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| {{#Wiki_filter:INITIAL L | | {{#Wiki_filter: |
| HARRIS EXAM
| |
| 50-4OO12OQ4-301
| |
| -
| |
| FEBRUARY 23 27,2004
| |
| & MARCH 4,2004 (WRITTEN)
| |
| | |
| Harris
| |
| Draft
| |
| SRO
| |
| Written
| |
| 2004
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| QBJESTION: 1
| |
| Given the following conditions:
| |
| While operating at 10004 power, a drop in 1'KZ pressure resulted in a Reactor Trip
| |
| and Safety Injection.
| |
| PRZ level is currently indicating ::~IO0%.
| |
| PRZ pressure has stabilized at 1400 p i g .
| |
| Containnicnt pressure is 3.6 psig and stable.
| |
| KCPs have been stopped.
| |
| RVLIS Full Range is indicating 209.0.
| |
| Core Exit Thermocouples are indicating 745°F.
| |
| RCS Wide Kange Hot Leg lemperatures are indicating 680°F.
| |
| Which of the following conditions currently exists'?
| |
| a. A PRZ steam space break has occurred and core heat removal is ADEQIJA'TE
| |
| b. A PRZ steam space break has occurred and core heat removal is INADEQITATE
| |
| c. An RCS hot leg break has occurred and core heat removal is ADEQUATE
| |
| d. An RCS hot leg break has occurred and core hcat removal is INADEQUATE
| |
| ANSWER:
| |
| b. A PRZ steam space break has occurred and core heat removal is INADEQUA'W
| |
| I'ost Validation Revision
| |
| | |
| IIams NKC Written Examiliation
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: I 'TIENGROUP: iii
| |
| KA riwowrmcx: RO SRO 3.4
| |
| 10CFR55 CONTENT: 41(b) 4303) 5
| |
| MA: 00000XAA2.30
| |
| Ability to determine and interpret the following as they apply to rhe Pressurizer Vapor Space Accident:
| |
| Iuadequate core cooling
| |
| OBJECTIVE: EOP-3.10-4
| |
| Given the following EOP steps, notes, and cautions, describe the associated basis
| |
| c. RVLIS level of 39 percent (C.1)
| |
| DEVELOPMENT REFERENCES: FOP-FRP-C:.1
| |
| CSFS'I-Core Cooling
| |
| REFERENCES SIJPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| OR SIGNIFICANTLY MODIFIED / DPWECT: New
| |
| NRC EXA,M m s m R Y : NOIE
| |
| DISTRACTOR JUSTIFICACTPON (CORRECT ANSWER d'd):
| |
| a. Plausible since the break is located in the PRZ steam space, but heat removal is not adequate
| |
| .! b. The R ( T is superheated and in excess of 70OoF,which indicates that inadequate t e a t r e m o d is
| |
| occurring. The break is in the PRZ stearn space as indicated by the pressurizer being fill.
| |
| e. Plausible since RCS temperatures are stable, hut the break is in the steam space and heat removal is
| |
| not adequate.
| |
| (8. Piausible since RCS heat removal is not adequate, but the break is in the steam space.
| |
| DIFI~ICXJLTYANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EX:PI.ANATIBN: Must analyze plant conditions to determine location of break, detemiine that
| |
| tetnperature indications support superheated conditions and that heat removal is
| |
| inadequate
| |
| Post Validation Revision
| |
| | |
| IIaris NRC Written Examination
| |
| Senior Reactor Operator
| |
| QUESTION: 2
| |
| Which of the following describes a condition which would require Emergency Boration
| |
| and the bases for taking this action?
| |
| a. e Twenty minutes following a Main Feedwater Pump trip, Control Rods are
| |
| determined to be below the rod insertion h i t
| |
| e Control the reactivity transient associated with a steam line break
| |
| b. a Twenty minutes following a Main Feedwater Pump trip, Control Rods are
| |
| determined to be below the rod insertion limit
| |
| e Control the reactivity transient associated with an inadvertent dilution
| |
| c. e During a reactor startup, the Reactor achieves criticality with Rank C rods at
| |
| 105 steps
| |
| e ('ontrol the reactivity transient associated with a steam line break
| |
| d. e During a reactor startup. the Reactor achieves criticality with Rank C rods at
| |
| 105 steps
| |
| 0 Control the reactivity transient associated with an inadvertent dilution
| |
| ANSWER:
| |
| c. e During a reactor startup, the Reactor achievcs critical@ with Bank C rods at
| |
| IO5 steps
| |
| 0 Control the reactivity transient associated with a steam Line break
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTION SUMBER: 2 TPEWGROUP: 112
| |
| MA IMPORTANCE: RO SRO 3.4
| |
| 80CFR55 CONTENT: 41(b) 43(b) 2
| |
| MA: 00002462.2.25
| |
| Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
| |
| (Emergency Boration)
| |
| OBJECTIVE: CVCS3.0-R4
| |
| Given a CVCS componeiitiparameter, state whether the componeiitiparameter is Tech Spec related
| |
| DEVELOPMENT REFERENCES: TS Bases 314. I . I
| |
| AOP-002 BD
| |
| GP-004
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTIQN SOURCE: NEW SIGNIFICANTLY MODIFIED
| |
| BANK NUMBER FOR SIGNIFICAXT1,Y MODIFIED i D1RE:C:T: AOP-3.2-RI 00 I
| |
| NRC EXAM IIISTORY: Kone
| |
| DISTRACTOR JUSIIFICACTION (CORRECT ANSWER Jd):
| |
| a. Plausible since if this condition existed for 2 hours, instead of 20 minutes, Emergency Boration would
| |
| be required. Additionally, in Modes I R: 2, SDM is required to control thc reactivity transient
| |
| associated with a steam line break. IIowever, it is not required during transient conditions, allowing
| |
| the 2 hours to restore rod position.
| |
| b. Plansibic since if this condition existed for 2 hours, instcad of 20 minutes, Emergency Boration would
| |
| bc required. However, it is not required during transient conditions, allowing the 2 hours to restore
| |
| rod position.
| |
| d s. Emergency boration is required if SDM is not met. Criticality at steady state conditions is considercd
| |
| to he a loss of SDM. In Modes I k 2 , SDM is required to control the reactivity transient associated
| |
| with a steam line break.
| |
| d. Plausible since Emergency boration is required if SDM is riot met. Criticality at steady state
| |
| conditions is considered to be a loss of SDM However, the concern for an inadvertent dilution is
| |
| related to a shutdown condition.
| |
| DIFFYCUI,TY ANALYSIS:
| |
| COMPREHENSIVE i ANALYSIS KNOWI.EDGE i RECALL
| |
| DIFFICULTY RtlTING: 2
| |
| EXPLANATION: Knowledge of the requirements for initiating Emergency Boration and the bases
| |
| for these actions.
| |
| Post Validation Kevision
| |
| | |
| Harris NRC Written Iixarnitiatiuii
| |
| Senior Keactor Operator
| |
| QUESTION: 3
| |
| Given the following conditions:
| |
| * The plant has been operating at 100% power for the past three (3) months.
| |
| e CSIP 1A-SA is operating.
| |
| e C'SIP IB-SB has just been restored to a normal alignment following maintenance on
| |
| the pump impeller.
| |
| e When CSIP IB-SB is started the operator notes that sucl~onpressure appears normal,
| |
| while discharge pressure, discharge flow. and pump current are oscillating.
| |
| Which of the following is the niost likely cause of these CSIF I R-SB indications?
| |
| a. Inadequate venting was performed during clearance restoration
| |
| b. 'Ihe CSIP 1B-SB discharge valve was inadvertently left closed during clearance
| |
| restoration
| |
| c, A failure of the CSIP 1R-SU miniflow isolation valve has resulted in gas binding
| |
| d. A failure ofthe <'SIP 1R-SR tniniflow isolation valve has resulted in all pump
| |
| flow being recirculated to the VCT
| |
| ANSWER:
| |
| a. Inadequate venting was performtvl during ciearanee restoration
| |
| Post Validation Revisiun
| |
| | |
| Hairis NRC Written Examitlation
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 3 TIEWGROUP: 2i I
| |
| L i IMPORTANCE: RO SRO 3.x
| |
| lOCFR55 CONTENT: 41(b) 43(b) 5
| |
| K4: 006.4204
| |
| Ability to (a) predict the impacts of the following malfunctions or operations on the EC'CS; and (b) based
| |
| on those predictions, use procedures to correct. control, or mitigate the consequences of those
| |
| malfunctions or operations: Improper discharge pressure
| |
| OBJECIIVE: AOP-1.2-4
| |
| Given a set of plant conditions and a copy of AOP-002, determine ifthe possibility of gas binding the
| |
| CSIPs exists and the corrective action to be taken
| |
| UEVEL.OPMENT REFERENCES: UP-107
| |
| SOEK 97-1
| |
| REFERENCES SUPPLIED TO APPLICANT': None
| |
| QUEKIION SQCRCE: SIGNIFICANTLY MODIFIED DPRECI
| |
| CANTLY MODIFIED / DIRECT:
| |
| NRC EXAM WISTBRY: None
| |
| DISTRACTOR JUSTIFICACTHON (CORRECT ANSWER +d):
| |
| d a. Gas binding of a pump results in lower than expected pressure, flow, and current. Likely cause is
| |
| improper venting of pump when restoring from post maintenance activities.
| |
| b. Plausible since improper alignment would result in low flow and current, but a closed discharge valve
| |
| would cause discharge pressure to be high.
| |
| c. Plausible since gas binding is cause ofthese indications, but will not occur as a result ofpump recirc
| |
| valve being open.
| |
| d. I'kausihle since a Fdiled open recirc valve wiil cause indicated flow to be low since flow is measured
| |
| dowstream of the recirc valve, hut discharge pressure and current would be at or nrar normal.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / AXALYSIS MNOWL.EDGEI RE.C:AB,L
| |
| DIFFICULTY ItlTINC: 3
| |
| EXPLANATION: Must analyze giveti pump conditions to determine hilure mode and then
| |
| determine likely cause of gas binding ofthe pump
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| QUESTION: 4
| |
| Given the following conditions:
| |
| e The unit is operating at 100% power, with Control Bank D rods at 2 15 steps.
| |
| e ALB 13-4-1, ROD CONTROL IIRGENT AI.ARM, is in ALARM due to a failure in
| |
| Power Cabinet 1AC.
| |
| e Rod ControI is in MAN
| |
| e A turbine trip occurs, but the Reactor fiils to trip either automatically or manually.
| |
| Which of the following actions should the Reactor Operator be directed to take?
| |
| a. Place the Rod Control K4NK SELECTOR in AUTO and allow rods to insert
| |
| b. Maintain the Rod Control BANK SELECTOR in MAN and manually insert rods
| |
| c. Place the Rod (hntrol BANK SEI,ECTOII in HANK D and manually insert rods
| |
| d. Maintain rods at 215 steps
| |
| 4WSWER:
| |
| d. Maintain rods at 2 I 5 steps
| |
| Post Validation Revision
| |
| | |
| Harris NRC' Written Examination
| |
| Senior Reactvr Operator
| |
| Data Slicets
| |
| QUESTION NUMBER: 4 THEWGROUP: 22
| |
| KA IMPORTANCE: RO SRO '4.0
| |
| llOCFR55 CONTENT: 41(b) J3@) 5
| |
| KA: 001G2.4.h
| |
| Knowledge of symptom based EOP mitigation strategies. (Control Rod Drive)
| |
| OBJECTIVE: EOP-3.19-4
| |
| Given a set of conditions during EOP implementation. determine the correct response or required action
| |
| based upon the EOP User's Guide general infonnation
| |
| z. IJse of "Bank Select" during an ATWS
| |
| DEVELOPMENT REFERENCES: IIOP-USERS GUIDE
| |
| EOP-FW'-S. i
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MQDIFIEI) DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| QHS'I'RACTOK .IUSTIFICACTION (CORRECT ANSWER +d):
| |
| a. Plausible since this is an RNO action for a failure of the reactor to trip. hut will not be successful due
| |
| to the urgent failure in rod control.
| |
| 19. Plausible since this is an RNO action for a failure of the reactor to trip, but will not be successful due
| |
| to the urgent failure in rod control.
| |
| c. Plausible since this will allow Bank D rods to move inward, and is the only nlcthod of inserting rods
| |
| with the rod control failure, but should not be used due to the potential to cause unanalyzed flux
| |
| shapes.
| |
| d d. Due to the urgent failure: rods will not move in AI.JT0 or MAE. Altlzough they will move in RANK
| |
| 1) with this particular failure, moving rods in individual banks may result in nnanalyzed flux shapes
| |
| which could result in fuel damage.
| |
| DIFFICI!LTY ANALYSIS:
| |
| COMPREIIENSIVE I AXALYSIS KNOWLEQGE I RECALL
| |
| DIFFICXLTY RtPTHNG: 3
| |
| EXPIANATION: Must analyze the effect of an urgent rod control failure and then apply the
| |
| failure results to the plant conditions to deteimine the proper actions
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Rcactor Operator
| |
| QUESTION: 5
| |
| Four Operators worked the following schedule in the Control Room over the past six
| |
| days:
| |
| IIOUKS WORKED (Shift turnover time not included. Do NOT assume any hours
| |
| worked before or after this period.)
