructor (Protech) (1) D. Landahl Emergency Planning Division Manager NRC Personnel R. Conte BWR Section Chief , T. Walker Senior Operations Engineer (1), (2), (3) S. Hansell Operations Engineer (1), (2) A. C. Cerne Resident Inspector (2) NOTES:
(1) Participated in Examination Preparation (2) Attended Entrance Meeting on November 5,1992 (3) Attended Exit Meeting on November 20,1992
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__ Ab W b U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE REGION 1 CANDIDATE'S NAME: FACILITY: Pilgrim 1 REACTOR TYPE: BWR-GE3 DATE ADMINISTERED: 92/11/16 INSTRUCTIONS TO CANDIDATE: Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheet Points for each question are indicated in parentheses after the questio The passing grade requires a final grade of at Joast 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE %
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9 .00 - % TOTALS FINAL GRADE All work done on this examination is my ow I have neither given nor received ai Candidate's Signature
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REACTOR OPERATOR Page 7 QUESTION: 001 (1.00) WHICH ONE of the following O Demand (OD) Programs would be used to obtain the effective readi. s from an LPRM string used in the most recent P1 power distribu on calculation? OD-6, Therma Data in a Specified Bundle b. OD-8, Pree nt LPRM Readings OD-9, ial Interpolation in a Specified LPRM String OD- , Periodic Core Performance Logs _ QUESTION: 002 (1.00) Conditions in a recently surveyed area are: 25 mR/hr general area radiation 100 dpm/100 cm2 alpha loose surface 500 dpm/100 cm2 beta-gamma loose surface 0.20 MPC airborne beta-gamma radioactivity WHICH ONE of the following describes the complete posting requirements for the area? " CAUTION RADIATION AREA"
' " CAUTION RADIATION AREA" and " CAUTION CONTAMINATED AREA" " CAUTION RADIATION AREA" and " CAUTION AIRBORNE RADIOACTIVITY AREA" " CAUTION RADIATION AREA," " CAUTION CONTAMINATED AREA," and
" CAUTION AIRBORNE RADIOACTIVITY AREA"
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i REACTOR' OPERATOR Page z 8
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QUESTION: 003 (1.00) A 23 year old-radiation worker with a current Form-NRC-4 needs to perform work in an area with general radiation levels of.75 mR/h The worker's exposure history is: Lifetime: 24.5 Rem Current year: 1400 mR Current quarter: 225 mR , WHICH ONE of the following is the maximum time that the worker can stay-in the area without exceeding any PNPS Exposure Control Levels? (Assure no special authorization.) minutes I hour and 20 minutes hours and 40 minutes hours QUESTION: 004 (1.00) WHICH ONE of the following describes proper procedures for handling sodium pentaborate?
a. Respiratory protection is required to prevent inhaling dust containing boron, b. Protective clothing is not required'because sodium pentaborate cannot be absorbed through the ski c. Report to the Chemistry Department immediately after leaving-the worksite to dispose of-protective clothing and' foot covering d. Direct any-liquid spillage to a floor drain, wipe and dry mop . the flocr,-and' discard mops and wipes _after use.
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REACTOR OPERATOR Page 9 QUESTION: 005 (1.00) WHICH ONE of the following situations is acceptable in accordance with PNPS Proc. No. 1.4.36, "High Pressure / Compressed Gas Cylinder Control?" Argon cylinders on a cart are staged for a non-active Maintenance Request that has been rescheduled for later in the week. The cart is secured to a fixed suppor Reserve nitrogen cylinders are secured at approximately 3/4 height to a permanently installed cylinder holding station, Acetylene cylinders on a cart are staged in the Cable Spreading _ Room for an active Maintenance Request. Work is expected to start next shift. The cart is secured to a fixed suppor Empty oxygen cylinders are secured at approximately 3/4 height to a fixed support during shift turnover. The cylinders wil] be removed next shif QUESTION: 006 (1.00) WHICH ONE of the following situations would require independent verification in accordance with PNPS Proc. No. 1.3.34, " Conduct of Operations?" (All components are safety related.) Fuses are pulled for maintenance in a cabinet in the Control - Roo b. An existing tagout is used to verify the position of an ' inaccessible componen Verification of a component's position is expected to take 15 minute Radiation levels in the area are 120 mr/h A large manually operated valve that requires two people to operate must be shut for maintenanc .y - . _ _ , .__.,_ _ .___ __ _ __ __ , REACTOR OPERATOR Page 10 QUESTION: 007 (1.00) Breaker 103 (Bus Al feeder from Unit Auxiliary Transformer) is open and , has a white tag with a green border hanging on the control switc , WHICH ONE of the following describes the meaning of the tag? a. A grounding device is installed on the breaker. The breaker may only be operated under the authority of the person for whom the tag was place b. The breaker is not in its normal operable status. The breaker may be operated by qualified operators when the precautions listed on the tag are followed, c. The breaker requires testing prior to restoring it to normal servic The breaker may be operated only under the authority of the person for whom the tag was placed, The breaker is open to protect personnel from injury or equipment from damage. The breaker may not be operated until > the tag is cleare QUESTION: 008 (1.00) WHICH ONE of the following accurately describes the NRC Overtime Guidelines? a. Operators may not work more than: 12 hours in 24 hours; 24 hours in 48 hours; or 84 hours in 7 days, b. Operators may not work more than: 16 hours-in 24 hours; 28 hours in 48 hours; or 84 hours in 7 days, c. Operators may not work more than: -16 hours in 24 hours; 24 hours in 48 hours; or 72 hours in 7 days, d. Operators may_not work more than: 12 hours in 24 hours; 28 hours in 48 hours; or 72 hours in 7 day ,-. _ -. - - -- .-. . . - .
? REACTOR OPERATOR Page-11 QUESTION: 009 (1.00)
kN p d4ce g WHICH ONE of the following j describes th electrical equipment controlled by REMVEC? a. All equipment 2 4.16 k All equipment 2 1 r kV c. All equipment _ 230 kV d. All equip nt 1 345 kV QUESTION: 010 (1.00) The paper in a Control Room chart recorder has jammed. WHICH ONE of the following describes the appropriate actions for replacing the recorder chart? Initial and date the recorder chart that was removed and the new recorder chart to be installe The removed recorder chart'will be stored in the Control Room for at least a wee Initial and date the recorder chart that was removed and the new recorder chart to be installed. The removed recorder chart will be sent to Records Management for storage at the earliest convenience, Initial and date the new recorder chart to be installed and log the recorder chart replacement in the. Operating Log. The removed recorder chart-will be stored in the control Room for at least a week, Initial and date the new recorder' chart to be installed and. log the recorder chart replacement in the. Operating Log.. The removed recorder chart will be-sent to Records Management-for storage at the earliest convenience.
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_ . REACTOR OPERATOR Page 12 QUESTION: 011 (1.00) WilICH ONE of the following describes the meaning of this icon: d kY when implementing the EOPs? a. The CS/RHR pumps may not be operated below the_ vortex limit, b. The CS/RilR pumps may be operated irregardless of the vortex . limi c. The CS/RHR pumps may not be operated below the NPSli limi d. The CS/RIIR pumps may be operated irregardless of the-NPSH limit , QUESTION: 012 (1.00) < WilICH ONE of the following Emergency Action Levels (EALs) is the minimum action level at which the emergency facilities (TSC, OSC, and EOF) must be activated? a. General Emergency b. Site Area Emergency
c. Alert d. Unusual Event
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REACTOR OPERATOR 1 Page 1 ; . QUESTION: 013 (1.00) WHICH ONE of the following communication systems can be used to transmit information from the Control Room directly to the EOF during an emergency? a. Digital Voice-Network (DVN) b. Emergency Notification System (ENS) Boston Edison Community Off-Site Notification System (BECONS)
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d. Plant Data Phone (PDP) QUESTION: 014 (1.00) WHICH ONE of the following primary containment parameters does NOT have any associated Technical Specification requirements? Drywell temperature b Torus water level Primary containment oxygen concentration d. Primary containment hydrogen concentration
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REACTOR OPERATOR Page 14 QUESTION: 015 (1.00) The plant was operating at 100% power when the main turbine first stage pressure transmitter (PT-652) failed downscale. WHICH ONE of the following describes the expected plant response? The Rod Worth Minimizer will enforce adherence to the rod sequence due to pressure below the Low Power Setpoin The Main Condenser Low Vacuum Scram will be bypacsed due to pressure below 600 psi The FW Control steam flow detectors will indicate lower than actual due to loss of density compensation, The " MAIN STEAM LINE LEAKAGE" alarm will annunciate due to the main steam flow / turbine steam flow mismatc QUESTION: 016 (1.00) During normal power operations, the operator at the controls accidently takes the control switch for the RWCU return isolation valve, MO-80, to the full closed position. WHICH ONE of the following describes the expected RWCU System response? a. As soon as MO-80 leaves the full open position, MO-2 and MO-5 will close causing the RWCU pumps to tri _ Once MO-80 reaches the full closed position, the RWCU pumps will tri As soon as MO-80 leaves the full open position, the RWCU pumps will tri Once MO-80 reaches the full closed position, no flow will exist in the system causing the holding pump to start.
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: REACTOR OPERATOR Page<15 QUESTION: 017 (1.00)
The reference legs to all of the FWLC level instruments have broke WHICH ONE of the following describes the signals that will be received-due to the failure of the level. instruments? a. Reactor Feed Pump tri RPV High Level alar b. Main Turbine, HPCI, and RCIC tri c. Reactor'Recirc Pump runbac RPV Low Level alar d. HPCI and RCIC initiatio EDG star PCIS isolation Reactor Recirc Pump tri Reactor scra QUESTION: 018 (1.00) WHICH ONE of the following conditions could cause indicated level on the Fuel Zone instruments to be higher than actual reactor vessel water level? a. Drywell temperature at 180* b. A break in the variable le c. No forced recirculation flow through the jet pump d. A rapid reactor depressurizatio v -- 9. , . - . . .. , - - - . -. . - . .. _ REACTOR OPERATOR Page 16-QUESTION: 019 (1.00) All of the Scram Discharge Volume (SDV) vent and drain valves were manually operated during maintenance. The valves were returned to their ' normal position, but the manual operators for the SDV valves were not returned to the NEUTRAL position. WHICH ONE of the following describes the concern related to this operation? a. The valves could fail to automatically reposition on a reactor scram, preventing drain down of the SD b. The valves could fail to automatically reposition on a reactor ceram, causing a direct discharge path from the RPV to the Reactor Building sum c. The valves could fail open after repositioning on a reactor scram, causing a breach of primary containment, d. The valves could fall closed after repositioning on a reactor-scram, preventing reset-of the scram due to high SDV leve QUESTION: 020 (1.00) A scram has occurred and the mode switch has been placed in SHUTDOW Scram Discharge Volume (SDV) level is high, but the scram has been reset by use of the SDV High Level Scram Bypass. RPS bus A is then transferred to its alternate power supply. WHICl! ONE of the following describes the expected response?' It is a dead bus transfer and a half scram will occur, It is a dead bus transfer and a full scram will occu c._It is a dead bus transfer, but is rapid enough so'that no_ scram will occur.
, It is a live bus transfer, so no scram will occu . _ . _ _ . _ . . . . _ . . _ . . _ . _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ . _ _ . _ _ . . _ _ . . _ _.
REACTOR OPERATO Page 17
QUESTION: 021 (1.00) A plant startup was underway when annunciator HMSIV NOT FULLY OPEN SCRAM AT RX PRESSURE >600 PSI" (905R, D4) illuminate A full scram occurre WHICH ONE of the following describes the expected status of the plant? a. One main steam line is isolated; reactor pressure is greater than 600 psi Two MSIVs have closed; reactor pressure is less than 600 psi c. Three main steam lines are isolated; reactor pressure is greater than 600 psig, Four MSIVs have closed; reactor pressure is less than 600 psig.
QUESTION: 022 (1.00) The plant was operating at 100% power when a loss of all feed caused reactor water level to decrease. The reactor scrammed and the mode switch was placed in SHUTDOWN. Reactor water level is -10 inches and reactor pressure is 900 psig. WHICH ONE of the'following sets of valves should have received isolation signals? Recirculation System process sample valves, Post Accident Sampling System isolation valves, and RHR reject to Radwaste valves RWCU isolation valves, RBCCW to Drywell Coolers isolation valves, and Primary Containment Atmosphere Control makeup purge-valves c. MSIVs, Drywell Equipment Drain ~ Sump isolation valves, and Radiation Leak Detection System isolation valves d. Main Steam Line drains, Drywell Floor Drain Sump isolation valves, and Hydrogen / Oxygen Analyzer System isolation valve _ _ ._ _ -. . _ ,
. . . REACTOR OPERATOR Page 18 QUESTION: 023 (1.00) The RCIC system is running following a valid initiation signal received 10 minutes ago. The following conditions develop: RCIC Steam Lino Flow: 230% RCIC Area Temperature: 205'F RCIC Pump Suction Pressure: 10" Hg vacuum RCIC Turbine Exhaust Preocure: 38 psig RPV Pressure: 150 psig RPV Water Level: +40 inches WHICH ONE of the following describes the expected response of the RCIC ~ system? Only the RCIC inboard and outboard isolation valves clos b. The RCIC inboard and outboard isolation valves close and RCIC trip throttle valve clone Only the RCIC trip throttle valvo close d. Only the RCIC steam to turbine supply valve close _ ___._______m . _ _ _ _ _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _
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k REACTOR OPERATOR Page-19-QUESTION: 024 (1.00) WHICH ONE of the following describes the restrictions on HPCI operation while implementing the EOPs? a. HPCI may NEVER be operated below 1000 RPM because this is the minimum speed required to maintain adequate cooling and lubricatio b. HPCI may NEVER be_ operated below 2000 RPM because_this is the minimum speed required to generate sufficient control oil pressure for control valve operatio c. HPCI may be operated below 1000 RPM only at NOS/NWE direction because low turbine exhaust pressure could create a cyclic steam hammer which could damage the exhaust check valv d. HPCI may be operated below 2000 RPM, but greater than 2000 RPM only as directed by_the EOPs because operation at these speeds could cause excessive turbine vibration or oscillating flow rate QUESTION: '025 (1.00) With the plant operating at 100% power, the Core Spray Loop A line break differential pressure indication on Rack 2207 reads +4.5 psi WHICH ONE of the following conditions is indicated? a. No core spray line break has occurre This indication is normal for full power operation b. A core spray line break has occurred inside the reactor vessel shroud, c. A core spray line break has occurred inside the reactor vessel, but outside the shrou . d. A core' spray line break has occurred inside the drywell or inside the reactor vessel, but outside the shroud.
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-The reactor is shutdown _and RHR loop B is in the shutdown cooling (SDC)
mode, A reactor coolant system leak causes vessel level to decrease to 0 inches and drywell pressure to increase to 2.8 psig. Loop A jet pump riser pressure is less than loop B jet pump riser pressure. . No operator action has been taken. WHICH ONE of the following describes the status of the Low Pressure Coolant Injection (LPCI) valves? a. Outboard injection valve 28A is open Inboard injection valve 29A is open Outboard injection valve 28B is closed-b. Outboard injection valve 28A is closed Inboard injection valve 29A is open Outboard injection valve 28B is closed Outboard injection valve 28B is open Inboard injection valve 29B is closed Outboard injection valve 28A is closed d. Outboard injection valve 28B is open Inboard injection valve 29B is open Outboard injection valve 28A is closed QUESTION: 027 (1.00) WHICH ONE of the following RHR loop lineups assures that the design limits are met for adequate drywell sprays and FUIR equipment operation?-
- RHR pump B is running, the RHR heat exchanger bypass valve is closed, and RHR heat exchanger flow is 4800 gp b. RHR pump C is running, the RHR heat exchanger bypass valve:is open, and RHR heat exchanger flow is 3750 gp RHR pumps B and D are running, the RHR heat exchanger bypass valve is closed, and RER heat exchanger flow-is 9100 gp RHR pumps.A and C are running, the RHR heat exchanger bypass valve is open, and RHR-heat exchanger flow is 5050 gp .
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QUESTION: 028 (1.00) LPCI initiated on a valid initiation signal. RPV water level =is -130 inches and increasing slowly. Drywell pressure is 2.0 psig and increasin WHICH ONE of the following describes the actions necessary to open MO-34 and MO-37 to spray the torus under these conditions? a. MO-34 and MO-37 cannot be opened until drywell-pressure increases above.2.5 psig, b. Place the keylock RPV level override switch in MANUAL OVERRID Place the pistol grip LPCI override switch in MANUAL OVERRID d. Place the keylock RPV level override switch in MANUAL OVERRIDE and place the pistol grip LPCI override switch in MANUAL' OVERRID QUESTION: 029 (1.00) An ADS blowdown was initiated on high drywell pressure and low-low reactor water level. The only low pressure CSCS pump running is Core Spray pump WHICH ONE of the following conditions will close the ADS valves? a. The ADS Inhibit switches are placed in INHIBIT, drywell pressure decreases to 2.0 psig, and the high drywell pressure reset pushbuttons are depressed, b. The ADS Inhibit switches are placed in INHIBIT, reactor water level increases to -35 inches, and Core Spray pump B trip Drywell pressure decreases.to 1.0 psig, the high drywell pressure reset pushbuttons are depressed, and the timer rese pushbuttons are depresse .d. Core Spray pump B trips, drywell pressure decreases to 1.8 psig,. and reactor water level increases to -40 inche < . L . _ . ._ . _ _- ~ - - - _ _ _ _ . _
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REACTOR OPERATOR Page 22 QUESTION: 030 (1.00)
: .The control switch for a safety relief valve is in-the REMOTE position on alternate shutdown panel 156. WHICH ONE of the following describes operation of the. safety relief valve in.this condition?
a. The valve will only operate.in the safety mode, b. The valve will only operate automatically in the ADS mode, c. The valve will only operate automatically in the ADS mode or the safety modo, d. The valve will operate automatically in the ADS mode or the safety mode and can be manually operated from the control roo QUESTION: 031 (1.00) Refueling operations are in progress. A fuel assembly is being removed
"from location 19-22 in-the core. Technical Specifications require an operable SRM in the quadrant where the fuel' assembly is being moved and an operable SRM in an adjacent quadrant. WHICH ONE of the following situations meets the Technical Specification core monitoring requirements? (A core map is attached.)
a. SRM 'A' is bypasse SRM 'B' is fully inserted and reading S cp SRM 'C' is fully inserted and reading 2 cp SRM 'D' is fully inserted and reading 8 cp b. SRM 'A'-is bypasse SRM 'B' is fully inserted and reading 10 cp SRM 'C' is bypasse SRM 'D' is fully inserted and reading 4 cps.
l c. SRM 'A' is fully inserted and reading 6 cps.
l SRM 'B' is bypassed.
l- -SRM-'C' is bypasse SRM 'D' is fully inserted and reading 9 cps.
j d. SRM-'A' is fully inserted and reading 12 cps.
j SRM 'B' is fully inserted and reading < cpc'. SRH fCJ is tully inserted and reading 7 cp SRM 'D' is bypassed.
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l REACTOR OPERATOR Page 23 QUESTION: ~032 (1.00) A normal plant startup is in progress with-the mode switch in STARTU IRM A is failed downscale and bypassed with-its function switch on panel
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936 in STBY. All the IRM range switches are on Range 2. WHICH ONE of the following describes the expected plant response if IRM A is taken out of bypass? a. No action b. Rod block c. Rod block and half scram , d. Rod block and full scram QUESTION: 033 (1.00) The Transverse Incore Probe-(TIP) detector was performing an_ automatic scan when power was lost to 120 V lighting panel 17L. Simultaneous with the loss of-power, the reactor scrammed and RPV water level dropped to 0 inches. WHICH ONE of the following describes the response of the TIP System? a. The TIP detector will remain in the core, the ball. valve will ' remain open, and the shear valve must be fired manually, b. The TIP detector will remain in the core, the ball valve will close, and the shear valve will fire automatically, c, The:TIP' detector will retract, the ball valve will remain open, and the shear valve must be: fired manuall d. The TIP detector will retract, the ball valve will_close, and it-is not necessary to fire the shear valv i w --
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- REACTOR OPERATOR Page 24 QUSSTION: 034 (1.00)
With the plant operating at 90%' power, the APRM Calibration section of the OD-3 Printout provided the following results:- lAPRM 1(A) 2(C) 3(E) 4 (B) 5(D) 6(F)- READING 9 .3 8 .0 8 .4 AGAF 0.975 1.102 0.994 1.003 0.987 1.058 WHICll ONE of the following identifies all the-APRMs that require calibration? a. APRM A, APRM D, and APRM E b. APRM D, APRM E, and APRM F c. APRM A, APRM C, and APRM F d. APRM B, APRM C, and APRM F
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REACTOR OPERATOR Page 25 QUESTION: 035 (1.00) The plant is operating at 95% power. The following are the indications received when the APRM meter function switches on Panel 937 are placed in the AVERAGE, COUNT, and FLOW positions: AVERAGE COUNT FLOW APRM A 98% 70% 85% APRM B 94% 60% 97% APRM C 96% 80% 85% APRM D 99% 55% 97% APRM E 93% 55% 85% APRM F 95% 70% 97% WHICH ONE of the following describes the expected plant response for these conditions? a. No action b. Rod block c. Rod block and half scram d. Rod block and full scram _
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D l REACTOR OPERATO Page 26 QUESTION: 036 (1.00) The plant-was operating at'100% power when the 'A' Recirculation pump trippe All of the appropriate actions were taken in response to the recirc pump; tri The following annunciators are lit:
" ROD WITHDRAW BLOCK" (C905L, A1) " ROD BLOCK MONITOR DOWNSCALE" (C905R, F3)
The RBM indicates 75% for the selected ro WHICH ONE of the following describes the-cause of these alarns? A voltage transient during the recirc pump trip caused the RBM to momentarily lower below 5/125 of scale causing the downscale tri b. As core power decreased, the local power around the selected rod also decreased. This caused the RBM output signal 'o decrease below 94% of the reference signal causing the downscale tri Prior to the transient, the RBM High Power Trip Setpoint was-activated for the selected rod. The transient caused power to decrease prior to leveling out at 75%. As reactor power lowered into the Low Power Trip Setpoint, the downscale trip was actuate d. The recirc pump trip caused 'A' recirc loop flow to decreas As a result, the 'A' flow converter input a downscale signal-into the 'A' RBM causing the downscale tri .- , . . , . - _ - . . .-- - REACTOR OPERATOR Page 27 QUESTION: 037 (1.00) Reactor power is at 16% during a reactor startup. Control rod 30-15 is at position 12 and has been selected to be withdrawn to position 4 The reed. switch for Rod 30-15, position 32 is faulty and will not actuate. WHICH ONE of the following describes the expected response if the rod movement control switch is taken to the NOTCH OUT position simultaneously with the " Emergency in/ notch override" switch being taken to the NOTCH OVERRIDE position? a. No rod motion will occur, A rod block will be initiated at position 32 and Rod 30-15 will settle at position 3 c. A rod block will be initiated at position 32 and Rod 30-15 will settle at position 3 d. Rod 30-15 will withdraw to position 4 _ _
REACTOR OPERATOR Page 28 QUESTION: 038 (1.00) Control rods are being withdrawn with reactor power on IRM range 4, in accordance with the attached Control Rod Sequence Sheet. Rod Group N is latche WHICH ONE the following manipulations will result in a RWM select error? a. Rod 14-39 is at position 08 Rod 38-39 is at position 06 Rod 38-15 is at position 04 Rod 14-15 is at position 14 Rod 14-39 is selected _ b. Rod 14-39 is at position 10 Rod 38-39 is at position 12 Rod 38-15 is at position 06 Rod 14-15 is at position 06 Rod 38-39 is selected Rod 14-39 is at position 04 Rod 38-39 is at position 06 Rod 38-15 is at position 04 Rod 14-15 is at position 12 Rod 38-15 is selected Rod 14-39 is at position 14 Rod 38-39 is at position 08 Rod 38-15 is at position 10 Rod 14-15 is at position 04 Rod 14-15 is selected QUESTION: 039 (1.00) A 1st point feedwater heater was just removed from service. There has been no change in turbine steam flo WHICH ONE of the following describes the effect on the plant? Main generator output decrease Plant efficiency increases, Main generator output decrease Plant efficiency decreases, c. Main generator output increase Plant efficiency increase Main generator output increase Plant efficiency decreases.
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REACTOR OPERATOR Page 29 LQUESTION: 040 (1.00) The plant was operating at 100% power when the-"RFP A LOW NPSH" .(C1L, t A1) annunciator is receive The RFP sequential trip selector switch is
"ON" and the reactor feed pump tripping sequence switch is in the " CAB" position. WHICH ONE of the following describes the potential cause of the alarm and the expected automatic actions?
a. The alarm was caused by a trip of one condensate pump. The condenser reject valves open. RFP A will trip if the low NPSH condition persists for 15 second b. The alarm was caused by a trip of one condensate pump. The condenser reject valves shut. RFP C will trip if the low NPSH condition persists for 15 second c. The alarm was caused by placing a condensate demineralizer in servic The condenser reject valves shu RFP A will trip if the low NPSH condition persists for 15 second ~ d. The alarm was caused by removing a condensate demineralizer from service. The condenser reject valves ope RFP C will trip if-the low NPSH-condition persists for 15 second QUESTION: 041 (1.00) The plant is operating at 100% power with the FWLC setpoint set at +28 inche WHICH ONE of the following describes the effect of a loss of one feedwater flow input to the FWLC System? Feedwater flow decreases, reactor water level decreases, and the reactor scram b. Feedwater flow decreases and reactor water level decreases, but the reactor should not. scram, Feedwater flow increases, reactor water level increases, and the reactor may scra Foodwater flow increases and reactor water level increases, but the reactor should not scram.
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REACTOR OPERATOR- Page-30' QUESTION: 042 (1.00) -The plant was operating at 100% power with the MHC System operating normally when the Bypass Opening Jack (BOJ) was taken to the RAISE position. WHICH ONE'of the following describes the expected plant response if the BOJ cannot be returned to the OFF position? a. The bypass valves will open fully. Reactor pressure will decrease and stabilize at a slightly. lower value than before the transien b. The bypass _ valves will open fully, then-the control valves will open until the control valve limit stop is reached. Reactor pressure will decrease and stabilize at a lower value than-before the transien c. The ccntrol valves will open until the control valve limit stop is reached, then the bypass valves will open fully. Reactor pressure will decrease until the MSIVs clos d. The control valves will open-until the control valve limit stop is reached, then the bypass valves will open until the reactor flow limit is reached. Reactor pressure will decrease until the MSIVs clos i _ __ _ ___-._ _ _ _
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REACTOR OPERATOR page 31 I QUESTION: 043 (1,00) A turbino runback was initiated due to high stator outlet coolant temperaturo at time zero. The following table shows tho generator load and stator outlet coolant temperaturo responco during the runback: TIME STATOR AMPS OUTLET TEMPERATURE
+0.5 minutos 25,300 88*C +1.0 minutos 21,800 86*C +1.5 minutos 18,300 84'c '+ minutos 14,000 8?.*C +2.5 minutes 11,300 80'C +3.0 minutes 7,800 '8'C +3.5 minutes 4,300 76*C i -+4.0 minutes 800 74'C WillCil ONE of tl 4 following correctly describes the plant response?
a. The turbine should have tripped at +2.0 minuto < b. The_turbino should have tripped at +3.5 minute . The runback should have stopped at +1.0 minutes, d. The runback should have stopped at 43.5 minute I QUESTION: 044 (1.00) The main condensor vapor valves (AO-3703,-A0-3704, AO-3710, and AO-3711)
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have shut automatically while the plant.was_ operating _at loot powe _ Wi!Icil ONE'of-the following conditions'could_have caused the isolation? i
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a. Offgas'high radiation caused by' abnormal carbon vault-temperature b. liydrogen concentration greater than l' percent caused by _ a recombiner malfunction c. Offgas high pressure caused by an explosion in the offgas system d. liigh- offgas flow caused Sv lowering condonner ' vacuum .l
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_ _ _ - _ _ _ _ _ . l 11EACTOR OPERATOR page 32 QUEST 10!i: 045 (1.00) The plant is in single loop operatio For WilICil OllE of the following conditions can the idle recirculation punp be started? (Consider administrative and functional limitations.)
a. The idle recirculation pump suction and discharge valves are fully open. The scoop tube lock light is off and speed control is in manual with an output setpoint >f zer b. The idle recirculation MG set generator field breaker is ope Lube oil pressure is 17 psig and lube oil temperature is 90* The operating pump is running at 40% of rated spee Core flow is 35 M1b/hr. Core thermal power is 1050 MW The idle pump suction temperature is 408'F. The operating pump suction temperature is 435' Bottom head drain temperatura in 492* Vt.asel dome temperature is 548' QUESTIo!1: 046 (1.00) The plant is operating at 50% power with both recirculation pumps in manua WilICil OllE of the following describes the expected response if all FWLC System input were lost to Recirc Pump B's flow controller? a. Recire Pump B runbacn to 65% due to a speed Limiter #2 runbac b. Recirc Pump B runback to 261 due to a Speed Limiter #1 runback, Recirc Pump B scoop tube will lockup due to a speed control signal failure, d. There will be no effect on Recire Pump B because reactor water % IcVel is normal and the pump discharge valve is full ope .
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QUESTloli: 047 (1.00) i A LOCA han occurred concurrent with a loss of AC power. Both diesel generatoro (DGs) started, but the electric governor for DG B malfunctioned immediately after the start signal was received. WilICli ollE of the following describes the expected response of the DG B output , breaker and DG B? a. The DG output breaker will not close becauno the DG will not
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come up to speed and voltage. The DG will continue to ru b. The DG output breaker will clone because the mechanical governor will take over.' The DG will continue to ru l ' The DG output breaker will clone, but will reopen as DG speed increason due to overcurrent. The DG will continue to ru l The DG output breaker will cloue, but the DG will trip on overspeed causing the output breaker to trip ope \d QUESTIG:: 048 (1.00) The Standbf Gas Treatment (SGT) System initip ed on a valid initiation signal 3 minuten ag Prior to the initia son, both SGT trains were in the normal standby lineu llo operator a ion has been taken and the initiation signal i s still present. Th SGT train A. heater just tripped due to high temperature. WillcIl OllE of the f ollowing describes the expected response of the SGT System? SGT train A fan will tri SGT train A inlet and outlet dampers will cl SGT train B an will continue to ru SGT train B inlet and outlet damperr will remain open, b. SGT train A fan will ip. SGT train A inlet and outlet dampers 1 will clos SGT train B fan will star SGT train B inlet and outlet dampers will ope c, SGT train A fan w'll continue to-run.- SGT train A inlet and outlet dampers w 11 remain open. SGT train B fan will continue to ru SGT tr.in B inlet and outlet damperG will remain ope SGT train A f.1 will continue to run. SGT train A inlet and - outlet dampe .3 will remain ope SGT train B fan will not star SGT rain B inlet and outlet dampers will, remain close , 2 8 e .~, , - . - . . . . - . _ , _ - . . _ ~ , , - - . , . _ , , . . . . . _ , , . . _ , , . , _ . , _ _ , . . - , , . . _ _ _ , _ , - . . _ , ,. , . . . . . .