| |
| QPEWATOR DAY1 DAY2 DAY3 DAY4 DAYS DAY6
| |
| 1 IO I4 off 12 12 12
| |
| 2 14 12 14 IO off 11
| |
| 3 off off off 13 11 14
| |
| 4 11 13 14 0ff I1 12
| |
| Which of the operators would be permitted to work a 12-hour shift on Day 7 WKTIPOUT
| |
| requiring permission to exceed nonnal overtime limits?
| |
| a. Operator I
| |
| b. Operator 2
| |
| c. Operator 3
| |
| d. Operator4
| |
| ANSWE R:
| |
| a. Operator 1
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| L)ata Sheets
| |
| QUESTION NUMBER: 5 TIEWGROUP: 3
| |
| KA IMPORTANCE: RO SRO 4.0
| |
| 10CPR55 CONTENT: 41&) 43(b) 5
| |
| KA: 2.1.2
| |
| Knowledge of operator responsihililies during all niodes of plant operation
| |
| OBJECTIVE: PP-2.0-SI
| |
| STATE the requirements contained in Administrative Controls Section, including requirements for
| |
| the following:
| |
| * Unit staff3including overtime limitations
| |
| DEVELOPMENT REFERENCES: AP-0 12
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESI'ION SOURCE: NEW SHGNIFICANTLU MO1)IFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Kobinson NRC 2001
| |
| NRC: EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTBON (CORRECT ANSWER d'd):
| |
| \' a. Working a 12 h c w shift on Day 7 would result in this operator working 24 hours out of48, and 72
| |
| liours in 7 days, both of which are permissible.
| |
| 12. Plausible since this operator would not exceed the 24 hours out of48 limit and has had a recent day
| |
| off, but would work 73 liours in 7 days which exceeds limit.
| |
| c. Plausible since this operator would not exceed the 72 hours in 7 day limit and has several recent days
| |
| off, but would work more than 24 hours in 48 which exceeds limit.
| |
| d. Plausible since this operator would not exceed the 24 hours out of 48 limit and has had a receut day
| |
| off; hut would work 73 hours in 7 days which exceeds limit.
| |
| D~FFIC~ILTY ANAL.YSHS:
| |
| COMPREIHENSIVE / ANALYSIS KNOWLEDGE / RECAI,I,
| |
| DIFFICULTY k4TINC;: 3
| |
| EXPLANATION: Required to compare glven data to adniinistrarive limits to determine which
| |
| operator would remain within acceptable overtime limits
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| QUESTION: 6
| |
| Given the following conditions:
| |
| e A Reactor Trip with SI occurs.
| |
| e The operators perform the immediate action steps, verify ECCS flow, and check
| |
| AFW flow.
| |
| * SG levels are 25% and the required AFW flow cannot be established, so the
| |
| operators enter EOP-FRP-11.1 Response to Loss of Secondary Heat Sink.
| |
| e KCS pressure is 175 psig.
| |
| e All SG pressures are between 300 psig and 350 psig.
| |
| Which of the following actions is to be taken!
| |
| a. Continue in EOP-FRF-11. I since EOP-FRP-11. I has a hib :r priority than PATI 1
| |
| and attempt to establish .4FW or Main Feedw,ater flow.
| |
| b. Continue in EOP-FRP-I1.I since EOP-FRP-H.1 has a higher priority than PATII-1
| |
| and initiate RCS feed and bleed.
| |
| c. Return to EOP-PATII-I at the step that was in effect since a secondary heat sink is
| |
| NOT required following a large break LOCA.
| |
| d. Return to EOP-PATH-1 at Entry Point C since a secondary heat sink is NOT
| |
| requird following a large break I.OCA.
| |
| IPNSWEK:
| |
| c. Return to EOP-PAI H-1 at the step that was in effect since a secondary heat sink is
| |
| NOT required following a large break LOCA.
| |
| Post Validatiori Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor (jperator
| |
| Data Sheets
| |
| QUESTION NUMBER: 6 THERKIPOUP: lil
| |
| KA IMPORTANCE: RO SRO 4.0
| |
| IOCFRSS CONTENT: 48(b) 43(b) 5
| |
| KA: 00001 l(i2.4.6
| |
| Knowledge of syniptoni based EOP mitigation strategies. (Large Break LOCX)
| |
| OBJE.ClWE: EOP-3.11-4
| |
| Given the following EOP steps, notes, and cautions, describe the associated basis
| |
| e. Requirements for a heat sink (11.1)
| |
| DEVEI,OPMENT REFERENCES: EOP-I'W-1.I
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIPICAXTLY MODIFIED
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED i DIRECT:
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| a. Plausible since these are actions that are taken upon entry into I'RP-11.1. but a secondary heat sink
| |
| would not he required with RCS pressure <' SG pressure.
| |
| h. Plausible since these are actions thar might be taken upon entry into FRP-H.1, but a secondary heat
| |
| sink would riot he required with RCS pressure -: SG pressure.
| |
| 4 c. Since RCS pressure is less than SG pressure, a secondary heat sink is not required since the SG would
| |
| act as a heat source rather than a heat sink. Return is to procedure and step in effect.
| |
| d. Plausihle since RCS pressure is less than SG pressure and a swondary heat sink is not required
| |
| Return is to procediire and step in effect, not Pintry Point C.
| |
| PCULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE !RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Must interpret first that a secondary lieat sink is not required based on RCS
| |
| pressure being greater than S G pressure and then must recognize the entry point
| |
| conditions for returning to PATH-I
| |
| Post Validation Revision
| |
| | |
| IIarrts NRC Written Examination
| |
| Senior Reactor Operator
| |
| QUES'I'IQN: 7
| |
| Given the following conditions:
| |
| e The Reactor has been taken critical and power is being increased.
| |
| e NIS IR channels N3S and N36 are both indicating 5 x IO-'' amps.
| |
| e NIS SK channel N3 1 is indicating 8 x 10" cps.
| |
| e Ilue to improper adjustment of the high voltage setting, NIS SR c.hanne1N32 is
| |
| indicating 7 x 10' cps.
| |
| Power should be stabilized ..
| |
| a. at or above amps, and the SR High Flux trip should then he blocked.
| |
| b. at the current power level, and the SR High Flux trip should then be blocked.
| |
| c. at or above IV'' amps. but the SR High Flux trip should NQT be blocked.
| |
| d. at the ciinrent power level, but the SR IIigh Flux trip should KOT be blocked.
| |
| ANSWER:
| |
| d. at the current power lever, but the SR High Flux trip should NOT he blocked.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Rcactor Operator
| |
| Data Sheets
| |
| QtJESTION NUMBER: 7 TIENGROUP: 112
| |
| KA IMPORTANCE: KO SRO 2.9
| |
| 10CFK55 CONTENT: 4I(b) 43(b) 5
| |
| KA: 0000324A2.09
| |
| Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear
| |
| Instrumentation: Effect of improper HV setting
| |
| OBJEC.TIVE: GP-3.4-1
| |
| Recognize off-normal conditions during a reactor start-up, including
| |
| a. Availability of excore nuclear instrumentation channels (SK,IR, I'R)
| |
| DEVELOPMENT REFERENCES: GP-004
| |
| ALB-0 12-4-5
| |
| REFERENCES SUPPLIED TO APP1,ICXNT: Xone
| |
| QIJESTION SOUIPCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANILY MODIFIED I DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| 1)IS'FRACTCPR JUSTIFICIACTION (CORRECT ANSWER v"d):
| |
| a. Plausible since power niust he increased above 10 l o amps before blocking trips, but increasing power
| |
| to this level will result in SR high flux trip.
| |
| b. Plausible since power cannot be increased above amps, but the block ofthe SR high flux trip is
| |
| interlocked at this power level.
| |
| c. Plausihle since the SI< high flux trip is not pennitted to be blocked without at Icast I decade of overlap
| |
| hetween SR and IK, but increming power above I O I" amps will result in a SR high flux trip.
| |
| I
| |
| 3 cl. Ixss than I decade of overlap exists hetureen SR and BR channel before trip would occur. Increasing
| |
| power to allow blocking SR would result in trip before reaching power level and attempting to block
| |
| at current power level will not be successful.
| |
| DIFFICULTY ANALYSIS:
| |
| CCPillPREIlENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULI'Y RATING: 3
| |
| EXPLANATION: Must determine that increasing power above IO."' amp will result in a rcactcir
| |
| trip due to SR high flux, and that attempting to block the SW high flux trip
| |
| below 10 I" amps will not be successful. Required to not block SR high flux trip
| |
| if < 1 decade of overlap exists.
| |
| Post Vaiidation Revision
| |
| | |
| Harris NRC Written Fxamination
| |
| Senior Reactor Operator
| |
| QIJESTION: 8
| |
| Given the following conditions:
| |
| EOP-FRP-S. I , Response to Nuclear Power Generatio~v.AT\h,is bcing
| |
| impleniented.
| |
| 4 An SI actuation has O C C U I T ~ .
| |
| 8 The Foldout page is applicabie.
| |
| Which of the following actions should be taken?
| |
| a. Continue with FJOP-FRP-S. I while verifying proper operation of safeguard
| |
| equipment
| |
| b. Continue with EOP-FRP-S. I until the reactor is tripped or niade subcritical, then
| |
| imniediately exit to EOP-PATH-I
| |
| c,. Transition to EOP-PATH-1 and verify all autornatic actions required for an SI
| |
| have occurred, then return to EOP-FRP-S. 1 only when directed by PATH-1
| |
| d. Reset SI and FW isolation as soon as possible to restore feed flow to the steam
| |
| generators, then continue with EOP-FRP-S. 1
| |
| ANSWER:
| |
| a. Continue with EOP-FRP-S. 1 while verifying proper operation of safeguard
| |
| equipment
| |
| Post Validation Revision
| |
| | |
| Harris NRC U'ritten Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTIOX NUMBER: 8 mcwGRouP: 2!'I
| |
| KA IMPORTANCE: RO SRO 4.0
| |
| 10CFR55 CONTENT: 41(b) 43(b) 5
| |
| ILa: 012G2.4.6
| |
| Knowledge of symptom based EOP mitigation strategies. (Reactor Protection)
| |
| OBJECTIVE: FOP-3. I S
| |
| Uescrihe the purpose of the following ROPs including tyFe of event for which they were designed and the
| |
| major actions performed
| |
| - FEW-S.1
| |
| DEVELOPMENT REFERENCXS: EOP-FW-S. 1
| |
| EOP User's Guide
| |
| REFERENCES SlJPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: 6] NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BASK NUMBER FOR SIGNIFICANTLY MOIXFIED I DIRECT: EOP-l.ls 021
| |
| NHC: EXAM HISTORY: Harris NRC 2000
| |
| DISTRACTOR .FUSTIFICACTI(4N (CORRECT ANSWER d'd):
| |
| -\I a. If a safety injection occurs while implementing FRP-S. 1. proper operation of SI equipment is verified
| |
| while continuing with FKP-S. I .
| |
| b. Plausible since PA?H-I provides instructions for a response to safety injection, but FKP-S.I niust be
| |
| performed until completion.
| |
| c. Plausible since PATH-I provides instiuctions for a response to safety injection, but FKP-S. 6 must be
| |
| performed until completion.
| |
| d. I'lriusible since a safety injection will result in a loss of'MFW, but AFW flow is capable of providing
| |
| minimum required flow.
| |
| PCULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE IRECALI.
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION: Knowledge of procedural requirements in EPP-FRP-S. I
| |
| Post Validation Rcvision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| QUESTION: 9
| |
| Given the following conditions:
| |
| a The plant is in Mode 3 with all Shutdown Rods withdrawn.
| |
| a All power is lost to the Digital Rod Position Indication display and CANNOT be
| |
| restored.
| |
| Which of the following actions is to be taken'?
| |
| a. Verify that a11 Shutdown Bank Rods are fully withdrawn using Demand Position
| |
| indication
| |
| b. Determine that all Shutdown Bank Rods are fully withdrawn using the movable
| |
| incore detectors
| |
| c. Commence a boration of the RCX to ensure adequate Shutdown Margin
| |
| d. Open the Reactor Trip Breakers
| |
| ANSWER:
| |
| d. Open the Reactor 'Trip Breakers
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Seniot Reactor Operator
| |
| Data Sheets
| |
| QtJESTION NUMBER: 9 TIEWGRQUP 2i I
| |
| Ktt IMPORTANCE: RO SRO 3.6
| |
| IOCFR55 CONTENT: 41(b) 43(b) 5
| |
| %(A: 014.42.02
| |
| Ability to (a) predict the impacts of the following malfunctions or operations on the WIS; and (b) based
| |
| on those on thosc predictions, use procedures to correct, control, or mitigate the consequences of those
| |
| malfunctions or operations: Loss of power to the IWIS
| |
| OBJECTIVE: RODCS-3.1-R4
| |
| Given a copy of 'Technical Spccitications and a plant mode, determine if rod position indication
| |
| components and actual rod positions meet their Limiting Conditions for Operation; if they do not, then the
| |
| applicable .4C'I'ION statements
| |
| DEVELOPMENT REFERENCES: 'rs 3.1.3.3
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOCRCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| a. Plausible since this would be required in the event o f a loss o f a single indication whiie operating in
| |
| Mode 1 or 2 , but with both indications lost in Mode 3 the Reactor Trip Breakers are to be opened.
| |
| b. I'lausible since this wouid be required in the event of a loss of a single indication while operating in
| |
| Mode 1 or 2, but with both indications lost in Mode 3 the Reactor 'I'rip Breakers are to be opened.
| |
| c. Plausible since Loss of indication of UKI'I may lead to belief that Sr)M cannot be verified; which
| |
| would require Emergency Ihration.
| |
| d d. With both DRPI indications inoperable in Mode 3,4, or 5, 'TSrequires that the Reactor Trip Breakers
| |
| be opened immediately.
| |
| DIFFICULTY ANALYSIS:
| |
| CQMPREIIENSHVE / ANALYSIS KNO\VLEDGE I RECALL
| |
| DIFFICBJLTY RATISG: 2
| |
| EXPLANATIQN: Knodedge of Tech Spec immediate action requirements in the event of a Loss
| |
| of both DRPI indications
| |
| Post Validation Revision
| |
| | |
| Harris NRC' Written Examination
| |
| Senior Reactor Opelator
| |
| QUESTION: 10
| |
| A licensed Reactor Operator has failed to meet the required number of hours this past
| |
| calendar quarter to maintain an active license.
| |
| Assuming all other requirements have been met to activate the license. which of the
| |
| following watches completed under instruction would satisfy the requirement to allow
| |
| activation of'the license?
| |
| a. 24 hours as the Control Operator during Mode 5 AND 36 hours as the Control
| |
| Operator during Mode 4
| |
| b. 48 hours as the Balance of Plant Operator during Mode 5 AND 12 hours as the
| |
| Control Operator during Mode 4
| |
| c. 40 hours as the Control Operator during Mode 5
| |
| d. 40 hours as the Balance of Plant Operator during Mode 4
| |
| AN sw Ew :
| |
| d. 40 hours as the Balance of Plant Operator during Mode 4
| |
| Post Validation Kevision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 10 TIEHIGROUP: 3
| |
| KA IMPORTANCE: RO SRO 3.8
| |
| 10CFR55 CONTENT: 41@) 43(b) 5
| |
| KA: 2.1.1
| |
| Knowledge of conduct of operations requirements
| |
| OBJECTIVE: PP-3.1-1
| |
| Given a situation, STATE whether or not an off-going operator may he reelieved during the shift turnover
| |
| process
| |
| DEVE1,OPMENT HEFERENCES: OMM-001
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QCKSTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| IMWK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
| |
| NRC EXAM IIISTOHY: None
| |
| DISTRACTOR JUSTIFBCACTION (COHRECT -4NSWER dd):
| |
| a. Plausible since this exceeds the required 40 hours for the (0or HOP position, but only those hours
| |
| when the plant is above 200°F are acceptahle.
| |
| b. Plausible since this exceeds the required 40 hours for the C:O or BOP position, but only those hours
| |
| when the plant is above 200F arc acceptable.
| |
| e. Plausible since this meets the required 40 hours for the CO or BOP position aud this has the most
| |
| hours in the CO position, but only those hours when the plant is above 200°F are acceptable.
| |
| t d. 40 hours are required in either the CO or BOP position when the plant is above 200°F.
| |
| DIFFICZl1,TY ANALYSIS:
| |
| 0 COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPL.ANATION: Must recall requirement for activating an inactive license from OMhl-00 I
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Kcactor Operator
| |
| QCESTIQN: 1I
| |
| Following a loss of off-site power during recovery from a SGTR, the crew is required to
| |
| transition from EPP-019, Post SGIR Cooldown Using Steam Dump, to either:
| |
| e EPP-017, Post SGTR Cooldown Using Hac.kfill
| |
| EPP-018, Post SGTR Cooldown Using Blowdown
| |
| Which of the following describe how RCS and S G pressure control in EPP-017 compares
| |
| to that in EPP-018?
| |
| a. o EPP-017 maintains KCS pressure below the ruptured S G pressure
| |
| e EPP-0 I8 maintains RCS pressure below the nlptured SG pressure
| |
| . o EPP-017 maintains KCS pressure below the ruptured S G pressure
| |
| 8 EPP-OI 8 maintains KCS pressure the same as the niptured SG pressure
| |
| c. 8 EPP-017 maintains KCS pressure the same as the ruptured SG pressure
| |
| e EPP-OL 8 maintains RCS pressure below the ruptured SG pressure
| |
| d. e EPP-017 maintains RCS pressure the same as the ruptured SG pressure
| |
| o EPP-018 maintains RCS pressure the same as the ruptured SG pressure
| |
| ANSWER:
| |
| . o
| |
| o
| |
| EPP-017 maintains RCS pressure below the ruptured SG pressure
| |
| EPP-018 tnaintains RCS pressure the same as the niptured SC; pressure
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Opemtor
| |
| Data Slicets
| |
| QUESTION NUMBER: 1I TIEWGROUP: 111
| |
| am m w r A N c E : RO SRO 4.4
| |
| IOCFR55 CONTENT: 41(b) 43@) 5
| |
| KA: 000038EA2.08
| |
| Ability to determine or interpret the following as they apply to a SGTR: Viable alternatives for placing
| |
| plant in safe condition when condenser is trot aliailable
| |
| OBJECTIVE: EOP-3.4- I
| |
| Describe the purpose of the following EOPs including the type of event for which they were desiped atid
| |
| the major actions performed
| |
| - EPP-0 17
| |
| - EPP-0 I 8
| |
| - wP-0 I9
| |
| DEVELOPMENT REFERENCES: EPP-0 17
| |
| EI'P-0 I 8
| |
| REFERENCES SUPPLIED TO APP1,ICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFKANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 8PP-3.4 0 I0
| |
| NRC EXAM HISTORY: Harris 2002
| |
| DISTRACX'OR JUSTIFICACTION (CORREC'I' ANSWER d'd):
| |
| a. Plausible since EPP-0 I7 maintains pressure below ruptured SG pressure, but EPP-0 IS mainpains
| |
| prcssurt: the same as the ruptured S G pressure.
| |
| d b. PP-017 maintains pressure helow SG pressure to allow backfill from the SG to the KCS,while EPP-
| |
| 018 maintains pressure the same as SG pressure to minimize SG leakage.
| |
| c. Plausible since either EPP-014 or EPP-018 maintains pressure below SG pressure and either EPP-014
| |
| or EPP-018 maintains pressure the same as Sci pressure, but this distracter has the correct conditioti
| |
| reveresed.
| |
| d. Plausible since EPP-0 18 maintains pressure the same as the ruptured SG pressure, hut EPP-0 17
| |
| maintains pressure below ruptured SG pressure.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / REC:AI,I,
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge of different mitigation strategies for EPP-0 I7 and EPP-0 I8
| |
| Post Validation Kevisioti
| |
| | |
| Harris NKC Written Examination
| |
| Senior Rcactor Operatoi
| |
| QUESTION: 12
| |
| A LOCA occurred several hours ago. Only one (1) Containment Spray Pump is running
| |
| due to actions taken in EPP-012, 1,oss of Emergency Coolant Recirculation.
| |
| A transition has just been made to FKP-J. 1 Response to High Containment Pressure.
| |
| Containment Pressure is 14 p i g .
| |
| Which of the following actions should be taken?
| |
| a. Start the second Containment Spray Pump if Containment pressure does NOT
| |
| decrease below I O psig before exiting FRP-J. 1.
| |
| b. Start the second Containment Spray Pump since pressure is above 10 psig.
| |
| C. Continue operation with one Containment Spray Pump regardless of any increase
| |
| in Containment pressure.
| |
| d. Continue operation with one Contaimnent Spray Pump unless Containment
| |
| pressure begins increasing, then start the second pump.
| |
| ANSWER
| |
| c. Continue operation with one Containment Spray Pump regardless of any increase
| |
| in Containment pressure.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Exdminatioll
| |
| Senior Reactor 0pe:ator
| |
| Data Sheets
| |
| QUESTION NUILgBEW: 12 TIEWGROCP: 1i2
| |
| #A IMPORTAPNCX: RO SRO 3.8
| |
| 10CPH55 CONTENT: 41(b) 43(b) 5
| |
| KA: WE14EA2.2
| |
| Ability to determine and interpret the following as they apply to the (High Containnient Pressure}
| |
| .4dherence to appropriate procedures and operation within the limitations in the faeilitys license and
| |
| amendments
| |
| OBJECTIVE: EOP-3.13-5
| |
| Given the following E.OP steps, notes, and cautions, describe the associated basis: h. CNMT spray
| |
| operation (EPP-012 or FW-J.1)
| |
| DEVELOPMENT REFERENCES: EOP-FW-B. 1
| |
| KEFERERCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED f D%RECI: EOP-3.13-R4 OOX
| |
| NRC EXAM HISTORY: None
| |
| DISTMCITOR SUSTIFICACTION (CIORRECT ANSWER dd):
| |
| a. Plausible since this \n.ould be a normal action directed by FIU-J. I
| |
| b. Plausible since this would be a nonnal action directed by FRP-J.1
| |
| d 6. EIP-012 directs the operators to run Containment Spray Pumps based upon Containment pressure and
| |
| Fan Cooler operation. lhese actions are taken to minimize RWST depletion. This configuration is to
| |
| be maintained even if FIZP-J. I is implemented.
| |
| d. Plausible since would better serve the intent of EPP-012, but would be contradictory to the intent of
| |
| FKP-J.i which has a higher priority concerning the operation of the Spray Pumps.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| nmwx;r,n~RATING: 3
| |
| EXPLANATIOK: Must compare the relative actions in the 2 procedures and make a judgement of
| |
| which condition takes precedent
| |
| Post Validation Revision
| |
| | |
| Elanis NRC Written bxamination
| |
| Senior Reactor Operator
| |
| QGESTION: I3
| |
| During operation at 100% power, an inadvertent SI occurs on 'R' Train ONLY.
| |
| Which of the following actions is required?
| |
| a. Maiiuaily actuate SI on 'A' Train and continue in PATII-I
| |
| b. Continue in PATH-I noting which 'A' Train ESF equipment is NOT running
| |
| c. Start ONLY the 'A' Train of ESF equipment for which the redundant 'B' Train
| |
| equipment failed
| |
| d. Transition directly to EPF-OOXI SI 'l'ermination
| |
| ANSWER:
| |
| a. ktanually actuate SI on 'A' Train and continue in I'AI'lB- 1
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QIJESTIONNU,MBER: 13 TIEWGROUP: 2i
| |
| KA IMPORTANCE: RO SRO 4.6
| |
| 10CPH55 CONTENT: 48(h) 43(b) 5
| |
| KA: 013A2.01
| |
| Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b)
| |
| based on those predictions, use procedunx to correct, control, or mitigate the conscquences of those
| |
| malfunctions or operations: L(3CA
| |
| OBJECTIVE: IE-3. IO-K4
| |
| Describe the expected operator actions associated with an imminent RPS or ESFAS actuatiun
| |
| DEVELOPMENT REFERENCES: EOP Users Guide
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: XEW SIGNIFICANTLY ILlODPFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: IE-3.10-R4 001
| |
| NHC EXAM HISTORY: Ilamis 2000
| |
| DISTRACTOR JUSTlFICACTION (CORRECT ANSWER dd):
| |
| d a. Prefcned method ofniannal actuation although it would be acceptable to start ;reposition all
| |
| equipment which would be actuated regardless ofthe perceived need since diagnostics have not yet
| |
| been perfornied.
| |
| b. Plausible since only a single train actuation is analyzed, but efforts are to be made to initiate both
| |
| trains.
| |
| E. Ilausible since starting equipment as needed would provide adequate protection, but since diagnostics
| |
| have not yet been completed the equipment required map not yet be known.
| |
| (8. Plausible since one of the goals following an inadvertent Si is to terminate SI as soon as criteria are
| |
| met to prcvent overfilling !pressurizing the KCS, but procedures are written assuming both trains
| |
| started.
| |
| DIFFHCULTY ANALYSIS:
| |
| COMPREHENSIVE I ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICIJI.TY RATING: 3
| |
| EXPLANATION: Required knowledge of procedural requirements for a single train of ESF
| |
| actuation
| |
| Post Validation Revision
| |
| | |
| Harris NKC' Written Examination
| |
| Senior Reactor Opetator
| |
| QUESTION: 14
| |
| Given the following conditions:
| |
| e 1CS-235, Charging Line Isolation, was closed to establish a clearance boundary for
| |
| maintenance on ICs-238.