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QUESTIO!!: a (3.00)
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WilIcIl 011r of the following conditions will causo a secondary containment isolation? a. Refuel floor rad monitor 1705-8A 25 mR/h ' Refuel floor rad monitor 1705-8B: 15 mR/hr Refuel floor rad monitor 1705-8C 20 mR/hr Hofuel floor rad monitor 1705-8D: 10-mR/hr b. Itofuel floor rad monitor 1705-8A: 10 mR/hr Refuel floor rad monitor 1705-8B: 15 mR/hr- l
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Refuel floor rad monitor 1705-8C: 20 mn/hr Rofuel floor rad monitor 1705-8D 0.1 mR/hr c. Refuel floor rad monitor 1705-8A: 0.1 mR/hr Refuel floor rad monitor 1705-813: 20 mR/hr Refuel floor rad monitor 1705-8C: 0.1 mR/hr 0.1 mR/hr
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Refuel floor rad monitor 1705-8D: Refuol floor rad monitor 1705-8A: 15 mR/hr l- Refuel floor rad monitor 1705-8B: 20 mR/hr Refuel floor rad monitor 1705-8C: 0.1 mR/hr - Refuel floor rad monitor 1705-8D: 0.1 mR/hr
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QUESTIoll: 050 (1.00) Reactor Building ventilation fans must be started in the proper order to l maintain reactor building pressure and.provent the spread of contamination. WilICil ONE of the following describes the proper sequence for starting reactor building ventilation fans? , a. Exhaust fans should be started before supply fan Contaminated exhaust (Zone 3) fans should be started before clean exhaust (Zono 2) fan ; b. Exhaust fans should be started before supply fans._ Clean , oxhaust (Zono 2) fans should be started beforo contaminated exhaust (Zone 3) fan c. Supply fans should be started beforo exhaust fans. . Clean exhaust (Zone 2) fans should be started before contaminated exhaust (Zone 3) fan d. Supply fans should be started before exhaust fan Contaminated exhaust (Zone 3) fans should be started before clean exhaust (Zone 2) fan . . _ . _. _ _ _ _ , _ _ _ _ _ _ , _ . _ _ _ . _ . . . . . . . . . . _ . _ . . _ _ _ _ _ _ . . . _ _ _ . . _ . _ _ _ . _ _ _ _ . . _. _ . .
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QUESTION! 051 (1.00) - Control room onvironmontal control supply f an its-77(VSF-103A) la in STBY and supply f an llS-78 (VSF-10313) 10 in AUTO. WilICil ONE of the following describes the Control Room llVAC System responso to a-Italon initiation? a.13oth control room environmental control supply f ans star Normal supply and exhaust fans trip. Italon exhaunt fan starta '; and damper Ao H-141 nhut b. Control room environmental control supply fan 118-78 start !!ormal supply and exhaust fans trip. Cable spreading room cupply and exhaunt dampero shut, e c. Both control room environmental control supply fanc start.- Normal air intake dampero clone and filtration system isolation-dampera open. Cable spreading room cupply and exhaust dampern '
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chu d. Control room environmental control supply f an 11S-70 starts.- Normal air intake dampero close and filtration nyctem isolation dampera open. Italon exhaust f an starto and damper A0 N-141 chuts.
QUESTION: 052 (1.00) If the Fuel Pool Cooling (PPC) System is unavailabic, WilICll ONE of the following nyntema can be cronatied for pool cooling? a. Fire Protection l i b. Residual llent Removal Reacter Building Cloned Cooling Water d. Condensato and Domineralized Water Storage and Trannfor i . _ - . . - _ __--__...___ - __ _ _ _ _ . _ . _ . _. .-- _ .. _ - _ . . .
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REACTOR OPERATOR Page 37 QUESTION: 053 (1.00) The reactor modo switch is in REFUE In WHICH ONE of the following conditions would control rod withdrawal be possiblo? a. One control rod withdrawn to position 0 The withdrawn control rod is selected. The bridge is over the core. -The main hoist is leaded with 600 lbs. The grapple is fully u b. All control rods are full in. The bridge in NOT-over or near the core. The monorail-hoist is loaded with 300 lbs. The grappio is NOT fully u ~ c. One control rod withdrawn to position 02. A second control rod is selected.- The bridge is over the cor The frame mounted-hoist is loaded with 200 lbs. The grapple lu--fully u d. All control rods are full in. The bridge is NOT over-or near-the core. The service platform hoist is loaded with 500 lb QUESTION: 054 (1.00) WHICH ONE of the following Fire Protection Systems would be utilized to suppress a fuel oil fire in the DG fuel oil pipo tronch? a. Dry Chemical System-b. Halon System c. Cardox System Fire Water System h,,
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REACTOR OPERATOR Page 38 ; QUESTION: 055 (1.00) IRM 'IT' is indicating 10 on Range WilICil OllE of the following values will IRM 'B' indicate if the range switch is placed on Range G?
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QUESTIOll: 006 (1.00) A startup is in progress with the modo switch in STARTUP. All SRM detectors are fully inserted. WilICil Ot1E of the following describes the conditions when SRM detector withdrawal should commence? a. When all IRMs are on Range 3
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b. When the heating range is reached
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c. When all the retract permissive indicators are illuminated d. When all required SRM/IRM overlap data has been taken QUESTIO!1: .057 (1.00) , WilICll ONE of the following 120 VAC loads would receive uninterrupted power on a loss of all offsite power? , a. CRD flow control valves (Y-1) _ b.-Rod Position Information System (Y-2) c. IIPCI system control (Y-3)- d. APRMs and LPRMs.(RPS)
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REACTOR OPERATOR Page 39 QUESTION: 058 (1.00) The plant is operating at 100% power with 4160 VAC bussen A-1 through A-G being supplied by the Unit Auxiliary Transforme The auto transfer switches for busses A-3 and A-4 are in the OFF positions. The auto transfer switches for all the other busses are in the ON position.
. Wi!ICll ONE of the following lists all of the busses that will automatically transfer to the Startup Transformer on a full reactor scram while the turbine is still on line? A-1, A-2, A-3, A-4, A-5, and A-6 A-1, A-2, A-5, and A-6 A-5 and A-6 None QUESTION: 059 (1.00) WilICH ONE of the following correctly completes the following statement concerning the procedure for synchronizing the main generator to the < grid and establishing initial loading?
-Turbine speed is adjusted so that the synchroscope is rotating slowly in the directio The ACB control switch is taken to close when the synchroscope pointer is on the mete _
-Initial load is picked up by going to on the governor speed control switc CLOCKWISE, AT 12 o' clock, LOWER CLOCKWISE, 5 degrees BEFORE 12 o' clock, RAISE COUNTERCLOCKWISE, 5 degrees BEFORE 12 o' clock, LOWER COUNTERCLOCKWISE, AT 12 o' clock, RAISE
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I l REACTOR OPERATOR Page 40 QUESTIOll: 060 (1.00) An ATWS has occurred and SLC has been initiate Reactor pressure is ' 1100 psig and reactor power is 25%. WilICil ONE of the following conditions would be indicative of a problem with SLC system operation? a. Annunciator " LOSS OF CONTINUITY TO SQUID VALVE" (C905R, B7) b. Annunciator "STAllDBY LIQUID CO!!T TAl1K llI/LO LEVEL" (C905R, A7) c. SLC pump discharge pressure 1075 psig l d. Squib valve amber lights on C905 out
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QUESTIOll: 061 (1.00) WilICll OllE of the following conditions satisfies the -logic for PCIS to recognize Ri!R in the Shutdown Cooling mode? a. MO-47 is ful,1 open, MO-50 is full open, and reactor pressure is 115 psi b. MO-47 is full open, MO-50 is partially open, and reactor pressure is 105 poi c. MO-47 is partially open, MO-50 is. partially open, and reactor pressure is 95 psi d. MO-47 is= partially open, MO-50 is full closed, and reactor pressure is 85 psi . b
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REACTOR OPERATOR Page 41 QUESTION: 062 (1.00) J A transient occurred which caused RPV water level to decreas The following conditions existed as the transient progressed: ! l T=0 minutes RPV water level dropped below the scram setpoint, ! but the reactor failed to scram T= 7 minutes RPV water level dropped below the low-low level setpoint T= 15 minutes ~ RPV water level is -100 inches RPV pressure is 800 psig , Drywell pressure is 2.0 psig I The Core Spray pump control-switches are placed-in l the stop position and the control switches for the inboard and outboard injection valves are placed in the close position ,- WHICH ONE of the following describes the expected status of the Core Spray system at T=15 minutes? a. The core Spray pumps were running and are now stoppe The inboard injection valves are open and the outboard injection valves are close b. The Core Spray pumps were running and are now stoppe The , inboard and outboard injection valves are close c. The Core Spray pumps did not start. The inboard injection valves are open and the outboard injection valves are close d. The Core Spray pumps did not star The inboard and outboard injection valves are close t - t i
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REACTOR OPERATOR Page 42
QUESTIOll: 063 (1.00) Wi!ICil Ofic of the following describes the operation of the RCIC steam supply isolation valves, MO-16 and MD-17? a. Both valves are full stroke valves for isolation signals and manual closing signals. MO-16 is a jog valve for manual opening and MO-17 is a full stroke valve for manual openin b. Both valves are full stroke valves for isolation signals and manual closing signals. MO-16 is a full stroko valve for manual opening'and Ho-17 is a jog valve for manual openin c. Both valves are full stroko valves for isolation signals and manual closing signal Both valves are jog valves for manual openin d. Both valves are full stroke valves for isolation signals, manual closing signals, and manual opening.
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4s mi p REACTOR OPERATOR Page 43 QUESTION: 064 (1.00) Following a loss of offsite power, llPCI was placed in service for pressure control. Prolonged IIPC1 opuration has caused torus water level to increase to +6 inche WilICll ONE of the following describes the expected response? IIPCI cannot be operated in pressure control mode because the
!!PCI/RCIC test return valve (MO-15) and llPCI full flow test valve (MO-10) automatically close when the llPCI torus suction valves (MO-35 and MO-36) automatically open, b. The llPCI CST suction valve (MO-6) will automatically close, but _
the llPCI torus auction valves (MO-35 and MO-36) will not automatically open because there is no initiation signal present. IIPCI will trip on low suction pressur c. The llPCI minimum flow valve (MO-14) will automatically close, if open, to prevent further increase in torus leve IIPCI will continue to operate in pressure control mode.
P d. The llPCI torus suction valves (MO-35 and MO-36) will automatically open and the llPC1 CST suction valve (MO-6) wil automatically close. IIPCI will continue to operate in pressure control mod QUESTION: 065 (1.00) No fuel movement is in progress. WillCil ONE of the following conditions would be a violation of Secondary Containment integrity? a. Reactor Building differential pressure is granter than 0-inches-of water in STARTU b. Both reactor building to torus vacuum breakers are inoperablo and stuck in the open position in RU c. Both Standby Gas Treatment trains are inoperable in COLD Sl!UTDOWN with the reactor coolant systen vented, Both reactor-building airlock doors are openLto remove a large piece of equipment from the reactor building in llOT- S110TDOW _ _ _ _ . _ _ _ _ _ .__ . _ _
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REACTOR OPERATOR page 44 >
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t QUESTIOll: 066 (1.00) WHICil ONE of the following conditions would be a safety limit violation? E
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a. While operating at 30% power, the MitC pressure regulators fai Reactor pressure drops to 800 psig before the MSIVs close and the reactor scram ,
b. While operating at 75% power, a malfunction in the master recirculation flow controller causes the speed of both recirc pumps to increase to the high speed stop. The Minimum Critical Power Ratio is 1.0 c. While operating at 90% power, a turbine trip occurs. The reactor falls to scram and the ATWS system logic trips the reactor feed pump d. While operating at 100% power, an inadvertent MSIV closure causes reactor pressure to increase. The reactor scrams _and ' reactor pressure increasco until both safety valves lif < QUESTION: 067 (1.00)
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WilICil OllE of the following describes the expected indications if Jet Pump 5 failed?
, a. Recire Loop A flow increase Total core flow decrease b. Recirc Loop A flow increase Total core flow increases, c. Recire Loop A flow decrease Total core flow decrease Recirc Loop A flow decrease Total core flow increases.
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-The plant was operating at 96% power on the 100% load lino for soveral months when Rocirc Pump A tripped due to operator erro Plant conditions following the trip are:
Reactor.Powert 60% Rocirc Loop A Flow 2.0 ML13/IIR (FI-263-107A) i Recirc Loop 11 Flow: 30. 5 MLil/IIR -'
( FI-2 63 -10713)
Total Core Flow: 32.5 MLB/llR (FR-263-110) : WilICll ONE of the following describes the appropriato action in accordance with PNPS Proc. No.-2.4.17, " Rocirculation Pump (s) Trip," 4 (attached)? l Increasc the spood of Rocirc Pump Il until loop flow is greater than 31.5 ML11/li b. Insert control rods to decrease reactor power below the 80% load lino, Restart the idle pump in accordance with-PNPS 2.2.84, "Reautor , Rocirculation System." Scram the reactor and concurrently perform PNPS 2.1.6, " Reactor Scram."
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- REACTOR OPERATOR Page 46 QUESTION: 069 (1.00)
The plant was operating at 100% power when a break developed on the , instrument air header. WHICH ONE of the following describes automatic l actions that should occur before air header pressure drops below 75 psig? I
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a. Service air header isolates Backup K104 air compressor starts Feedwater control valves lock up b. Non-essential instrument air isolates ! Feedwater control valves lock up Scram pilot valve air header depressurizes Drywell pneumatic supply header isolates Lagging compressor loads Service air header isolates d. Low pressure service air crossconnect isolates (if open) Backup K104 air compressor starts Non-essential instrument air isolates QUESTION: 070 (1.00) < The plant was operating at 100% power when a complete loss of instrument-air occurre Power was also lost to the Ac solenoid operated air cutoff valves for the inboard MSIVs and to the DC solenoid operated' air cutoff valves for the outboard MSIVs. The primary containment is inerte WHICH ONE of the following describes the response of the e MSIVa? a. Inboard MSIVs remain ope Outboard MSIVs remain open, s b. Inboard MSIVs close due to air pressure. _ Outboard MSIVs close-due to air pressure.-
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c.__ Inboard _MSIVs close due toLspring pressur Outboard MSIVs remain ope ; Inboard MSIVs remain ope Outboard MSIVs close due to spring-pressur .__
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_ _ _ REACTOR OPEllATOR Page 47 QUESTIO!1: 071 (1.00) An inadvertent containment isolation signal was received and immediately cleare Ilo operator action was taken. WilICl! OllE of the following sets of valves do llOT require a manual reset to reopen the valves if they isolated on the specified signal? Yi1l_v.RD LI193311R11_S111Ilill Recirc sample lines Main Steam Line low pressure b. IIPCI exhaust vacuum liigh drywell pressure breaker isolation valves (w/ low reactor pressure) __ RCIC inlet steam valves Low reactor pressure RiiR/LPCI Injection valves liigh reactor pressure (in Shutdown Cooling mode) QUESTIO!1: 072 (1.00) Three itBCCW pumps are unavailabl WilICil OllE of the following describes the appropriate action that niust be taken with the specified pumps unavailable? RBCCW pulups A, B, and C are unavailable. The RBCCW loops must be crosstied to prevent a reactor scram on high drywell pressure due to a loss of drywell coolin _ RBCCW pumps A, C, and E are unavailabl The RBCCW loops must be crosstied to prevent a trip of the recirculation pumps due to loss of coolin RBCCW pumpa B, D, and F are unavailabl It is not necessary to crosstie the RBCCW loops because at least one pump in each loop is still available, RBCCW pumps D, E, and F are unavailabl It is not necessary to crosstic the RBCCW loops because at least one pump in each loop is still availabl .
._____ _ _ __ _ _ . _ _ . _ _ ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ _ .. _ REACTOR OPERATOR Page 48 i QUESTION 073 (1.00) 1 The Salt Service Water (SSW) System was operating with SSW Pumps A, C, and D running and SSW Pumps B and E in AUTO. The loop selector switch is in its normal positio A complete losu of normal AC power occurs
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simultaneously with a loss of coolant accident (LOCA). - WilICll ONE of tho following describes the response of the SSW System with no operator action? : a. TBCCW heat exchanger outlet valves throttle 90% close ' RBCCW heat exchanger outlet valves open full SSW Pumps A and D restart after load sheddin b. TBCCW heat exchanger outlet valves throttle 90% close RBCCW heat exchanger outlet valves open full SSW Pumpa B and E start after load sheddin , ' c. TBCCW heat exchanger-outlet valves open full RBCCW heat exchanger outlet valves throttle 90% close SSW Pumps A and C restart after load sheddin , d. TBCCW heat exchanger outlet valves open full RBCCW heat exchanger outlet valves throttle 90% close SSW pumps B and C start after load sheddin * QUESTION: 074 (1.00) The plant is operating at 800 MWt. Core flow is 35 MLB/ilR. Condenser vacuum is 18" li WilICil ONE of the following describes tne appropriate actions to be taken under these conditions? a. Reduce recirculation pump speed,-maintaining core flow-above 31.5 MLB/l!R, until the vacuum decrease is terminated' . b. Reduce-recirculation pump-speed until the vacuum decrease is terminated OR the pumps reach minimum spee ' c. Insert control. rods in reverse order of the pull-sheet as necessary to stop the vacuum decreas . d. Immediately scram the reactor and trip the turbin i
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REACTOR OPERATOR Page 49 QUESTION: 075 (1.00) The plant was operating at 95% power when the Train A 4th point feedwater heator isolated due to a tubo ruptur Reactor power increased to 100%. WilICil ONE of the following describes the appropriato immediato actions? a. Hunback recirculation flow and insert control rods as necessary to maintain reactor power below 95%. b. Runback recirculation flow and insert control rods as necessary to maintain reactor power below 25%. c. Runback recirculation flow until reactor power is below 75% or core flow reaches 31.5 Mlh/h Runback recirculation flow until reactor power reaches 70% or core flow reaches 31.5 Mlb/hr.
QUESTION: 076 (1.00) The plant is operating at 100% power when the in-servico CRD flow control valve malfunctions causing CRD system flow to oscillato. WilICil ONE of the following subsequent CRD system failures would require an immediato reactor ucram?_ _ a. Trip of both CRD pumps b. Loss of all rod position indication c. More than one control rod in a 9-rod array drifts d. More than one CRD mechanism high-temperature alarm in a 9-rod array _ _ _ . _ _ -
REACTOR OPERATOR Page 50 QUESTIoll: 077 (1.00) Coro offloading is in progress. A spent fuel bundle is grappled over the coro, when the Refueling Floor Area Radiation Monitor (ARM) alarm lio other alarms have been roccived on the refuel floor or in the Control Roo WilICil OllE of the following describos the appropriato actions? a. Lower the fuel bundio into its designated location in the spent fuel poo b. Lower the fuel bundle into the nearest open location in the spent fuel pool, Lower the fuel bundle into the nearest open location in the reactor vesse i d. Lower the fuel bundle into the location from which it was removed in the reactor vesne QUESTIO!1: 078 (1.00) An electrical failuro has occurred causing multiple alarms to annunciato. One of the illuminated alarms is annunciator " VITAL IllST SYS LOSS OF DC POWER," (C3 Conter, window DS). Based on this , information, WilICl{ OllE of the following describen the amorgency procedures that are applicabic for this situation? a. P!1PS Proc. Ilo . 5.3.6, " Loss of Vital AC (Y-2)" and PilPS Pro . 5.3.13, " Loss of Essential DC Bus-D6" s P!1PS Proc. Ilo . 5.3.6, " Loss of Vital AC (Y-2)" and PllPS Pro lio . 5.3.30, " Loss of 250V DC Bus D-10" PNPS Proc. 11 .3.7, " Loss of Instrument Power Bus (Y-1) " and PliPS Proc li .3.13, " Loss of Essential DC Bus DG" PNPS Proc. No. 5.3.7, " Loss of Instrument Power Bus (Y-1)" an PNPS Proc. No. 5.3.30, " Loss of 250V DC Bus D-10"
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REACTOR OPERATOR Page 51 QUESTION: 079 (1.00) A station blackout has occurred and the SBO diesel failed to star RCIC and HPCI were being used to control RPV level and pressure when both systems tripped and could not be restarte PHPS Proc. No. 5.3.31,
" Station Blackout," directs use of PHPS Proc. No. 5.3.26, "RPV Injection During Emergencies," for alternate methods. WHICH ONE of the methods could be used for RPV injection in this situation?
a. SSW crosstied to RHR b. Fire Water crosstied to RHR
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c. Condannate Transfer crosstled to ECCS fill lines Domineralized Water Transfer crosstled to SBLC _ _ _ QUESTION: 080 (1.00) The Control Room has been evacuated due to a fire in the Cable Spreading Roo At WHICH ONE of the following locations would you find an Alternate Shutdown toolbox? ' Aux Bay b. 23' RPS MG Room ' Switchgear Room ' Turbine Building . . . ___________________________j
_ _ - _ . REACTOR OPERATOR Page 52 QUESTIOll: 081 (1.00) Following a reactor scram, the operator is required to determino whether or not all control rods are inserted to or beyond position 02 in order to determine the appropriate actions to be taken in accordance with the EOPc and P!1PS Proc. 11 o . 2.1.6, " Reactor Scram." WilICil OllE of the following is an acceptable method for determining whether all rods are fully inserted? a. Observing that all the blue scram lights are illuminated on the full-core display, b. Verifying that all APRMs are downscale and reactor power is _ trending down on the IRM Prior to scram reset, verifying that the rod drift annunciator will not clea With the mode switch in HEFUEL, taking the rod select power off, then back on, and observing if the Refuel Mode Select Permissive light illuminate QUESTIO!1: 082 (1.00) E0P-02, "RPV Control, Failure-to-Scram," directs actions based on whether or net the reactor is shutdown. For WilICl! Ol1E of the following conditions would the reactor be considered shutdown in accordance with E0P-02? _ a. Reactor pressure is O psig. The !!ot Shutdown Boron Weight of sodium pentaborate has been injected into th2 RP b. Reactor power is 20 on Range Reactor period is +100 and stable, Reactor power is 50 on Range Reactor period is negative, Reactor power is 30 on Range Reactor period is negative.
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_ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ .____ __ . - REACTOR OPERATOR Page 53 QUESTIOll: 003 (1.00) A reactor scram has occurred and all control rods are not inserte WilICli ONE of the following indications would be expected if the f ailure of the rods to insert was due to an electrical failure of the Reactor Protection System (RPS)? a. Alarm " SCRAM VALVE PILOT llEADER LO PRESSURE" (C905R, A6) illuminated, b. Alarm " SCRAM DISCll VOLUME III LEVEL SCRAM" (C905R, I4) ! illuminate c. Group Scram logic lights on C905 i lluminated, d.-Blue Scram lights on C905 i lluminate .;
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QUESTION: 084 (1.00) i A reactor scram has occurred and all control rods are not inserte In accordance with-PNPS-Procedure No. 5.3.23, " Alternato Rod Insertion," WilICl! ONE of the following methods for inserting control' rods requires the scram to be reset? a. Venting the overpiston areas of the control rod drive b. Individually scramming control rods from panel C91 c. Venting the Scram air header and deenergizing the Scram solenoids, Inserting a manual Scram from panel C90 .
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REACTOR OPERATOR Page 54_ i
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QUESTIOll: 085 (1.00)
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Wli1Cil OllE of the following actions should be taken to maximize drywell cooling in accordance with E0P-03, " Primary Containment Control?" a. Initiate drywell spra , b. Secure the recirculation pump Increase the RBCCW system temperature, d. Increase the RBCCW system flow rat , QUESTIOll: 086 (1.00) WiiICil O!1E of the fol. lowing lineups would maximize torus cooling in accordance with E0P-03, " Primary Containment Control?" Both RilR loops in service with one pump por loop operating and the HilR heat exchanger bypass valves shut, b. One RllR loop-in service with both pumps-operating and the RllR
heat exchanger bypass valve shu ; c. Both Ri!R loops in service with one pump per loop _ operating and the RilR heat exchanger bypass valves:ope d. One Ri!R loop in service with both pumps operating and the RilR' heat exchanger bypass valve open.
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. _ . . REACTOR OPERATOR page 55 QUESTION: 087 (1.00) A station blackout and complete loss of instrument air has occurre Plant conditions are as follows:
-Drywell pressure is 3 psig and increasing slowly-Torus water level is 220 inches-Neither Standby Gas Treatment (SGTS) train is operabic because the outlet dampers are stuck in the closed position WilICH ONE of the following describes the reason that the Direct Torua Vent (DTV) cannot be used in this situation?
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a. SGTS must have an operable vont path for use of the DT b. Primary Containment conaitions do not meet the criteria for use-of the DT c. The DTV path is not available during a station blackout, d. The DTV path is not available during a loss of instrument ai QUESTION: 088 (1.00) Both torus water level indicators are upscale and the primary containment water level indicator is downscal For WHICH ONE of the following conditions is this an accurate indication? (PNPS Proc. N .3.27, " Determining Primary Containment Water Level," is attached.) - a. Drywell Wide Range Pressure: 25 psig Torus Bottom Pressure: 10 psig b. Drywell Wide Range Pressure: 5 psig Torus Bottom Pressure: 35 psig c. Drywell Wide Range Pressure: 20 psig Torus Bottom Pressure: 40 psig d. Drywell Wide Range Pressure: 26 psig Torus Bottom-Pressure: 20 psig
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REACTOR OPERATOR Page 56 t QUESTIOll: 089 (1.00) . - Following a major loss of coolant accident and loss of off-sito power, ;
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RPV level cannot be maintained abovo TA The operating crew determinos that an alternato injection system noodo to be aligned in order to recover RPV level. Reports from the field indicate that the aux. bay in ! inacconcible. WlIICll OllE of the following lineups could be used to deliver a high capacity flowrate to the voucal given the current plant
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conditions? , USW crocotled to R!lR l b. Fire water cronstled to RilR Fire water crosntied to Foodwater Domineralized water tranufer crountied to SDLC QUESTIO!1: 090 (1.00) A reactor startup is in progron The turbino is at 600 RPM when the TURBIllE IIIGli VIHRATIOli (C2 Loft, window D4) alarm 10 roccive At WillCll a OllE of the following vibration levels should the turbino be tzipped? mils . h. 12 milo mila l' mils . !
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_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ . _ _ _ . _ _ . . . _ _ , . REACTOR OPERATOR Page 57-QUESTION: 091 (1. 00) ' The plant was operating at-100% power when all service water was lost ; due-to an unisolablo pipe brea WHICH ONE of the following describes l the appropriate immediate actions in accordance with PNPS Proc. N l 5.3.3, " Loss of All Service Water?" l " a. Isolate the RWCU System and monitor CRD temperature l b. Reduce reactor power as necessary to maintain equipment temperatures within operating limit c. Reduce reactor power and trip all but one Feedwater pum l Scram the reactor and trip all the Feedwater pump J l
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l l } QUESTION: 092 (1.00) ;
WHICH ONE of the following would be a concern if Jet Pump 5 failed? a. The cross-sectional flow area for blowdown following a doubic-ended recirculation line break would decrease.
t i b. It could preclude the capability of maintaining 2/3 core coverage during a loss of coolant accident.
' ' Failure of_the jet pump body could cause a-subsequent failure of g the jet pump nozzle.
} d. Resulting inaccuracies in indicated core flow would affect
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thermal limit calculations nonconservativel . _ . . . . _ _ . __
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REACTOR OPERATOR Page 58
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. QUESTION: 093 (1.00)
WHICH ONE of the following is the desired sequence for operation of the - Safety Relief Valves (SRVs)? a. A-B-C-D t b. B-C-D-A c. A-C-B-D- B-D-C-A ,
' QUESTION: 094 (1.00)
The plant was operating at 60% power when the MSIVs closed on low pressur Several rods failed to insert and all the APRMs are downscale. WHICH ONE of.the following describes the-appropriate actions with respect to operation of the recirculation pumps? a. Leave the recirc pumps operating at the present spee b. Runback the recire pumps to-minimum.- c. Runback the recirc pumps to minimum, then trip the recirc pumps, d. Trip the recirc pumps immediately.
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REACTOR OPERATOR Page 59 QUESTION: 095 (1.00) The plant was operating at 100% power when the " MAIN STEAM LINE !!I RADIATION SCRAM," (C905R, Ab) alarm was received due to a valid signa The MSIVs closed, but the reactor failed to scram. WilICl! ONE of the following describes the appropriate actions for RPV pressure control? Depressurize the RPV at less than 100aF/hr using the main turbine bypass valves, Depressurize the RPV at less than 100aF/hr using IIPCI in full flow test and the SRV _ Maintain reactor pressure less than 1085 psig using the main turbine bypass valve Maintain reactor pressure less than 1085 psig using llPCI in full ' flow test and the SRVs.
s QUESTION: 096 (1.00) The isolation Gwitches for DG A on the alternate shutdown panels are in the LOCAL positio A transient occurs which causes drywell pressure to increase to 5.0 psi Subsequently, the DG A jacket water temperature increases to 200"F. WilICll ONE of the following describes the expected response of DG A to this event?" DG A does not star _ DG A starts, then autor.atically trips on hiqh jacket water temperatur DG A starts, continues to run with high jacket water temperature, but can be tripped manually by depressing the local stop pushbutton on panel C10 DG A starts, continues to run with high jacket water temperature, and cannot be tripped manually by depressing the local stop pushbutton on panel C10 ._ _ _ _ _ _ _ _
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QUESTION: 097 (1.00) Reactor Building HVAC is operating and no vacuum breakers are ope Current plant conditions are:
- 'Drywell pressure (PID-5067A): +1.7 psi -
Torus pressure (PID-50678): -0.5 psi
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Reactor Building pressure (manometer on Panel C-61):- 0 inches of water (pegged high)
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Reactor Building dp - north (DPI-8117 on Panel C-7): 0 inches of water (pogged high)
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Reactor Building dp - south _
(DPI-8128 on Panel C-7): 0 inches-of water .(pegged _high)
WHICH ONE of the following lists the documents that should be referenced
'for these conditions?" EOP-03, " Primary Containment Control" and EOP-04, " Secondar Containment Control" b. EOP-03, " Primary-Containment Control" and suppression chamber t drywell vacuum breaker Tech Specs EOP-04, " Secondary Containment Control" and reactor building to torus vacuum breaker Tech Specs Suppression chamber to drywell-vacuum breaker Tech specs and
- reactor building to torus vacuum breaker Tech Specs QUESTION: 098 (1.00) WHICH ONE of the following alarms would require entry into EOP-04,
" Secondary Containment Control?"
3 a. " STACK CAS HI-HI RADIATION" (903R,-C2) " REFUELING FLOOR: AREA-HI RADIATION" (903R, B4)- " BUILDING EXHAUST HI RADIATION" (904L, F3) " STANDBY GAS TREATMENT DISCHARGE HI RADIATION" (904L, A3).
- - , , . . m-, -
, - .-
- .. ._-. -- . . . ,-- .- - . - - - - -- - - . _ - . . _ - - . .
.
REACTORLOPERATOR' Page161-- QUESTION: 099 (1.00)
-WHICH ONE of the following systems would be considered a primary system when implementing EOP-04, " Secondary Containment Control?" vuol Pool Cooling Reactor Building Closed Cooling Water c. Reactor Water Cleanup Primary Containment Cooling QUESTION: 100 (1.00)
A fuel cladoing failure has occurred, causing offgas radiation levels to increas Keylock switch 17A-S12 on panel CP600 is in POSITION-2. The Offgas Adsorber System is in AUTO. WHICH ONE of the following describes the expected response of the Offgas System? a. The charcoal adsorber bypass valve shuts and the inlet valve opens when the SJAE Offgas radiation monitors trip on1high radiation. The Offgas isolation and holdup line_ drain valves close 13-minutes after the AOG post treatment radiation monitors trip on high radiation, b. The charcoal adsorber bypass valve shuts and the inlet valve opens when the AOG post treatment radiation monitors trip on _ high radiation. The Offgas isolation and holdup line drain o valves close 13 minutes after the SJAE Offgas radiation monitors trip on high radiatio c. The charcoal adsorber bypass valve shuts and the inlet . valve opens-when the AOG post treatment radiation monitors. trip on-high radiation. The Offgas isolation and holdup line drain valves close when the Main Stack radiation monitors trip on high radiatio d. The charcoal adsorber bypass valve shuts and'the inlet valvo 1 opens when the'SJAE Offgas radiation monitors trip on high radiation. The Offgas isolation and holdup line. drain valves close 13 minutes after the. Main Stack radiation monitors trip on high radiation, i~ l l I-t '
(********** _
END OF EXAMINATION **********) l l p I ,- -
REACTOR OPERATOR Page 62 yk& APSWER: 001 (1.00) REFERENCE: IG: 0-RO-O -11-01, " Control Room Computer System (EPIC /SPDS)," ELO (3.2 3.4] 29 001A115 ..(KA's) ANSWER: 002 (1.00) REFERENCE: PNPS Proc. No. 6.1-024, " Radiological Posting of Areas of the Station." Module C-GT-02-02-01, " Introduction to Radiation Protection," ELO (3.3/3.8]
- .
294001K103 ..(KA's)
-ANSWER: 003 (1.00)
b g.
l l ! l l < - . _
--. _ .- . - . . - . . . - ~ . . - . - . . . . - . . . . .- .. ...~. .~ . . . -. .-
REACTOR OPERATOR- ' Page 63
. REFERENCE:
1. . Module C-GT-02-02-02, " Exposure Limits," ELO' (3.3/3.8) i l 294001K103 ..(KA's)
.
ANSWER: 004 (1.00) ; REFERENCE:
' - ' 1' . PNPS Proc. No. 1.4.9, " Storage, llandling, and-Disponal of Sodium Pentaborate," page 6, UG: 0-RO-04-04, " Emergency Tasks," Task 9 [3.1/3.4)
294001K110 ..(KA's) ANSWER: 005 (1.00) REFERENCE: , PNPS Proc. No. 1.4.36, "lligh Pressure / Compressed Gas-Cylinde Control." IG: C-GT-01-01-03, " Industrial Safety," LO 3 [3.4/3.8] 294001K109 ..(KA's) .
-
1 Y
. _ . _ _ . _ _ _ . _ _ _ _ _ . . . . . - _ . _ __ _ .. . . . _ . . .__ . - . ._ __. .. . . .
;
REACTOR: OPERATOR _ . Page 64
.
-
ANSWER: 006 (1.00) -;
-i ~ REFERENCE: PNPS Proc. No. 1.3.34, " Conduct of Operations," pages 38 and 39.- (3.7/3.7]
294001K101 ..(KA's)
'
,
J "- ANSWER: 007 (1.00)
- REFERENCE: PNPS Proc. No. 1.4.5, " PNPS Tagging Procedure."
. UG: 0-RO-04-02, " Administrative Tasks," Task [3.9/4.5) 294001K10 ..(KA's) ANSWER: 008 (1.00) _ k
4-(- .
.31-. ,- < -r w wrwr v w-m-,w- - - , -mm n --%e,- , * -----n=m+- -w- -
-_ _- ._ _ _ . . _ _ _ . . _ _ _ _ . _ . __ . .,_ _ __
'
REACTOR OPERATOR Page 66 ANSWER: 011 -(1.00) s d. de REFERENCE: IG: 0-RO-03-04-02, "EOP Development and Use," ELO 2 [4.2/4.2) 294001A102 ..(KA's) ANSWER: 012 (1.00) REFERENCE: IG: T-ER-01-01-11, " Emergency Plan Training for Licensed Operators and shift Technical Advisors," ELO to be added (per A._Shiever).
[2.9/4.7] 294001A116 ..(KA's) l ' ANSWER: 013 (1.00) d.
i ( l l I
,.
REACTOR OPERATOR Page 67 REFERENCE: IG: T-ER-01-01-11, " Emergency Plan Training for Licensed Operators and Shift Technical Advisors," ELO (2.9/4.7) 294001A116 ..(KA's) ANSWER: 014 (1.00) \ REFERENCE: PNPS Technical Specifications 3.2.11 and 3.7. . OJT Guide 10, Technical Specification Objectiv {3.3/4.2] 223001G011 ..(KA's) ANSWER: 015 (1.00) d.
_- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ -
REACTOR OPERATOR Page 68 REFERENCE: 1. . IG: 0-RO-02-04-02, " Condensate and Feedwater System," ELOs 61 and-8 . IG: 0-RO-02-04-01, " Main Steam System," ELO 1 . IG: 0-RO-02-06-03, " Control Rod Drive System," ELO 2 . IG: 0-RO-02-07-02, " Reactor Protection System and Anticipate Transient Without Scram System," ELOs 14 and 2 . Modified question from 11/26/90 NRC Exam (modified / replaced distractors and reworded correct answer).
[3.2/3.0) 239003G008 ..(KA's) ANSWER: 016 (1.00) b.
REFERENCE: IG: 0-RO-02-06-05, " Reactor Water Cleanup System," ELO [3.4/3.3] 204000G007 ..(KA's) ANSWER: 017 (1.00) a.
-_ _ . . _, _ . . _ , . _ . ,_
~ . . - - - - - . . . - . - . . . . ~ ~ . - . . . . . - . . . -. ..~. - . - -
REACTOR OPERATOR -Page.69 REFERENCE: 1.- IG: 0-RO-02-06-01, "Non-Nuclear Instrumentation and Reactor Vessel Internals," ELO . Modified question from 11/26/90 NRC Exam (changed conditions) .
[3.6/3.8) -
216000K122 ..(KA's)
;
ANSWER: 018 (1.00)
, ,
REFERENCE: IG: 0-RO-02-06-01, "Non-Nuclear Instrumentation and Reactor Versel Internals," ELO 1 .. Systems Reference Text: Nuclear Boller Instrumentatio . LO Requalification IG: Reference Leg Perturbations Due to Non-Condensable . Modified questions from 11/26/90 and 12/2/91 NRC Exams (combined concepts of questiens and changed correct answer).
[3.3/3.5] 216000A210 ..(KA's) ANSWER: 019 (1.00) b.
, REFERENCE: L IG: LO-RO-02-07-02, " Reactor Protection System and_ Anticipated Transient Without Scram System," ELO 2 (3.8/4.2): 212000G010 ..(KA's)
=. _- - ._
. - .
. . _ . . . _ ~ . . . . - . . _ - - _ . _ . - _ _ . . _ . . _ _ _ _ . _ _ . _ . - _ . . _ _ - . _ .. _ _... _ _ .
.
. - . . .
h-REACTOR. OPERATOR Page-70 h ANSWER: 020 (1.00) REFERENCE: PNPS Proc. No. 2.2.79, " Reactor Protection System," page 1 . IG: 0-RO-02-07-02, " Reactor Protection System and Anticipated Transient Without Scram System," ELO 1 . Facility Question TYPA-13 (from Requal Retake Exam).
[3.9/4-1) 212000K412 ..(KA's) ANSWER: 021 (1.00) c.
, REFERENCE: PNPS: ARP-905R-D4, Rev. . IG: 0-RO-02-04-01, " Main' Steam System," page IG-19-7/90, ELos 18 and 21 . Modified question from- 11/26/90 NRC Exam (changed conditions).
[4.0/4.1) 239001K127 ..(KA's) , ANSWER: 022 (1.00) a.
2-
, - , .a , -
REACTOR OPERATOR Page 71 REFERENCE: IG: 0-RO-02-08-01, " Primary Containment System," ELos 33 and 34, Modified questions from 11/26/90 and 12/02/91 NRC Exams (changed conditions, distractors, and correct answer).
[3.5/3.5) 223002A302 ..(KA's) ANSWER: 023 (1.00) REFERENCE: IG: 0-RO-02-09-04, " Reactor Core Isolation Cooling System," ELO 1 . Modified question from 11/26/90 NRC Exam (changed conditions, distractors and correct answer).
[3.8/3.8) 217000A215 ..(KA's) ANSWER: 024 (1.00) REFERENCE: IG: 0-RO-03-04-02, "EOP Development and Use," ELO 2 {3.9/3.8] 206000G010 ..(KA's) _ _ _ _ - _ _ _ _ _ _ _ _ -
. . - -- .- .. . _ . _ .. .- - . . . . . - - - . . _ - . . -
.
,
REACTOR-OPERATOR- ;Page-72
,
' ANSWER: 025 (l'. 0 0) -
' .;
_ REFERENCE: IG: 0-RO-02-09-02, " Core Spray Systent," ELO_1 [3.3/3.6] 209001A205 ..(KA's) ANSWER: 026 (1.00) c REFERENCE:. IG: 0-RO-02-09-01, " Low Pressure Coolant Injection and Residual Heat Removal," ELOs 9 and 1 . Modified questions from 11/26/90 and 12/2/91 NRC Exams (changed' conditions).
,. [4.2/4.3) 203000A101 ..(KA's) l 027 (1.00) ' , ANSWER: l d.
!
. ,,= . . . .- ,
.. _. .._ .. _- _._.___ __ . _..._ _ - _.... _
_ _ _ .. _ ._ _ _ _.. _ ._ .. .__ _ . . . +
^
iREACTOR OPERATOR _ Page.73_ REFERENCE: PNPS Proc. No.-2.2.19, " Residual Heat Removal," page 19, 'UG: 0-RO-04-04, " Emergency Tasks," Task 89 .- Modified question from 11/26/90 NRC Exam (changed values in distractors).
[3.2/3.4) 226001G010 ..(KA's) ANSWER: 028 (1.00)
, REFERENCE: PNPS Proc. 2.2.19, " Residual Heat Removal," page 2 . IG: 0-RO-02-09-01, " Low Pressure Coolant Injection and Residua Heat Removal System," ELO 1 . Derived from Facility Questions 10-B and RHR/SDC-07.
[4.0/3.9]
230000A406_ ..(KA's) ANSWER: 029 (1.00) c.
i~ t-REFERENCE: . IG: 0-RO-02-09-05, " Automatic Depressurization System,j" ELOs. 5 'and-15 . - '. [ 4' . 2 / 4 . 3 ] 218000A206 ..(KA's) P
L I ' . . -. - _
. - . .
__. . REACTOR OPERATOR.:
.
Page 74
,
ANSWER: 030 (1.00) REFERENCE:
, IG: 0-RO-02-09-05, " Automatic Depressurization System," ELO-1 ' [3.6/3.7)
239002K405 ..(KA's) ANSWER: 031 (1.00) ,
= REFERENCE: PNPS Technical Specification 3.1 . OJT Guide N , SRM System Objective [3.2/3.9)
215004G011 ..(KA's) ANSWER: 032 (1.00) '
-
J
._
i y- - - ..- .,yv- -
REACTOR OPERATOR Page 75 REFERENCE: IG: 0-RO-02-07-01, " Neutron Monitoring Systems," ELOs 18, 23, and 2 [3.5/3.7) 215003A202 ..(KA's) ANSWER: 033 (1.00)
-_ REFERENCE: IG: 0-RO-02-07-01, " Neutron Monitoring Systems," ELos 40, 42, 44, ,
and 4 [3.4/3.7) 215001A207 ..(KA's) ANSWER: 034 (1.00) _ REFERENCE: PNPS Proc. No. 2.1.15, " Daily Surveillance Log (Tech Specs and Regulatory Agencies) ," Daily Log Test #2 . UG: 0-RO-04-02, " Administrative Tasks," Task . Modified question from 12/2/92 NRC Exam (changed conditions, distractors, and correct answer).
[3.0/3.4) 215005A107 ..(KA's) . _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. _ ~ . ___ . . _ _ . . _ . - . _ _ _ . . _ _ _ _ . . .--. . - . . - . . . . _ . . _ __._ ___ ,_ - - _ _ _ -_._._. _ _ _ . . - . _ . _ . .
'
REACTOR OPERATOR -Page 76 ANSWER: 035 (1.00)
. <
REFERE!1CE:
' IG: 0-RO-02-07-01, "floutron Monitoring Systems," ELos 63, 64, and 6 [3.2/3.4]
215005A207 . . (KA's) ANSWER: 036 (1.00) a REFERENCE: IG: O-RO-02-07-01, "Noutron Monitoring Systems," ELO 5 . Facility Question PSU-24 (from Requal Retake Exam).
[3.3/3.3] . 215002A202 . . (KA's) ANSWER: 037 (2.00) 1 C.
! i l l
-! ,
w "w 6 v-- <= e n- _aos-w , - -v, rv v- --e,e , n y m-:*g w, :- ** -mm -- u- m T s'+ - t
.-. - ._ ....- . .... - . - . . . ~ ~ - . ~ , . . . - . . . . . - - . - - - - ~ . - - -- _-. .. .. -,
-REACTOR OPERATOR Page 77
'
' REFERENCE:
-
- 1 '. IG: 0-RO-02-06-04, " Rod Position Information System," ELO 11.d.
.,
, [3.1/3.3)
214000A201 ..(KA's) ANSWER: 038 (1.00) -
,
a , cr~ ) . , REFERENCE: IG: 0-RO-02-06-03, " Control Rod Drive' System," ELO 29.
,_
(3.5/3.5)-
. 201006K403 ..(KA's) !' ANSWER: 039 (1.00) i
" REFERENCE:
- IG: 0-RO-02-04-02, " Condensate-and Feedwater System," ELO:2 s
. .
[3.3/3.4)
I _259001A204 ..(KA's) _- ._. .
.. .._._ .,_ . _ ... _ .-- _ ~, ._-.._.- _ _ _ _. _ _ _ _ _.- - - .. . .>_ _ m_.--
: REACTOR OPERATOR - Page-7 !
!
I ANSWER: 040 (1.00) C.
I REFERENCE: IG: 0-RO-02-04-02, " Condensate and Feedwater System," ELO 15.-
[3.6/3.7]
256000K304 ..(KA's) ,
.
ANSWER: 041 ( 1. 00) REFERENCE: IG: 0-RO-02-06-02, " Condensate and Feedwater System," ELO 83; Modified question from 11/26/90 NRC Exam (modified distractors and _ correct answer).
[3.1/3.1] 259002K604 ..(KA'a) ANSWER: -042 (1.00) d.
I ,
...y , y, , _--- - - , _ , - - -- - ---- . , - , - . . -.-- - --- - - - - - - - - - . -
.. ._ _ . _ . _ . _ . _ . _ . _ _ _ _ _ . . . . . . _ . _ _ . . . . _ . _ _ _ _ _ _ . _ - . _ _ _ _ _ - . . . _ _ _ . . - . REACTOR OPERATOR- --
Page 79-
-REFERENCE: IG: 0-RO-02-05-01,: " Main Turbine System," ELos 70, 79, and'8 , [4.1/4.1)
241000K306 ..(KA's)
!
ANSWER: 043 (1.00) REFERENCE: '
. IG: 0-RO-02-01-03, " Main Generator,"~ELO 19, IG: 0-RO-02-05-01, " Main Turbine-System," page IG-34-5/89, Modifled questions from 11/26/90 NRC Exam (combined 2 questions). _ (2.7/2.8)
245000A304 ..(KA's).
ANSWER: 044 (1.00) REFERENCE: IG: 0-RO-02-04-03, " Main Condenser Vacuum and Augmented Off Ga .
= Systems," ELOs 2, 91, 18d, and 18e.
,
[3.5/3.9)
271000A206 ..(KA's)
5'- 1 .c...,
. . _ _ _ ._
~ . _
. - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
l ' REACTOR OPERATOR Page 80 ANSWER: 045 (1.00) REFEREllCE: IG: 0-RO-02-06-02, " Recirculation System," ELOs 9, 14, 15, 18, and 2 [3.3/3.4] __ 202001K410 ..(KA's) ANSWER: 046 (1.00) REFERENCE: IG: 0-RO-02-06-02, " Recirculation System," ELO 3 [3.5/3.5) - 202002K604 ..(KA's) ANSWER: 047 (1.00) . REACTOR OPERATOR Page 81 REFERENCE: IG: 0-RO-02-09-06, " Diesel Generator System," ELO 2 [3.8/3.7] 264000K408 ..(KA's) N
>
ANSWER: 048 (1.00) REFERENCE: IG: 0-RO- 2-08-03, " Standby Gas Treatment System," ELos 9, 13, and 15 {2.9,[.2] 2r 000A203 ..(KA's) ANSWER: 049 (1.00) - t REFERENCE: , IG: 0-RO-02-08-05, " Plant Ventilation Systems," ELO . Modified question from 11/26/90 NRC Exam (changed correct answer and conditions on all distractors).
[3.6/3.9) 272000K403 ..(KA's)
,
REACTOR OPERATOR Page 82 I ANSWER: 050 (1.00) REFERENCE: IG: 0-RO-02-08-05, " Plant Ventilation Systems," ELos 9, 10, and 11, Expanded Previous Question fron 12/02/91 NRC Exa [3.3/3.4) __ 290001G010 ..(KA's) ANSWER: 051 (1.00) i b, REFERENCE: IG: 0-RO-02-08-05, " Plant Ventilation Systems," ELOs 5 and . Modified question from 11/26/90 NRC Exam (added concept and changed answers).
[3.1/3.3] _ 290003A204 ..(KA's) ANSWEP: 052 (1.00) _- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ - _ _ - _ _ _ _ _ - _ _ _ - _ - _ . .. . . ~. ._- .. . _ . ~ . . . . . - . . . . .. -. - . - -.~.-. -. . - . . . . . . . -
- REACTOR' OPERATOR: Page 83 i REFERENCE: , IG: 0-RO-02-09-01, " Low Pressure Coolant Injection and Residual lleat Remova1 System,"-ELO 1 . IG: 0-NL-03-11-01, " Fuel Pool Cooling," ELO [2.9/3.0]
233000K102 ..(KA's) ANSWER: 053 (1.00) REFERENCE: IG: 0-RO-02-08-06, " Fuel Handling Equipment," ELO 8.
,
[3.3/4.1]
234000K402 ..(KA's)
ANSWER: 054 _ (1. 0 0) a.
I _ REFERENCE: 1.- IG: 0-RO-02-10-01, " Fire Protection System,"-ELOs 15,-16, 17,-and 1 [3.8/3.9]- 286000G004 ...(KA's)
.
J
, -r,4 ,-- - .
. .
5 REACTOR _ OPERATOR Page184-ANSWER: 055 (1.00). REFERENCE: IG: 0-RO-02-07-01, " Neutron Monitoring Systems," ELos-16 and 1 [3.6/3.4] _ 215003A403 ..(KA's) ANSWER: 056 (1.00) d.
. REFERENCE: IG: 0-RO-02-07-01, " Neutron Monitoring Systems," ELO . PNPS Proc. No. 2.2.64, " Source Range Monitoring System," page .- PNPS Proc. No. 2.1.1, "Startup From Shutdown," page 2 (3.6/3.9) _ 215004G001 ..(KA's) ANSWF.R:- OS7 (1.00) . . . , . . . .
. _ _ _ . _ _ _ _ _ _ . . - . _ . , _ ..m.. . . .~.._..-....- -- ._ . _._. . ._._ . . - - . . ~ _ - . . _ . . - - _ _ . _ . _
,
REACTOR OPERATOR 'Pagel8 _ REFERENCE:
- IG: .O-RO-02-01-02, "AC Electrical Distribution," ELOs.34, 35, and 3 [2.7/2.9]
262002K103 . . (KA's)
~ ANSWER: 058 (1.00)
,
' REFERENCE: IG: 0-RO-02-01-02, "AC Electrical Distribution," ELos 14b, 15, and 1 [3.2/3.3) ;
'
262001A302 . . (KA's) ANSWER: 059 (1.00) REFERENCE: l IG: 0-RO-02-01-03, " Main Generator," ELO _ I' [3.1/2.9] 245000A402 . . - ( KA ' s )
-< ' .-, , - - - - - _ _ . . ~ . . _
_. . . _ _ _ . _ . . . . . _ _ . . . . _ _ . _ . _ - . _ . . . . _ _ . . _ _ . _ -_ . _ _ _ . _ . - -
. _ . _
. REACTOR OPERATOR Page 86-ANSWER: 060 (1.00)
,
c.
REFERENCE:
' IG: 0-RO-02-06-06, " Standby Liquid Control System," ELO I
,
[3.5/3.5)
211000A301 ..(KA's) ANSWER: 061 (1.00) REFERENCE: IG: 0-RO-02-08-01, "Prinary Containment Systems," page 34, IC: -O-RO-02-09-01, " Low Pressure Coolant Injection and Residual lleat Removal System," ELO 1 . Modified question from-12/2/91 NRC Exam (changed-distractors and correct answer).
[3.6/3.5) f 205000A402- ..(KA's) ANSWER: 062 '(1. 00) _ . . _ _ _ . _ _ . _ , - _ . , _ . . - . ,. . . . _ , ,. . . . _ . . . _ , _ . , , . . _ . . ~ . _ , _ . . . .
. _ . _ _ _ - . . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ . _ _ _ _ - . - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . l i l REACTOR OPERATOR Page 87 i
,
REFEREllCE: I IG 0-RO-02-09-02, " Coro Spray System," ELos 4 and . Modified question from 11/26/90 liFC Exam (changed conditions and answers).
(3.7/3.7) ,
209001A104 . . (KA's) , l A!1SWER: 063 (1.00) -; i a,
,
REFERE!1CE: ,
'
, PIlPS Proc. 11 o . 2.2.22, " Reactor Coro Isolation Cooling System " .
.
'
~ IG: 0-RO-02-09-04, " Reactor Core 1colation Cooling System," ELO . Facility Question TYPA-4 (from proposed Requal Retake Exam). ,
[3.4/3.3)
!
217000A403 . . (KA's) _ _ _ -_ - A11SWER: 064 (1.00) -, , REFEREllCE: IG: 0-RO-02-09-03, " lligh Pressure Coolant Injection," ELO<1 . I (3.7/3.8) 206000A104 . . (KA's) ._ - g ) t T '$' -T-F*-t'TrvT-y-y m y w yy'upipy -g y -, yrw-ivy =r- y am-M q w g z m t--g--=wygrm..r--- e +-~)et M-w er e 7 -r --yyM9- === iwe .g-y-%,-.- e- gm - *-
_ _ _ - - . _ _ _ _ _ . . _ _ _ . - -. _ _ . . . . _ _ . _ _ .._ _ -_ .._ . _ . _ . _ _ _ _ _-..-__ _ _ .~. _
,
,
REACTOR OPERATOR Pago 88
,
F
;
}
AllSWER: 065 (1.00) j : REFERE11CE:
' PliPS Technical Specifications, sections 1.0.11 and 3. . IG: 0-RO-06-01-01, " Technical Specification Dofinitions " ELO , r IG! O-RO-04-09, " Technical Specification Overview," ELO [3.9/4.2) l
.
295035K101 . . (KA's) AllSWER: OG6 (1.00) i REFERE!1CE:
. P!1PS Technical Specifications, section . IG: 0-RO-06-01-02, " Safety Limits and Limiting Safety System <
Settings," ELO ,
(3.5/4.3)
295025G003 - . . (KA's)
,
p A!1SWER: 067 (1.00) a.- ,
,
I-
,
a i
.
y _ .--- - - - - -- - yr - c -c y- 2i,wie--,ir r*rir2-,,->.-+my-,=w-r = 'eirie"- r-w+-' -- t- -+w,-+-- - * -= - - =-e---m= e.v-i- w w-r -av-= ~w-
+
rw -e se m- * - - = - - - = = - - -v='-cr*--o-=w--p w e+- - - *
.. REACTOR OPERATOR Page 89 REPEREliCE: PilPS Technical Specifications, 3.6.E t.nd 4.6.E Bases, page 147 . OJT Guide ll , llecirculation System Objective 2 [3.9/4.2) 295001G011 ..(KA's) A!1SWER: 068 (1.00)
- REFERE!JCE: Pill'S Proc. 11 0 , 2.4.17, "ltecirculation Pump (s) Trip," page . OJT Guide 4, Recirculation System Objective 2 [3.5/3.8)
29S001A201 ..(KA's) A!1SWER: 069 (1.00) _ REFEREllCE: P!1PS Proc. 11 o . 5.3.8, Rev. 16, " Loss of Instrument Air," page . IG: 0-RO-02-02-04, " Instrument and liigh Pressurn Air," ELO . Modified question from 12/2/91 11RC Exam (changed conditions).
[3.6/3.7) 295019A202 ..(KA's) l B_ _ _ _ _ _ _
.. .- -. - . - . . - - .~.. - . - .. -. . .--..~.. . ~ . .-.-. . - _ - . . . . - ~.~._, .
.
REACTOR OPERATOR Page 90 l i
,
i ANSWER: 070 (1.00)
.i ,
i REFERE!1CE: !
" Main IG 0-RO-02-04-01, Steam System," ELO 2 I (3.6/3.7) .
i e 295020K201 . . (KA's) . A11SWER: 071 (1.00) i REFERENCE:
" IG: 0-RO-0208-01, Primary Containment System," ELO 3 [3.6/3.6)
295020A101 . . (KA's) > T h A11SWER: 072 (1.00) , f i
>
?
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.- - -. - - - - _ .... _ --- - - - . _ . .-_- . . .- ~ - . . - . ,
!
ItEACTOR OPERATOR Page 91 REFERE!1CE: PilPS Proc. 110. 2.4.42, " Loss of RUCCW," pages 4 and . IGt 0-RO-02-02-06, "Roactor Building closed Cooling Water," ELos 2, 3, 8, and 1 . OJT Guido li , RBCCW System Objective 2 (3.3/3.4] ] 295018K201 .
.(KA's)
:
A11SWER: 073 (1.00) REFERE!!CE: IG: 0-RO-02-02-02, " Salt Service Water System," ELO (4 4/4.4]
,
295003A103 . .(KA's) ; ANSWER: 074 (1.00)
. t
_ _ _ . _
4
9'
'
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.__ _ ____ m._ _ _ _ .- _ . _ . _ _ _. _ . _ . . . _ _ . . ._ . _. _ _ _ _ _ . _ _ _ . ____ .. _ _ . - _ _ ,
REACTOR OPERATOR Page 92 l REFEREllCE
" Procedure PilPS Proc. 11 .3.4, ," page 1 l P11PS Proc. flo. 2.4.36, " Decreasing Condonsor Vacuum," page i IGt 0-RO-03-04-02, " EOP Development and Use," ELos 13 and 1 . IG: 0-RO-02-07-02, " Reactor Protection System and Anticipated Transient Without Scram System," ELO 1 . IG: 0-RO-02-05-01, " Main Turbino System," ELO 2 . Modified quantions from 11/26/90 and 12/2/91 11RC Exama (changed conditions).
[3.8/3.7) i 295002G010 ..(KA'a) !
.I
A11SWER: 075 (1.00) i
! REFERE!1CE: PilPS Proc. Ilo . 2.4.150, "Loso of Feedwater lleating," page . OJT Guide 110. 8, Feedwater lleating System Objective 2 r
-
[4.0/3.9)
295014G010 ..(KA's) , ANSWER: 076 (1.00) b
+ -
_ _ __ _ _ r--sy- .-m+y ,as-c ,ywq -- .
--
.. _ _ _ . . . ._ . . _ . _ . _ . _ _ - _ _ _ .____.___._ _ me__._. . _._ . . _ . . ..._. ... REACTOR OPERATOR Page 93 -I REFERENCE: -5 l- PHPS Proc. No. 2.4.4, " Loss of CRD Pumps." PHPS Proc. No. 2.4.11, " Control Rod Pooltionjng Malfunctions." PHPS Proc. No. 2.4.11.1, " CRD System Malfunctions." OJT Guido N , CRDM Syctom Objectivo 2a and CRD liydraulic System Objectivo 2 l (3.7/3.5) 295022G010 ..(KA'n)- ANSWER: 077 (1.00) l REFERENCE:
. ,
; PHPS Proc. No. 5.4.3, Rev. 10, Refueling Floor liigh Radiation,"
page . IG: 0-RO-06-04-01, " Fuel llandling Operations and Supervision," ELO [3.8/3.9) 295023G010 ..(KA's) ANSWER: 078 (1.00) r __ __ ._. _. _ _ _ _ _ ._ _ _ W*ry*w "-'"4 * ' ' '*'T' "* * * '* "I'
- '*N P* ' ' '
_ _ _ _ _ _ _ _ _.. _ _ ._.. _ _ _ _ _ _ _ _ _ _ _ _ .._ _-.- -. _ _ . _ .. - . _. _ . . . . _ _ _ _ . _ i
?
REACTOR OPERATOR Page 94 * i
,
REFERENCE: PNPS Proc. No. 5.3.6, " Loss of Vital AC (Y-2)." PNPS Proc. No. 5.3.30, " Loss of 2:3V DC Bus D-10." OJT Guido No. 1, 120/240 VAC System Objectivo 2b and 250 VDC System Objective ;
(3.5/3.5)
295004G005 ..(KA's) ; ANSWER: 079 (1.00) b J REFERENCE: PNPS Proc. No. 5.3.31, " Station Blackout," page . PHPS Proc. No. 5.3.26, "RPV Injection During Emergencies," page . OJT Guide N , 480/208 VAC System Objective 2 (4.4/4.4) * _ _ 295003A103 ..-(KA's) ANSWER: 080 (1.00) ; REFERENCE:
-1.- PNPS Proc. No.'2.4.143, Rev.=12,- " Shutdown-from--Outside Control'
Room," page 1 _ _ 2.- OJT-.Guido O-RO-04-04, " Emergency Tasks," Task 6 [4.1/4.1) 295016G006 ..(KA's)
.
.. y
- w ,. . r a n deewwee....wE-,.e,-+r-,-,-.5,c,.we~,--.w-_,..e~....m.,v---.,.i,,,,,-*-..y,-o--
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___ _ _ _ _ _ - - - . ._ _ .-- -- - _ - .
. . - ~ .
i REACTOR OPERATOR Page 95 i i
;
j AllSWER: 081 (1.00) l . l d.
- REFEREllCE:
(Dorived
'
1 UG: 0-RO-04-04, " Emergency Tasks," Tasks 1, 2, and ; from facility question PRO-1 from proposed roqual retake exam.)
! [4.3/4.4] i A202 ..(KA's) , I L . i AllSWER:. 082 (1.00)
'
l l ' c.
!
! ~ REFERENCE: l i- IG: 0-RO-02-07-01, "lieutron Monitoring Systems," ELOs 14 and 16.
' IG: 0-RO-03-04-04, "EOP-02, Failure to Scram,"-ELO 16.
! -
(4.2/4.3]
.
. 295037A201 ..(KA's) l ANSWER: _083 (1.00) i F' e t,
!
l.
l
'
..
- - -
. _ _ _ _ . _ _
_
.---_ --.:-.-. ...:. :- -- .- - -.- - -. ._ _.: >
REACTOR OPERATOR Page 96 REFERENCE: PNPS Proc. No. 5.3.23, Rev. 9, " Alternate Rod Insertion," pago . OJTPG: 0-RO-04-04, " Emergency Tasks," Task _ [4.0/4.1) 295015K204 . ..
(KA's)
ANSWER: 084 (1.00) REFERE!1CE:
" Alternato Rod Insertion," pages 6 - PHPS Proc. No. 5.3.23, Re ,
1 OJTPG: 0-RO-04-04, " Emergency Tasks," Task . Modified question from 12/2/91 NRC Exam (modified distractors).
(3.8/3.9) 295015A101 . . (KA's)
.
At1SWER: 085 (1.00) ] i , REFERENCE: i IG: 0-RO-03-04-05, Primary Containment Control,", page IG-5, ELO
- 1 . OJTPG: - 0-RO-04-04,- " Emergency _ Tasks,"-Task--89 . Modified question from 12/2/91 NRC-Exam (replaced -1- distractor) . .g (3.6/4.3)
,
295012G012 . . - (KA's)
!
l: _ 9- pt= ygap= -genyAT ,mMi-W r jagaem as- w*,e e<y,m.ee-irm=- gy, wm-r.-y--r-*- p-,w- - .--w-esw.m,m,.ee,e-g#,u-me,--am,,.w-wwsws.w_ep-e.-= w un wwm-1'*aww- m
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-
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. _ _ . . _ .. _.~ . _ _ _ _ ___ _ . _ _ _ _ _ _ _ _ . ... _ ___ _ .__. -, _ .. _ _ _ _ _ _ _ _ _ ._ _ __ _ . _ _ _ .
- REACTOR OPERATOR Page 97
.
,
AllSWER: 086 (1.00) , l REFEREllCE: IG: 0-RO-03-04-05, Primary Containment Control,", page IG-1 . OJTPG: 0-RO-04-04, " Emergency Tacks," Task 89 [3.8/4.5) 295026G012 . . (KA's) A11SWER: 087 (1.00) REFEREliCE:
" Primary IG: 0-RO-02-08-01, Containment System," ELO 39 (3.6/3.9)
295024G007 . . (KA's) e A11SWER: 088 (1.00) C.--
. _ -
l l
.-. , _ _ .. . _-
up -- v 1 ,m.,- r we ---'ir-New -'rbe6-'- W- m --yev'* 'r+=-'4 m' a'h."vt-v- W- -- -9m Y y eilip-- ev - '-- - g'- r wie my&-geemetwg
. .. _ _ _ _ _ _ ___ . - . . . .. _ .. _ ..-_____.__..__.____.m_ ___._. .