| |
| e 1CS-235 had to be inanuailg torqued shut.
| |
| e 1CS-235 is a Liinitorque SMR-OO!SR-OO motor-operated valve.
| |
| Prior to declaring 1CS-235 operable after the clearance is removed, the valve must be ...
| |
| a. verified to have the torque switch c.alibrated correctly
| |
| b. stroked with the control switch.
| |
| c. monitored for seat leakage.
| |
| d. manually stroked full open.
| |
| ANSWER:
| |
| b. stroked with the control switch.
| |
| Post Validation Revision
| |
| | |
| IIarris NRC Written Exarninatioii
| |
| Senior Rcactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER. 14 TIEWGROUP: 3
| |
| Iw IMPORTANCE: RO SRO 3.1
| |
| 1QCFR55CONTENT: 4I(b) 43Eb) 5
| |
| KA: 2.2.19
| |
| KnowIedge of maintenance work order requirements
| |
| OBJECTIVE: PP-2.4-1
| |
| Identify the primary functions and explain the responsibilities of the Work Coordination Centet
| |
| DEVELOPMENT REFERENCES: OMh.1-014
| |
| REFERENCES SIJPPLIED TO APPLICANT: None
| |
| QUMTIOX SOURCE: SIGNIFICA?XI,Y MODIFIED DIRECT
| |
| CANTLY MODIFIEI) / DIRECT: 028
| |
| NRC EXAM HISTORY: Hareis 2000
| |
| DISTKACTOR JIWTFHFICACTION (CORRECT ANSWER 4'61):
| |
| a. Plausible since the valve has been nianually torqued onto the seat, hut the requirement is that the valve
| |
| must he stroked electrically from the control switch.
| |
| d b. All Lirtiitorque SMI3-00!SB-00 motor operated valves, if manually operated, are required to be stroked
| |
| electrically from the control switch to he declared operable.
| |
| c. %"ible since over torqueing a \tahe may result in seat leakage, but the requirement is that the valve
| |
| must he stroked electrically from the control switch.
| |
| d. Plausible since the valve was manually torqued closed, hut the requirement is that the valve must be
| |
| stroked elecrricaliy from the control switch.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS I(NOWLE1)GE / RECALI,
| |
| DIFFICULTY RATING: 3
| |
| EXPLAKATION: Kriowlc.dge of administrative post-work practices required
| |
| Post Validation Revision
| |
| | |
| Harris NR( Written Examination
| |
| Senior Rcdctor Operator
| |
| QUESFHBN: 15
| |
| Given the following conditions:
| |
| o Following a Reactor Trip and Safety Injection, a transition has eventually been made
| |
| to EOP-EPP-015, Uncontrolled Depressurization of All Steam Generators.
| |
| o Both Main and Auxiliary Feed Flow h a w been isolated to all SGs.
| |
| o Directions have just been given to locally isolate steam flows from all SGs.
| |
| o SG A pressure appears to have stabilized at approximately 100 psig, while the other
| |
| X i s have completely depressurized.
| |
| Which of the following actions should be taken?
| |
| a. Transition to EOP-EPP-014. Faulted S G Isolation, since this is indication that
| |
| SG A has been isolated.
| |
| b. Continue in EOP-EPP-015 and re-establish AFW flow to SG A at minimum
| |
| flow.
| |
| c. Transition to EOP-IATK-2 if lwa1 radiation surveys indicate priinary-to-
| |
| secondary leakage is occurring.
| |
| d. Transition to EOP-T;PP-OOR, SI Termination, to prevent overprcssurizing the
| |
| RCS.
| |
| AWSWEK:
| |
| c. Transition to EOP-PATH-2 if local radiation surveys indicate primary-to-
| |
| secondary leakage is occurring.
| |
| Post Validation Rei2ision
| |
| | |
| Harris NRC' Written Fxamiiiation
| |
| Senior Keactor Operator
| |
| Data Sheets
| |
| QI!ESTION NUMBER: 15 TIERIGROUE 1/1
| |
| IC$ IMPORTANCE: RO SRO 3.8
| |
| 80CPR55 COYIENT: 41(b) 43(b) 2
| |
| Iw: 000040Ci2.I.32
| |
| Ability to explain and apply all system limits and precautions. (Steam 1 . k Rupture - Excessive Heat
| |
| Transfer)
| |
| OBJECTIVE: EOP-3.9-7
| |
| Given a step, caution, or note from an emergency procedure, slate its purpose
| |
| DEVELOPMENI' REFERENCES: EC)P-EPP-015
| |
| REVERENCES SUPPLIED TO APPI,ICANT: None
| |
| BANK NUMBER FOR SIGNIFICANT1,Y MODIFIEI) / DIRECT: New
| |
| NRC EXAM IIISTORY: None
| |
| DlSTRACTOR JUSTHFYCAC.TION(CORRECT ANSWER d'd):
| |
| a. Plausible since once a SG is confirmed to be isolated in EPP-0 15, a foldout page item directs a
| |
| transition to EPP-014.
| |
| b. Plausible since without indications o f a SC; tube leak, actions would he taken to remain in EPP-015
| |
| and maintain feed flow at minimum.
| |
| 4 c. A SCi may he suspected to he iuptured if it fails to dry out following isolation of feed flow. Local
| |
| checks for radiation can be used to confirm ~ r i n i a r ~ ~ t o - s ~ cleakage.
| |
| otida~
| |
| 41. Plausible since a desired goal after isolating a faulted S G is to terminate SI as soon as conditions are
| |
| met to prevent overfilling and overpressurizing the IICS.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNQWLEDGE I RECALL
| |
| DIFFICULTY KhTlllL'G: 3
| |
| EXPLANATION: Must analyze the cause ofthc failure of the SG to depressurize and then
| |
| determine the correct actions based on the analysis.
| |
| Post Validation Revision
| |
| | |
| I h - i s NRC Written Exaniination
| |
| Senior Keactor Operator
| |
| QUESTION: 16
| |
| The unit has tripped due to a I.OCA and ESF equipment has failed to start. As a result,
| |
| EOP-FRP-C.2, Response to Degraded Core Cooling, has bcen entered.
| |
| A depressuridon of the Steam Generators (SGs) to 80 psig is being performed, in
| |
| accordance with the procedure, when the STA reports that a Red Path condition for Integrity
| |
| has occurred.
| |
| Which of the following actions slriould be taken?
| |
| a. Inirnediately transition to EOP-FRP-P. 1, Response to imminent Pressurized
| |
| Thermal Shock Conditions
| |
| b. Stop the S!G depressurization and, if the red path does not clear, transition to EOP-
| |
| FRP-P. I , Response to Imminent Pressurized Thermal Shock Conditions
| |
| c. Coniplete FOP-FRP-C.2 and then transition to EOP-FRP-P. I , Response to
| |
| Imminent Pressurized Thermal Shock Conditions, if the red path still exists
| |
| d. Complete the S!G depressurization and then transition to EOP-FRP-P. 1. Response
| |
| to Imminent Pressurized Thermal Shock Conditions, if the red path still exists
| |
| ANSWER:
| |
| c. Complete FOP-FW-C.2 and then transition to FOP-FRY-P. I , Response to
| |
| Imminent Pressurized Thermal Shosk Conditions, if the red path still exists
| |
| Post Validation Revision
| |
| | |
| Harris NKC Written Examination
| |
| Senioi Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 16 TIEWGROUP: 1!3
| |
| M A IMPORTANCE: RO SRO 3.8
| |
| 1QCFR55CONTENT: 41(b) 43(b) 2
| |
| KA: WEObG2.1.32
| |
| Ability to explain and apply a11 system limits and precautions. (Degraded Core Cooling)
| |
| OBJECTIVE: IiOP-3.10-4
| |
| Given the following EOP steps, notes, and cautions, describe the associated basis
| |
| g. Stopping SG depressurization at 80 p i g (C.2)
| |
| DEVEEOPMENT REFERENCES: EOP-FRP-C.2
| |
| REFERENCES SUPPLIED TO APP1,ICANT: Kone
| |
| QUESTION SOBJRCE: SIGNIFICANTLY IL1B)DBFIED DIRECT
| |
| CAXTLY MOMFIED /DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DETRACTOR JI!STIFHCACTION (CORRECT ANSWER %Id):
| |
| a. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
| |
| into I<OP-FKI-C.2, but under these particular conditions a transition should not occur until completion
| |
| of the EOP-FKP-C.2.
| |
| b. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
| |
| into EOP-FKP-C.2, but under these particular conditions a transition should not occur until conipletion
| |
| of the EOP-FKP-C.2.
| |
| d c. During the depressurization, a red path may occur due to injecting the accumulators. A transition
| |
| should not be matie until the entire procedure has been completed.
| |
| d. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
| |
| into EOP-IXP-C.2. but uiidcr these particular conditions a transition should not occur uritil completion
| |
| of the EOP-FKP-C.2.
| |
| DIFFICULTY AWALYSHS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPIANATION: Must analyze plant cotiditions to determine that the cause of the red path is the
| |
| depressurization and that, under these specific conditions, an immediate
| |
| transition is not warranted
| |
| Post Validation Kevision
| |
| | |
| Harris NRC Written Exammation
| |
| Senior Reactor Operator
| |
| QUESTION: 17
| |
| Given the following conditions:
| |
| e The unit is in Mode 3.
| |
| 0 Instrument I3uses IDP-IB-SI1 and IDP-IB-SIV are both de-energized.
| |
| 0 Maintenance reports that Instrument Bus IDP-IB-SI1 is ready to be re-energized.
| |
| In order to prevent an inadvertent Safeguards Actuation, which of the following tnust be
| |
| verified prior to re-energizing the bus and why?
| |
| a. Train A Logic Input Error inhibit must be verified to he in INHIBIT due to the
| |
| proper coincidence for an actuation being available
| |
| b. Train A Logic Train Output must be verified to be in T E S l to prevent an
| |
| inadvertent Safeguard Actuation due to the loss of the SI BLOCK Signals
| |
| c. Train R Logic Input Error Inhibit must be verified to be in INHIBII due to the
| |
| proper coincidence for an actuation being available
| |
| d. Train B Logic Train Output must be vcrified to be in TESI to prevent an
| |
| inadvertent Safepard Actuation due to the loss ofthe SI BLOCK Signals
| |
| ANSWER:
| |
| d. Train 3 Logic Train Output must be verified to be in TFST to prevent an
| |
| inadvertent Safeguard Actuation due to the loss ofthe SI BLOCK Signals
| |
| Post Validation Revision
| |
| | |
| Waris NRC Written Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTION NWMBER: 17 IIEWGROIJP: 2: I
| |
| KA IMPORTANCE: RO SRO 3.4
| |
| IOCPRS5 CONTENT: 4I(b) 43b) 2
| |
| KA: 062G2.2.22
| |
| Knowledge of limiting conditims for operations and safety limits. (.4CElectrical Distribution)
| |
| OBJECTIVE: ESFAS-3.0-4
| |
| Given applicable logic diagrams and a set of plant conditions. predict how loss of any of the four
| |
| instrument buses will affect the ESFAS output functions of each SSPS train.
| |
| DEVELOPMENT REFERENCES: OP-156.02
| |
| REFERENCES SGTPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY b1ODIFIED 0 DIRECT
| |
| BANK NUMBER FOR SIGNIPICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM TORY: None
| |
| DPSTRACTOW .JUSTYFICAC:TlQSN(CORRECT ANSWER kd):
| |
| a. Plausible since the loss of hoth trains of power will provide the proper coincidence, but power must he
| |
| available to the output relays to actuate. Placing the input error inhibit in INHIBIT at this time will
| |
| not prevent an actuation since the logic is already made up. Also the incorrect Train.
| |
| b. Plausihle since the loss of both trains of power causes the SI Block signals to he lost and when either
| |
| ofthe supplies is restored, power will he available to the output relays to cause an actuation, however
| |
| this occurs on Train B for this event.
| |
| c. Plausihie since the loss of both trains of power will provide the proper coincidence, hut power must he
| |
| available to the output relays to ac.tuate. Placing the iuput error inhihit in INHIBIT at this time will
| |
| riot prewnt an actuation since the logic is already rnade up.
| |
| 4 d. The loss of both trains of power causes the SI Block siguais to he lost. When either of the supplies is
| |
| restored, power will he available to the output relays to cause an actuation.
| |
| I)IFFICI!LTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS 0 KNOWIXDGE /RECALL
| |
| DIFF%CI!LTYILITING: 3
| |
| EXPLANATION: Must deteimine train of SSPS affected by the 10% of power and then analyze the
| |
| effect of partially restoring power
| |
| Post Validation Revision
| |
| | |
| f Iarris NKC Written Examination
| |
| Senior Reactor Operator
| |
| QUESTION: 18
| |
| The Unit-SCO and Superintendent-Shift Operations are discussing invoking
| |
| 1OC'FR50.54(x) during the implementation of the F'mergency Operating Procedures due
| |
| to a condition arising which is NVQ'I'addressed by the procedures or Technical
| |
| Specifications.
| |
| \J%ich of the foilowing conditions must be met when invoking 1OCFR50.54(x)?
| |
| a. The action must be approved by an additional licensed Senior Keactor Operator
| |
| when the action is necessary to prevent equipment damage.
| |
| b. The action must be approved by the Superintendent-Shift Operations prior to
| |
| taking the action.
| |
| c. The NRC must concur with the action to be taken prior to the action actually being
| |
| taken.
| |
| d. The action must be approved by the Manager-Operations when the action is
| |
| necessary to protect plant personnel.
| |
| ANSWER:
| |
| b. The action must be approved by the Superintendent-Shift Operations prior to
| |
| taking the action.
| |
| Post Vaiidation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| Dzta Sheets
| |
| QUESTION NUAMBER: 18 TIEW<;ROUP: 3
| |
| KA IMPORTANCE: RO SRO 3.3
| |
| 10CFR55 CONTENT: 41@) 43(b) 3
| |
| KA: 2.2.10
| |
| Knowledge of the process for determining if the margin ofsafety, as defined in the basis ofany t e c h n i d
| |
| specification is reduced by a proposed change, test or experiment
| |
| OBJECTIVE: PP-2.0-S2
| |
| LIST the actions required by the individual who authorizes a deviation from the Technical Specifications
| |
| or license conditions
| |
| DEVELOPMENT REFERENCES: PRO-NGGC~-0200
| |
| REFERENCES SUPPLIED TO APPLICANT: XOIK
| |
| QUESrXON SOURCE: SIGNIPICANTLY MODIFIED DIRECT
| |
| CANTLY MODIFIED / DIRECT: 0 23318
| |
| NRC EXAM HISTORY: None
| |
| DIS~RAC~TOK JUSTIPIC.4CTION (CORRECT ANSWER >I'd):
| |
| a. Plausible since lOCFRS0.54(x) esquires that a licensed SRO approve any actions which deviate from
| |
| license conditions prior to perfomlance. but the actions must be to protect the health and safety of the
| |
| public.
| |
| 4 b. The niinimurn level of approval per PRO-NGGCX200 is the Supcrintendent-Shifr Operations, but it
| |
| can be approved by any personnel holding an SRO license above this position also.
| |
| c. Plausible since the XRC must be notified, but the notification requirements are within 1 hour per AP-
| |
| 617.
| |
| d. Plausible since the Manager-Operations can approve a deviation if he holds an SKO license, but the
| |
| actions must be to protect the health and safety of the public.
| |
| m r L m ANALYSIS:
| |
| COMPREHENSIVE./ ANALYSIS KNOWLEDGE /RECAI,I.
| |
| DIFFlCULTY RATING: 2
| |
| EXPLANATION: Requires knowledge of requirements for process of perfoiming actions not
| |
| described in any licensing basis documents.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| QBJESTYON: I9
| |
| Given the following conditions:
| |
| e t d partiall: loaded.
| |
| Following a INSS of All Power, EDG 1A-SA ha b :n r e s ~ ~ r and
| |
| e A transition has been made to FOP-EPP-003. "Loss of All AC Power Recovery with
| |
| SI Required."
| |
| e EDG IA-SA is currently loaded to 4.5 MWe and 3.5 MVAR.
| |
| Which of the following would result in an UNACCEPTABLE loading condition for EDG
| |
| 1A-SA?
| |
| a. 0 Pick up an additional 0.5 MWe
| |
| e Pick up an additional 0.1 MVAK
| |
| b. e Pick up an additional 1 .O MWe
| |
| e Pick up an additional 0.5 MVAR
| |
| c. 0 Pick up an additional I .5 MWe
| |
| 0 Pick up an additional 1 .O MVAR
| |
| d. e Pick up an additional 2.0 MWe
| |
| 0 Pick up an additional 1.2 MVAR
| |
| ANSWER:
| |
| c. e Pick up an additional 1.5 MR7e
| |
| 0 Pick up an additional 1.O MVAR
| |
| Post Validation Revision
| |
| | |
| IIarris NRC: Written Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: I O TIEWGROUB: lil
| |
| KA IMPORTANCE: RO SRO 4.6
| |
| 10CFR.55 CONTENT: 41(b) 43(b) 5
| |
| KA: 0011056.4A2.14
| |
| Ability to deterniine and interpret the following as they apply to the Loss of Offsitc Power: Operational
| |
| status of EDKis (A and B)
| |
| OBJECI'IVE: EOP-3.7-6
| |
| Given a step, caution, or note from K?P-001, EOP-002, or EOP-003, state its purpose
| |
| DEVIELOPIIPENT REFERENCES: OP-155, Attachment 9
| |
| EOP-!PI'-003
| |
| REFERENCES SUPPLIEI) TO APPLICANT: OP-155, Attachment 9
| |
| QUESTION SOURCE: XEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR S'IGNIFICANT1,Y ILIODIPIEI) I DIRECT: N~\v
| |
| NRC EXAM I m m R Y : NOIK
| |
| DISTRACTOR .IUSTIFICAC:TIION (CORRECT ANSWER +d):
| |
| a. Plausible since new loading will be 5.0 MWe and 3.6 MVAR, which is just within acceptable limits.
| |
| b. Plausible since new loading will he S.5 MWe and 4.0 MVAR, which is just within acceptable limits.
| |
| 4 c. New loading will he 6.0 MWe and 4.5 MVAR, which is outside acceptable limits
| |
| d. Plausible since new loading will be 6.5 MWe and 4.7 MIVAK, which is just within acceptable limits.
| |
| DIFFICUL'I'Y ANALYSIS:
| |
| COMPREHENSIVE I ANALYSIS KNOW'IXDGPS I RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPEANATIOY: Must analyze EDG operability curve tu determine whether additional MWe and
| |
| MVAK loading is within acceptable limits
| |
| Post Validation Revision
| |
| | |
| Iarris NKC Written Examination
| |
| Senior Reactor Operator
| |
| QUESTIOK: 20
| |
| h reactor trip occurred due to a loss of offsite power. The plant is being cooled down on
| |
| RIIK per EPF-006. Natural Circulation Cooldown with Steam Void in Vessel with
| |
| RVLIS.
| |
| e RSS cold leg temperatures are 190'F.
| |
| e Steam generator pressures are 50 ps~g.
| |
| e RVLIS upper range indicates greater than 100%.
| |
| e Three CRDM fans have been running during the entire cooldown.
| |
| Steam should be dumped from all SGs to ensure ..
| |
| a. boron concentration is equalized throughout the RCS prior to taking a sample to
| |
| verify cold shutdown boron conditions.
| |
| b. all inactive portions of the RCS are below 200°F prior lo complete KCS
| |
| depressurization.
| |
| c. RCS and SG temperatures are equalized prior to any subsequent RCP restart.
| |
| d. RCS ternperatures do not increase during the required 29 hour vessel soak period.
| |
| AFJSWER:
| |
| b. all inactive portions ofthe RCS are below 200'F prior to complete RCS
| |
| depressurization.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Esainination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTION NIJMBER: 20 TIEWGROUP Ii2
| |
| KA IMPORTASCE: RO SRO 3.8
| |
| lOCFRS5 CONTEXT: 41(b) 436b) 2
| |
| K k . WE09G2.1.32
| |
| Ability to explain and apply all system limits and precautions. (Natural Circulation Operations)
| |
| OBJECTIVE: EOP-3.8-2
| |
| Dcmotistrate the helowassumed operator knowledge from the SHNPP Step Ihkation 1)ocument and the
| |
| WOG EI1Gs that support perhniance of EOP actions: Determining that upper head and SG IJ-tube
| |
| temperatures are below 200 F
| |
| DEVELOPMENT REFERENCES: FOP-EPP-MJB
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QIJESTION SOURCE: SIGNIFICANTLY ?&ODIFI[EU [3DIRECT
| |
| CANTLY MODIFIED / DIRECT: EOP-3.S 006
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTHON (CORRECT ANSWER ?Id):
| |
| a. Plausible siiice this action would kave been performed in this procedure, but must he completed prior
| |
| to depressurizing the RCS below 1900 psig.
| |
| 4 b. SG pressure above (I psig indicates that the SGs are above 200°F. Depressurizing the RCS under this
| |
| condition will result in additional void formation in the SG u-tubes.
| |
| c. Plausible since KCP operation throughout NC Cooldown is desirable, hut will not be performed at this
| |
| point in the procxdure.
| |
| d. Plausible since a soak period is addressed, but only if continued operation of CKDM h u s had not been
| |
| maintained.
| |
| DHFFICWLTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPI.ANATIOX: Must analyze the conditions and determine that the entire RCS is not below
| |
| 200°F and the effect of depressurizing uiider these conditions.
| |
| Post Validation Revision
| |
| | |
| Harris NKC Written Examination
| |
| Senior Reactor Opcrator
| |
| QUESTION: 21
| |
| During an emergency, a worker has been directed to enter a high radiation area and
| |
| perform a repair necessary for the protection of valuable property.
| |
| In accordance with PEP-330, Radiological Consequences, the workers exposure
| |
| should be limited to ., .
| |
| a. 10 Rem TEDE and the entry does NOT require specific Site Emergency
| |
| Coordinator authorization.
| |
| b. 10 Rem TEDE and the entry requires specific Site Emergency Coordinator
| |
| authorization.
| |
| c. 25 Rem TEDE and the entry does NOT require specific Sitc Emergency
| |
| Coordinator authorization.
| |
| d. 25 Rem TEDE and the entry rtyuires specific Site Emergency <:oordinator
| |
| authorization,
| |
| ANSWER:
| |
| b. 10 Rem TEDE and the entry requires specific Site Emergency Coordinator
| |
| authorization.
| |
| Post Validation Revision
| |
| | |
| Harris KRC Written Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 2 I TIEWGROUP: 3
| |
| KA IMPORTANCE: RO SRO 3.3
| |
| BOCFR55 CONTENT: 41(b) 43(b) 4
| |
| %<A: 2.3.7
| |
| Knowledge of the process for preparing a radiation xvork permit
| |
| OBJECTIVE: EPZO-2A
| |
| Identify the types of protective actions for HiVP persotmrl (both on and off-site) and who is responsible
| |
| for directing them.
| |
| DEVELOPMENT REFERENCES: PEP-330
| |
| ROFEMENCES SIJPPLIEIP TO APP1,ICANT:
| |
| ~~ None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNWIICANTLYMODIFIED I DLRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR .IUSTFIF%CACTION (CORRECT .4NSWER +d):
| |
| a. Plausible since 10 rem TEDE for protecting valuable company property, hut S-SO approval is
| |
| required.
| |
| d h. Exposure is limited to 10 rem TEDE is the limit required for this activity and S-SO approval is
| |
| required.
| |
| c. Ildusibk since 25 rem T E W is the limit required for lifesaving efforts, but the limit to protect
| |
| equipment and property is 10 rein IEDE.
| |
| d. Plausible since 25 rem TEDE is the limit required for lifesaving efforts, but the Limit to protect
| |
| equipment and property is 10 rem TELX.
| |
| DIFFICCLTY ANALYSIS:
| |
| CY3MPREHENSIVE 1 ANN,YSIS KNOWLEDGE / RECAI.1,
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Requires knowledge of the emergency exposure limits and approval
| |
| requirements
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| Given the folhwing conditions:
| |
| m Power i s currently at 32% during a plant startup.
| |
| e Instrument Bus IDP- I R-SIV deenergiied as a result of a fault in PIC CAB-4.
| |
| * PIC ('AB-4 has been isolated from Instrument Bus SIV atid will be deenergiied fix
| |
| approximately eight (8) hours while repairs are being made.
| |
| Which ofthe following actions must be taken'?
| |
| a. Place all PIC CAB-4 Reactor Trip instruments in the tripped condition
| |
| b. Place all PIC C A B 4 ESP instruments in the tripped condition
| |
| c. Place all MFW Regulating Valves in MANUAL
| |
| d. Perform a plant shutdown
| |
| ANSWER:
| |
| d. Perform a plant shutdown
| |
| Post Validation Revision
| |
| | |
| Harris N I X Written Examination
| |
| Senior Reactor Operator
| |
| L)ata Sheets
| |
| QUESTION NUMBER: 22 TIEWGROUP: l/l
| |
| K4 IMPORTANCE: RO SRO 4.1
| |
| 10CFR55 CONTENT: 41(b) 43(b) 2
| |
| K h : 000057Ci2.2.22
| |
| Knowledge of limiting conditions for operations and safety limits. (Loss of Vital AC Instrument Bus)
| |
| OBJECTIVE: AOP-3.24-4
| |
| Dcterniine the following: a. Consequences of the loss of all power to PIC CAR-4
| |
| DEVELOPMENT REFERENCES: AOP-024
| |
| TS Table 3.3-3, pg 3-18 and 3-27
| |
| TS 3.0.3, pg 0-1
| |
| REFERENCES SUPP1,IE.DTO APPLICANT: None
| |
| QUESTION SBtJRCE: NEW SIGNIFICANTLY MODIFPED DIRECT
| |
| RANK IVLMBER FOR SIC; CANTEY MODIFIED / DIRECT: AOP-3.24-R4 00 I
| |
| NRC EXAM HISTORY: None
| |
| IPISTR4CrOR .JUSTIIFICACTION (CORRECT ANSWER Jd):
| |
| a. Piausible since instrunlent failures require bistables tripped, hut they are deenergized to actuate and
| |
| are already tripped since no power is available.