'
REACTOR OPERATOR Page 98-REFERE! ICE: , i PHP5 Proc. No. 5.3.27, " Determining Primary Containment Water ; Level." OJT Guido No. 10, " Primary Containment System Structure Objective ' 2 '
' IG: 0-RO-02-08-01, " Primary Containment System," ELOs 13 and 1 (3.4/3.5)
295029A203 ..(KA's) ANSWER: 089 (1.00) c REFERENCE: , PHPS Proc. No. S.3.26, " RPV Injection During Emergencies." Facility Question 5.3.26-1 (Proposed Requal Retake Exam).
[4.1/3.9) 295031G006 ..(KA's) ANSWER: 090 (3.00) ' REFERENCE: PNPS Proc. No. 2.4.46, " Turbine Bearing Malfunction "-page . ,
- OJT Guide No.-7, Turbine Supervisory Instrumentation Objective 2 [3.8/3.6)
295005G010 ..(KA's) l'
.
. _
e- - --+yw .Wgi,e.'tTT- r T W--P mm W D rfr 4-T rg
. _ . _ . . . . _ _ _ . _ _ - _ . . _ . . . _ . . _ . - _ _ _ . . _ . _ - - - _ _. _ . . ~ . . _. _ . _ _ _ , _ . . - _ - _ _ .
REACTOR OPERATOR Page 99
,
' ANSWER: 091 (1.00) REFERENCE: PNPS Proc. No. 5.3.3, " Loss of All Service Water," page . IG: 0-RO-02-02-02, " Salt Service Water System," ELO . OJT Guido No. 2, Service Water System Objective 2 [J.4/3.3) 295018G010 .
.(KA's)
ANSWER: 092 (1.00) ' REFERENCE: PNPS Proc. No. 2.4.23, " Jet Pump Flow-Failure," page , IG: 0-RO-02-06-01, "Non-Nuclear Instrumentation and Reactor Vessel Internals," ELOa 11 and 1 (3.3/3.6) 295001G007 . .(KA's) ANSWER: 093 (1.00) b.
l-l 1.
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- - - _ _ _ - _
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. _ . _ . . . _ _ _ . _ _ . . _ . _ _ _ _ _ _ . _ - . . _ . - _ . _ _ - _ . . . _ _ . _ _ _ _ . _ _ _ _ . _ _ .
-
REACTOR OPERATOR Page100
,
i REFERE!1CE: < IG: 0-RO-03-04-03, "EOP-01, RPV Control," ELO i l L
(4.4/4.4)
295025A103 ..(KA's)
.
A!1SWER: 094 (1.00) b.
REFEREllCE: , IG: 0-RO-03-04-04, "EOP-ON, Failure to Scram," ELO 1 [4.1/4.2) 295037K301 ..(KA's) . AliSWER : . 095 (1.00) d.
REFERENCE: IG: .O-RO-03-04-04, "EOP-02, Failure to Scram," pages IG-33 and'IG-41, ELO (3.9/4.6)
'
-295037G012 ..(KA's)
..
-..r,ew w a + fm' Te% -< ='-w * em 'ee '- eem*e*W4 #sp-we m, e e T- --l'w-*A' "'r - --
. - ~ . _ . - . _ _ ~ - ~ . . - - . . - . - . - . _ - . . _ - - - - - - .-. . ~ . - . .--. . . - . . -
I HEACTOR OPERATOR Page101
'
s ANSWER: 096 (1.00) i
. REFEREllCE: IG: 0-RO-02-09-06, " Diesel Generator System " ELos 9 and 2 ,
[3.7/3.7)
295024A106 . . (KA's) A11SWER: 097 (1.00) REFERENCE: IG: 0-RO-03-04-06, "EOP-04, Secondary Containment Control," ELO . IG: 0-RO-02-08-01, " Primary Containment System," ELO 1 [3.9/4.2)
,
'
295035G011 . . (KA's) T ANSWER: 098 (1.00) . _ _
-- t .9p -g_-- -w--g .y. tv ,w._-,,e ,e.,, __..p. , 9 y gpys-y #-4yym-- *1rt'---t**pmM+v8Tc1-*' e.ee-M.wa,w w %- g ey pqy ------
_. .._.__.-.._ --_._ _ _____. _ _ .__.._ -_____ _ ___ _. _ ._. _. .._ . _ _ _ . . . . . . . . . _ _ _ _ _ . REACTOR OPERATOR Page102 REFERENCE: IG: 0-RO-03-04-06, "EOP-04, Secondary Containment Control," ELO j l i (4.2/4.3) l t i 295034G011 ..(KA's) , ANSWER: 099 (1.00) REFERENCE: IG: 0-RO-03-04-06, "EOP-04, Secondary Containment Control," ELO (3.6/4.4) 295036G012 ..(KA's) ' ANSWER: 100 (1.00) REFERENCE: IG: 0-RO-02-03-02,- " Process Radiation Monitoring System,_" ELos-5 and 1 (3.6/3.8) , 295038K202 ..(KA's) l ! l (********** END OF EXAMINATION **********)
. - __ -_ __
_ _ _ l -- i
- - . _ . _ . . - . ~ = . - - . - . . - _ - . . - . - . . - - . - - .- . - - _ ,- - .. . - . - - _ . .--
l f REACTOR OPERATOR Page 1 [ A ti S W E R KEY :
MULTIPLE C1101CE 023 b , 001 p MN 024 c 002 b 025 c ' 003 gb 026 c 004 a 027 d 005 b 028 c 006 b 029 c 007 c 030 d , t 008' c 031 c 009 fr' ,l.dd<d 032 c ; 010 a 033 a 011 d orb 034 d 012 c 035 b 013 d 036 b 014 d 037 c-015 d '038 a 0" d 016 b 039 d
'
l 017 a 040 c 018 d 041 c 019 b 042 d' ,. 020 b 043 c l 021 c 044 c 022 a '045 d r !.
- _ _ . . . . . . _ , _ _ . . - ., .. . . - - - - ,
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REACTOR OPERATOR Page 2
..
)
ANSWER KEY - l 046 b 069 d , 047 b 070 d 048 , h' M 071 b !
>
049 c 072 c 050 a 073 a
'
051 b 074 d 052 b 075 d , 053 b 076 c 054 a 077 c 055 c 078 b 056 d 079 b 057 b 080 c 058 c 081 d 059 b 082 c 060 c 083 'c 061' c 084 d-062 d 085 d 063 a' 086 a-064 a 087 'b 065 d 088 c
~ 066 c- 089 c ' 067 -
a - 090 a-068 b 091 d
, I
_ _ _ _ _ _ _.. _ r wg>-T-t-r-'- 9g-is. vue speMw- W% .c .-nwig e ge asaw- WwA-o-g., g,pe, .m e --g --F w y er a ug g-.mywg 7- ,i--+twprey-M w- h k7'W'-f rf*TV F 'T'N""T PVNWE M *T' TF VY W T*'W'** wor' T Y'T77 7~ TTTT FT-'**'-"
REACTOR OPERATOlt Page 3 A 11 S W E R KEY 092 b 093 b 094 b 095 d 096 a 097 c 098 c 099 c 100 b (********** E!1D OF EXAMIllATIO!1 **********)
. - _ _ _ . _ -- .-. ______-_-_---__ _ __-_ ___________ _ ____ _ __ _ _ _ -
_ _ _ _ _ - -__ . . . . . . . . TEST CROSS REFEREllCE Page 1 RO Exam 13 W R Reactor Organized by Quostion 11 u m b o r QUESTIOli VALUE REFEREllCE 001 1.00 9000629- Md 002 1.00 9000630 003 1.00 9000631 004 1.00 9000632 005 1.00 9000633 006 1.00 9000634 007 1.00 9000635 008 1.00 9000636 909 1.00 ^ O00644-- Mlob 010 1.00 9000638 011 1.00 9000639 012 1.00 9000640 013 1.00 9000641 014 1.00 9000651 015 1.00 9000652 016 1.00 9000G53 017 1.00 9000654 018 1.00 9000655 019 1.00 9000656 020 1.00 9000657 021 1.00 9000658 022 1.00 9000659 023 1.00 9000660 024 1.00 9000661 025 1.00 9000662 026 1.00 9000663 027 1.00 9000664 028 1.00 9000665 029 1.00 9000666 030 1.00 9000667 031 1.00 9000669 032 1.00 9000670 033 1.00 9000671 034 1.00 9000672 035 1.00 9000673 036 1.00 9000674 037 1.00 9000675 038 1.00 9000676 039 1.00 9000677 040 1.00 9000678 041 1.00 9000679 042 1.00 9000680 043 1.00 9000681 044 1.00 9000682 045 1.00 9000683 046 1.00 9000684 047 1.00 9000685 448 1. OG- 900n6a4. 4 W 049 1.00 9000687
_ _ _ . __ _.._. _ .. _
-
_ _ . _ _ _ . . _ _ _ _ _ _ . _ . _ . _ _ . _ , _ _ _ _ _ _ _ _ . - _ _ _ _ _ i
!
TEST CROSS REFERENCE Pago 2 RO Exam BWR Roactor organizod by Quootion H u m b o r- ) i P QUESTION VALUE REFERENCE ! 050 1.00 9000688 , 051 1.00 9000689 052 1.00 9000690 053 1.00 9000691 l 054 1.00 9000692 -1 055 1.00 9000693 056 1.00 9000694 057 1.00 9000695 I 058 1.00 9000696 059 1.00 9000697 1 060 1.00 9000698 , 061 1.00 9000699 062 1.00 9000700 063 1.00 9000701 064 1.00 9000702 l 065 1.00- 9000712 # 066 1.00 9000713 067 1.00 9000714 i 068 1.00 9000715 069 1.00 9000716 ; 070 1.00 9000717 ; 071 1.00 9000718 Ei
'
072 1.00 9000719
-'
073 1.00- 9000720 074 1.00 9000721 075 1.00 9000722 076 1.00 9000723 077 1.00 9000724 078 1.00 9000725 079 1.00 9000726 080- 1.00 9000727 -l 081 1.00 9000728-082 1.00 9000729 083 1.00 9000730
'
084 1.00 9000731-. 085 1.00 -9000732 086 1.00 9000733 - 087 1.00 9000734 088 1.00 9000735 089- 1.00 9000736
- 090- 1.00 9000737 091 1.00- 9000738 I 092- 1.00 9000739 (
093 1.00 9000740 094 1.00 9000741 b _ 095 1.00 9000742 a 096 1 00 -9000743 097 1.00 9000744 098 1.00 9000745
-
-:
'
.._:-- ..,s.,,. ,ma, ,,,a._.,.--__ .s.,--,~.w .~._,,._,-,,,_.~,,=,m=.,,--,,,..-.,..~ . - - - - . . , , - . , ,
. _ _ _ . ~ . . . _ _ _ _ _ _ . . . . _ . - . . . . . . . _ . _ _ _ _ _ . . . . _ , _ . _ - _ . _ .
_ . _ _ _ . _ . _ _ _ _ . - _ _ _ _ _ _ _ _ _i
-
l i I l TEST CROSS REFEREllCE Pago 3 ; i i l RO Exaa BWR Roactor , Organ 1 zed by Quoation 11 u m b o r l
i QUESTION VALUE REFEREllCE i 099 1.00 9000746 100 1.00 9000747 { ______
,
ecc.oc=97 0 ______ mWW6mW 10e.00 *i .
.
A L ..._b'_ _ __ _ - .-
, ..L-,_- L'.,_ ,.,_.4m _,_ ,;~_ ,,.,m.ymm,,_,,,,,__,, ,,.,._,,.,,m,.,,.,,_,._.,_,__, , , . _ _ _ . , , , , _ _ ,
._ _ _ _ . _ _ _ . . _ . . _ _ - . _ _ . _ . _ . _ _ _ - - _ _. _ ___- _ _ _ _ . TEST CROSS REFERE!1CE Page 4
.
R0 Exam BWH Reactor organized by KA Group PLAllT WIDE GEllERICS QUESTIOli VALUE RA 011 1.00 294001A102 008 1.00 294001A103 010 1.00 294001A106
*04 b- 1.vu 2 9 4 0 V itu 15 012 1.00 294001A110 013 1.00 294001A116 006 1.00 294001K101 007 1.00 294001K102 003 1.00 294001K103 002 1.00 294001K103 GO':
005 1.00 1.00
'9A001K10'7 294001K109
@
004 1.00 294001K110 ______ PWG Total 13.00= H.ca PLA!1T SYSTEMS Group I QUESTIOli VALUE KA 046 1.00 202002K604 026 1.00 203000A101 064 1.00 206000A104 024 1.00 206000G010 062 1.00 209001A104 025 1.00 209001A205 060 1.00 211000A301 019 1.00 212000G010 020 1.00 212000K412 032 1.00 215003A202 055 1.00 215003A403 056 1.00 -215004G001 031 1.00 215004G011 034 1.00 215005A107 035 1.00 215005A207 010 1.00 216000A210 017 1.00 216000K122 023 1.00 217000A215 063 1.00 217000A403 029 1.00 218000A206 014 1.00 223001G011 022 1.00 223002A302 030 1.00 239002K405 042 1.00 241000K306
. . _ - - . . --
TEST CROSS REFEREllCE Page 5 RO Exam BWR Reactor Organized by KA Group PLANT SYSTEliS Group I QUESTIOli VALUE KA 039 1.00 259001A204 041 1.00 259002K604 G40 -1 .-0 0 2MOOOA343-- M W 047 1.00 264000K408 ______ PS-I Tota 1 MrOO-- 17.do Group II QUESTIO!1 VALUE KA 038 1.00 201006K403 045 1.00 202001K410 016 1.00 204000G007 061 1.00 205000A402 037 1.00 214000A201 036 1.00 215002A202 027 1.00 226001G010 028 1.00 230000A406 021 1.00 239001K127 043 1.00 245000A304 059 1.00 245000A402 - 040 1.00 256000K304 058 1.00 262001A302 057 1.00 262002K103 044 1.00 271000A206 049 1.00 272000K403 054 1.00 286000G004 050 1.00 290001G010 051 1.00 290003A204 ______ PS-II Total 19.00 Group III QUESTIOli VALUE KA 033 1.00 215001A207 052 1.00 233000K102 053 1.00 234000K402 015 1.00 239003G008 ______ PS-III Total 4.00 ______ ______ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ . _ _ _m
.. . . .___ _ _ . ._ -m ... _ _ . _ . _ _ _ _ . _ . _ _ _ _ _ _ _ . . _ . _.._. ___
TEST CROSS REFEREllCE Page 6 i RO E :: am BWR Roactor , Or9anizod by KA Group
->
PLANT SYSTEMS t I , QUESTIOli VALUE KA . PS Total - MM 90.o0 ,
;
E!4ERGEliCY PLAliT EVOLUTIONS Group I
.
QUESTIOli VALUE KA 090 1.00 29S005G010 081 1.00 295006A202 075 1.00 295014G010 ' 084 1.00 295015A301 083 1.00 295015K204 096 1.00 295024A106 087 1.00 295024G007 093 1.00 295025A103 295025G003
'
066 1.00 089 1.00 295031G006-082 1.00 295037A201-095 1.00 295037G012 094 -1.00 . 295037K301
......
;
EPE-I Total 13.00 Group II QUESTION VALUE KA 068 1.00 295001A201 ' 092 1.00 - 295001G007 067 1.00 295001G011 . 074 1.00 295002G010 079 1.00 295003A103 073 1.00 295003A103 078 1.00 295004G005 085 1.00 295012G012 080 1.00 295016G006 ' 0911 1.00 ' 295018G010 072 1.00 295018K201
. 069 1.00 295019A202 071- 1.00 295020A101 070 1.00 295020K201-076 1.00- 295022G010 086 1.00- 295026G012 088 1.00 295029A203 098 1 00 295034G011
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TEST CROSS REFERENCE Page 7 RO Exam BWR Reactor organized by KA Group
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EMERGENCY PIANT EVOI,UTIONS ,
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Group II QUESTION VALUE KA l 100 1.00 295038K202 ______ EPE-II Total 19.00 Group III , QUESTION VALUE KA 077 1.00 295023G010 097 1.00 295035G011 065 1.00 295035K101 099 1.00 295036G012 ______ EPE-III Total 4.00 ______ ______ EPE Total 36.00 ______ ______ ______ Test Total +00T00* 47. co ., 4 e -g,m,--m _--i_o-,y,w- f.ww,-T -mer--.-r- w w t-e p-7m--yrmwm * u*wy 9-r%w-+ge-----4ts-- -f:dy r wsycr-py 3- -4'T r a y-W v m-r.w---$7e g wep m-g a e t1BV "4. t &cT+ g g
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+ - - 1 A x: .u 4 , _m, BOSTON EDISON PILGRIH NUCLEAR P0HER STATION Procedure No. 2.4.17 RECIRCULATION PUHP(S) TRIP
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INFORMATION ONLY Use restricted to reference REVIEHERS AND APPROVERS 0 hk 0 Procedufs Hriter c kf Date VA x 32 2
/ / / Technical Reviewer Ohte 4 ?// 0 s A A-Validator Date Wm w A. $,too>w."* sb Hz Procedure Owner Date tu/W OAD Manacer Date Y 3/Y/f SAFETY REVIEH REQUIRED I/ ORC (hairman Date ORC REVIEH REQUIRED E #dd- 3 PMnt Manaaer Date Effective Date: S4-%
2.4.17 Rev. 14
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. 1.0 SYMPTOHS ALARMS ANNUNCIATOE EAEf1 HINDOH
[1] RECIRC. HG SET GENERATOR DIFF. OVERCURRENT 904C G4 904R A3 [2] RECIRC. HG SET DRIVE HOTOR TRIP 904C B4 904R E2 [3] RECIRC. HG SET DRIVE HOTOR OVERLOAD 904C C4 904R F2 [4] RECIRC. HG SET GENERATOR LOCK 0UT 904C E4 904R H2 [5] RECIRC. PUHP LOCKED ROTOR TRIP 904C A4 904R D2 [6] RECIRC. HG SET LUBE OIL LOH PRESS 904C B3 904R E1 [7] RECIRC. HG SET FLUID DRIVE HI OIL TEM C 03 904R G1 PLANT INDICATIONS [1] Reduction in recirculation flo [2] Sudden decrease in reactor powe _
2.0 AUTOMATIC ACTIONS None 3.0 IMMEDIATE OPERATOR ACTIONS
[1] HONITOR alarms and instrumentation &MQ DETERMINE the type of system malfunction that has occurre [2] LE both recirculation pumps trip Ilim RANUALLY SCRAM the reactor &EQ CONCURRENTLY PERFORM PHPS 2.1.6, " Reactor Scram", with this Procedur .4.17 Rev. 14 Page 2 of 9 .. .. .. . ._ _ - _ _ _ _ _ _ _ - _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _
. . __ _ - . - 4.0 $11HSIQU14T OPERATOR ACTION $'
[1] TURN to the page of this Procedure for the malfunction that has occurred 6HD PERFORM the inoicated steps:
Eigt Section (a) Trip of one recirculation pump 3 (b) Trip of both recirculation pumps 6 .1 TRIP Of ONE RECIRCULATION PUHP lQ The reactor shall not be operated with one recirculation loop out of service for more than 24 hours. With the reactor operating, if one recirculation loop is out of service, the Plant shall be placed in a Hot Shutdown condition within 24 hours unless the loop is sooner returned to servic ............................................................................... CAUTION If power level is less than 30%, stratification may occur; refer to PNPS 2.4.24, " Reactor Vessel Cold Hater Stratification".
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[1] CLOSE affected HO-202-5A or B, PUMP DISCH VL (a) HUEH 5 minutes have elapsed IHEN REOPEN the discharge valv [2] CHECK speed on the in-service pump to ensure it has not increase .4.17 Rev. 14 Page 3 of 9
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. TRIP OF ONE RECIRCULATION PUHp (Continued)
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[3] DETERMINE Total Core Flow (TCF) -[NRC Inspection Report 91-25]-
(a) DETERMINE direction of flow through idle jet pump (1) ADD in-service and idle jet pump loop flow rete:
+ =
In-Service Idle Sumed FI-263-107A(B) FI-263-107A(B) Value (2) MJLTIPLY idle jet pump loop flow by 0.95 MQ SUBTRACT-this multiple from in-service jet pump loop flow:
- [0.95 X _ ] -
In-Service Idle Subtracted FI-263-107A(B) FI-263-107A(B) Value (3) USE current reactor power MQ PLOT both of_ the - calculated flow values on the Power-To-Flow Ha HQIf The change in reactor power due to the recirc pump trip will-result: ' in a Xenon transient. This transient will cause the previous load line to lower. The amount the load line shifts is dependent on the time after the retirc pump trip and previous equilibrium reactor power conditions.
'-
(4) E the Subtracted Value-falls to the left of the expected load line (i.e., on or to-the left of the minimum pump speed line) MQ theLSumed Value- falls -
approximately on the-expected Load Line IHER Forward-Flow exists through the_ idle jet; pump' loo (5) - E the Subtracted Value-falls approximately on-.th expected load line MQ the Sumed Value falls- below and-to the right, IHEM Reverse Flow-exists through the idle; jet pump loo .
(b) CALCULATE Total Core Flow (TCF) (1) E Forward Flow through the idle jet pump loop exists, THEN Total Core Flow equals-the Sumed Valu (2) E Reverse Flow through the idle jet pump loop exists,
, IEEN Total Core Flow equals the Subtracted Valu .4.17- Rev.-14-Page-4 of 9- - : _
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4.1 TRIP OF ONE RECIRCULATION PUMP (Continued)
. [4] If the reactor is operating AT OR ABOVE the 80% load line &MQ total core flow decreases below 31.5 Hib/hr, IREM PERFORM the following steps: [IEB 88-07, SUPP 1: BHROG 8879) (a) HONITOR the APRMs and LPRHs for neutron flux instability oscillation (b) INCREASE speed of the operating recirculation pump until any flux instability ceases and core flow is greater than 31.5 Hib/h (c) INSERT control rods to decrease reactor power below the 80%
load lin (d) LE APRM oscillations of greater than 10% peak-to-peak QB periodic LPRM upscale or downscale alarms are observed, IHEN SCRAM the reactor 6HQ CONCURRENTLY PERFORM PNPS 2.1.6 with this Procedur [5] &EIER the recirculation pump is secured, ADJUST total core flow to greater than 27.6 Hlb /hr. [GE SIL 517]
(a) VERIFY that the recirculation loop flow of the active loop on FI-107A or B is less than 36.9 Hib/hr. [GE SIL 517) [6] SEND an operator to the 4kV breaker and to the MG Set Room to record all relay targets to determine cause of tri [7] ENSURE that power is available as follows: (a) Instrument power Panel Yi (b) Vital service Panel Y2 (c) 4160V load centers A3 or A4 (d) Power Centers B17, B18, B20 and 09 (e) Power Panels 04, D5 and 06 l [8] IE the cause of the trip can be determined and corrected &EQ the l reactor is operating outside of the area of the power flow map bounded by the 80% load line and 31.5 Hib/hr, IHEN the pump may be restarted in accordance with PNPS 2.2.84, " Reactor Recirculation System". [IEB 88-70, Supp 1; BHROG 8879) [9] REFER to EPIP-100, " Emergency Classification", to determine whether an Emergency Action Level (EAL) has been exceede .4.17 Rev. 14 Page 5 of 9
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4.2 TRIP OF BOTH RECIRCULATION PUMPS
[1] CLOSE both H0-202-5A and B. PUMP DISCH VLV (a) BdEM 5 minutes have elapsed, IHIH REOPEN the discharge valve [2] SEND an operator to the 4kV breaker and to the HG Set Room to record all relay targets to determine cause of tri , [3] ENSURE that power is available as follows: (a) Instrument power Panel Y1 (b) Vital service Panel Y2 (c) 4160V load centers A3 or A4 (d) Power Centers B17, B18, B20 and 09 (e) Power Panels D4, 05 and D6 [4] IE the cause of the pump trips can be identified and corrected, IHER RESTART the pumps in accordance with PNPS 2.2.84, " Reactor Recirculation System".
[5] REFER to PNPS EPIP-100, " Emergency Classification", to determine whether an Emergency Action Level (EAL) has been exceeded.
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f 2.4.17 Rev. 14 l Page 6 of 9 !
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3' - BOSTON EDISON PILGRIH HUCLEAR POWER STATION Procedure No. 5.3.27 DETERMINING PRIMARY CONTAINMENT HATER LEVEL -
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INFORMATION ONLY-Use restricted to reference Approved M ku / - If fA A b/ /# , Plant Hartiger (.) = b Date 5.3.27 Rev. 3 Page 1 of-8
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: "* I .
u; PURFOSE
&}I (Tr - This procedure provides instructions for determining Primary i
.
* Containment Hater Level when flooding of the Primary Containment 'is
, directed by the Emergency Operating Procedures'(EOPs). .
* Primary Containment (PC) Hater Level values are referenced to plant elevatio .0 ACTIONS FOR CONTAINHENT LEVEL LESS THAN 47 FEET [1] CALCULATE the existing differential pressure (sensed) between the drywell air space and the bottom of_ the torus, as follows (Figure 1):
TORUS BOTTOM PRESS (PI-1001-69)
(Panel C903) 0s19-minus: DRYHELL HIDE RANGE PRESSURE (PI-1001-600A/B) (Panel C170/171) osig equals: CONTAINHENT TO TORUS BOTT0H dP esig (2) For the calculated CONTAINHENT TO TORUS BOTT0H dP, the corresponding Primary Containment Hater Level is obtained from the Primary Containment Water Level curve (Figure-2).
. FOR CONTAINHENT LEVEL GREATER THAN OR EQUAL TO 47 FEET
[1] READ containment water level on DRYHELL LEVEL Gauge LI-5008 on r Panel C90 '
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e i 5.3.27 Rev. 3' Page 2 of 8 >
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CONTAINHENT TO TORUS BOTTON dP CALCULATION TORUS BOTT0H PRESS (PI-1001-69)
(Panel C903) esig ~
ORYWELL HIDE RANGE PRESSURE (PI-1001-600A/B)
(Panel C170/171) 0519 - CONTAINHENT TO TORUS BOTT0H dP esig FIGURE 1 .
l i i ! 5.3.27 Rev. 3 l Page 3 of 8 ! , I a's ., . a ,8 ..'ga t g **,e *
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A b h m \- 3 U. S. NUCLEAR REGULATORY COMMISSIO SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 1 CANDIDATE'S NAME: FACILITY: Pilgrim 1 REACTOR TYPE: DWR-GE3 DATE ADMINISTERED: 92/11/16 INSTRUCTIONS TO CANDIDATE: Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheet Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE % _ 4 G.0&- % TOTALS FINAL GRADE All work done on this examination is my ow I have neither given nor received ai Candidate's Signature
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SENIOR REACTOR OPERATOR Page 7 - -QUESTION: 001 (1.00) WHICH ONE of the following On D nd (OD) Programs would be used to obtain the effective readings rom an LPRM string used in the most recent P1 power distributio calculation? a. OD-6, Thermal ta in a Specified Bundle b. OD-8, Pres t LPRM Readings c. OD-9, ial Interpolation in a Specified LPRM String d. OD- , Periodic Core Performance Logs QUESTION: 002 (1.00) Conditions in a recently surveyed area are: 25 mR/hr general area radiation 100 dpm/100 cm2 alpha loose surface 500 dpm/100 cm2 beta-gamma loose surface 0.20 MPC airborne beta-gamma radioactivity WHICH ONE of the following describes the complete posting requirements for the area? " CAUTION RADIATION AREA" b. " CAUTION RADIATION AREA" and " CAUTION CONTAMINATED-AREA" " CAUTION RADIATION AREA" and " CAUTION _ AIRBORNE RADIOACTIVITY AREA"
" CAUTION RADIATION AREA," " CAUTION CONTAMINATED AREA," and ~ " CAUTION AIRBORNE RADIOACTIVITY AREA"
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. ,- .. . . . . . . . - . . - - . --. . . .. = SENIOR' REACTOR-OPERATOR Parin 8 QUEST 10N:.003 (1.00)
A-23 year old. radiation worker with a current Form NRC-4 needs to-- perform work in an area with general radiation levels of-75 mR/h The worker's exposure history is: Lifetime: 24.5 Rem Current year: 1400 mR Current quarter: 225 mR WHICH ONE of the following is the maximum time that the worker can stay in the area without exceeding any PNPS Exposure Control Levels? (Assume no special authorization.) minutes b. 1 hour and 20 minutes hours and 40 minutes hours QUESTION: 004 (1.00) WHICH ONE of the following describes proper procedures for handling sodium pentaborate? , Respiratory protection is required to prevent inhaling 1 dust containing boro Protective clothing is not required because sodium pentaborate cannot be absorbed through the ski Report to the Chemistry Department _immediately after leaving the worksite to dispose of protective clothing and foot coverings, Direct any liquid spillage to a floor drain, wipe'and dry mop , the floor, and-discard mops and wipes after us _ . . _ _ _ _ _ _ - - . . . _ . . _ _ . ._ _ _ _ _ _ .. _ _ _. _ _ . . . .__ _ __ SENIOR-REACTOR OPERATOR- Page 9-QUESTION: 005 (1.00) WHICH ONE of the following situations is acceptable in accordance with PNPS Proc. No. 1.4.36, "High Pressure / Compressed Gas Cylinder Control?"- a. Argon cylinders on a cart are staged for a non-active Maintenance Request that has been rescheduled for later in the week. The cart is secured to a fixed suppor Reserve nitrogen cylinders are secured at approximately 3/4 * height to a permanently installed cylinder-holding station ~. c. Acetylene cylinders on a cart are staged in the cable Spreading Room for an active Maintenance Request. Work is expected to start next shif The cart is secured to a fixed suppor Empty oxygen cylinders are secured at approximately 3/4 height to a fixed support during shift turnover. The cylinders will be removed next shif QUESTION: 006 (1.00) WHICH ONE of the following situations would require independent verification in accordance with PNPS Proc. No. 1.3.34, " Conduct o Operations?" (All components are safety related.) Fuses are pulled for maintenance in a cabinet in the Control Room.
' An existing tagout is used to verify the position of an-inaccessible component, Verification of a component's position is expected to take 15-minute Radiation levels in the area are 120 mr/h d. A large manually operated valve that requires two people to operate must be shut for maintenance.
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SENIOR REACTOR OPERATOR- Page 10
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. QUESTION: 007 (1.00) Breaker 103 (Bus Al feeder from Unit Auxiliary Transformer) is open and has a white tag with a green border hanging on the control switc WHICH ONE of the following describes the meaning of the tag? a. A grounding device is installed on the breaker. The breaker may only be operated under the authority of the person for whom the tag was placed, b. The breaker is not in its normal operable status. :The breaker may be operated by qualified operators when the precautions listed on the tag are followe c. The breaker requires testing prior to restoring it to normal service. The breaker may be operated only undar the authority of the person for whom the tag was placed, d. The breaker is open to piotect personnel from injury or equipment from damage. The breaker may not be operated until the tag is cleared.
' QUESTION: 008 (1.00)
'
WHICH ONE of the following accurately describes the NRC Overtime Guidelines? a. Operators may-not work more than: 12 hours in 24 hours; 24 hours-in 48 hours; or 84 hours in 7 days, b. Operators may not work more than: 16 hours in 24 hours; 28 hours-in 48 hours; or 84. hours in 7 day c. Operators may not work more than: 16 hours in 24 hours; 24 hours in 48 hours; or 72 hours in 7 day , d. Operators may not work more than: 12 hours in 24 hours; 28 hours in 48 hours; or 72 hours in 7 day .
. _ _ _ > _ . _ _ _ . . - -_ . _ _ . ___ . _ - - . _ _ _ . - _ . . _ > SENIOR-REACTOR OPERATOR- Page 11 QUESTION: 009 (1.00)
An initial entry is being made into Primary Containment in accordance with PNPS Proc. No. 1.4.12, " Primary Containment Entry." A nitrogen pocket causes-the oxygen analyzer drywell entrance light to illuminat WHICH ONE of the following describes the entry team's awareness of this situation? a. They would not be aware that the light had' illuminate b. The alarm at the drywell entrance is relayed to a siren in the drywell which would alert the tea c. The backup team stationed outside the drywell entrance would alert the entry team of the condition by the plant paging syste d. H.P. stationed outside the drywell entrance would alert the entry team of the condition by walkie talki QUESTION: 010 (1.00) The plant is operating at 100% with the Technical' Specification Minimum Operating Shift Crew Composition on duty. The STA is filling a licensed SRO position. A fire has been reported in the turbine building. WHICH ONE 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 and the STA b. 1 unlicensed operator, 1 licensed SRO, and the STA l c. 1 unlicensed operator and 1 licensed RO j- -d. 2 unlicensed operators w -
SENIOR REACTOR OPERATOR . Pago 121
' QUESTION: 011 (1.00)
The Control Room has been evacuated due to a fire in the Cable Spreading Room.- There is a shortage of licensed operators. WHICH ONE of the following systems can be operated by a non-licensed operator from the Alternate Shutdown Panel? HPCI b. RBCCW c. ADS ' RHR QUESTION: 012 (1.00) WHICH ONE of the following situations would require an entry into the-Lifted Lead / Jumper (LLJ) Log? a. A jumper is installed and leads are lifted in an emergency condition in accordance with PNPS Proc. No. 5.3.21, " Bypassing Selected Interlocks." The emergency condition is expected to last longer than the current NWE working shif b. Motor leads are lifted for motor replacement. The leads fall within the Maintenance Request (MR) tagout boundary and are identified with colored tape. The MR-provides for relanding the leads prior to clearing the isolation boundary. The motor-replacement is expected to last longer than-the current NWE working shif c. A jumper _is installed for a surveillance test. Installation and removal of the jumper are signoff steps in the procedure. .The jumper will be removed prior to the end of the current NWE working shif d. A gagging device is installed'for maintenance'in accordance with an approved-station procedure. The gagging device is~within_the-Maintenance Request (MR) tagout boundary and the MR'provides for removing the gagging-device prior to clearing the isolation boundary. The gagging device-will not be removed prior to the end of the current NWE working shif . - .. .
SENIOR REACTOR OPERATOR Page 13 QUESTION: 013 (1.00) The plant is at 100% power with IRM G and APRM B inoperable. An I&C supervisor reports, after reviewing surveillance data, that the power supplies for APRM D and IRM A need replacement and the instruments are technically inoperable. WHICH ONE of the following deceribes the entries that should be made in the Limiting condition for Operation (LCO) Log? A tracking LCO is entered for APRM D and a tracking LCO is entered for IRM b. A tracking LCO is entered for APRM D and an active LCO is entered for IRM _ An active LCO is entered for APRM D and a tracking LCO is entered for IRM An active LCO is entered for APRM D and an active LCO is entered for IRM QUESTION: 014 (1.00) WHICH ONE of the following conventions for annotating flowcharts is preferred when executing the EOPs? Place an 'X' through any entry conditions that have been me When a step is completed, place an 'X' through the step and draw a line to the next ste While performing steps in another EOP or another flow chart branch, draw arrows to indicate the stopping point in the current branc Periodically update values of the parameters controlled by the EOP next to the step that is being performed when the update is provide . - _ _ .
SENIOR REACTOR OPERATOR Page 14 QUESTION: 015 (1.00) WHICH ONE of the following describes the appropriate action to be taken when an Alarm Statement is reached when implementing the EOPs? Review the overrides of all flowpaths being performe Take the action specified in any applicable overrides, Review the overrides of the flowpath containing the Alarm Statement that was reached. Take the action specified in any applicable override Exit the flowpath being executed when the Alarm Statement was reached and enter the contingency procedure applicable to the Alarm Statemen Exit all the flowpaths containing the Alarm Statement that was reached and enter the contingency procedure applicable to the Alarm Statemen QUESTION: 016 (1.00) A double ended shear of Recirculation System piping has occurre Secondary containment has isolated and SBGTS is operating. WHICH ONE of the following describes the expected release? SBGTS removes most of the noble gases and halogen Discharge of SBGTS is an elevated rsleas SBGTS removes most of the halogens, but almost none of the noble - gase Discharge of SBGTS is an elevated releas SBGTS removes most of the noble gases, but almost none of the halogen Discharge of SBGTS is a ground level release, SBGTS removes most of the halogens and noble gase Discharge of SBGTS is a ground level releas _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ - _ _ _ - _ - -
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SENIOR REACTOR OPERATOR Page 15 QUESTION: 017 (1.00) ! A LOCA has occurred and a General Emergency has been declare Emergency venting is in progress. Dose projection is not possible at this tim Plant conditions are: j Drywell radiation level: 3.0EE4 R/hr and increasing Torus radiation level: 2.8EE4 R/hr and increasing
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Drywell hydrogen conc: 7.2% and decreasing Drywell oxygen conc: 5.4% and decreasing Torus hydrogen conc: 5.8% and decreasing Torus oxygen conc: 4.5% and decreasing Drywell pressure: 44 psig and decreasing Torus bottom pressure: 40 psig and decreasing WHICH ONE of the following describes the appropriate Protective Action Recommendation (PAR) for these conditions? (EP-IP-400, Attachment 1 is attached.)- a. Consider evacuation of 2 mile ring and shelter 5 miles downwind, b. Consider evacuation of 2 mile ring and shelter 5 miles downwin Shelter other affected subarea c. Consider evacuation of 5 mile ring and 10 miles downwin d. Evacuate 5 mile ring and 10 miles downwin QUESTION: 018 (1.00) WHICH ONE of the following conditions could cause indicated level on the
' Fuel Zone instruments.to be higher than actual reactor vessel water-level?
a. Drywell temperature at'180"F.
l b. A break in the variable ~ le c. No forced = recirculation flow through the jet pumps.
l l d. A rapid reactor depressurization.
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SENIOR. REACTOR OPERATOR Page 16 . . QUESTION: 019 (1.00) All'of the Scram Discharge Volume (SDV) vent and drain valves were manually operated during maintenance. The valves were returned to their normal' position, but the manual operators for the SDV valves were-not returned to the NEUTRAL position. WHICH ONE of the following describes-the concern related to this operation? a. The valves could fail to automatically reposition on a reactor scram, preventing drain down of the SD b. The valves could f ail to automatically reposition on a reactor scram, causing a direct discharge path from the RPV to the Reactor Building sum c. The valves could fail open after repositioning on a reactor scram, causing a breach of primary containment, d. The valves could fall closed after repositioning-on a reactor scram, preventing reset of the scram due to high SDV leve QUESTION: 020 (1.00) A scram has occurred and the mode switch has been placed in SHUTDOW Scram Discharge Volume (SDV) level is high, but the scram has been reset-by use of the SDV High Level Scram Bypass. RPS bus A is then transferred to its alternate power supply. WHICH ONE of the following describes the expected response? It is a dead bus transfer and a half-scram will occur, b. It is a dead bus transfer and.a full scram _will occu It is a dead bus transfer, but is rapid enough so-that no scram will occu d. It is a-live-bus transfer,- so no scram will occu ,. _ . . _ . - -_ - _ _
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-SENIOR REACTOR-OPERATOR PageJ17
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y-QUESTION: 021 (1.00) A plant startup was underway when annunciator "MSIV-NOT FULLY OPEN SCRAM AT RX PRESSURE >600 PSI" (905R, D4) illuminated. A full scram' occurre WHICH ONE of the following describes the expected status of the plant? a. One main steam line is isolated; reactor pressure is greater than 600 psig, b. Two MSIVs have closed; reactor pressure is less than 600 psi c. Three main steam lines are isolated; reactor pressure is greater than 600 psi d. Four MSIVs have closed; reactor pressure is less than 600 psi QUESTION: 022 (1.00) The plant was operating at 100% power when a loss of all feed caused reactor water level to decrease. The reactor scrammed and the mode switch was placed in SHUTDOWN. Reactor water level is -10 inches and reactor pressure is 900 psig. WHICH ONE of the following sets of valves should have received isolation signals? a. Recirculation System process sample valves, Post Accident Sampling System isolation valves, and RHR reject to Radwaste valves b. RWCU isolation valves, RBCCW to Drywella Coolers: isolation valves, and Primary Containment Atmosphere Control makeup purge valves c. MSIVs, Drywell Equipment Drain Sump isolation valves, and Radiation Leak Detection System isolation valves d. Main-Steam Line drains, Drywell Floor Drain Sump isolation valves, and Hydrogen / Oxygen-Analyzer System isolation valves
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SEtt1OR REACTOR OPERATOR Page 18 l' QUESTIOti: 023 (1.00) The RCIC nystem is running .following a valid initiation signal roccived 10 minuten ago. The following conditions develop: RCIC Steam Line I' low: 230% RCIC Area Temperature: 2 0 S a l' RCIC Pump Suction Prensure: 10" lig vacuum RCIC Turbine Exhaunt Pronaure: 38 poig RPV Presnure: 150 poig RPV Water Level: 440 inchen WilICil OllE of the following deccribes the expected responce of the RCIC system?
-_ Only the RCIC inboard and outboard inolation valveo clon The RCIC inboard and outboard inolation valven clone and RCIC trip thrott le valve clonen, Only the Reic trip throttic valve clone Only the RCIC steam to turbine supply valve clocc _
_ _ . _ _ _ _ _ _ _ . _ _ _ _ _ . . _ _ . _ _ , _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ __ ______ __ _ _
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QUESTION: 024 -(1.00) WHICH ONE of the following describes the restrictions on HPCI operation while implementing the EOPs? a. HPCI may NEVER be operated below 1000-RPM because this is'the minimum speed required to maintain adequate cooling and lubricatio b. HPCI may NEVER be operated below 2000 RPM because this is the minimum speed required to generate sufficient control oil pressure for control valve. operatio HPCI may be operated below 1000 RPM only at NOS/NWE direction because low turbine exhaust pressure could create a cyclic steam hammer which could damage the exhaust check valv d. HPCI may be operated below 2000 RPM, but greater than 1000 RPM only as directed by the EOPs because operation at these speeds could cause excessive turbine vibration or oscillating flow rate QUESTION: 025 (1.00) With the plant operating at 100% power, the Core _ Spray Loop A.line break differential pressure indication on Rack 2207 reads +4.5 psi WHICH ONE of the following conditions is indicated? a. No core spray line break has-occurre This indication is normal for full power operation b. A core spray line break has occurred inside the reactor vessel shrou c. A core spray line break has occurred inside the reactor vessel, but outside the shroud, d. A core spray line break has o'ccurred-inside the drywell or inside-the reactor. vessel, but outside the shroud.- l -.
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SENIOR. REACTOR OPERATOR Page 20-l QUESTION: 026 -(1.00) The reactor is shutdown and RHR loop B is in the shutdown cooling (SDC) mode. A reactor coolant system leak causes vessel level to decrease to-0 inches and drywell pressure to increase to 2.8 psig. Loop A jet pump riser pressure is less than loop B jet pump riser pressure.- No operator action has been taken. WHICH ONE of the following describes the status of the Low Pressure Coolant Injection (LPCI) valves? a. Outboard injection valve 28A is open Inboard injection valve 29A is open Outboard injection valve 28B is closed Outboard injection valve 28A is closed Inboard injection valve 29A is open Outboard injection valve 28B is closed c. Outboard injection valve 28B is open Inboard injection valve 29B is closed ' Outboard injection valve 28A is closed Outboard injection valve 28B is open Inboard injection valvo 29B is open Outboard injection valve 28A is closed QUESTION: 027 (1.00) WHICH ONE of the following RHR loop lineups assures that the design-limits are met for adequate drywell sprays and RHR equipment operation? a. RHR pump B is running, the RHR heat exchanger bypass valve is closed, and RHR heat exchanger flow is 4800 gp b. RHR pump C is running, the RHR heat exchanger bypass valve is-open, and RHR heat exchanger flow is.3750 gp c. RHR pumps B and D are running, the RHR-heat exchanger-bypass valve is closed, and RHR heat exchanger flow is 9100 gp RHR pumps A and C are running,.the-RHR heat exchanger bypass valve is open, and RHR heat exchanger-flow is 5050 gp .
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SENIOR REACTOR OPERATOR Page 21
QUESTION: 028 (1.00) LPCI initiated on a valid initiation signal. RPV water level is -130 inches and increasing slowl Drywell pressure is 2.0 psig and increasing. WHICH ONE of the following describes the actions necessary to open MO-34 and MO-37 to spray the torus under these conditions? a. MO-34 and MO-37 cannot be opened until drywell pressure increases above 2.5 psig, Place the keylock RPV level override switch in MANUAL OVERRID Place the pistol grip LPCI overridc switch in MANUAL OVERRID Place the keylock RPV level override switch in MANUAL OVERRIDE -- and place the pistol grip LPCI override switch in MANUAL OVERRID QUESTION: 029 (1.00) An ADS blowdown was initiated on high drywell pressure and low-low reactor water level. The only low pressure CSCS pump running is Core Spray pump B. WHICH ONE of the following conditions will close the ADS valves? a. The ADS Inhibit switches are placed in INHIBIT, drywell pressure decreases to 2.0 psig, and the high drywell pressure reset pushbuttons are depressed, b. The ADS Inhibit switches are placed in INHIBIT, reactor water - level increases to -35 inches, and Core Spray pump B trip Drywell pressure decreases to 1.0 psig, the high drywell pressure reset pushbuttons are depressed, and the timer reset pushbuttons are depresse Core Spray pump B trips, drywell pressure decreases to 1.8 psig, and reactor water level increases to -40 inche _ _ _ _ - - _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ _ - _ _ - - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _
SENIOR REACTOR OPERATOR Page 22 QUESTION: 030 (1.00) The control switch for a safety relief valve is in the REMOTE position on alternate shutdown panel 156. WHICH ONE of the following describes operation of the safety relief valve in this condition? a. The valve will only operate in the safety mod b. The valve will only operate automatically in the ADS mode, c. The valve will only operate automatically in the ADS mode or the safety mod d. The valve will operate automatically in the ADS mode or the safety mode and can be manually operated from the control roo QUESTION: 031 (1.00) Refueling operations are in progress. A fuel assembly is being removed from location 19-22 in the core. WHICH ONE of the following situations meets the Technical Specification core monitoring requirements? (A core map is attached.)
a. SRM 'A' is bypasse SRM 'B' is fully inserted and reading 5 cp SRM 'C' is fully inserted and reading 2 cp SRM 'D' is fully inserted and reading 8 cp SRM 'A' is bypasse SRM 'B' is fully inserted and reading 10 cp SRM 'C' is bypasse SRM 'D' is fully inserted and reading 4 cps, SRM 'A' is fully inserted and reading 6 cp SRM 'B' is bypasse SRM 'C' is bypasse SRM 'D' is fully inserted and reading 9 cp SRM 'A' is fully inserted and reading 12 cp SRM 'B' is fully inserted and reading 4 cp SRM 'C' is fully inserted and reading 7 cp SRM 'D' is bypasse _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _
f3EllIOil itEACTolt OPEllAToll Page 23 QUESTIO!1: 032 (1.00) A normal plant startup is in progress with the modo cwitch in GTAltTU A in f ailed downr.cale and bypassed with its function switch on panol 936 in f3TD All the 1104 range uwitchen are on Itange 2. WillCIl Olit of the following deceriben the expected plant recponce it 1104 A in taken out of hypanu? action Itod block Itod block and half scram Itod block and Iul1 neram _ QUESTIoll: 033 (1.00) The Trannverne Incore Probe (TIP) detector was perf orming an automatic scan when power wan loot to 120 V lighting panel 17L. filmultaneous with the loan of power, the reactor scra.nmed and itPV water level dropped to 0 inches. WillCil Ol1E of the following dencribec the recponse of the TIP flyotem? a. The TIP detector will remain in the core, the ball valve will remain open, and the chear valve munt be ! ired manually, b. The TIP detector will remain in the core, the ball valve wi.ll clone, and the chear valve will 1 ire automatically, The TlP detector wi11 retract, the ha11 valve wi11 remain open, and the shear valve must be fired manuall d. The TIP detector will retract, the ball valve will clone, and it iu not necennary to 1 ire the shear valv _ , . . . SEllIOR REACTOR OPERATOR Pago 24 QUESTIOll: 034 (1.00) With the plant operating at 90't power, the APRM Calibration section of the OD-3 Printout provided the fallowing results: APRM 1(A) 2(C) 3(E) 4(B) 5(D) 6(F) READIliG 9 .3 8 .0 8 .4 AGAP 0.975 1.102 0.994 1.003 0.987 1.058 WilICl! 011C of the following identifion all the APRMs that require calibration? APRM A, APRM D, and APRM E _ APRP D, APRM E, and APRM F APRM A, APRM C, and APRM F APRM li, APRM C, and APRM F
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- . . . . SE1170R REACTOR OPERATOR Pago 35 QUESTIO!!: 035 (1.00) The plant is operating at 95% power. The following are the indications roccived when the APRM motor function switches on Panel 937 are placed_ in the AVERAGE, COUliT, and FLOW positions: AVERAGE COU11T FLOW APRM A 98% 70% 85% APRM B 94% ~7% 97% APRM C 96% .0% 85% APRM D 99% 55% 97% APRM E 93% 55% 85% APRM F- 95% 70% 97% _ WilICil OllE of the following describes the expected plant response for-those conditions? a. 11o action b. Rod block c. Rod block and half scram d. Rod block and full scram
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i i QUESTIOll: 036 (1.00) ;
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The plant was operating at 100% power when the 'A' Rocirculation pump ' tripped. All of the appropriato actions were taken in response to the recirc pump trip. The following annunciators are lit: , " ROD WITilDRAW BLOCK" (C905L, A1)
" ROD 13 LOCK HollITOR DOWllSCALE" (C905R, F3)
The RBM indicaten 75% for the selected ro WilICil OllE of the following ! doccriben the cause of these alarma? a. A voltage transient during the recirc pump trip caused the RBM to momentarily lower below 5/125 of scale causing the downncale trip, b. Au core powor_ decreased, the_ local power around the selected _ rod also decreaned. This caused the RBM output signal to decrease . below 94% of the reference signal causing the downscale tri !
, Prior to the transient, the R11M liigh Power Trip Sotpoint was i
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activated for the colected rod. The transient caused power to decreace prior to levoling out at 75%. As reactor power lowered into_the Low Power Trip Sotpoint, the downscale trip was actuate d. The recirc pump trip cauced 'A' recire loop flow to decrean As a result, the 'A' flow converter input a downncale signal into the 'A' RBM causing the downscale tri , _ -- s h h , Y
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SDlIOR REACTOR OPERATOR Page 27 l QUESTIOlit 037 (1.00) Reactor power is at 16% during a reactor startup. Control rod 30-15 is at position 12 and has been selected to be withdrawn to position 4 The reed switch for Rod 30-15, position 32 is faulty and will not actuat WilICll OllE of the following describes the expected response if I the rod movement control switch is taken to the llOTCil OUT position i
! aimultaneously with the " Emergency in/ notch override" switch being taken to the llOTCl! OVERRIDE position? lio rod motion will occu ,
b. A rod block will be initiated at position 32 and Rod 30-15 will settle at position 3 c. A rod block will be initiated at position 32 and Rod 30-15 will i settle at position 3 Rod 30-15 will withdraw to position 4 ;
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SENIOR REACTOR OPERATOR Page 28
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. QUESTION: 038 (1.00) '., control rods are being withdrawn with reactor power on IRM range 4, in accordance with the-attached Control Rod Sequence Sheet. Rod Group N is latched. WHICH ONE the following manipulations will result in a ; - RWM select error? ! t a. Rod-14-39 is at position 08 .
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Rod 30-39 is at position 06 Rod 38-15 is at position 04 Rod 14-15 is at position 14 Rod 14-39'is selected
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b. Rod 14-39 is at position 10 Rod 38-39 is at position 12 Rod 38-15 is at position 06 Rod 14-15 is at position 06 Rod 38-39 is selected Rod 14-39 is at position 04 s Rod 38-39 is at position 06 Rod 38-15 is at position 04 Rod 14-15 is at position 12 Rod 38-15 is selected d. Rod 14-39 is at position 14 Rod 38-39 is at position 08 Rod 38-15 is at position 10 Rod 14-15 is at position 04 Rod 14-15 is selected
. QUESTION: 039 (1.00)
A 1st' point feedwater heater was just removed from servic There has been no change in turbine steam flo WHICH ONE of the following describes the effect on the plant?
- a. Main generator output decrease Plant officiency-increases.- -1 b. Main generator output decrease Plant efficiency decreases, c. Main generator output increase Plant efficiency increase d. Main generator output-increase Plant efficiency decrease *
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QUESTION: 040 (1.00) , The plant was operating at 100% power when the "RFP A LOW NPSil" (C1L, , A1) annunciator is received. The RFP sequential trip selector switch is ,
"ON" and the reactor feed pump tripping sequence switch is in the " CAB" position. .WilICli ONE of the following describes the potential cause of the alarm and the expected automatic actions?
a. The alarm was caused by a trip of one condensate pump. The ,
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condenser reject valves open. RFP A will trip if the low HPSil condition persists for 15 seconds, b. The alarm was caused by a trip of one condensate pump. - The ! condenser reject valves shut. RFP C will trip if the low NPSil condition persists for 15 second ! c. The alarm was caused by placing a condensate domineralizer i servic The condenser reject valves shut. RFP A will trip-if the low NPSil condition perolsts for 15 second , d. The alarm was caused by removing a condensate domineralizer from service. The condenser reject valves ope RFP C will trip if the low HPSil condition persists for 15 seconds.
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- QUESTION
- 041 (1.00)
The plant is operating at 100% power with the FWLC setpoint set at +28 inche Wi!ICil ONE of the following describes tho'effect of a loss of one feedwater flow input to the.FWLC System? a. Feedwater flow decreases, reactor water level decreases, and the reactor scram : b. Feedwater flow decreases and reactor water 1cvel decreases, but~ -r the reactor should not scram, c. Feedwater flow increases, reactor water level increases, and-the reactor may scra d. Feedwater flow increases and reactor water level increases, buti the reactor should-not scra :
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_ . _ _ _ . _ _ _ _ _ .- . . . _ _ SEl4IOR REACTOR OPERATOR Page 30 P QUESTIOll: 042 (1.00) , The plant was operating at 100% power with the MitC System operating !
.normally when the Bypass opening Jack (B0J) was taken to the RAISE ,
i positio WilICl! OllE of the following describes the expected plant response if the B0J cannot be returned to the OFF position? a. The bypass valves will open fully. Reactor pressure will decrease and stabilize at a slightly lower value than before the transien b. The bypass valves will open fully, then the control valves will , open until the control valve limit stop is reached. Reactor ' pressure will decrease and stabilize at a lower value than before the transien c. The control valves will open until the control valve limit stop_ is reached, then the bypass valves will open full Reactor pressure will decrease until the HSIVs clos d. The control valves will open until the control valve limit stop is reached, then the bypass valves will open until the reactor flow limit is reached. Reactor pressure will decrease until the MSIVs close.
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i SENIOR REACTOR OPERATOR Page 31 ; i
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QUESTION 043 (1.00) A turbine runback was initiated due to high stator outlet coolant temperature at time zero. The following table shows the generator load ; and stator outlet coolant temperature responso during the runback: TIME STATOR AMPS OUTLET TEMPERATURE 40.5 minutas 25,300 88'c
+1.0 minutes 21,800 86'C +1.5 minutes 18,300 84*C +2.0 minutes 14,800 82*C +2.5 minutes 11,300 80'C +3.0 minutes 7,800 78aC 43.5 minutes 4,300 76*C +4.0 minutes 800 74*C WilICl! ONE of the following correctly describes the plant response?
a. The tarbine should have tripped at 42.0 minute ! b. The turbine should have tripped at +3.5 minutes, c. The runback should have stopped at +1.0 minute d. The runback should have stopped at +3.5 minute QUESTION: 044 (1.00) The main condenser vapor valves (AO-3703, AO-3704, AO-3710, and AO-3711) have shut automatically while the plant was operating at 100% powe WillCIl ONE of the following conditions could have caused the isolation? a. Offgas high radiation caused by abnormal carbon vault temperature b. liydrogen concentration greater than 1 percent caused by_a recombiner malfunction c. Offgan high pressure caused by an explosion in the offgas' system d. High offgas flow caused by lowering condenser vacuum
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_ _ - - . . . _ _ _ . _ . . _ __. __...__ - _ _ _ . _ _ . . - - . _ _ _ _ _ . . _ - . .. SENIOR REACTOR OPERATOR Page 32 QUESTION: 045 (1.00) 1 i . The plant is in single loop operation. For WilICil ONE of the following I conditions can the idle recirculation pump be started? (Consider ! administrative and functional limitations.) ; a. The. idle recirculation pump' suction and discharge valves are fully open. The scoop tube lock light is off and speed control is in manual with an output setpoint of zero, b. The idle recirculation MG set generator field breaker-10 ope Lube oil pressure in 17 psig and lube oil-temperature is 90* i c. The operating pump in running at 40% of rated spee Core flow is 35 M1b/hr. Core thermal power in 1050 MW d. The idle pump suction temperature i s 408*F. The operating pump suction temperature in 435* Bottom head drain temperature is 492* Vessel dome temperature is 548' l QUESTION: 046 (1.00) The plant is operating at 50% power with both recirculation pumps in manual. WillCli ONE of the following describes the expected response if all FWLC System input were lost to Recirc Pump D's flow controller? a. Recirc Pump B runback to'65% due to a Speed Limiter /2-runbac Recirc Pump 11 runback to 26% due to a Speed Limiter #1 runbac c. Recirc Pump B scoop tube will lockup.due to a speed control , signal failure, , ' d. There will be no effect on Recirc Pump B because reactor water level ~is normal and the pump discharge valve is full open.
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SENIOR REACTOR OPEllATOR , Page 33 QUESTION: 047 (1.00) A BOCA has occurred concurrent with a loca of Ac power. Both diocol generatorn (DGo) started, but the electric governor for DG B malfunctioned immediately after the start signal was received. WillCH ONE of the following doncribes the expected response of the DG B output breaker and DG B? , a. The DG output breaker will not close because the DG will not i come up to speed and voltage. The DG will continue to run, b. The DG output breaker will close because the mechanical governor will take ove The DG will continue to run, The DG output breaker will clone, but will reopen as DG speed increanen due to overcurrent. The DG will continue to ru The DG output breaker will close, but the DG will trip on overspeed causing the output breaker to trip ope ' QUESTION: 048 (1.00) f The Standby Gas Treatment (SGT) System initiat on a valid initiation signal 3 minutes ag Prior to the initiati i, both SGT trains were in the normal standby lineu No operator ac on has been taken and the initiation nignal is still present. The oGT train A heater just tripped due to high temperature. WNIcli ONE of he following describes the expected responne of the SGT System? a. SGT train A fan will tri SGT train A inlet and outlet dampers will clos SGT train B an will continue to ru SGT train il inlet and outlet damper > will remain ope b. SGT train A fan will tri SGT train A inlet and outlet dampers will clos outlet'dampera SGT w t7ain B fan will start. SGT train 13 inlet and 1 open, SGT train A fap will continue to ru SGT train A inlet and outlet damper- will remain ope SGT train B fan will continue to ru SGT train 11 inlet and outlet dampers will remain ope d. SGT train A fan will continue to run. SGT train A inlet and outlet _ c}. mpers will _ remain open. SGT train B fan will not start. 'SGT train B inlet and outlet dampers will remain close ~~
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SElllOR REACTOR OPERATOR * Pago 36
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QUESTION: 049 (1.00) WilICll ONE of the following conditions will cause a secondary containment isolation? a. Refuel floor rad monitor 1705-8A: 25 mR/hr Refuel floor rad monitor 1705-8B: 15 mR/hr Refuel floor rad monitor 1705-8C: 20 mR/hr ! Refuel floor rad monitor 1705-8D: 10 mR/hr l b. Refuel floor rad monitor 1705-8A: 10 mR/hr Refuel floor rad monitor 1705-8B: 15 mR/hr -;~ Refuel floor rad monitor 1705-8C: 20 mR/hr Refuel floor rad monitor 1705-8D: 0.1_mR/hr ; c. Refuel floor rad monitor 1705-8A: 0.1 mR/hr Refuol floor rad monitor 1705-8B: 20-mR/hr Refuel floor rad monitor 1705-8C: 0.1 mR/hr ' Refuel floor rad monitor 1705-8D: 0.1 mR/hr ; i d. Refuol floor rad monitor 1705-8A: 15 mR/hr Refuel floor rad monitor 1705-8B: 20 mR/hr Refuel floor rad monitor 1705-8C: 0.1 mR/hr Refuel floor rad monitor 1705-8D: 0.1 mR/hr i
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f3E!110R REACTOR OPERATOR Page 35 QllCSTIO!1: 050 (1.00) Reactor 13uilding ventilation f ano must be started in the proper order to maintain reactor building pressure and prevent the spread of contaminatio WilICl! OllE of the following describes the proper sequence for starting reactor building ventilation fann? a. 1:xhaunt fann should be started before supply fan Contaminated exhaunt (Zone 3) fans chould be started before clean exhaust (Zone 2) fan b. Exhaunt Iann should be started betore supply fan Clean exhaunt (Zone 2) fann uhould be started before contaminated exhaunt (Zone 3) fan _ c. Supply fann should be started before exhaust fan Clean exhaunt (Zone 2) fann uhould be started before contaminated exhaunt (Zone 3) fan d. Supply lana nhould be otarted before exhaunt fan Contaminated exhaunt (Zone 3) fann chould be started before clean exhaust (Zone 2) fan _.
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' QUESTION: 051 (1.00)
Control room environmental control supply fan HS-77(VSF-103A) is in STBY and supply fan HS-78(VSF-103B) is in AUTO. WHICH ONE of the following describes the Control Room HVAC System response to a Halon initiation? ,
a. Both control room environmental control supply fans star l Normal supply and exhaust fans trip. Halon exhaust' fan starts i l and damper AO N-141 shut b. Control room environmental control supply fan HS-78 start Normal supply and exhaust fans trip. Cable _ spreading room supply and exhaust dampers shut, Both control room environmental control supply fans star Normal air intake dampers close and filtration system isolation dampers open. Cable spreading room supply and exhaust dampers l-shu Control room environmental control supply fan HS-78 start Normal air intake dampers close and filtration system isolation ; dampers open. Halon exhaust fan starts and damper AO N-141 shut > QUESTION: 052 (1.00) If the Fuel Pool Cooling (FPC) System is unavailable, WHICH ONE of the following systems can be crosstied for_ pool cooling? a. Fire Protection Residual Heat Removal :1 c.' Reactor Building Closed Cooling Water ' d. Condensate and Demineralized Water Storage and Transfer
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QUESTIOll: 053 (1.00)
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The reactor modo switch is in REFUE In WilICil OllE of the following conditions would control rod withdrawal be possible? a. One control rod withdrawn to position 02. The withdrawn control , rod is selected. The bridge is over the core. The main hoist .
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is loaded with 600 lbs. The grapple is fully u b. All control rods are full in. The bridge is 110T over or near the cor The monorail hoist is loaded with 300 lbs. The grapple is 110T fully u , One control rod withdrawn to position 02. A second control rod is selecte The bridge is over the core. The frame mounted hoist is loaded with 200 lbs. The grapple is fully u . d. All-control rods are full in. The bridge is 110T over or -near the cor The service platform hoist is loaded with 500 lb QUESTIOll: 054 (1.00) WillCl! OllE of the following Fire Protection Systems would be utilized to suppress a fuel oil fire in the DG fuel oil pipe trench?