| |
| b. Plausible since instrument failures require bistables tripped, but they are dcenergized to actuate and
| |
| are already tripped since no power is available.
| |
| c. Piausible since this is the immediate operator action for a loss of Instrument Bus SUI, not SIV.
| |
| \ d. Loss of all power to PIC CAB-4 will result in 3 bistable channels of Steam Line Pressure becoming
| |
| inoperable. The TS action is to trip the bistables within one hour, but the bistables are energized to
| |
| actuate. Without power available, this xtion cannot be perfomicd and TS 3.0.3 becomes applicable
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREIIENSIVE / AXALYSIS KNOWLEDGE / RECALL
| |
| DIFFICUI.1Y RATING: 4
| |
| EXPLANATION: Must recognize that energized to actuate bistables cannot be placed in tripped
| |
| condition without power, thus an entry into TS 3.0.3 is required, arid must
| |
| determine the required TS 3.0.3 actions
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Opermr
| |
| QUESTION: 23
| |
| During the performance of FOP-PATH-2, the S'1 A reports that the following two (2)
| |
| YEI.I.OW path Critical Safety Function Status Trees (CSFST) exist:
| |
| e Integrity
| |
| e Ileat Sink
| |
| Which ofthe following describes how these YEILOVV paths are to he addressed and !or
| |
| implemented?
| |
| a. Both must be addressed and implemented, with Heat Sink having a higher priority
| |
| than Integrity, as soon as FOP-PATII-2 actions are completed provided no other
| |
| higher priority CSFST conditions exist
| |
| b. Both must be addressed, hut implemented at the discretion of the Superintendent-
| |
| Shift Operations, prior to exiting from the N I P network
| |
| c. Both must be addressed and implemented, with Heat Sink having a higher priority
| |
| than Integrity, prior to exiting from the EOP network
| |
| d. Both must be addressed, hut implemented at the discretion ofthe Superintendent-
| |
| Shift Operations, as soon as EOP-PATH-2 actions are completed provided no
| |
| other higher priority CSFST conditions exist
| |
| ANSWER:
| |
| b. Both must be addressed, but implemented at the discretion of the Supsrintendent-
| |
| Shift Operations. prior to exiting from the FOP network
| |
| Post Validation Kevision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Opcrator
| |
| Data Sheets
| |
| QUESTION NUMBER: 23 'I'IEWGROUP: 3
| |
| KA IMPORTANCE: RO SRO 4.0
| |
| 10CFR55 CONTENT: 41@) 43(b) 5
| |
| Kti: 2.4.22
| |
| Knowledge of the bases for prioritizing safety functions during abnornial!eiiiergeney operations
| |
| OBJECTIVE: EOP-3.19-2
| |
| Describe C:ontrol Room usage of status trees as it relates to the following
| |
| a. Priority of status trees
| |
| b. Rules of usage
| |
| DEVELOPMENT REFERENCES: FOP 1Jser's Guide
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRE.CT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DISTRAC~'ORJUSTIFICACTION (C:ORRECTANSWER +d):
| |
| a. Plausible since they are to he addressed, but only prior to leucing the 1;OP network and are not
| |
| required to be implernented.
| |
| d b. All YELLOW-condition CSFSl's should be addressed prior to exiting the E,OP network. However, the
| |
| operator is allowed to decide if and urhen to iniplernent. and whether to complete any YELLOW-
| |
| condition FRP.
| |
| c. Plausible since they are to be addressed, but only prior to leacing the EOP network and arc not
| |
| required to be implemented.
| |
| d. Plausible since they are to he addressed, but only prior to leaving the HOP network and are not
| |
| required to be implemented.
| |
| DWFHCULTY ANALYSIS:
| |
| COR'IPREMENSIVE / ANALYSIS KNOWIXDGE /RECALL
| |
| DIFPICXJLTY HATING: 2
| |
| EXPIANATION: Knowledge of the iniplenientation criteria for yellow CSFSTs as directed by
| |
| plant procedures
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reac.tor Operator
| |
| QUESTION: 24
| |
| Following a loss of all h C power, how long are the safety-related 125 VDC batteries
| |
| DESIGNED to allow equipment operation'!
| |
| a. 2 hours, assuming DC h a d shedding occurs within 30 minutes ofthe loss ofail
| |
| AC power
| |
| b. 2 hours, assuming DC load shedding occurs within 60 minutes of the loss of all
| |
| AC power
| |
| c. 4 hour?, assuming DC load shedding occurs within 30 minutes ofthe loss of311
| |
| AC power
| |
| d. 4 hours, assuming DC load shedding occurs within 60 minutes of the loss of all
| |
| AC power
| |
| ANSWER:
| |
| d. 4 hours, assuming I)C load shedding occurs within 60 minutes of the ioss of all
| |
| AC power
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QUEST ION NUMBER: 24 TIER/GROUP: iil
| |
| KA IMPORTANCE: RO SRO 3.7
| |
| 10CPR55 CONTENT: 4101) 43(b) 2
| |
| K.4: 000058G2.2.25
| |
| Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
| |
| (L.oss of DC Power)
| |
| OBJECTIVE: EOP-3.7-6
| |
| Given a step, caution, or note from EOP-001, EOP-002. or ECP-003, state its purpose
| |
| DEVELOPMENT REFERENCES: Tech Spec Bases 3.8.2, pg 8-2
| |
| mP-EPP-00 I
| |
| ADEL-LP-2.6
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTKON SOIIRCIF,: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: ADI!.L2-6-S I 00 I
| |
| NRC EXAM HISTORY: Kone
| |
| DISTRACTOR SUS'IIFICACTION (CORRECT ANSWER d'd):
| |
| a. Plausible since this is the time limit which requires actions being taken in accordance with Technical
| |
| Specifications, but the design of the batteries is 4 hours.
| |
| b. PlausihIe since this is the tinie limit which requires actions being taken in accordance with Technical
| |
| Specifications: hut the design of the batteries is 4 hours.
| |
| c. Plausible since the design of the hatteries is 4 hours, but the design assumes that DC Load shedding
| |
| occurs within 60 minutes, not 30.
| |
| d d. Batteries are designed to carry required safety related loads for up to 4 hours without AC input to
| |
| cany bus or charge battery, assuming that required load shedding occurs within I hour.
| |
| DIFFIC:UI.TY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE:/ RECALL
| |
| DIFFICULTY RA'TING: 3
| |
| EXPLANATION: Knowledge of tech spec basis and design of safety-related batteries
| |
| Post Vaiidation Revision
| |
| | |
| Harris KRC Written Examination
| |
| Senior Rcactor Operatot
| |
| QUESTION: 25
| |
| Which of the following actions would be INAPPROPRIATE to perfomi prior to
| |
| direction in an EOP?
| |
| a. Isolating AFW' tlow to a single faulted SG
| |
| b. Throttling AFW flow to control a ruptured SG level within the required level hand
| |
| c. Securing a CSIP to prevent overfilling the pressurizer foliowing an inadvertant SI
| |
| d. Shutting the MSIVs to isolate a steamline break which has not resulted in an SI
| |
| ANSWER:
| |
| c. Securing a CSIP to prevent overfilling the pressurizer following an inadvertant SI
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Senior Reactor Operator
| |
| Data Sheets
| |
| QWESTION NUMBER: 25 TIEWGROUB: 3
| |
| KA IMPORTANCE: RO SRO 3.9
| |
| lQCFR55CONTENT: 41(b) 43(b) 5
| |
| KA: 2.4.14
| |
| Knowledge of general pidclines for EOP flowchart use
| |
| OBJECTIVE: EOP-LP-3.19-1
| |
| Describe Control Room wage of the EOP network as it relates to the following: a) Peilbrming steps out
| |
| of sequence
| |
| DEVELOPMENT REFERENCES: KOP Users Guide
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| RANK NUMBER FOR SIGNIFICANTLY hIODIFIE1) / DIRECT: EOP-3.19-RI 0 18
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER dd):
| |
| a. Plausible since this is a numbered step in P N H - I which are normally required to he performed in
| |
| sequence, but the EOI IJsers Guide addresses this as being acceptable.
| |
| b. Plausible since this is a numbered step in PATH-I which are nonnaily required to be perfonned in
| |
| sequence, hut the EOP Users Guide addresses this as being acceptable.
| |
| v E. Performing steps out of sequence is allowed. but must be done with caution to prevent masking
| |
| symptoms or defeating the intent of the JiOP being used. Although terminating SI early might be
| |
| heneficial to prevent filling the pressurizer if the only event is a spurious SI, this may result in further
| |
| degradation of the piant if another undiagnosed eveut is in progress.
| |
| d. Plausible since this is a nnmbered step in PAIH-1 which arc nonnally required to be perfixmed in
| |
| sequence, but the EOP Users Guide addresses this as being acceptable.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREIIENSLVE / ANALYSIS 0 KNOWLEDGE /RECALL
| |
| DIFFICU1,TY RATIXG: 3
| |
| EXPLANATION: Must differentiate between those actions which could potentially result in
| |
| degradation ofthe plant if taken out of sequence and those actions which would
| |
| likely have little impact on the operators abilities to diagnose other events.
| |
| Post Validation Revision
| |
| | |
| . -
| |
| HARRIS EXAM
| |
| 50-40812004-304
| |
| -
| |
| FEBRUARY 23 27,2004
| |
| & MARCH 4,2004 (WRITTEN)
| |
| 'I
| |
| | |
| Harris WRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 1
| |
| Following a Reactor Trip, the RCS temperature is being controlled by the S t e m Dump
| |
| Control System at 540°F. FOP-EPP-004, Reactor Trip, directs that the WCS be
| |
| maintained at 557°F.
| |
| Given the following range of instruments, if the Steam Dump Control System is placed in
| |
| the Sleam Pressure mode, what approximate setpoint is required to maintain RCS
| |
| temperature at 557F?
| |
| Steam header pressure full range: 0-1300 psig
| |
| * Steam generator pressure full range: 0-1300 psig
| |
| * Turbine main steam pressure full range: 0-1500 psig
| |
| a. 16%
| |
| b. 24%
| |
| C. 73%
| |
| d. 84%
| |
| ANSWER:
| |
| d. 84%)
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Datr Sheets
| |
| QUES~IQNNJMRER: I TIEWGROUP: 111
| |
| KAIMPORTANCE: RO 3.7 SRQ
| |
| 10CFR55 CONTER'I': 41@) 4 43w
| |
| KA: 000004EA1.10
| |
| Ability to operate and monitor the following as they apply to a reactor trip: SIG pressure
| |
| OBJECTIVE: SDCS-3.0-4
| |
| Explain how the steam dump.valves are automatically modulated in the steam pressure control mode,
| |
| including control alignments, setpoint determination and adjustment, and the normal setpoint at power
| |
| DEVE1,OPMENT REFERENCES: EOP-EPP-004
| |
| OP-126
| |
| REFEIUZNCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY ,~.ZODIF'PP:D /DIRECT: S D C S - ~ 4004
| |
| NRC EXAM HISTORY: None
| |
| I)PSTR4CTOH JUSTIFHCACTION (CORRECT AKSWER d'd):
| |
| a. Plausible if the incorrect instrument is used to determine the range of the instrument and the
| |
| calculation is performed incorrectly (1500 - 1092 i 1500).
| |
| b. Plausible ifthe correct instrument is used to determine the range ofthe instrument, but the calculation
| |
| is performed incorrectly (1300 - 1092 I 1300).
| |
| c. Plausible if the incorrect instrument is used to determine the rangc of the instrument (1092 I 1500).
| |
| d d. The equivalent steam pressure for the required RCS temperature is approximately 1092 p i g . This
| |
| calculates to be a setpoint of 84% (1092 / 1300).