' Dry Chemical System b. llalon System Cardox System Fire Water System
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/ N QUESTIOll: 055 (1.00)
The plant is operating at 100% power. WilICil Oli of the following conditions would flOT require initiation of a lant shutdown in accordance with Technical Specifications? a. One ADS valve is inoperable. T to llPCI inverter tripped on high voltage. Voltage returned to ormal, but the inverter failed to automatically reset. Manua reset of the inverter-is expected to take 15 minute b. RCIC is inoperable. T 4. !!PCI gland seal condenser is inoperabl Repairs e expected to take 12 hour c. IIPCI is inoperable RilR valve MO-1001-37A (Torus Spray Valve) is stuck in the c osed positio Repairs are expected to take 2 day d. LPCI B is ino erable. The 11PCI flow indicator controller ramp generator is inoperable. A replacement part had to be ordered and will n^ be available for 5 day ,
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QUESTION: 056 (1.00) t The plant is operating at 100% powe PNPS Proc. No. 8.4.1, " Standby Liquid Control Pump Oper and Flow Test has just been performed." The surveillance data sheets are attached. ' Current SLC System status is as follows: Boron Enrichment: 55'.0 Atom Percent ' SLC Tank Concentration! 8.72% by weight SLC Tank Volume (C905): 1900 gallons SLC Tank Temperature: 45'F
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WHICH ONE of the following describes the required actions in accordance l with Technical Specifications? ! Continued reactor operation is permitted for seven day l
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b. The reactor shall be placed in COLD SHUTDOWN with all operable ' control rods inserted within the next 24 hour Flow test the SLC System to verify a flow path. If the test is successful, there are no restrictions on reactor operatio d. There are no restrictions on reactor operation.
. QUESTION: 057 (1.00) The-plant is operating at 100% power with the shutdown transformer out of service for maintenance. The offsite 345 kV transmission line from Bridgewater Station is unavailable due to a problem at Bridgewater i Statio It has-just been reported that Diesel Generator A failed its !
-monthly operability tes WHICH ONE of the following is the' appropriate action in accordance with Technical Specifications?
a. Continued reactor operation is permissible for 7 days.
t b. Continued reactor operation is permissible for 72 hours, Reduce reactor power level to 25% and notiny the NRC within 1 . hou ' d.. Initiate an-orderly shutdown and-be.in COLD SHUTDOWN within 24 hour .
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SENIOR REACTOR OPERATOR Page 40 QUES TION: 058 (1.00) The plant is operating at 100% power. On the most recent P-1 computer printout MAPRAT in 1.02. WHICH ONE of the following is the appropriate action that should be taken in accordance with Technical Specifications? a. Within two hours, restore MAPRAT > 1.04 and insert all insertable control rod b. Within two hours, restore MCPR within limits and insert all insertable control rods, c. Within 15 minutes, initiate action to restore APLHGR within limits, d. Within 15 minutes, initiate action to restore MCPR within limit l l , i
QUESTION: 059 (1.00) The-Salt Service Water (SSW) System is operating with SSW Pumps A, C, and E running with the plant at 100% power. SSW Pump B is in AUTO and SSW Pump D is tagged out for maintenance. A fault on MCC B-10 causes SSW Pump C to tri WHICH ONE of the following describes the requirrid actions in accordance with Technical Specifications while the MCC is < being repaired?
' No action is require b. Reactor operation may continue for 7 day Reactor operation may continue for 72 hour Initiate a shutdown and be in COLD SHUTDOWN within 24 hour b r '
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SENIOR REACTOR OPERATOR Page 41 QUESTION: 060 (1.00) Core offloading is in progres Standby Gas Treatment (SGT) Train A has been inoperable for four days. A spent fuel bundle is grappled over the core, when it is reported that the SGT Train B fan is operating at 3500 cfm during surveillance testing. WilICil ONE of the following describes the appropriate actions? a. Lower the fuel bundle into its designated location in the spent-fuel pool, then cease fuel movemen b. Lower the fuel bundle into the nearest open location in the reactor vessel, then cease fuel movemen c. Fuel movement may continue for seven days provided all other components of SGT B are operabic, d. Fuel movement may continue for three days provided secondary containment integrity is maintaine QUESTIONi 061 (1.00) PNPS Proc. No. 2.4.29, " Stuck Open Relief Valve," directs.the operator to initiate a manual scram if.a safety / relief valve (SRV) cannot be closed. WilICl! ONE of the following= describes the reason for this requirement? a. Suppression pool temperature could increase and cause an increase in drywell temperature and pressur b. Rapid shutdown, cooldown, and depressurization of the reactor minimizes the coolant loss through the stuck open SR c. Technical Specifications require that the self actuating relief-mode for all SRVs be operabl d. Sustained pressure and flow will damage the relief valve-downstream piping.
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l QUESTION: 062 (1.00) During preparations for refueling operations, with the plant in HOT , SHUTDOWN, an unisolable leak in the Spent Fuel Pool caused fuel storage ' . pool level to decrease below the top of the fuel bundles in the pool.
i General area radiation levels on the refuel floor are 250 nR/h : Radiation levels in the FPC pump room and skimmer surge tank area are , 750 mR/hr. Water levels in the SE and SW quadrants are 8 inches above , the floo WHICH ONE of the following is the appropriate emergency , action level for this situation? a. None b. Unusual Event c. Alert d. Site Area Emergency +
QUESTION: 063 (1.00) The. Control Room was evacuated due to a fire 5 minutes ago. All the immediate actions of PHPS Proc. No. 2.4.143, " Shutdown from Outside the Control Room," were performed with the exception of tripping the Reactor- l' Feed Pumps. The Feed Pumps were tripped at their 4160 VAC breaker current plant conditions are:
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The reactor has been verified to be shutdow RPV level is > +50 inches (pegged high).
- RPV pressure is 700 psig and decreasing slowl The MSIVs are open and cannot be close The turbine bypass valves appear to be functioning normall .
- An operator is stationed to control RPV level and pressure with 1 HPCI as necessar WHICH ONE of the following is the appropriate emergency action level for this situation? Unusual Event - b. Alert Site Area Emergency General Emergency
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QUESTIOll: 064 (1.00) l
An Anticipated Transient Without Scram (ATWS) has occurred. Reactor power is 5%. Reactor pressure is 1175 psig and reactor water level'is
+25 inches. Torus water temperature is 105'F and torus water level is 178 inche WilICil OllE of the following is the appropriate emergency action level for this situation? '(Assume no operator action.) -
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a. Unusual Event ,
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b. Alert c. Site Area Emergency d. General Emergency QUESTION: 065 (1.00) No fuel movement is in progren WilICll ONE of- the following conditions would be a violation of Secondary Containment integrity? a. Reactor Building differential pressure is greater than 0 inches of water in STARTU b. Both reactor building to torus vacuum breakers are inoperable and stuck in the open position in RU c. Both Standby Gas Treatment trains are inoperable in COLD SilUTDOWN with the reactor coolant system vented.- Both reactor building airlock doors are open to remove a large picco of equipment from the reactor building in llOT SilOTDOW t
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i l QUESTIOll: 066 (1.00) l WilICli ONE of the following conditions would be a safety limit violation?
a. While operating at 30% power, the MilC pressure regulators fai Reactor pressure drops to 800 poig before the MSIVs close and the reactor scram b. While operating at 75% power, a malfunction in the master recirculation flow controller causes the speed of both recirc pumps to increase to the high speed stop. The Minimum Critical * Power Ratio is 1.0 ! k
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' c. While operating at 90% power, a turbine trip occurs. The reactor fails to scram and the ATWS system logic trips the reactor Icod pumps, d. While operating at 100% power, an inadvertent MSIV closure causes reactor pressure to increase. The reactor scrams and reactor pressure i ncreases until both safety valves lift, i
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QUESTION: 067 (1.00) Wi!IcIl-ONE of the following describes the expected indications if Jet Pump 5 failed? , a. Recirc Loop A flow increase Total core flow decreases,
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b. Recire Loop A flow increase Total core flow increase c. Recirc Loop A flow decrease Total core flow decreases, d. Recire Loop A flow decrease Total core flow increase . L $ i
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SENIOR REACTOR OPERATOR Page 45 [
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QUESTION: 068 (1.00)
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The plant was operating at 96% power on the 100% load line for several months when Recirc Pump A tripped due to operator error. Plant conditions following the trip ate:
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Reactor Power 60% Recirc Loop A Flow: 2.0 MLB/IIR (FI-263-107A) Reciro Loop B Flow: 30.5 MLB/llR ,
(FI-263-107B)
Total Core Flow: 32.5 MLB/llR (FR-263-110) WilICil ONE of the following describes the appropriate action in accordance with PNPS Proc. No. 2.4.17, " Recirculation Pump (s) Trip," ,
(attached)?
i a. Increase the speed of Recire Pump B until loop flow is greater than 31.5 MLB/H b. Insert control rods to decrease reactor power below the 80% load lin Restart the idle pump in accordance with PHPS 2.2.84, " Reactor Recirculation System." , Scram the reactor and concurrently perform PNPS 2.1.6, " Reactor Scram."
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_ _ ._._._.. _._ ._ _ _. _ _ _ __ _-_ _ . _ _ . . _____._ _ SENIOR REACTOR OPERATOR page 46 QUESTION: 069 (1.00) The plant was operating ~at 100% power when a break-developed on the ! instrument air header. WilICil ONE of the following describes automatic ; < actions that should occur before air header pressure drops below 75 psig? a. Service air header isolates Backup K104 air compressor starts l Feodwater control valves lock up b. Non-essential instrument air isolates Feedwater control valves lock up Scram pilot valve air header depressurizes c. Drywell pneumatic supply header isolates . Lagging compressor loads ! Service air header isolates d. Low pressure service air crossconnect isolates (if open) Backup K104 air compressor starts r Non-essential instrument air isolates
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_ QUESTION: 070 (1.00) The plant was operating at 100% power when a complete loss of instrument air occurre Power was also lost to the AC solenoid operated air cutoff valves for the inboard MSIVs and to the.DC' solenoid operated air cutoff valves for the outboard MSIVs. The primary containment-is inerte WilICil ONE of the following describes the response of the MSIVa? a. Inboard MSIVs remain ope Outboard MSIVs remain open.- b. Inboard MSIVs close due to air pressur Outboard'MSIVs close due to air pressure.
l c. Inboard MSIVs close due to spring pressure.. Outboard MSIVs ' remain _open.
I d. Inboard-MSIVs remain ope Outboard MSIVs close-due to spring pressure.
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I QUESTION! 071 (1.00) An inadvertent containment isolation signal was received and immediately cleare No operator action was taken. WilICl{ ONE of the following sets of valves do NOT require a manual roset to reopen tho valvos if they isolated on the specified signal?
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Valvos Isolation Signal
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a. Recirc sample lines Main Steam Lino low pressure b. IIPCI exhaust vacuum liigh drywoll pressure breaker isolation valves (w/ low reactor pressure) RCIC inlet steam valves Low reactor pressure RilR/LPCI Injection valves liigh reactor pressure (in Shutdown Cooling modo) QUESTION: 072 (1.00)
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Three RBCCW pumps are unavailable. WilICll ONE of the following describes the appropriate action that must be taken with the specified pumps unavailable? a. RBCCW pumps A, B, and.C are_ unavailable. The RBCCW loops'must be crosstied to prevent a reactor scram on high drywell pressure due to a loss of drywell coolin b. RBCCW pumps A, C, and E are unavailabl The RBCCW loops must be crosstied to' prevent a trip of the recirculation pumps due to loss of cooling,
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, I c. RBCCW pumps B, D, and F are unavailable. It is not necessary to I crosstie the RBCCW loops bocause at least one pump in each loop j- is still availabl d. RBCCW pumps D, E, and F are unavailable. It is not necessary to crosstic the RBCCW loops because at least one pump in each loop , ' is still availabl _
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SEMIOR REACTOR OPERATOR Page 48 QUESTION: 073 (1.00) The Salt Service Water (SSW) System was operating with SSW Pumps A, C, and D running and SSW Pumpo B and E in AUTO. The loop selector switch is in its normal positio A complete loss of normal AC power occurs simultaneously with a loss of coolant accident (LOCA). WilICII ONE of the following describes the response of the SSW System with no operator action? e a. TBCCW heat exchanger outlet valves throttle 90% close RBCCW heat exchanger outlet valves open full SSW Pumps A and D restadt after load shedding, TBCCW heat exchanger outlet valves throttle 90% close RBCCW heat exchanger outlet valves open full SSW Pumps B and E start after load shedding, TBCCW heat exchanger outlet valves open full RBCCW heat exchanger outlet valves throttle 901 close SSW Pumps A and C restart after load sheddin d. TBCCW heat exchanger outlet valves open fully.
" RBCCW heat exchanger outlet valves throttle 90% close SSW Pumps B and C start after load sheddin QUESTION: 074 (1.00) The plant is operating at 800 MWt. Core flow is 35 MLB/llR. Condenser vacuum in 18" lig . WilICll ONE of the following describes the appropriate actions to be taken under these conditions? - Reduce recirculation pump speed, maintaining core flow above 31.5 MLB/IIR, until the vacuum decrease is terminated, b. Reduce recirculation pump speed until the vacuum decrease is terminated OR the pumps reach minimum spee Insert control rods in reverse order of the pull sheet as necessary to stop the vacuum decrease, Immediately scram the reactor and trip the turbin _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
SENIOR REACTOR OPERATOR _ page 49 QUESTION: 075 (1.00) The plant was operating at 95% power when the Train A 4th point feedwater heater isolated due to a tube rupture. Reactor power increased to 100%. WHICH ONE of the following describes the appropriate immediate actions? Runback recirculation flow and insert control rods as necessary to maintain reactor power below 95%. b. Runback recirculation flow and insert control rods as necessary to maintain reactor power below 25%. Runback recirculation flow until reactor power is below 75% or core flow reaches 31.5 M1b/h Runback recirculation flow until reactor power reaches 70% or core flow reachec 31.5 Mlb/h QUESTION: 076 (1.00) The plant is operating at 100% power when the in-service CRD flow control valve malfunctions causing CRD system flow to oscillate. WHICH ONE of the following subsequent CRD system failures would require an immediate reactor scram? ! a. Trip of both CRD pumps Loss of all rod position indication More than one control rod in a 9-rod array drifts More than one CRD mechanism high temperature alarm in a 9-rod array l { l: ! '
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_..._.___ _ _ _ _ ______ _ _ ____ .m___ SENIOR REACTOR OPERATOR Page 50 QUESTION: 077 (1.00) Core offloading is in progross. A spent fuel bundle is grappled over " the core, when the Refueling Floor Area Radiation Monitor (ARM) alarm No other alarms have been received on the refuel floor or in the control Roo WilICll ONE of the following describes the appropriate actions? a. Lower the fuel bundle into its designated location in the spent i fuel poo b. Lower the fuel bundle into the nearest open location in the spent fuel pool.
' Lower the iuol bundle into the nearest open location in the - reactor vessel, Lower the fuel bundle into the location from which it was removed in the reactor vesse . QUESTION: 078 (1.00)
An electrical failure has occurred causing multiple alarms to-annunciate. One of the illuminated alarms is annunciator " VITAL INST SYS LOSS OF DC POWER," (C3 Center, window DS). . Based on this information, WilICll ONE of the following describes the emergency procedures that are applicable for this situation? a. PNPS Proc. No. 5.3.6, " Loss of Vital AC (Y-2)" and PNPS Pro :' No. 5.3.13, " Loss of Essential DC Bus D6" b. PNPS Proc. No. 5.3.6, " Loss of Vital AC (Y-2)" and PHPS Pro Ilo . 5.3.30, " Loss of 250V DC Bus D-10" PHPS Proc. No. 5.3.7, " Loss of Instrument Power Bus:(Y-1)" and PNPS Proc. No. 5.3.13, " Loss of Essential DC Bus D6" d. PNPS Proc. No. 5.3.7, " Loss of Inu ti:ument Power Bus (Y-1)" and * l- PNPS Proc. No. 5.3.30, " Loss of 250V DC. Bus D-10" t , i-
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SENIOR REACTOR OPERATOR Page 51 QUESTION: 079 (1.00) A station blackout has occurred and the SBO diesel failed to star RCIC and itPCI were being used to control PPV level and pressure when both systems tripped and could not be restarted. PNPS Proc. No. 5.3.31,
" Station Blackout," directs use of PNPS Proc. No. 5.3.26, "RPV Injection 'cring Emergencies," for alternate methods. WillCil ONE of the methods ould be used for RPV injection in this situation?
- SSW crosstled to Ri!R Fire Water crosstied to RilR Condensate Transfer crosstied to ECCS fill lines
' Domineralized Water Transfer crosstied to SBLC
Y QUESTION: 080 (1.00) The Control Room has been evacuuLed due to a fire in the Cable Spreading P,o At WHICil ONE of the following locations would you find an * Alternate Shutdown toolbox? ' Aux Bay
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23' RPS MG Room
' ' Switchgear Room ' Turbine Building ~ .-__ _ _ _ - _ - _ _ - - _ _ _ _ _ _ _ _ _ - - - _ - _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ - _ _ - _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ - _ -- -
fiENIOR REACTOR OPERATOR Page 52 QUESTION: 081 (1.00) Following a reactor scram, the operator is required to determine whether or not all control rods are inserted to or beyond position 02 in order to determine the appropriate actions to be taken in accordance with the EOPs and PNPS Proc. No. 2.1.6, " Reactor Scram." WHICH ONE of the following is an acceptable method for determining whether all rods are fully inserted? Observing that all the blue scram lights are illuminated on the full-core displa Verifying that all APRMs are downscale and reactor power is trending down on the IRM a Prior to scram reset, verifying that the rod drift annunciator will not clear, With the mode switch in REFUEL, taking the rod select power off, then back on, and observing if the Refuel Mode Select Permissive light illuminate QUESTION: 082 (1.00) EOP-02, "RPV Control, Failure-to-Scram," directs actions based on whether or not the reactor is shutdown. For WHICH ONE of the following conditions would the reactor be considered shutdown in accordance with EOP-02? - Reactor pressure is O psi The Hot Shutdown Boron Weight of sodium pentaborate has been injected into the RP b. Reactor power is 20 on Range Reactor period is +100 and stabl c. Reactor power is 50 on Rangc Reactor period is negative, Reactor power is 30 on Range Reactor period is negativ I _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ __ _ _ _ _ _ _ - _ - _ _ _ _
SENIOR REACTOR OPERATOR Page 53 QUESTION: 083 (1.00) A reactor scram has occurred and all control rods are not inserte WilICll ONE of the following indications would be expect ,1 if the f ailure of the rods to ir.scrt was due to an electrical failure of the Reactor Protection System (RPS)? a. Alarm " SCRAM VALVE PILOT !!EADER LO PRESSURE" (C905R, A6) illuminate Alarm " SCRAM DISCH VOLUME III LEVEL SCRAM" (C905R, 14) illuminate Group Scram logic lights on C905 illuminate _ Blue Scram lights on C905 illuminate QUESTION: 084 (1.00) A reactor scram has occurred and all control rods are not inserte In accordance with PNPS Procedure No. 5.3.23, " Alternate Rod Insertion," WilICll ONE of the following methods for inserting control rods requires the scram to be recet? a. Venting the overpiston areas of the control rod drives, Individually scramming control rods from panel C91 Venting the Scram air header and deenergizing the Scram solenoid Inserting a manual Scram from panel C90 __-_-__ _ _ - _ - _ _ _ _ - _ _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
_ _ _ _ . . . _ _ _ _ . . . _ . . _. . ~ _ ... . . =._ _ _._ .. _ m . .~ ...__. _ SENIOR REACTOR OPERATOR Page 54 QUESTION: 085 (1.00) WHICH ONE of the following actions should be taken'to maximize drywell cooling in accordance with'EOP-03, " Primary Containment Control?" Initiate drywell spra * b. Secure the recirculation pump . Increase the RBCCW system temperatur Increase the RBCCW system flow rat QUESTION: 086 (1.00) WHICH ONE of the following. lineups would maximize torus cooling in accordance with EOP-03, " Primary Containment Control?" Both RHR-loops in service-with one pump per loop-operating and the RHR heat exchanger bypass valves shut.- One RHR loop in service with both pumps operating and the RHR heat exchanger bypass valve shu , Both RHR loops in service with one pumpLper loop operating and the RHR heat exchanger bypass valves ope ' d. One RHR loop in service with both pumps operating and the RHR heat exchanger bypass valve open.
.
+
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_ _ . . _ _ _ . _ _ _ _ _ _ - - . _ _ _ . . . . _ -__ _ __ -. SENIOR REACTOR; OPERATOR Page.55-QUESTION: 087 -(1.00) A station blackout and complete loss of instrument air has occurre Plant conditions are as follows:
-Drywell pressure is 3 psig and increasing slowly-Torus water level is 220 inches-Neither Standby Gas Treatment (SGTS) train is operable because the outlet dampers are stuck in the closed position WilICH ONE of the following describes the reason that the Direct Torus-Vent (DTV) cannot be used in this situation? SGTS must have an operable vent-path for use of the DTV.-
b. Primary containment conditions do not meet the criteria for use of the DTV.
' The DTV path is not available during a-station blackout, d. The DTV path is not available during a loss of instrument ai QUESTION: 088 (1.00) Both torus water level indicators are upscale and the primary containment water level indicator is downscale. For WHICH ONE of the-following conditions is this an accurate indication? (PHPS. Proc. N .3.27, " Determining Primary Containment Water Level," is attached.)
a. Drywell Wide Range Pressure: -25 psig Torus Bottom Pressure: 10 psi b.-Drywell Wide Range Pressure: 5 psig Torus Bottom Pressure: 35 psig Drywell Wide Range Pressure: 20 psi Torus Bottom Pressure: 40 psi Drywell-Wide Range Pressure: 26 psig-Torus Bottom Pressure: 20 psig , s, . -- , ,_, >
...- - . . . - . _ - . - - - . - - . . .. . .
SENIOR REACTOR OPERATOR Page 56 i l
.
QUESTION:-089 (1.00)
~
Following-a major loss-of coolant accident.and loss of off-site power, RPV_ level cannot be maintained above TAF. The operating crew-determines that an alternato injection system needs to be aligned in order-to recover RPV leve Reports from the field indicate that the aux. bay is inaccessible. WHICH ONE of the following lineups could be'used to deliver a high capacity flowrate to the vessel given the current plant conditions? a. SSW crosstled to RHR b. Fire water crosstled to RHR c. Fire water crosstied to Feedwater d. Domineralized water transfer crosstied to'SBLC QUESTION: 090 (1.00) A primary system leak has made the Reactor Building inaccessibl The following conditions exist:
-
Reactor pressure is 200 psig
-
RPV water level (indicated) is +5 inches
-
Drywell temperature is 300*F WHICH of the following instruments, if any, can be used to determine RPV water level? a. Narrow Range, Fuel Zone, and FW Control b. Narrow Range and FW Control c. FW Control d. None ,
, , , . - , .
SENIOR REACTOR OPERATOR Page 57 QUESTION: 091 (1.00) EOP-02, " Failure to Scram," arep L-10 directs you to stop and prevent all injection into the RPV except for boron, CRD, and RCIC. WHICH ONE of the following is the reason for securing injection in this step? Securing injection sources helps to reduce the pressure in the reactor vessel while performing Alternate RPV Depressurizatio Securing injection sources prevents the addition of a large volume of cold, unborated water into the cor c. The injection sources are not necessary, because adequate core cooling is assured by core submergenc ___ The injection sources are not necessary, because adequate core cooling is assured by steam cooling without injectio QUESTION: 092 (1.00) EOP-02, "RPV Control, Failure-to-Scram," is being implemented. The reactor is shutdown, but it has not yet been determined that it will remain shutdown under all conditions. WHICH ONE of the following describes the appropriate actions that should be taken if RPV level cannot be determined? Enter EOP-1 Exit the RPV level leg of EOP-0 Enter EOP-1 Exit the RPV level leg and the RPV pressure leg of EOP-0 Enter EOP-2 Exit the RPV level leg of EOP-0 Enter EOP-2 Exit the RPV level leg and the RPV pressure leg of EOP-0 l
. - - - - - - - _ __ _ _____ _ ______ _ __ _ _ _ _ _ ______ ___________ __
_. . . _ ~ ., .-_ _ . - _ - _ _ . _ . - . _ . - _ _ _ _ _ . _ _ _ . - _ _ _ . .m. . _ - 4 ,
._ .
,
: SENIOR REACTOR OPERATOR Page 58 i
QUESTION: 093_ . (1.00) Tbo--plant was operating at 100% power when the " MAIN STEAM LINE HI-RADIATION SCRAM," (C905R,-A5). alarm was received due to a' valid signa , The expected automatic actions did NOT_ occur and have NOT been "{ accomplished manually. -Torus water temperature is 115*F and reactor power _is 18%. Main Stack radiation levels are.75000 cps and have been; steady for the last 20 minutes. WHICH ONE of the following describes the appropriate actions for RPV pressure control? Perform Alternate RPV Depressurization using the SRV Depressurize the RPV at less than 100*F/hr using the main turbine bypass valve c. Maintain reactor _ pressure less than 1085 psig using the main ' turbine bypass valve Maintain reactor pressure lens than 1085 psig using HPCI in full flow test and the SRVs.
s QUESTION: 094 (1.00) Step TT-4 of EOP-03, " Primary Containment Control," directs concurrent performance of EOP-01, "RPV Control," before torus water temperature reaches the Boron Injection Initiation Temperature-(BIIT). WHICH ONE of-the following~is the reason for performing EOP-01 before torus water temperature reaches the BIIT? a. To reduce the potential for a-condition that would compromise adequate core cooling which could preclude the use of RHR for torus _ coolin b. To ensure that the torus will be able to accept the-heat addition from.the RPV while.the reactor is being shutdown 1 without exceeding the Heat Capacity Temperature Limit.
' c. To reduce the potential heat load on.the torus by ensuring that * the turbine bypass valves are used for-pressure control and ADS
- is inhibite d.-To assure cyclic condensation (chugging) does not occur when'the RPV is depressurized by Alternate RPV Depressurization if the Heat Capacity: Temperature Limit is exceede !.
1 y f y w w% y w w . e +- -, +~- --
: SENIOR REACTOR OPERATOR Page 59 QUESTION: 095 (1.00)
Following a LOCA, HPCI is-the only-injection-source available. HPCI.is-maintaining RPV level at -50 inches with' suction from the CST. The HPCI torus suction valves are stuck closed. For WHICH ONE of the following conditions would HPCI have to be secured irrespective of adequate' core cooling in accordance with EOP-p,3, " Primary Containment Control?" a. Torus water level: 80 inches RPV pressure: 600 psig b. Torus water level: 250 inches RPV pressure: 1000 psig c. Drywell water level: 30 feet RPV pressure: 400 psig d. Torus water level: 200-inches RPV pressure: 800 psig l $ .
. = -, :_
-
SENIOR REACTOR. OPERATOR- Page-60 QUESTION: 096 (1.00) WilICll-ONE of the following conditions would require Alternate RPV Depressurization? a. RPV Pressure: 1050 psi Torus Bottom Pressure: 15 psig Torus Water Temp: 120 *F Torus Water Level: 100 inches DW: Temperature: 200 *F RPV Pressure: 900 psig Torus Dottom Pressure: 10 psig Torus Water Temp: 140 "F Torus Water Level: 85 inches DW Temperature: 170 'F RPV Pressure: 850 psig Torus Bottom Pressure: 45 psig Torus Water Temp: 175 'F Torus Water Level: 220 inches DW Temperature: 275 *F d. RPV Pressure: 1000 psig Torus Bottom Pressure: 20 psig Torus Water Temp: 160 'F Torus Water Level: 190 inches DW Temperature: 225 *F
~. ;
!
SENIOR REACTOR OPERATOR Page 61 QUESTION: 097 (1.00) The following conditions exist in primary containment:
- Torus bottom pressure is 50 psig - Torus water level is 200 inches - Drywell hydrogen concentration is 5% - Drywell oxygen concentration is 7% - Torus hydrogen concentration is 3% - Torus oxygen concentration is 4%
WHICil ONE of the following describes the app'.3priate actions in accordance with EOP-03, " Primary containment control?" i a. Vent the drywell and espablish a nitrogen purg Vent the torus and establish a nitrogen purg Vent the drywell and establish an air purge, Vent the torus and establish an air purg _
.. -- _ __ _ ___ _m __.._ _._____-_-_ __m-___-__._.__ _ _ _ _ . _ _ . _ _ _ . - _ _ _ _ . _ - _ . _ . - _ - _ - _ _ _ _ _ . _
.. _ . - _ . . . . _ . - _ . . __ . _ _ , SENIOR REACTOR OPERATOR Page 62
QUESTION: 098 (1.00)
'
Following a Loss of Coolant Accident (LOCA), current plant conditions-are:
. - Torus bottom pressure is 10 psig - !!ydrogen concentration is 5% - Oxygen concentration is 6%
Chemistry reports that projected offsite dose rates are 2 R/yr whole bod WHICH ONE of the following describes the correct action that should be taken to lower containment pressure-and hydrogen and oxygen levels? (Attachment 1 of PNPS Proc. No. 5.4.6, " Primary Containment Venting and Purging Under Emergency Conditions," is attached.)
a. Do not vent the toru b. Vent the torus through SGTS using the 2" torus exhaust valves onl Vent the torus through SGTS using the 2" torus exhaust valves and the 8" torus exhaust valve Vent the torus, bypassing SGTS.
l l l [ ! ! !
SENIOR REACTOR OPERATOR Page 63 QUESTION: 099 (1.00) The reactor is shutdown with all control rods inserted. WHICH ONE of the following conditions would require Primary Containment Flooding? RPV Pressure: 800 paig RPV Water Level: -170 inches Torus Pressure: 10 psig No SRVs Open b. RPV Pressure: 80 psig RPV Water Level: -140 inches Torus Pressure: 40 psig 1 SRV Open
- RPV Pressure: 90 psig RPV Water Level: Cannot Be Determined Torus Pressure: 45 psig 2 SRVs Open RPV Pressure: 60 psig RPV Water Level: Cannot Be Determined Torus Pressure: 25 psig 3 SRVs Open ,
QUESTION: 100 (1.00) A primary system is discharging into secondary containmen WHICH ONE of the following conditions would require Alternate RPV Depressurization? _ a. HPCI Turbine Area temperature: 200 *F HPCI Piping Area - 23 ft El. temperature: 310 *F HPCI Torus Piping Area temperature: 200 *F RCIC Turbine Area temperature: 230 *F RCIC Piping Area - 23 ft El. temperature: 200 F RCIC Torus Piping Area temperature: 270 *F RWCU Backwash Tank Area temperature: 210 *F RWCU Heat Exchanger Area temperature: 220 *F RB West Area - 51 ft El. radiation level: 1200 mr/hr RWCU & RHR Piping Area - 23 ft El. temp: 260 F CRD Quadrant water level: 8 in, above floor CRD Pump Room - 17 ft 6 in El. rad level: 1200 mr/hr (********** END OF EXAMINATION **********)
SENIOR REACTOR OPERATOR Page 64 ANSWER: 001 (1.00) c.
' REFERENCE: IG: -RO-02-11-01, " Control Room Computer System (EPIC /SPDS)," ELO (3.2/3.4] _ 294001A115 ..(KA's) ANSWER: 002 (1.00) REFERENCE: PNPS Proc. No. 6.1-024, " Radiological Posting of Areas of the Station." Module C-GT-02-02-01, " Introduction to Radiation Protection," ELO '
[3.3/3.8]
294001K103 ..(KA'c) ANSWER: 003 (1.00) b
, c'.
_ _ - _ _ _ _ _ . . . _ .
_ . - ._ _ . _ _ _ . _ . _ . . _ . . . - . _ . _ . - . _ . _ . . .- . . _ _ . ~_
- ' SENIOR-REACTOR OPERATOR , Page 65 REFERENCE:' l Hodule C-GT-02-02-02, " Exposure Limits," ELO .i (3.3/3.8) .
294001K103 ..(KA's) U ANSWER: 004 (1.00) , REFERENCE: PNPS Proc. No. 1.4.9, " Storage, Handling, and Disposal of Sodium Pentaborate," page . UG; O-RO-04-04, " Emergency Tasks," Task 9 [3.1/3.4] 294001K110 ..(KA's)
ANSWER: 005 (1.00) REFERENCE: , ' 1.: PNPS Proc. No. 1.4.36, "Iligh Pressure / Compressed Gas Cylinder Control."- i IG: C-GT-01-01-03, " Industrial Safety," LO 3 , j _ {3.4/3.8] 294001K109 . . ( '.(A ' s ) i
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. REFEREllCE: PilPS Proc. tio . 1.3.34, " Conduct of Operations," pages 38 and 39, {3.7/3.7)
294001K101 . . (KA's) AtiSWER: 007 (i.00) C.
. REFEREtiCE: PilPS Proc. tio . 1.4.5, "PilPS Tagging Procedure. " UG: 0-RO-04-02, " Administrative Tasks," Task 8.
!
(3.9/4.5)
294001K102 . . (KA's) l . AtiSWER: 008 (1.00) ! ; I-l::
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SEllIOR - REACTOR' OPERATOR- : Page'67-
.. REFEREllCE: PflP8 Proc. !!o. 1.3.67, " Control of overtimo."- . 2, UG: 0-RO-04-20, "5RO On-Shift Tanko," Tank 2 [2.7/3.7)
294001A103 . . (KA's)
-. AllSWER: 009 (1.00)
'i i REFl:HEllCE: P!lPS Proc. 11 .4.12, " Primary Containment Entry." liodirim! Facility Quention ADfif tl-3 2 ( f rom - Hequal itetake Exam) . >
,
[3.2/3.6)
294001K113 . . <MA'n) AtlSWER: 010 (1.00) a.
t . REFElmilCE:
' ' PflPS Technical Specificationa, Tablo 6.2- . -JG: 0-RO-06-01-04, " Technical Specif'ication Donign Factora and -
Admin Controlo," - ELO ti.
- [3.!I/3.8)
294001K11 . . (VA'u) , 1'4- w 4 E. - .-.g.[e~ _ _ , - < . ..,...,--r--. ,,,..u-,m--- E y - ,d c y -c .g-,, w- w r r -- -+ir'e-~ - * * * - '* -+ Y
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Page 68~
,
ANSWER:. 011 -(1.00) b.- REFERENCE:
- P11PS Proc. No. 2.4.143, Rev. 12, " Shutdown from Outside Control-Room," page . OJT Guido 0-RO-04-10: " SRO On-Shift Tasks," Task 7 [3.6/4.2]
294001A110 ..(KA's) ANSWER: 012 (1.00) REFERENCE: PNPS Proc. No. 1.5.9.1, " Lifted Leads and Jumpers," page 6.
? Modified question from 12/2/91 NRC Exam (replaced correct answer, reworded distractors).
[3.3/3.6]
'294001K107 ..(KA's)
ANSWER: 013 _(1. 0 0 ) ' _ c_.
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.l SENIOR' REACTOR OPERATOR Page 6 l l
REFERENCE:
- 1 '. . PNPS Proc. No. 1.3.34.2, " Limiting Condition for Operations Log,"
page . -. PNPS Technical Specifications, Table 3. . UG: 0-RO-04-10, "SRO On-Shift Tasks," Task 8 [3.4/3.6]- 294001A106 ..(KA's) ,. ANSWER: 014 (1.00) REFERENCE:
- IG: 0-RO-03-04-02, "EOP Development and Use," ELO [3.4/3.6)
294001A106 ..(KA's) ANSWER: 015 '( 1 .00) ' REFERENCE: IG: 0-RO-03-04-02, "EOP Development and Use," page IG-20,- ELO 3n.
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[4.2/4.2]
294001A102 ..(KA's)
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REFERENCE: 6 IG: T-ER-01-01-80, " Dose Assessment / Protective Action Recommendations," ELOs 4 and ('e .9/4.7]
,
294001A116 ..(KA's) , i ANSWER: 017 (1.00) REFERENCE: IG: T-ER-01-01-80, " Dose Assessment / Protective Action Recommendations," ELO 1 [2.9/4.7) 294001A116 ..(KA's) ANSWER: 018 (1.00) d.
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SENIOR REACTOR. OPERATOR' Page 71
. REFERENCE: IG: 0-RO-02-06-01, "Non-Nuclear Instrumentation and Reactor Vessel Internals," ELO 1 . Systems Reference Text: Nuclear boiler Instrumentatio . LO Requalification IG: Reference Leg Perturbations Due to Non-Condensable . Modified questions from 11/26/90 and 12/2/91 NRC Exams (combined concepts of questions and changed correct answer). , [3.3/3.5]
216000A210 ..(KA's) ANSWER: 019 (1.00) REFERENCE: IG: 0-RO-02-07-02, " Reactor-Protection System and-Ant'icipate Transient Without Scram System," ELO 2 [3.8/4.2] 212000G010 ..(KA's) ANSWER: 020 (1.00) b.
l
. _ . . ._ _ ._ _ , ,
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SEllIOR REACTOR OPERATOR Pt.go 72 REFEREf1CE: 1._ - P11PS Proc. Ilo. 2.2.79, "Itonctor Protection System," pago 1 .1 -IG: 0-RO-02-07-02, "Roactor Protection System and Anticipated Transient Without Scram System," Eth-1 . Facility Question TYPA-13 (from Requal Retake Exam).
- ( 3. 9 / 4 .1 ) 212000K412 ..(KA'n) ! AllSWER: 021 (1.00) f RI'FEREllCE:
- Pill'S
- A RP-9 0 511-D 4 , Rev. 8, IG: 0-RO-02-04-01, "!4ain Steam System," page IG-19-7/90, ELos 18 and 211, tiodi fied question f rom 11/26/90 t1RC Exam (changed conditionn).
.
[4.0/4.1)
239001K127 ..(KA'n) ,: AllSWER: 022 (1.00) a.
l l REFl:REllCI:: !- IG; 0-RO-02-08-01, '_' Primary Conta inment System, " ELOn 33 and 3 . tiodified questions-from 11/26/90 and 12/02/91 liRC Examn (changed conditionn, dintractors,-and correct-answer).
.[3.5/3.5)
-223002A302 ..(KA's)
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SENIOR REACTOR OPERATOR -Page 73 _
' ANSWER: 023 (1.00) REFERENCE: IG: 0-RO-02-09-04, " Reactor Core Isolation Cooling System," ELO 1 . Modified question from 11/26/90 NRC Exam (changed conditions, distractors and correct answer) . [3.8/3.8]
217000A215 ..(KA's) ANSWER: 024 (1.00) . REFERENCE: IG: 0-RO-03-04-02, "EOP Development and Use," ELO 22.
.
[3.9/3.8]
206000G010 ..-(KA's) ANSWER: 025 ( 1. 0 0)- C.-
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SENIOR IREACTOR OPERATOR Page'74- t
: REFERENCE: * :D3: 0-RO-02-09-02, " Core Spray System," ELO 1 '
(3.3/3.6) i 209001A205 ..(KA's)
,
ANSWER: '026 (1.00) c REFERENCE:
- IG: 0-RO-02-09-01, " Low Pressure Coolant Injection and Residual Heat Removal," ELOs 9 and 1 . Modified questions from 11/26/90 and 12/2/91 NRC Exams (changed conditions).
[4.2/4.3] 203000A101 ..(KA's) ANSWER: 027 (1.00) d.
' REFERENCE: PNPS Proc. No. 2.2.19, " Residual Heat Reinoval," page 1 . UG: .0-RO-04-04, " Emergency Tasks," Task 89 ._ Modified question from 11/26/90 NRC Exam (changed values in distractors).
[3.2/3.4] 226001G010 ..(KA's)
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SENIOR REACTOR OPERATOR !Page 75-ANSWER: 028 (1.00) REFERENCE: PNPS Proc. 2.2.19, " Residual licat Removal,"_page 2 . IG: 0-RO-02-09-01, " Low Pressure Coolant Injection and Residual IIcat Removal System," ELO 14, Derived f rom Facility Questions 10 -B and RilR/SDC-0 [4.0/3.9] 230000A406 . . (KA's) , ' ANSWER: 029 (1.00) c.
I REFERENCE: 1.
( IG: 0-RO-02-09-05,." Automatic Depressurization System," ELOs 5 and ' 15.
l- [4.2/4.3] 218000A206 . . ( K.' ' s ) ANSWER: 030 (1.00) __
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. CREFERENCE: -
*
,
IG:. O-RO-02-09-05,-" Automatic Depressurization-System," ELO 1 ~ 1.= -
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; ~239002K405 -. .(KA's)- ;
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' ANSWER: 031 (1.00)
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- REFERENCE: . PNPS Technical Specification 3.10.B.-
-2., .OJT Guide N , SRM System Objective (3.2/3.9).
, '
.
-215004G011
- . . ( KA _' s )
- ANSWER: 032 (1.00) -;
.
c.- JREFERENCE: . 1-2 J -IG: ; O -RO-02-07-01, " Neutron . Monitoring Syst. ems, "' ELOs 18, - 23, and - Et
' 2 -4 . [3.5/3.7]- @ 215003A202 . .(KA's): -
.
4
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SENIOR REACTOR OPERATOR Page 77 ANSWER: 033 (1.00) a.
REFERENCE: IG: 0-RO-02-07-01, " Neutron Monitoring Systems," ELOs 40, 42, 44, and 4 [3.4/3.7]
-
215001A207 ..(KA's) ANSWER: 034 (1.00) REFERENCE: PNPS Proc. No. 2.1.15, " Daily Surveillance Log (Tech Specs and Regulatory Agencies) ," Daily Log Test #2 . UG: 0-RO-04-02, " Administrative Tasks," Task . Modified question from 12/2/92 NRC Exam (changed conditions, distractors, and correct answer). -
[3.0/3.4)
215005A107 ..(KA's) ANSWER: 035 (1.00) .. - - _ . . . - _ . . _ _ - . . . _
. . . _ _ . - . . . _ _ . . . . __ -_ . _ _ _ - . . . . SENIOR REACTOR OPERATO Page.78 REFERENCE:
1.- IG: 0-RO-02-07-01, " Neutron Monitoring Systems," ELOs 63', 64, and-6 [3.2/3.4) 215005A207 ..(KA's) ANSWER: 036 (1.00) REFERENCE: IG: 0-RO-02-07-01, " Neutron Monitoring Systems," ELO 5 . Facility Question PSU-24 (from Requal Retake Exam).
(3.3/3.3) 215002A202 ..(KA's) ANSWER: 037 (1.00) REFERENCE: IG: 0-RO-02-06-04, " Rod Position Information System,'d ELO 11.d.
.
[3.1/3.3)
214000A201 ..(KA's)
-- . . .- - .
. - .
. . . . . . . . - . . . - - . - - - . - . . - _ . - . - _ . = . . .
. _ - . . . ..
SENIOR REACTOR OPERATOR Page 7 ~ ANSWER: 038 (1.00) 1 a fb REFERENCE: IG: 0-RO-02-06-03, " Control Rod Drive System," ELO 2 '
> (3.5/3.5]
201006K403 ..(KA's)
l' ANSWER: 039 (1.00) REFERENCE: IG: 0-RO-02-04-02, " Condensate and Feedwater System," ELO 2 [3.3/3.4] 259001A204 ..(KA's) ANSWER: 040 (1.00) c.
i i l i l l l l l
- .m, -. - -
. . . . - . .. ...-_. - - . . - . , ,
SENIOR REACTOR OPERATOR Page 80 REFERENCE: IG: 0-RO-02-04-02, " Condensate and Feedwater System," ELO 1 !
[3.6/3.7)
256000K304 ..(KA's) ANSWER: 041 (1.00) , REFERENCE: IG: 0-RO-02-06-02, " Condensate and Feedwater System," ELO 8 . Modified question from 11/26/90 NRC Exam (modified distractors and correct answer).
[3.1/3.1) 259002K604 ..(KA's) ANSWER: 042 (1.00) I REFERENCE: IG: 0-RO-02-05-01, " Main Turbine System," ELOs 70, 79, and 8 [ 4 .1/ 4 .1 ) 241000K306 ..(KA's)
SENIOR REACTOR OPERATOR Page 81 ANSWER: 043 (1.00) C.
REFERENCE: IG: 0-RO-02-01-03, " Main Generator," ELO 1 . IG: 0-RO-02-05-01, " Main Turbine System," page IG-34-5/8 . Modified questions from Id/26/90 NRC Exam (combined 2 questions).
[2.7/2.8) l
--
245000A304 ..(KA's) ANSWER: 044 (1.00) c.
REFERENCE: IG: 0-RO-02-04-03, " Main Condenser Vacuum and Augmented off Gas Systems," ELos 2, 9i, 18d, and 18 [3.5/3.9] _ 271000A206 ..(KA's) ANSWER: 045 (1.00) e
, . - . . ~ ,- - . - - - . - , . ~ . _ - - . ~ - . - . - - . . . . - . . - -
SENIOR' REACTOR OPERATOR 'Pago U2
~ REFERENCE: IG: 0-RO-02-06-02, " Recirculation System," ELOs 9, 14, 15, 18, and ' 2 0. . [3.3/3.4)
202001K410 ..(KA's) ANSWER: 046 (1.00) REFERENCE: IG: 0-RO-02-06-02, " Recirculation System," ELO 3 [3.5/3.5] 202002K604 ..(KA's) l l ! ANSWER: 047 (1.00) l- , b.
l ' REFERENCE: l i-l IG: 0-RO-02-09-06, " Diesel Generator-System," ELO 2 [3.8/3.7) 264000K408 ..(KA's) i i
+ ,. - , - - , . -, .,+ ,
- - . - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _
SENIOR REACTOR OPERATOR Page 83 ANSWER: 048 (1.00) pr i REFERENCE: IG: 0-RO-02-0 , " Standby Gas Treatment System," ELos 9, 13, and 15 [2.9/3.% .
~_
2610 A203 ..(KA's) ANSWER: 049 (1.00) c.
< REFERENCE: IG: 0-RO-02-08-05, " Plant Ventilation Systems," ELO . Modified question from 11/26/90 NRC Exam (changed correct answer and conditions on all distractors).
-
[3.6/3.9] ,
272000K403 ..(KA's) ANSWER: 050 (1.00) h._. .
SENIOR REACTOR OPERATOR Page 84' REFERENCE: IG: 0-RO-02-08-05, " Plant Ventilation Systems," ELOs 9, 10, and 1 . Expanded Previous Question from 12/02/91 NRC Exa [3.3/3.4) 290001G010 ..(KA's) ANSWER: 051 (1.00) REFERENCE: IG: 0-RO-02-08-OS, " Plant Ventilation Systems," ELOs 5 and 9, Modified question from 11/26/90 NRC Exam (added concept and changed answers).
(3.1/3.3) 290003A204 ..(KA's) ANSNER: 052 (1.00) .
-
REFERENCE: IG: 0-RO-02-09-01, " Low Pressure Coolant Injection and Residual lleat Remova1 System," ELO 1 . IG: 0-NL-03-11-02, " Fuel Pool Cooling," ELO [2.9/3.0) 233000K102 ..(KA's)
set 1IOR REACTOR OPERATOR Page 85 AllSWER: 053 (1.00) , REFERE!1CE: IG: 0-RO-02-08-06, " Fuel llandling Equipment," ELO [3.3/4.1) 234000K402 ..(KA'n)
-
ANSWER: 054 (1.00) REFERE!1CE: IG: 0-RO-02-10-01, " Fire Protection System," ELon 15, 16, 17, and 1 [3.8/3.9) 286000G004 ..(KA'n) k@b * ANSWER: 055 p(-1[00)
/ /
i . .
. . _ _ . _ . _ _ _ . _ . . . _ _ .
SENIOR REACTOR-OPERATOR Page 86 , l REFERENCE:-- - IG: 0-RO-02-09-03, ."Hi Pressure Coolant Injection," page 3 ; PNPS- Technical Speci ' cations, section 3. . IG 0-RO-06-01-03 ' Limiting Conditions for Operations," ELO [3.6/4.3] 206000G00r ..(KA's) t ANSWER: 056 (1.00) REFERENCE: PNPS Technical Specifications, Section . IG 0-RO-06-01-03, " Limiting Conditions _for Operations," ELO [3.6/4.4] 211000G005 ..(KA's) ANSWER: 057 (1.00)
' REFERENCE: PNPS Technical Specifications, Sections 3.9-and 3. . IG: 0-RO-06-01-03, " Limiting Conditions for Operations," ELO [2.9/3.9]
- . .
l 262001G005 ..(KA's) l ! l.
I-
._
. ....__ -._ _. . . . _ . . _ .-_ _- .. . .._... . . . . . . _ . _ . . _ . - . . . . 4 i SENIOR REACTOR OPERATOR Page 8 T ' ANSWER: 058 (1.00) REFERENCE: PNPS Technical-Specifications,_ sections 2.0 and 3.1 . PNPS Proc. No. 1.3.6, " Technical Specification _ Adherence and Clarification," page . OJT Guide No. 13,_ Process Computer System Objectis. 2 [3.5/4.3)
295014G003 .
.(KA's)
ANSWER: 059 (1.00) REFERENCE: PNPS Technical Specification 3. . IG: 0-RO-06-01-03, " Limiting Conditions for Operations," ELos 3 andL [3.4/4.1] . 295018G008 . .(KA's) ANSWER: 060 (1.00)
.a.
4 e v e e
- . - . . ..
SENIOR REACTOR _ OPERATOR Page 88 REFERENCE: PHPS Technical Specification 3. . IG: 0-RO-06-01-03, " Limiting Conditions for Operations,"-ELO (2.9/3.8) 295023G003 ..(KA's)
>
ANSWER: 061 (1.00) REFERENCE: I PNPS Proc. No. 2.4.29, " Stuck Open Safety / Relief Vavle," page . OJT Guide N , ADS System Objective 2 [2.9/4.1] 295026K305 ..(KA's) ANSWER: 062 (1.00) , REFERENCE: i EP-IP-100, " Emergency Classification," sections 4.1 and . IG: T-ER-01-01-30, " Classifications," ELO 5.
l l (3.1/4.5)
295033G002 ..(KA's) L l-l
' SENIOR REACTOR.OPERATORl Page 89-ANSWER: 063 (1.00) REFERENCE: .
- EP-IP-100, " Emergency Classification," sections 6.2, 7.1, and . IG: T-ER-01-01-30, " Classifications," ELO [3.1/4.5)
295016G002 ..(KA's) _ ANSWER: 064 (1.00) REFERENCE: EP-IP-100, " Emergency Classification," sections 2.2, 2.3, 3.2, and . IG: T-ER-01-01-30, " Classifications," ELO (2.9/4.2] 295029G002 ..(KA's) ANSWER: 065 (1.00) .- _ - _ _ .- - - - - -
SENIOR REACTOR OPERATOR Page 90
=
REFERENCE: I t- PNPS Technical Specifications, sections 1.0.N and 3. . IG: 0-RO-06-01-01, " Technical Specification Definitions," ELO . IG: 0-RO-04-09, " Technical Specification Overview," ELO [3.9/4.2) 295035K101 ..(KA's) ANSWER: 066 (1.00) _ _ _ REFERENCE: PNPS Technical Specifications, section . IG: 0-RO-06-01-02, " Safety Limits and Limiting Safety System Settings," ELO [3.5/4.3) 295025G003 ..(KA's) - ANSWER: 067 (1.00) _ REFERENCE: PNPS Technical Specifications, 3.6.E and 4.6.E Bases, page 147 . OJT Guide N , Recirculation System objective 2 [3.9/4.2] 295001G011 ..(KA's)
m ___ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . - .
!
SENIOR REACTOR OPERATOR Page 91 ANSWER: 068 (1.00) b.
REFERENCE: PNPS Proc. No. 2.4.17, "Repirculation Pump (s) Trip," page . OJT Guide 4, Recirculation System Objective 2 [3.5/3.8) 295001A201 ..(KA's) ANSWER: 069 (1.00) d.
REFERENCE: PNPS Proc. No. 5.3.8, Rev. 16, " Loss of Instrument Air," page' . IG: 0-RO-02-02-04, " Instrument and liigh Pressure Air," ELO . Modified question from 12/2/91 NRC Exam (changed conditions).
[3.6/3.7] 295019A202 ..(KA's) ANSWER: 070 (1.00) - .
, -, , . - - - -
SENIOR REACTOR OP"RATOR Page 92 REFERENCE: IG: 0-RO-02-04-01, " Main Steam System," ELO 2 [3.6/3.7) 295020K201 ..(KA's) ANSWER: 071 (1.00) ___ REFERENCE: IG: 0-RO-0208-01, " Primary Containment System," ELO 3 [3.6/3.6) 295020A101 ..(KA's) ANSWER: 072 (1.00) REFERENCE:
- PNPS Proc. No. 2.4.42, " Loss of RBCCW," pages 4 and . IG: 0-RO-02-02-06, " Reactor Building Closed Cooling Water," ELos 2, 3, 8, and 1 . OJT Guide No. 2, RBCCW System Objective 2 [3.3/3.4)
295018K201 ..(KA's) - . . _ .
>
SENIOR REACTOR OPERATOR Page 93
.
ANSWER: 073 (1.00) REFERENCE: IG: 0-RO-02-02-02, " Salt Service Water System," ELO (4.4/4.4) 295003A103 ..(KA's)
-
ANSWER: 074 (1.00) REFERENCE: PNPS Proc. No. 1.3.4, " Procedures," page 17, PNPS Proc. No. 2.4.36, "Docreasing Condenser Vacuum," page . IG: 0-RO-03-04-02, "EOP Development and Use," ELos 13 and 1 . IG: 0-RO-02-07-02, " Reactor Protection System and Anticipated Transient Without Scram System," ELO 1 . IG: 0-RO-02-05-01, " Main Turbine System," ELO 2 . Modified questions from 11/26/90 and 12/2/91 NRC Exams (changed - conditions).
[3.8/3.7) 295002G010 ..(KA's) ANSWER: C75 (1.00)
,
d.
-
SENIOR REACTOR-OPERATOR Page 94 REFERENCE: PNPSLProc. No. 2.4.150, " as of Feedwater Heating," page ._ OJT Guido N , Feedwate lleating System Objective 2 (4.0/3.9) 295014G010 ..(KA's) ANSWER: 076 (1.00) REFERENCE: PNPS Proc. No. 2.4.4, " Loss of CRD Pumps." PNPS Proc. No. 2.4.11, " Control Rod Positioning Malfunctions." PNPS Proc. No. 2.4.11.1, "CRD System Malfunctions." . OJT Guido No. 3, CRDM System Objective 2a and CRD llydraulic System Objective 2 p (3.7/3.5] 295022G010 ..(KA's) ANSWER: 077 (1.00) C.
.
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'
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- .. . - . ~ - - _ - - . ~ . . . - . - _ . - . . . . - - . - . - .. .. [ SENIOR REACTOR [ OPERATOR Page 95
= REFERENCE: 1, .- PHPS Proc.-No. 5.4.3, Rev. 10, Refueling Floor liigh Radiation," page . IG: 0-RO-06-04-01, " Fuel llandling Operations and Supervision," ELO - (3.8/3'.9] 295023G010 ..(KA's) . ANSWER: 078 (1.00) , REFERENCE: PNPS Proc. No. 5.3.6, " Loss of Vital AC (Y-2)."
' PNPS Proc. No. 5.3.30, " Loss of 250V DC Bus D-10." OJT Guide N , 120/240 VAC System Objective 2b and 250 VDC System Objective [3.5/3.5) 295004G005 ..(KA's) ANSWER: 079 (1.00) b
. . - . - . _ =. .
. . .- , . _ . .. . . . . .. - ... _ . .- - .. . .. . . ..-.. . . _ . ~ . -.. - - . . . - . $ENIOR REACTOR OPERATOR Page 96 - REFERENCE:
' PNPS-Proc. No. 5.3.31, " Station Blackout," page . PNPS Proc. No. 5.3.26, "RPV Injection During Emergencios," page 2.
. 3.- OJT Guide No'. 1, 480/208. VAC System Objective 2 [4.4/4.4] 295003A103 .
.(KA's)
ANSWER: 080 (1.00) REFERENCE: PNPS Proc. No. 2.4.143, Rev. 12, " Shutdown from outside Control
'
Room," page 1 . OJT Guide O-RO-04-04, " Emergency Tasks," Task 6 [4.1/4.1) 295016G006 .
.(KA's)
ANSWER: 081- (1.00) REFERENCE: " UG: 0-RO-04-04, " Emergency Tasks," Tasks 1,-2, and (Derived from facility question PRO-1'from proposed requal retake exam.)
,
[4'.3/4.4]
295006A202 . .(KA's) ~
.
? i
- .., ..e .
.w_ , -
. . . . . - . - . - - . . . - . . . - . . . . - . ~ . - . - - . . _ .
. . - .
-SENIOR REACTOR OPERATOR -Page 97 +
.
ANSWER: 082 '( 1. 0 0)
- REFERENCE:
' IG: 0-RO-02-07-01, "Noutron Monitoring Systems," ELOs;14 and 1 . IG: 0-RO-03-04-04, "EOP-02, Failure to Scram," ELO 1 [4.2/4.3]
295037A201 ..(KA's) ANSWER: 083 (1.00) REFERENCE: PNPS Proc. No, S.3.23, Rev. 9, " Alternate Rod Insertion," pago . OJTPG: 0-RO-04-04, " Emergency Tasks," Task [4.0/4.1] 295015K204 ..(KA's) ANSWER: 084 (1.00) _
-
l SENIORfREACTOR OPERATOR .Page_98 i REFERENCE: PNPS Proc. No. 5.3.23, Re , "Alternato Rod Insertion," pages 6 ~ 1 ! OJTPG: 0-RO-04-04, " Emergency Tasks," Task H Modified question from 12/2/91 NRC Exam (modified distractors). q (3.8/3.9)
-l l
295015A101 ..(KA's) ANSWER: 085 (1.00) d.
REFERENCE: IG: 0-RO-03-04-05, Primary Containment Control,", page IG-5, ELO 1 . OJTPG: 0-RO-04-04, " Emergency Tasks," Task 89 . Modified question from 12/2/91 NRC Exam (replaced 1 distractor).
'
[3.6/4.3]
295012G012 ..(KA's) ANSWER: 086 (1.00) a.
REFERENCE: IG: 0-RO-03-04-05, Primary Containmept Control,", pago IG-1 . OJTPG: 0-RO-04-04, " Emergency Tasks," Task 89 [3.8/4.5] 295026G012- ..(KA's)
,
_ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
'
SE1410R REACTOR OPERATOR Page 99 ANSWER: 087 (1.00) REFERENCE: IG: 0-RO-02-08-01, " Primary Containm ant System," ELO 39 [3.6/3.9) 295024G007 ..(KA's) - ANSWER: 088 (1.00) REFERENCE: PNPS Proc. No. 5.3.27, " Determining Primary Containment Water Level." OJT Guide No. 10, " Primary Containment System Structure Objective 2 . IG: 0-RO-02-08-01, " Primary Containment System," ELOs 13 and 1 [3.4/3.5) 295029A203 ..(KA's) ANSWER: 089 (1.00) c l . __-_ _ _ - - _ - - _ _ - _ - _ _ _ - _ - _ _ _ _ _ - _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ - - - _ _ -
. - . .. . ..
LSENIOR REACTOR OPERATOR Page100 REFERENCE:
- PNPS Proc. No. 5.3.26, "RPV Injection During Emergencies."
- ' Facility Question 5.3.26-1 (Proposed Requal Retake Exam).
[4.1/3.9) 295031G006 ..(KA's) ANSWER: 090 (1.00) REFERENCE: IG: 0-RO-03-04-07., "EOP Development and Use," ELO 2 . Modified question from 12/2/91 NRC Exam (changed conditions).
[3.8/4.3) 295028G012 ..(KA's) ANSWER: 091 (1.00) REFERENCE: IG: 0-RO-03-04-02, "EOP-02, Failure to Scram," page IG-15, ELO 1 [4.1/4.3] 295037K101 ..(KA's) ANSWER: 092 (1.00) . - -
. - _ _ _ - _ . _.._.... _ _ _ _ _ _ _ . . . _ --_ _ - - _ . .
SENIOR-REACTOR OPERATOR 'Page101
. REFERENCE:-
1.- IG:-0-RO-03-04-08, "EOP-16 and EOP-26, RPV Flooding," ELOs-2 and 1 [3.9/4.6] 295037G012 ..(KA's) b ANSWER: 093 (1.00) REFERENCE: IG: 0-RO-03-04-04, "EOP-02, Failure to Scram," pages IG-33 and IG-41, ELO 22 . IG: 0-RO-03-04-07, "EOP-05, Radioactivity Release Control," ELO 6.
'
(3.9/4.6)
295037G012 ..(KA's) ANSWER: 094 (1.00) l REFERENCE: IG: 0-RO-03-04-05, "EOP-03, Primary Containment Control," page IG-1 . IG: 0-RO-03-04-03, "EOP-01, RPV Control," ELO l Modified question from 12/2/91 NRC Exam (changed correct answer and ' modified distractors).
(3.9/4.1] 295026K305 ..(KA's) ! { l L l
._ , ,.
SENIOR REACTOR OPERATOR Page102 ANSWER: 095 (1.00)- a.- I REFERENCE: EOP-03, " Primary Containment Control." , IG: 0-RO-03-04-05, " Primary Containment Control," ELO 18, Modified question from 12/2/91 NRC Exam (modified to require use of EOP flowchart).
[4.6/4.7) _ 295031K101 ..(KA's) ANSWER: 096 (1.00) REFERENCE: EOP-03, " Primary Containment Control." IG: 0-RO-03-04-05, "EOP-03, Primary Containment Control," ELos 1 and 1 . r
[3.7/4.4]
295030G012 ..(KA's) ANSWER: 097 (1.00) d -. i l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . -
. . . . - . , . .. . - . - _ . - . . . .
_ SENIOR REACTOR _ OPERATOR _ Page103- ,
)
REFERENCE:c 1.: 'IG: 0-RO-03-04-05, "EOP-03, Primary Containment Control," ELO 1 [3.9/4.5) 295024G012 ..(KA's) ANSWER: 098 (1.00) REFERENCE: IG: 0-RO-03-04-05, "EOP-03, Primary Containment Control," ELO 18, IG: 0-RO-03-04-07, "EOP-05, Radioactivity Release Control," ELos 3 and . LP-IP-100, " Emergency Classification," Attachment 1, section 5, [3.9/4.5] 295038G012 ..(KA's) ANSWER: 099 (1.00) Or 6 REFERENCE: EOP-01, "RPV Control." EOP-16, "RPV Flooding."
3.- IC: 0-RO-03-04-03, "EOP-01, RPV Control," ELO 2 . IG: 0-RO-03-04-08, "EOP-16 and EOP-26, RPV Flooding," ELO [3.9/4.5] 295031G012 ..(KA's)
. . . . _ _ _ . . . _ . ~ _ _ _ . . _ . . _ _ _ _ . . .._m.______~.._ _ .
. _ _ _ _
, SENIOR REACTOR OPERATOR Patje104 :
!
,
i
+
ANSWER: 100 (1.00) ! i l i a.
1 REFERENCE: ; i . EOP-04, " Secondary Containment Control." , IG: 0-RO-03-04-06, "EOP-04, Secondary Contajnment Control," ELO 1 !
!
,
!
(3.0/4.4) ;
-
i 295032G012 ..(KA's) i
!
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(**********'END 0F_ EXAMINATION.**********)
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. . . . _ . . . . . _ . _ _ _ _ . . _ _ _ . _ . _ . . . _ - .
-
. . . _ . . , .
~ _ _ _ _ _ _. _..-. _ ..._.._ -.__ _ _.. _ SENIOR REACTOR OPERATOR Paga 1 AHSWER KEY
!
, I_ t
.
HULTIPLE CilOICE 023 b
001 yd 024 c 002- b 025 c ; I 003 /b 0;!6 c ,
- 004 a 0;!7 d 005 b 028 c 006 b 029 c !
*
007 'c 030 d 008 c 031 c
,
009 c 032 c [ 010 a 033 a 011 b 034 d 012 d 035 b '[ 013 c 036 b 014 b 037 c 015- a 038- a er d 016 b 039 d 017 d 040 c 018 d 041 c 019- b 042 d 020 b 043- c < 021' c 044 c 022 a 045 d
.
. !