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREEIENSWE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY wrmc;: 3
| |
| EXPLAX.4TION: Must detennine required steam pressure for RCS temperature and then calcuiate
| |
| setpoint
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 2
| |
| With the plant at 100 percent steady-state condition, the following occurs:
| |
| m AtB-06-7-3. TOTAL MAKEUP WATER FLOW DEVIATION, alanns.
| |
| m A1.B-06-8-4. BORIC ACID FLOW DEVIATION, a i m s .
| |
| 0 VCT level is at 19.5% and decreasing at the same rate it has beer1 for the last few
| |
| days.
| |
| Which ofthe following procedures should be addressed?
| |
| a. AOP-002, Emergency Roration
| |
| b. AOP-003, Malfunction of Keactor Makeup Control
| |
| e. AOP-016, Excessive Primary Plant Leakage
| |
| d. AOP-017, Loss of Instrument Air
| |
| ANSWER:
| |
| b. AOP-003, Malfimction of Reactor Makeup Control
| |
| Post Validation Revision
| |
| | |
| H a m s NRC Written Excamination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QCES'rION NUMBER 2 TIEWGHOUP: 1/1
| |
| KA IMPORTANCE: RO 4.0 SRQ
| |
| 10CFR55 CONTENT: 41(b) 10 43(W
| |
| 0: 00002262.4.4
| |
| Ability to recognize ahnonnal indications for system operating parameters which are entry-level
| |
| conditions for emergency and abnormal operating procedures. (Loss of Reactor Coolant Makeup)
| |
| OBJECTIVE: AOP-3.341
| |
| IDENTIFY symptoms that require entry into AOP-003. Malfunction of Reactor Makeup Control
| |
| DEVELOPMENT RJ3FERENCES: AOP-003
| |
| REFERENCES SUPPLIED TO APPLICANT: Kone
| |
| QUESTION SOURCE: NEW SIGNHFICAXTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: AOP-3.3-KI 002
| |
| NRC EXAM HISTORY: Hams NRC 2000
| |
| DISTRACTOR JVSTIFPCACTHON (CORRECT ANSWE.R d'd):
| |
| a. Plausible since Emergency Roration entry conditions include any condition which is a result of an
| |
| unexplained reactivity addition, which candidate may consider this to he.
| |
| 4 b. 'I'hese are entry conditions for Reactor Makeup Control malfunction
| |
| e. Plausible since CVCS leakage, if suspected, would cause entry into AOP-016.
| |
| d. Plausible since nontial horation flowpaths are not available during a loss of instrument air event.
| |
| I)IFFICUI,TY ANALYSIS:
| |
| COMPREIIENSPVE I ANALYSIS KNQWLE.DCEI RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATIOX: Knowledge of entry requirements for loss of reador makeup
| |
| Post Validation Revision
| |
| | |
| Harris NKC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 3
| |
| Given the foilowing conditions:
| |
| e The plant is operating at 50% power.
| |
| e PT-457, Channel I I I Pressurizer Prcssure, has failed and all associated bistables are in
| |
| the tripped condition.
| |
| e Power is subsequently lost to UPS Bus IDP-1A-S1.
| |
| Which of the following describes the effkct of this loss of power on the Phase A
| |
| Containment Isolation valves?
| |
| a. NO Phase A Containment Isoration valves will close
| |
| h. ONLY Train A Phase A Containment Isolation valves will close
| |
| c. ONLY lrain B Phase A Containment Isolation valves wit1 close
| |
| d. Ail Phase A Containment Isolation valves will close
| |
| ANSWER:
| |
| c. ONLY Train B Phase A Containment Isolation valves will close
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 3 TIER/CROUP 2i I
| |
| KA IMPORTANCE: NO 4.3 SRO
| |
| 10CFR55 CONTENT: 41(b) 5 43(W
| |
| Eka: 013K3.03
| |
| Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: Containment
| |
| OBJECTIVE: ESFAS- 3.0-4
| |
| PREDICT how loss of any ofthe four instrument buses will affect the ESFAS outpiit functions of each
| |
| SSPS train
| |
| DEVELOPMENT REFERENCES: AOP-024
| |
| SD-103
| |
| REFERESCES SUPPLIED TO APPLICANT: None
| |
| QUESTHON SOURCE: NEW SHGNIFICA"T1,Y MODIFIED DIRECT
| |
| BANK NUMBER FOR SHGSIFHCANTIiUMODIFIED i DIFUCCT: ESFAS-3.0-K4 00 1
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFHCXCTIOS (COmECT ANSWER d'd):
| |
| a. Plausible since Train SA slave relays will not actuate, hut Train SB relays will still actuate.
| |
| b. Plausible since one train of Phase A will not actuate. hut the train that will not actuate is Train SA.
| |
| d e. A loss of Bus II)P-lA-SI under these conditions will result in a 2!3 sisnal to both trains of ESFAS,
| |
| resulting in an SI and Phase A signal. lrain SA slave relays; however, are powered from IDP-IA-SI
| |
| and are energized to actuate, so 'l'rain SA slaves will not perform their function.
| |
| d. Plausible since SK and Phase A signals will be generated on both trains ofE.SFAS, but Train SA slave
| |
| relays wiil not actuate due to not having power.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPRRIEIIEXSIVF:1ANALYSIS KNOWLEDGE I RECALL
| |
| DIFFICUIJY RATING: 3
| |
| EXpI.AXATION: Analyze the effect of a loss ofpower on the actuation signals and determine
| |
| which power supplies power which output relays
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 4
| |
| Given the following conditions:
| |
| e The unit is operating at 30% power.
| |
| e A dropped Control Rank 'C' rod has just been re-aligned.
| |
| a While attempting to operate the ROD CCPN'I'KOL ALARM RESET, the operator
| |
| inadvertently operates the ROD CXNTROL START-BJP RESET.
| |
| Which of the following describes the effect of operating the incorrect reset?
| |
| a. ,411 Control Bank 'C' rods drop into the core, causing an automatic reactor trip
| |
| b. All rods, including Control Hank and Shutdown Bank rods, drop into the core,
| |
| causing an automatic reactor trip
| |
| c. All rods remain in their current position and there is NO effect on the Rod Control
| |
| System circuitry
| |
| d. All rods remain in their current position, but the Rod Control System circuitry
| |
| senses all rods are lb-ullyinserted
| |
| ANSWER:
| |
| d. All rods remain in their current position, but the Rod Control System circuitry
| |
| senses all rods are fiiily inserted
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QIJESTION NUMBER: 4 TIENGROUP: I12
| |
| MA IMPORTANCE: RO 3.6 SRO
| |
| 10CFR55 CONTENT: $I@) 6 W O
| |
| KA: 000003AA1.02
| |
| Ability to operate and i or monitor the following as they apply to the Dropped Control Kod: Controls and
| |
| components necessary to rec~verrod
| |
| OBJIETA'E: ROI?CS-3.0-K7
| |
| DISCUSS the effects of manipulating each of the following rod control-related switches
| |
| 0 ROD CONTKOI, START-UP RESET switch
| |
| e ROD CONTROH.AI.AKM RIISET switch
| |
| DEWXOPMENT REFERENCES: AOP-OO1
| |
| ROL)CS-3.0
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QtJEST'nON SOURCE: SHGNHFICANTLY MODIFIED c]DIRECT
| |
| CANTLY MODIFIED /]DIRECT: RQDCS-3.0-R7 001
| |
| KRC EXAM HISTORY: None
| |
| DISTRACTOR K'STHFICACTION (CORRECT ANSWER +d):
| |
| n. Plausible since improper operation of correct switch could result in rods dropping into core, but
| |
| operated switch only resets starting points for rod control circuitry.
| |
| b. Plausible since improper operation of correct switch could result in rods dropping into core, hut
| |
| operated switch only resets startina points for rod control circuitry.
| |
| e. Plausible if misconception that effect is nothing if perfonned at power since switch is normally only
| |
| operated prior to withdrawing any rods: but operated switch resets starting points for rod control
| |
| circuitry.
| |
| d d. Operating switch at power does not affect actual rod position, but resets rod control such th.at circuitry
| |
| senses rods are at "full inserted" position.
| |
| ICIJLIY ANALYSIS:
| |
| COMPREHENSIVE I ANALYSIS KNOWLEDGE I RECALL
| |
| DIFFICULTY R4TING: 3
| |
| EXPLANATION: Knowledge of the function of rod control system controls
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 5
| |
| FRP-J. I, Response to High Containment Pressure, monitors the status of the ESW
| |
| Booster Pumps.
| |
| Which of the following is the concern if ESW Booster pumps fail to start while high
| |
| containment pressure conditions exist?
| |
| a. ESW Pump mnout
| |
| b. Flooding of safety equipment in containment
| |
| c. Loss of containment cooling capability
| |
| d. Radioactivity release to the environment
| |
| ANSWER:
| |
| d. Radioactivity release to the environment
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 5 TIER/GRQPUP: 1I2
| |
| KAHMPORIANCE: RQ 3.3 SRO
| |
| 10CFR55 CONTENT: 41(10) 8:9 4300
| |
| KA: WT14EA1.2
| |
| Ability to operate and i or monitor the following as they apply to the (High Containment Pressure)
| |
| Operating behavior characteristics of the facility
| |
| OBJECTIVE: Bl3-3.13-R3
| |
| Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
| |
| m ESW booster pump operation
| |
| DEVELOPMENT WEFERFNCES: FIW-J. 1
| |
| FW-J.1 Step Deviation Basis
| |
| REFERENCES SI!PPLIED TO APPLICANT: None
| |
| QUEsTION SOURCE: NEW SIGNIFICANTLY IkZODHFIED DHKECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIWECT: EOP-3.13 012
| |
| NRC EXAM HISTORY Harris NRC 2002
| |
| D r s m . 4 c r o R JUSTIFICACTION (CORRECT AKSWER Jw):
| |
| a. Plausible since ESW flow is supplying containment as well as other loads, but actions taken if an
| |
| ESW booster pump is not running arc not for ninout concerns?but rather to raise ESW pressure inside
| |
| containment.
| |
| b. Plausible since an ESW rupture inside containment could result in flooding of containment, but the
| |
| booster pump is checked running to ensure ESW pressure is adequate inside containment.
| |
| e. Plausible since the ESW booster pump supplies containment cooling units, but the ESW pump is
| |
| capable of supplying the loads without the booster pump, hut not at the required pressure for inside
| |
| containment.
| |
| 4 d. ESW Booster Pumps are required to be running to ensnre ESW pressure is > containment pressure
| |
| following a LOCA to prevent any leakage in the ESW system to cause the leakage to be to
| |
| contninnient and not to the ESW system.
| |
| DIFFICXJLTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS MNOWI.EBPGE I WECALL
| |
| DIFFICULTY RATING: 3
| |
| E.XPIANAIION: Knowledgc of EOP procedural bases
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 6
| |
| Given the following conditions:
| |
| e EOI-FRP-II.I, Response to a Loss of Secondary Heat Sink, is being implemented.
| |
| o RCS bleed and feed has been initiated when Auxiliary Feedwater (AFW) capability is
| |
| restored.
| |
| e All SGs are completely dry and depressurized.
| |
| Which of the following describes which SGs are to be fed under these conditions?
| |
| a. Feed ONLY one (1) SG to enstire RCS cooldown rates are established within
| |
| Technical Specification limits
| |
| b. Feed ONLY one (1) SG to limit the possibility of a SG tube rupture to a single SG
| |
| c. Feed ALL SGs to establish subcooling conditions in the RCS as soon as possible
| |
| d. Feed ALL SGs to allow termination of RCS bleed and feed as .soon as possibk
| |
| ANSWER:
| |
| b. Feed ONLY one (1) SG to limit the possibility o f a SG tube rupture to a single SG
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Kenctor Operator
| |
| Data Sheets
| |
| QUESTION NCJMBER: 6 TIEWGWOUP: 1/1
| |
| KAIMPORTANCE: MO 3.6 SRO
| |
| IOCFW55 CONTENT: 41(b) 4/10 43(b)
| |
| KA: 000054AK1.02
| |
| Knowledge of the operational implications ofthe following concepts as they apply to I m s of Main
| |
| Feedwater (MFW): Effects of feedwater introduction on dry S/G
| |
| OBJECTIVE: EOP-3.11-4
| |
| Given the foilowing EOP steps, notes. and cautions, DESCRIRE the associated basis
| |
| e Feed restoration
| |
| BPEVELOPhTENT REFERENCES: EOP-FRP-11.1
| |
| EOP-LP-3.11
| |
| REFERENCES SUPPLIED TO .UPLLICXNT: None
| |
| QLTESTHON SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NHC EXAX IIISTORR None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER .Id):
| |
| a. Plausible since feed is established to only one dry SG, but the reason is to ensure any subsequent
| |
| failures due to thermal shock are limited to a single S G .
| |
| d b. Flow should only be established to one dry SG so that ifexcess thermal shock causes Pdilure, the
| |
| failure is limited to one SG.
| |
| e. Plausible since RCS subcooling is a desirable condition to achieve, but only one SG at a time is fed.
| |
| d. Plausible since terminating KCS bleed and feed is a desirable condition to acl~ieve,hut only one SG at
| |
| a time is fed.
| |
| 1CLTIW ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICIJLTY RATING: 3
| |
| EXPLANATION: Knowledge of the requirements for feeding a dly S G arid the reasons for these
| |
| actions
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Exaniination
| |
| Reactor Operator
| |
| QUESTION: 7
| |
| Given the following conditions:
| |
| PZR level is 53% and stable.
| |
| a VCT level is 23 and stable.
| |
| a 1,etdou.n flow is 45 gpm (FI-150.1).