-l l
'
.
l
. - _..,, . - . , , - , _ - . , .-..-. -_ . . . . . - . - - . - . = . - . - .-
,__ _ _ _ _ _ .._____ _.._._.._._.__ __ _ _._ . . _ _ __ _ _._. . _ _-__ ._. _ _ _
TEST CROSS REFERENCE Page 1-SRO Exam BWR Roactor o r g a n i-z o d by Quoation Humbor QUESTION VALUE REFERENCE l Uvl 1.v0 3000100 - (b'US*d l 002 1.00 9000101 l 003 1.00 9000102 I 004 1.00 9000103 005 1.00 9000104 006 1.00 9000105 007 1.00 9000106 000 1.00 9000107 009 1.00 9000113 010 1.00 9000114 011 1.00 9000115 . 012 1.00 9000116 013 1.00 9000117
'
014 1.00 9000118 015 1.00 9000119 016 1.00 9000120 , 017 1.00 9000121 ' 018 1.00 9000126 019 1.00 9000127 020 1.00 9000128 021 1.00 9000129 022 1.00 9000130 023 1.00 9000131 024 1.00 9000132 , 025 1.00 9000133 026 1.00 9000134 . 027 1.00 9000135 028 1.00 9000136 029 1.00 9000137 030 1.00 9000138 031 1.00 9000139 ' 032 1.00 9000141 ,
'033 1.00 9000142 034 1.00 9000143 035 1.00 9000144 036 1.00 9000145 037 1.00 9000146 .
038 1.00 9000147 039 1.00 9000148 040 1.00 9000149 041 1.00 9000150 042 1.00 9000151 043 1.00 9000152 ! 044 1.00 9000153 7 045 1.00 9000154 046 1.00 9000155 ' 047 1.00 9000156 045 1.05 ^00G137 vb'Islcl 049-- 1.00 9000158 l-l1 .. . - . - - . - . - -. .-. - - - .. - . - -- ..
. -- .. - . .__ . . _ _ ~ - . - . .. . .. _ _ _ . TEST CROSS REFERENCE Pago 3 SRO Exam BWR Reactor Organirod by Queation Numbor l QUESTION VALUE REFERENCE j
050 1.00 9000159 051 1.00 9000160 052 1.00 9000161 053 1.00 9000162 054 1.00 9000163 I' 4 65 1.00 9004144--. M 056 1.0C 9000175 , 057 1.00 9000176 058 1.00 9000177 059 1.00 9000178 060 1.00 9000179 061 1.00 9000180 062 1.00 9000181 063 1.00 9000182 064 1.00 9000183 065 1.00 9000184 066 1.00 9000185 067 1.00 9000186 068 1.00 9000187 069 1.00 9000188 070 1.00 9000189 071 1.00 9000190 072 1.00 9000191 073 1.00 9000192 074 1.00 9000193 075 1.00 9000194 076 1.00 9000195 077 1.00 9000196 078 1.00 9000197 079 1.00 9000198 080 1.00 9000199 081 1.00 9000200 082 1.00 9000201 083 1.00 9000202 084 1.00 9000203 085 1.00 9000204 086 1.00 9000205' 087 1.00 9000206 088 1.00 9000207 089 1.00 9000208 090 1.00 9000220 091 1.00 9000221 092 1.00 9000222 093 1.00 9000223 094 1.00 9000224 095 1.00 9000225 096 1.00 9000226 097 1.00 9000227 098 1.00 9000228
. . - - ._ _ - . _ _ -.-
TEST CROSS REFERENCE Pago 3 SRO Exam DWR R o a c t'o r organized by Q u o a t.1 o n it u m b e r QUESTION VALUE REFEREllCE o 099 1.00 9000229 100 1.00 9000230
,
______ 100. 00= 77.do
. . - .--___
100.00- f7.co
,
k
'
;,-.. . . . - - - - , , . - - ~ ~ . - , , , - . .-vm- .x. - -
-t-----. --,,--..re -- - , - , - -,,we ,
. ., . - ._ __ _ _ _ ._-_._ _ . __ . _ . _ _ _ _ _ _ _ _ _ _ _ - - . . _ _ _
TEST CROSS RCFERENCE Page 4 SRO Exam BWR Roactar ' Organized by KA Group
!
N -'
</IDE GENERICS QUESTION VALUE KA 015 1.00 294001A102 008 1.00 294001A103 013 1.00 294001A106 014 1.00 294001A106 011 1.00 294001A110 402 1.- G 0 294001Ali5: (I a I a N 016 1.00 294001A116-017 1.00 294001A116 006 1.00 294001K101 007 1.00 294001K102 003 1.00 294001K103 002 1.00 294001K103 012 1.00 294001K107 005 1.00 294001K109 004 1.00 294001K110 009 1.00 294001K113 '
010 1.00 294001K116
......
PWG Total in n et 14 0 o PLANT SYSTEMS Group I QUESTION VALUE KA 046 1.00 202002K604 026 1.00 203000A101 05E Iv00 2 00000G00Er Mk 024 1.00 206000G010 025 1.00 209001A205 056 1.00 211000G005 019 1.00 212000G010 020 1.00 212000K412 031 1.00 215004G011 034 1.00 215005A107 035 1.00 215005A207 018 1.00 216000A210 023- 1.00- 217000A215 029 1.00 218000A206 022 1.00 223002A302 027 1.00 226001G010 030 1.00 239002K405 042 1.00 241000K306 041 1.00 259002K604 043 1.00 2G.1000A?OS Md
- - .- . . . . - -
-
_ _ . . . -... _ .m.-._ _ _ .-...._ ___..- ..___.._._.___m_ . _ _ _ . _ _ _ _ . _ _ f TEST CROSS REFERENCE Phgo 5 ! t i SRO Exam BWR Reactor
'
OrganiZ ed by KA Group i PLANT SYSTEMS Group I QUESTION VALUE KA [ 057 1.00 262001G005 047 1.00 264000K408 050 1.00 290001G010
------
PS-I Tota 1 A-3-CO * Z.l. c o Group II , l QUESTION VALUE KA t I 038 1.00 201006K403 045 1.00 202001K410 . 037 1.00 214000A201 036 1.00 215002A202 032 1.00 215003A202 028 1.00 230000A406 053 1.00 234000K402 043 1.0 A304 039 1.00 259001A204 044 1.00- 271000A206 049 1.00 272000K403 054 1.00 286000G004 , 051 1.00 290003A204
------
PS-II Total 13.00 _ _ __ Group.III
,
QUESTION VALUE KA 033 1.00 215001A207 < 052 1.00 233000K102 021 1.00 239001K127-
.040 1.00 256000K304 ------
PS-III Total 4.00
------ ------
PS Tota 1 +0rotr 3).00
'
EMERGENCY PLANT EVOLUTIONS Group I-
'
_ . - _ _ - - . . _ __ _ _ . . . _ _ . _ .- _ , _ - - . . _ . - . - . . , . - . _ . . . . _ . . - . . . _ _ _ . _
. __ _ ____ - . .._ . . - . . . _ . _ _. . _ . - _ . . _ . _ _ _ _ . _ _ _ .. . _ _ _ . _ --._ . _ _ _ .
.
TEST-CROSS REFERENCE Pago 6
,
SRO Exam BWR Roactor Organized by KA Group EMERGENCY PLANT EVOLUTIONS i Group I QUESTION VALUE KA 079 1.00 295003A103 073 1.00 295003A103 '
-
081 1.00 295006A202 058 1.00 295014G003 ; 075 1.00 295014G010 084 1.00 295015A101 ' 083 1.00 295015K204 063 1.00 295016G002 080 1.00 295016G006 i 060 1.00 295023G003 077 1.00 295023G010 087 1.00 295024G007 097 1.00 295024G012 066 1.00 295025G003 i 086 1.00 295026G012 061 1.00 295026K305 094 1.00 295026K305 096 1.00 295030G012 089 1.00 295031G006 i
-
099 1.00 295031G012 095 1.00 295032K101 082 1.00 295037A201 092 1.00 295037G012 093 1.00 295037G012 091 1.00 295037K101 098 1.00 295038G012 ______ EPE-I Total 26.00 Group II I QUESTION VALUE RA 068 1.00 295001A201 067 1.00 295001G011 074- 1.00 295002G010 078 1.00 295004G005 085 1.00 295012G01 .00 295018G008 072 1.00 295018K201 069 1.00 295019A202 ; 071 1-00
.
295020A101 070 1.00 295020K201 076 1.00 295022G010 090 1.00 295028G012 J aw~ - ~ . . . ~.,.--.-a...-+y.+.~...4-_,t...,, ..#-r--...,-.n.,._v .. r -,c,w m--,, ,,,,,,..w,-.,-
-
,.,-r.., y e v.rv-,--w.,y,-g.y ,.<w-- ~
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. ,-..-.. - .- . ._ _ . . . - .. . .. - - .-- _.- .. __ . . _ _. - _ . . . ~ . . ..
I TEST CROSS REFERENCE Page 7 l SRO Exam DWR Roactor organized by XA Group EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA' i l
,
088 1.00 295029A203 ) 064 1.00 295029G002 100 1.00 295032G012 062 1.00 295033G002 > 065 1.00 295035K101 ,
<
_. ___ EPE-II Total 17.00 ______ ______ EPE Total 43.00 ______ ______ ______ Test Total itfr-ee-470o f h ,-
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.
17 g 3 PAGE N .
) NUCLEAR OPERATIONS SUPERVISOR DATE 9.13 REV.11 Forn RE.12 Pace 9 of 12
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DOS 10N EDISON PILGRIH NUCLEAR POWER STATION Procedure No. 2.4.17 s RECIRCULATION PUMP (S) TRIP
.
INFORMATION ONLY ~~ Uso rostricted to referenco REVIENERS AND APPR0"ERS hd hl 0 M -
$ VA Procedufi Hriter Date t Ax L- 329 / // Technical Reviewer 'Date
&woA- Validator
>Mx Date i 7t44 A- to- 3 lA l92 Procedure Owner Date Nfk OAD Mannaer Date FDbY/
I/ ORC thairman 3/Y/f2 _ Date SAFETY REVICH REQUIRED ORC REVIEH REQUIRED r #4 3 d PM nt Manaaer Date Effective Date: S-4-h-2.4.17 Rev. 14 i . . .
. .. . .. ---_--_-___-__----_--_-__-___---__-_--.-.-A
_ . - _ _ _ _ . -- - .__ _ _---
.
1.0 51MPJ0MS ALARMS ANNUNCIATOR EANEL HINDOH
[1] RECIRC. HG SET GENERATOR DIFF. OVERCURRENT 904C G4 904R A3 [2] RECIRC. MG SET DRIVE HOTOR TRIP 904C S4 904R E2 [3] RECIRC. MG SET DRIVE HOTOR OVERLOAD 904C C4 904R F2 [4] RECIRC. HG SET GFHERATOR LOCK 0UT 904C E4 904R H2 [5] RECIRC. PUHP LOCKED ROTOR TRIP 904C A4 904R D2 [6] RECIRC. HG SET LUBE OIL LOH PRESS 904C 03 904R E1 [7] RECIRC. HG SET FLUID DRIVE HI OIL TEM C D3 904R G1 PLANT INDICATIONS [1] Reduction in recirculation flo [2] Sudden decrease in reactor power, 2.0 AUTOMATIC ACTIONS None 3.0 IMMEDIATE OPERATOR ACTIONS [1] MONITOR alarms and instrumentation 6MD DETERMINE the type of system malfunction that has occurre [2] lE both recirculation pumps trip, IREN KANUALLY SCRAM the reactor 6MD 00NCURRENTLY PERFORM PNPS 2.1.6, " Reactor Scram", with this Procedur .4.17 Rev. 14 Page 2 of 9 . _ - . . . . _ . , - . _ _ _ - .
_ _ . _ _ _ _ __ ___ _ _ . _ _ _ . _ _ . _ _ _ .___ _ _ _ - _ .. _ _
;
4.0 SUBSEOUENT OPERATOR ACTIONS 1
'
.
[1] TURN to the page of this Procedure for the malfunction that has i occurred 6HD PERf0RN the indicated steps:
'
East Section (a) Trip of one recirculation pump 3 (b) Trip of both recirculation pumps 6 .1 TRIP OF ONE RECIRCULATION PUMP HQII The reactor shall not be operated with one recirculation loop out of service for more than 24 hours. Hith the reactor operating, if one recirculation loop is out of service, the Plant shall be placed in a-Hot Shutdown condition within 24 hours unless the loop _is sooner returned to servic t
...............................................................................
CAUTION
:
If power level is less than 301, stratification may occur; refer to PHPS 2.4.24. " Reactor Vessel Cold Hater Stratification".
............................................................................... ,
[1] CLOSE affected HO-202-5A or B. PUMP DISCH VL (a) HUfH 5 minutes have_ elapsed. IHER REOPEN the. discharge valv [2] CHECK speed on the in-service pump to ensure it has not increased. -
2.4.17 Rev. 14 Page 3 of19 , - -__. _ _ . . . _ _ .- - _ _ - _ . , _ - ___; __ _ _ ._ . . _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
4.1 TRIP OF ONE RECIRCULATION PUHP (Continued)
[3] DETERMINE Total Core flow (TCF) (NRC Inspection R3 port 91-25]
(a) (K ' ERMINE direction of flow through idle jet pump l (1) ADD in-service and idle jet pump loop flow rate:
+ -
In-Service Idle Summed FI-263-107A(B) FI-263-107A(B) Value (2) MLILTIPLY idle jet pump loop flow by 0.95 m SUBTRACT this multiple from in-service jet pump loop flow:
- [0.95 X ) .
In-Service Idle Subtracted ' FI-263-107A(B) FI-263-107A(B) Value
. (3) USE current reactor power E PLOT both of the .
calculated flow values on the Power-To-Flow Ha , RQlE The change in reactor power due to the recirc pump trip will result in a Xenon transient. This transient will cause the previous load line to lower. The amount the load line shifts is dependent on the time after the recirc pump trip and previous equilibrium reactor power condition (4) 1E the Subtracted Value falls to the left of the ' expected load line (i.e., on or to the left of th minimum pump speed line) M the Summed Value falls approximately on the expected Load Line, M Forward Flow exists through the idle jet pump loo (5) IE the Subtracted Value falls approximately on the expected load line M the Summed Value falls below and to the right, M Reverse Flow exists through' the idle jet pump loo (b) CALCULATE Total Core flow (TCF)
-(1) If Forward Flow through the idle jet pump loop exist M Total Core Flow equals the Summed Valu .
'
(2) If Reverse Flow through the idle jet rump loop exists, M Total Core Flow equals the Subtracted Valu .4.17 Rev.- 14 bge 4 of 9
, .,- . - - .-. - .., . . - . - . ~ .-. - - ~ - - TRIP OF ONE RECIRCULATION PUHP (Continued)
. [4] E the reactor is operating 61R ABOYE the 80% load line 6tH1-total core flow decreases below 31.5 Hlb /hr. Ilif!( PERFORM the following steps: [IEB 88-07. SUPP 1: BHROG 8879) (a) NONITOR the APRHs and LPRMs for neutron flux instability oscillation (b) INCREASE speed of the operating recirculation pump untti any flux instability ceases and core flow is greater than 31.5 Hlb /h (c) INSERT control rods to decreasc reactor power below the-80%
load lin (d) LE APRH oscillations of greater than 10% peak-to-peak QB ; periodic LPRH upscale or downscale alarms are observed, Illf!( SCRAM the reactor 6!iQ CONCURRENTLY PERFORM PNPS 2.1.6 with this Procedur ,
[5] AEIER the recirculation pump is secured, ADJUST total core flow to greater than 27.6 H1b/hr. IGE SIL 517] (a) VERIFY that the recirculation loo) flow of the active loop on FI-107A or B is less than 36.9 H13/hr. [GE SIL 517) [6] SEND an operator to the 4kV breaker and to the MG Set Room t record all relay targets to determine cause of tri [7] ENSURE that power is available as follows: (a) Instrument power Panel Y1 (b) Vital service Panel Y2 (c) 4160V load centers A3 or A4 (d) Power Centers B17, B18, B20 and D9 (e) Power Panels D4, 05 and 06 [8] If the cause of the trip can be determined and. corrected 6!!Q the reactor is o)erating outside of the area of the power flow map ;
bounded. by- tie 80% load line and'31.5 Hib/hr, IllEli the pump may be restarted in accordance with PNPS 2.2.84, " Reactor _ Recirculation System". [IEB 88-70.-Supp 1; BHROG 8879) l
[9] REFER to EPIP-100, " Emergency Classification", to determine- ;
i whether an Emergency Action Level (EAL) has been exceeded.- ' l ,
'
l L 2.4.17 Rev. 14- i l Page 5 of 9
, , -- .n . :~ ,- - - ._- ._ - , - - ~ , . . - - - - - - - - -
_ - . - - . - - - - - . - . _- --. .. - . - _ . - - -. - - .. _-.
, - TRIP Of DOTH RECIRCULATION PUHPS . (1) CLOSE both H0-202-5A and B PUHP D1501 VLV I (a) HUIN 5 minutes have elapsed, IHLH REOPEH the discharge valves, (2) SEND an operator to the 4kV breaker and to the HG Set Room to
. record all relay targets to determine cause of tri [3] ENSURE that power is available as follows:
(a) Instrument power Panel Y1 (b) Vital service Panel Y2 (c) 4160V load centers A3 or A4 (d) Power Centers B17, B18, B20 ard 09 (e) Power Panels 04, 05 and 06 [4] IE the cause of the pump trips can be ident1fted and corrected, lulH RESTART the pumps in accordance with PHPS 2.2.84, " Reactor Recirculation System".
(5) REFER to PNPS EPIP-100, " Emergency Classification", to determine whether an Emergency Action Level (EAL) has been exceeded.
i L l
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2.4.17 Rev. 14 l Page 6 of 9
, . _ _ _ _ _ _ _ . _ _ . , _ . . - _ . . _ _ _
, ' N .' Ir' I' [Q5104 [DISM PILGRIM NUCLEAR POWER STATION Procedure No. 5.3.27 DETERMINING PRIMARY CONTAINHENT HATER LEVEL INFORMATION ONLY Uso restricted to referenco Approved b /Im /, du fA A b/7k70
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PlantMargger() l' Date 5.3.27 Rev. 3 Page 1 of 8
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N This procedure provides instructions for determining Primary Containment Water Level when flooding of the Primary Containment is directed by the Emergency Operating Procedures (EOPs).
- Primary Containment (PC) Hater Level values are referenced to plant elevatio .0 ACTIONS FOR CONTAlhMEXT LEVEL LESS THAN 47 FEET
[1] CALQJLATE the ext: ting differential pressure (sensed) between the drywell air space and the bottom of the torus, as follows (Figure '):
10$I SOTTCH f'RESS (P:-RJi-60)
(Panu C903) osig minu:: URYHELL HIDE RANGE PRESSURE (PI-1001-000A/B) (Panel C170/171) osig equals: CONTAlWHENT TO TORUS BOTTOH dP esig (2) for the calculated CONTAINHENT TO TORUS BOTTOH dP, the corr 6sponding Primary Containment Hater level is obtained from the Primary Containment Hater Level curve (Figure 2). FOR CONTAINHENT LEVEL GREATER THAN OR EQJAL TO 47 FEET [1] READ containment water level on DRYHELL LEVEL Gauge LI-5008 on Panel C903, t
5.3.27 Rev 3 Page 2 of 8
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iT CONTAINHENT TO TORUS BOTTOH dP cal.CULATION TORUS BOTT0H PRESS (PI-1001-69)
(Panel C903) osig ~
DRYWELL HIDE RANGE PRESSURE
- (PI-1001-600A/B) (Panel C170/171) osig - CONTAINHENT TO TORUS BOTT0H dP osig FIGURE 1 .
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. .3.27 Rev. 3 Page 3 of 8
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Number g-) PHPS Emergency Plan Implementing ; 1}- [-) p Procedure Manual EP-IP-400 ., ;
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Tide Revision Protective Action Reconnendations 1A . ATTACHMENT 1
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' Protecttve Action Recommendations Based on Plan * Conditions
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2) Indicatbne of large amoums ed fission products in containment include:
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3) Releases from containment, intentional or unittentional, mey be to be imminent based on indsestions such as:
. senames passee epioenereio e m esmee a re PcPL 8" . Prenery senternere omvensene gas sensoreemans asene setsymmene leve e 4) Releases from containment, intentional or unintentional, may be g ed likeV but not imminent based on lndiostione such as:
. enne m en p esse wonene>=eenmo (.,,. . Pr ,wy sweevnere swiensene ses earnerossens espe news esseremme tm o M w.. E o va es e ) Dose pro}ections een be performed using:
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6) Inimum Protecttve Action Recommendation when in a General
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Page 14 of 21 Attach, ment
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o%.... _ ATTACHMENT 1 Sheet 6 of 6 LN1ERVICE PUMP TESTING DATA SHEET (P-207A) REFERENCE ACCEPTABLE ALERT RANGE REQ. ACTION RANGE MEASURE 3 TEST PARAMETER /ALUE RANGE LOH I HIGH LOW F HIGH VA(UE~ 41.6 to 4 < 41.6 to > 45.6 to < 3 > 4 FLOW RATE (GPM) 4 .9 4 YZ 7 DISCH. PRESS. (PSI) 1275 NA NA NA NA 17 83
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TANK LEVEL AT_ START (INCH) 10 1/2 NA NA NA NA NA et 49 TANK LEVEL AT FINISH (INCHES) 41 3/8 NA NA NA NA NA l/2.Et (H) (V) > 1.82 to > 2.73 ( VIB. DISPLACEMENT (MILS) NA 0.91 0 to 1.82 NA 2.73 NA (H) (V) 1 0.314 > 0.314 ( VIB. VELOCITY (IN./SEC) NA 0.15 NA NA NA O 16 LUBRICANT - STOPPED VISIBLE VISIBLE NA NA NOT VISIBLE NA Ues.* v
* NA NA NA NA NA NA LUBRICANT - RUNNING NA _ #
PERFORMED BY: (SIGNATURE) V " N- DATE // / / 3 /7z- TIME /V30 IST REVIENED BY: (SIGNATURE) [ N
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DATE // / O/#2 TIME / / CC CALCULATIONS: TEST EWIPMENT: TEif PARAMEllR INST. N ~~ CAL. DATE F CAL. C#3E_QA
[1] (Final Tank Level [ Step 17] d2.Yz- inches - Starting Tank Level [ Step 11] 12 577 inches) x 4.3 gal / Inch DISCHARGE PRESS P 1 2.1 A 9 /m/92- Ir/re/ft - t #E . 5' gallons FLOH RATE 1~ 1%D 9 /27/n It / 77/Fr"- [2] 127. f ga11ons/3 minutes - Flow rate 42 6 gpm VIBRATION V34S iO / st /fz- 1ht/r3 NOTES: * Lubricant level Runnino Cannot be observed _due to Dumo desio _
Reference values were obtained durino cerformance of this Procedure on 2/20/90 (vibration) and 6/20/90 (hYdraU11C).
J 8. Rev. 27 Page 22 of 28 _ _ _ - _ _ _ _ _ _
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PILGRIM NUCLEAR POWER STATIDH Procedure No. 5. PRIMARY CONTAINHENT VENTING AND PURGING
. UNDER EMERGENCY CONDITIONS
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INFORMATION ONLY Uso restricted to reference
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REVIDiERS AND APPR0"ERS
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_ 7tda .A - Jim <c-- 3ll4 l42 Procedure Witer Date A?fd SkYN lechnitd1 Reviewer Date_
(7116 0 A- suxst.- 3b6 lf 2 .
Validator Date-7fknssks.na, bt ?b snJh 3 ll6 /(2 Procedure Owner Date t0 //r SAFEYY REVIEH REQUIRED . GAD Hanaaer Date ORC REVIEH RE0JIRED/ . J! / M/ 9' ~ NOT-REQ 0HED IRC' Chairman Date ff 7 Ar ?/>r
' Plant Manacer 5/M/97 Date Effective Date: S4M3 <
5. Rev. 23
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TABLE OF CONTENTS ,
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I PJL2f PURPOSE 3
, ACTIONS 3 DISCUSSION 5 , REFERENCES 5 ATTAOtHENTS 5 ^
ATTACHHENT 1 - TORUS VENTING UNDER EHERGENCY CONDITIONS 6
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ATTACHMENT 2 - DRYHELL VENTING UNDER EHERGENCY CONDITIONS 11 ATTACHMENT 3 - DRYHELL NITROGEN PURGE UNDER EHERGENCY CONDITIONS 13 ATTACHMENT 4 - DRYHELL AIR PURGE 18 ATTACHMENT 5 - DISCUSSION 19
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5.4.6 Rev. 23 Page 2 of 20
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( lSheetIof8 l , i 1.0 IQBMS VENTING UNDER EHfRGENCY CMDITIONS ( [1] If primary containment pressure is below 2.5 psig,
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Ill@ VENT the torus through the SGTS System as follows:
(a) IE while performing Step 1.0[1], primary containment pressure increases above 2.5 psig, IllM VENT the torus according to Step 1.0[2]. (b) INITIATE Oft VER!fY INITIATED SGT (c) OPEN the minimum number of the following valves required to malatain Primary Containment pressure below the Primary Containment pressure limit or to maintain hydrogen / oxygen below the limits of E0P-03 " Primary Containment * Control":
E01I Hhen RPV water level is above low-low level, isolation interlocks for A0-5041A and AD-50418 are def eated by placing talve control switches to the "EHERG OPEN" positio I (1) 2" Torus Normal Exhaust Valves (C7):
* A0-5041 A, TORUS NORML EXHAUST ISOL VLV (Key: CR-33)
e A0-5041B, TORUS NORML EXHAUST ISOL VLV (Key: CR-32)
(2) Containment Atmospheric Dilution System vent valves (Cl?0 and C171) Special Key CR- * SV-5083A, TORUS ISOLATION VALVE = SV-5084A, TORUS ISOLATION VALVE * SV-5083B, TORUS ISOLATION VALVE * SV-50848. TORUS ISOLATION VALVE (3) 8 in. Torus Purge Exhaust Valves (C7): * A0-5042A, TORUS PURGE EXHAUST ISOL VLV * A0-50420 TORUS PURGE EXHAUST ISOL VLV i 5. Rev. 23 Page 6 of 20 . - . .. . . -__ ._ .- - ,
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ATTACP.F.ERT 1 f Sheet 2 of 5 I (2) tMEH primary containment pressure is at or above 2.5 psi I IHEM VENT the Torus through the SGTS as follows:
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CAUTION Actions performed in this step have the potential to rupture the duct work between the torus vent valves and SGTS resulting in a direct release
of primary containment atmosphere to the Reactor Building. All ' non-essential personnel should be evacuated from the Reactor Buildin ...............................................................................
(a) INITIATE 0E VERIFY IRITIATED SGT (b) DEFEAT isolation interlocks as necessary 6_NJ OPEN the minimum number of the following valves required to maintain Primary Containment Pressure below the Primary Containment Pressure Limit or to maintain hydrogen / oxygen below the limits of E0P-3, * Primary Containment Control":
i E Isolation interlocks for A0-5041 A and A0-5041B are defeated by: Placing valve control switches in the "EHERG OPEN" position if RPV water level is greater than low-low leve Q8 In accordance with PNPS 5.3.21 " Bypassing Selected Interlocks".
Attachment 5 if RPV water level is less than or equal to low-low leve (1) 2" Torus Normal Exhaust Valves (C7):
* A0-5041A, TORUS NORRAL EXHAUST ISOL VLV (Key: CR-33) . A0-50418 TORUS NORMAL EXHAUST ISOL VLV (Key: CR-32)
i 5.4.6 Rev. 23 Page 7 of 20
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: Sheet 3 of 5-
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(2) Containment Atmosphere Dilution Systeo Torus Isolatten valves:
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(C170 and C171) Special Key CR- I * SV-5083A, TORUS ISOLATION VALVE
* SV-5084A, TORUS ISOLATION VALVE
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* SV-5083B, TORUS ISOLATION VALVE
* SV-5084B, TORUS ISOLATION VALVE
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CAUTION Prior to performing the following, refer to Emergency Classifications I AH EP-IP-100. Reference EAL 3.4. EAL 3.5.1.4. and EAL 3.5. ...............................................................................
(c) 11 additional venting capacity is required, M primary containment pressure is J. 30 psig,
, IIH VENT the Torus, bypassing SGTS as follows:
(1) CLOSE E VERIFY CLOSED the following valves:
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* A0-5042A, TORUS PURGE EXHAUST ISOL VLV * A0-5041 A, TORUS NORMAL EXHAUST ISOL VLV * AO-5041B, TORUS NORMAL EXHAUST ISOL VLV e SV-5083A, TORUS ISOLATION VALVE * SV-5084A, TORUS ISOLATION VALVE * SV-5083B, TORUS ISOLATION VALVE * SV-5084B, TORUS ISOLATION VALVE (2) PLACE the fan control switches for SGTS Exhaust Fans VEX-210A M VEX-210B to "0FF".
(3) CLOSE E VERIFY CLOSED the following valves:
* A0-N-112, TRAIN B OUTL DHPR * A0-N-108 TRAIN - A OUTL DMPR (4) OPEN A0-5042B, TORUS PURGE EXHAUST ISOL VLV, bypassing isolation interlocks according to PNPS 5.3.21 " Bypassing f
Selected Interlocks",= Attachment 1 .4.6 Rev.-23 Page 8 of 20
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: Sheet 4 o9 5 ,
( ED1E The fuse holders for the'two 3 amp fuses are designated "UQQ" in Panel C7.
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(5) INSTALL two 3 amp fuses for valve A0-5025, DIRECT TORUS VENT ISOL YLV, in the back of Panel C7 (fuses art stored in the Control Room "Q" Fuse Box).
(6) OPEN AD-5025, DIRECT TORUS VENT ISOL VLV (Key: CR-H).
[3] Bd[H venting is no longer required, IHfE TERMINATE Torus venti,ng as follows:
(a) If Torus venting is through A0-5025, DIRECT TORUS VENT ISOL VLV, IBIE 00NTINUE with Step (4). . (b) LE not required to be operating by E0P-4, " Secondary Containment Control",
THEN SHUT DOWN SGTS in accordance with PNPS 2.2.50, " Standby Gas Treatment".
( (c) CLOSE QB VERIFY CLOSED the following valves:
* A0-5042A, TORUS PURGE EXHAUST ISOL VLV e AD-5042B, TORUS PURGE EXHAUST ISOL VLV * A0-5041 A, TORUS NORMAL EXHAUST ISOL VLV * A0-5041B, TORUS NORMAL EXHAUST ISOL VLV * SV-5083A, TORUS ISOLATION VALVE * SV-5084A, TORUS ISOLATION VALVE * SV-5083B, TORUS ISOLATION VALVE * SV-50848, TORUS ISOLATION VALVE (d) LE control circuits for the following valves were bypassed, IHEM RESTORE them to normal in accordance with PHPS 5.3.21. " Bypassing:
Selected' Interlocks". Attachment 5:
* A0-5041A, TORUS NORMAL EXHAUST ISOL VLV a AD-5041B, TORUS NORMAL EXHAUST ISOL VLV I 5.4.6 Rev. 23-Page 9 of 20 - -
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l'S55"eI"5S5l / E The following step (Step [4]) is only to be performed if SGTS has been bypassed and the Torus was vented through the Direct Torus Ven [4] tilDi venting through the Direct Torus Vent is no longer required, 111Di TERMINATE Torus venting as follows:
(a) CLOSE the following valves: * AO-5025 DIRECT TORUS VENT ISOL VLV '
e AO-50428, TORUS PURGE EXHAUST ISOL VLV (b) OPEN the Outlet Valve of the SGTS fan that will be placed in servic * AO-N-108, TRAIN A OUTL DHPR
* AO-N-ll2, TRAIN B 00TL DHPR (c) RETURN the fan control switch for the SGTS Exhaust Fan (VEX-210A QB VEX-210B) that will be placed in service to "AUT0"
6.NQ N the fan control switch for the standby fan to " STANDBY".
(d) REK)VE the fuses from Panel C7 for valve AO-5025, DIRECT TORUS VENT ISOL VLV, 6N_Q RETURN the fuses to the NH (e) RESTORE the control circuit for A0-5042B, TORUS PURGE EXHAUST ISOL VLV, to normal in accordance with PNPS 5.3.21. " Bypassing Selected Interlocks", Attachment 1 ' 5.4.6 Rev. 23 Page 10 of 20 }}