| |
| a RCP seal injection flows are:
| |
| RCP A at 8.3 gpm
| |
| RCP B at 7.9 gpm
| |
| RCP C at 7.8 gpm
| |
| a RCP seal return f h v s are:
| |
| RCP A at 2.8 gpm
| |
| RCP B at 3.1 gpin
| |
| RCP C at 2.9 gpm
| |
| Which of the following would be the expected flow indication on FI-122A. 1, Charging
| |
| Fl(w.*, assuming NQ RCS leakage?
| |
| a. 21 g p
| |
| b. 30gpm
| |
| d. 54gpm
| |
| ANSWER:
| |
| b. 30gpm
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operatur
| |
| Data Sheets
| |
| QUESTION NBTICEWER: 7 TIEHPIGROUP: 2il
| |
| KA IMPORTANCE: RO 3.3 SR8
| |
| 10CFR.55 COYIENT: 48(b) 3 43w
| |
| Ktk 003A4.01
| |
| Ability to manually operate and/or monitor in the control room: Seal injection
| |
| OBJECTIVE: CVCS-3.0-Rl
| |
| Given appropriate C 3 T S information, PERFORM a CVCS flow balance without reference to procedures
| |
| DEVELOPMENT REFERENCES: SD-107
| |
| CPL-2165-Sl305
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNHFICANIIJY MODIFIED DIRECT
| |
| BANK NEWBEH FOR SIGNIFICANT1,Y MODIFIED / DIRECT CVCS-K 1 00 1
| |
| NRC EXAM MIS1ORk: Noire
| |
| DISTRACTOR SUSTIFICACTION (CORRECT ANSWER $d):
| |
| a. Plausible if misconception is that seal leakoff flow is ignored, but leakoff flow is not.required to be
| |
| made up (45 24 = 21). However, seal leakoff flow is required to be included.
| |
| ~
| |
| 1 b. Charging flow should equal letdown flow (45 m m ) less seal injection flow (24 gpni) plus seal return
| |
| flow (9 a m ) (45 - 24 + 9 = 30).
| |
| c. Plausible if misconception that seal injection flow is measured as part of charging flow and seal
| |
| leakoff must be subtracted, but seal injection is required to be inciuded (45 - 9 = 36).
| |
| d. Plausible if misconception that seal injection flow is measured as paIt of charging flow, but seal
| |
| injection is required to be included (45 + 9 = 54).
| |
| ICULTY .W&ISPS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Calculation of expected charging flow indication based on given CVCS
| |
| parameters
| |
| Post Validation Revision
| |
| | |
| Hams NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 8
| |
| EOP-EPP-008, SI Termination, directs resetting SI.
| |
| Which of the following describes the effect of operating only ONE (1) of the two (2) SI
| |
| RESET switches at this step instead of both?
| |
| a. e Bypass - Permissive Light Panel light 4-1, SI ACTUAlE, would blink due to
| |
| only one train of SSPS having an SI signal
| |
| e Bypass - Permissive Light Panel light 5-1,SI RESET - AUTO SI BLOCKED,
| |
| would biink due to only one train of SSPS having SI reset
| |
| b. e Bypass - Permissive Light Panel light 4-1, SI ACTUAIX, would extinguish
| |
| due to neither train of SSPS having an Si signal
| |
| e Bypass -Permissive Light Panel light 5-1, SI RESET - AtrTCP SI BLOCKED,
| |
| would light due to both trains of SSPS having Si reset
| |
| c. 0 Bypass - Perniissive Eight Panel light 4-1, SI ACTUATE, would blink due to
| |
| only one train of SSPS having an SI sibma1
| |
| e Bypass - Permissive Light Panel light 5-1, SI RESET - AUTO SI BLOCKED,
| |
| would light due to both trains of SSPS having auto SI blocked
| |
| d. e Bypass -Permissive Light Panel light 4-1, SI ACTUATE, would extinguish
| |
| due to neither train of SSPS having an SI signal
| |
| e Bypass -Permissive Light Panel iight 5-1, SI RESET AUTO SI BLOCKED,
| |
| s-
| |
| would light due to both trains of SSPS having auto SI blocked
| |
| ANSWER:
| |
| a. e Bypass -. Permissive Light Panel light 4- I , SI ACTUATE, would blink due to
| |
| only one train of SSPS having an SI signal
| |
| e Bypass - Pemiissivc Light Panel light 5-1, SI RESET . AUTO
| |
| ~ SI BLOCKED,
| |
| would blink due to only one train of SSPS having SI reset
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Exsmination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTKON NUMBER: 8 TIEWGROUP: 211
| |
| KA IMPORTANCE: RO 3.9 SRO
| |
| 1QCFR55CONTENT: 41(b) 7 Wb)
| |
| IM: 006K4.11
| |
| Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Reset of SIS
| |
| OBJECTIVE: SIS-3.0-R4
| |
| DETERMINE SIS status from the following
| |
| * Eypass-Permissive Light Box
| |
| DEVELOPMENT REFERENCES: SD-103
| |
| REFERENCES SUPPLIED TO APPIKANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR s I m I F I c m r I , Y MODIFIED DIRECT: INPO 1073
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTPFICACPHON (CORRECT ANSWER +d):
| |
| d a. Operating only one switch only resets SI in a single train of SSPS. This would result in a disparity
| |
| between the two trains of SSIS for both the reset and the actuation signals so both lights would blink.
| |
| h. Plausible since the SI Actuation switch only requires a single switch to actuate SI, but the reset
| |
| switches are train-related.
| |
| c. Plausible since only train of SI would be reset so window 4-1 would be responding correctly, but
| |
| window 5-1 would also be hlinking due to the disparity between trains.
| |
| d. Pkdusibk since the SI Actuation switch only requires a single switch to actuate SI, but the reset
| |
| switches are train-related.
| |
| DIFFICULTY ANALYSIS:
| |
| C:OlllPREHENSNE I ANALYSIS LVOW1,EDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Comprehend the effect of only operating a single train switch on SSPS and how
| |
| the indications would be affected
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 9
| |
| Given the following conditions:
| |
| e Containment Pressure Channel I, PT-950A, is in TEST for surveillance testing
| |
| purposes.
| |
| e Containment Pressure Channel 111, PT-952A, is failed low.
| |
| e A large break &OCAoccurs and actual Containment Pressure reaches 21 psig.
| |
| Which of the following describes the response of the Containment Spray system?
| |
| a. NEITHER train of Containment Spray will automatically actuate
| |
| b. ONLY Train 'A'of Containment Spray will automatically actuate
| |
| c. ONLY Train 'B' of Containment Spray will automatically actuate
| |
| d. BOTH trains of Containment Spray will automatically actuate
| |
| ANSWER
| |
| d. BOTH trains of Containment Spray wiil automatically actuate
| |
| Post Validation Revision
| |
| | |
| Harris NKC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 9 TPERIGROGP: 2)1
| |
| UIMPBRTANCE: RO 2.7 SRO
| |
| POCFH55 CONTENT: 41(b) 7i9 43@)
| |
| Ktf: 013K6.01
| |
| Knowledge of the effect o f a loss or malfunction on the following will have on the ESFAS: Sensors and
| |
| detectors
| |
| OBJECTWE: CSS-Rl
| |
| STATE the conditions that will cause a containment spray actuation signal (CSAS) including coincidence
| |
| and setpoints
| |
| DEVELOPMENT REFERENCES: SD-103
| |
| REFERENCES SUPPLIED TO APPL.ICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIFOZCT
| |
| BANK NUI\IBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CSS-RI 003
| |
| NRC EXAM HISTORY Harris NRC 2002
| |
| DISTRACTOR SIJSTHFICACTION (CORRECT ANSWER dd):
| |
| a. Plausible since CSAS is energized to actuate and 2 of the channels are in a deenergized condition, hut
| |
| the remaining 2 channe1.s will cause an actuation of both trains of Spray.
| |
| b. Plausible since CSAS is energized to actuate and 2 ofthe channels are in a deenergized condition, hut
| |
| the chamnek input both trains of SSPS.
| |
| c. Plausible since CSAS is energized to actuate and 2 ofthe channels are in a deenergized conditi.cn, but
| |
| the channels input both trains of SSPS.
| |
| d d. CSAS is energized to actuate and although 2 of the channels are in a deenergized condition, the
| |
| remaining 2 channels will cause an actuation of both trains of Spray.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREIIEWSIVE I ANALYSIS KNOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Compreherision of the cffcct of failed channels and testing on ESF signals
| |
| Post Vslidation Revision
| |
| | |
| Harrix NKC Written Examination
| |
| Reactor Operator
| |
| QUESTION: io
| |
| Given the following conditions
| |
| e The plant is operating at 43% power.
| |
| o i20VAC Vital Bus IDF-1B-SI1 deenergizes,
| |
| Outward rod motion is inhibited by ...
| |
| a. C- 1 ,Intermediate Range rod stop.
| |
| b. C-2, Power Range rod stop.
| |
| c. C-3, QTAT rod stop.
| |
| d. C-4, QPAT rod stop.
| |
| ANSWER:
| |
| b. C-2, Power Range rod stop.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 10 TIEWGHOUP: 212
| |
| KAIILBPORTANCE: RO 3.4 SRO
| |
| 10CFR55 CONTENT: 48(b) 6 43w
| |
| KA: OClK4.07
| |
| Knowledge of CRDS design feature(s) anct'or interlock(s) which provide for the following: Rod stops
| |
| OBJECTIVE: NIS-3.0-9
| |
| DISCUSS the operation of the following NI trip-related Cunetions:
| |
| b. SR, IR and PR (low) trip blocks
| |
| DEVELOPMENT REFERENCES: OF-105
| |
| AOP-024
| |
| KEFEHENCES SUPPLIED 1'0APPLICANT:
| |
| QUESTION SOURCE: NEW a None
| |
| SIGNIFICANTLY MODIFIED
| |
| BANK NKJMRER FOH SIGNlFICANTLY LMODIFIED1%)ERECT: NHSR6 003
| |
| DIIPECT
| |
| NHC EXAM HISTORY: Nonc
| |
| DISTRACTOR .KJSTIFICAC'HON (CORRECT ANSWER d'd):
| |
| a. Plausible since this causes 3 rod stop, and coincidence is 1!2, but 1R rod stop is blocked above P-10 by
| |
| manual operator action. Must have 2/4 PK below P-IO io reset.
| |
| .\, b. PR rod stop is 1/4 coincidence. With $2-SB deenergized, PR N-42 is tripped.
| |
| c. Plausible since causes rod stop. but coincidence is Z 4 instead of 1/4
| |
| d. Plausible since causes rod stop, but coincidence is U4 instead of 1!4.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY wrnw: 3
| |
| EXp1,ANATION: Analyze effect of loss of power on NIS and rod control and detennine effect of
| |
| single channel tripped
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 11
| |
| The hasis for the operation of the Electric Hydrogen Recombiners is to minimize
| |
| hydrogen concentration build up in Containment following a LOCA due to the ...
| |
| a. zirc-water reaction and release of hydrogen from the PRT.
| |
| b. corrosion ofmetals in Containrrient and release of hydrogen from the RCDT.
| |
| c. release of hydrogen from the PRT and the radiolytic decomposition ofwater.
| |
| d. radiolytic decomposition of water and the corrosion of metals in Containment.
| |
| ANSWER:
| |
| d. radiolytic decomposition of water and the corrosion of metals in Containment.
| |
| Post Validation Revision
| |
| | |
| Harris NRC X7rittenExamination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTIOK NI:MBEM: i I THIERIGRBtJP: 2!2
| |
| KA IMPORTANCE: RO 3.4 SRO
| |
| BOCFRSS CONTENT: 48(b) 10 43w
| |
| IKA: 028G2.2.22
| |
| Knowledge of limiting conditions for operations and safety limits. (Hydrogen Recombiner and Purge
| |
| Controi)
| |
| OBJECTIVE: KR-3.0-1
| |
| STATE the purpose and fimction of the Hydrogen Recombiner System, including the following
| |
| components:
| |
| 6 Electric hydrogen recombiner
| |
| DEVELOPMENT REFERENCES: 'IS 3.6.4.2 Basis
| |
| SD-125
| |
| IP-BR-3 .O
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: b]NEW
| |
| p.9
| |
| u
| |
| R
| |
| SIGNIFICANTLY MODIFIED DPRFXT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIPIED / DIRECT: IIR 0 1
| |
| NMC E m ¶ HIS'I'ORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER +d):
| |
| a. Plausible since Electric Hydrogen Recombiners are designed to remove hydrogen in containment
| |
| folIowing a 1,OCA due to generation from the zirc-water reaction, but not due to release from the
| |
| I'KT.
| |
| b. Plausible since Electric Hydrogen Reconibiners are designed to remove hydrogen in containment
| |
| following a LOCA due to generation from the corrosion of metals in containnient! but not due to
| |
| release from the RCDT.
| |
| c. Plausible since Electric Hydrogen Recombiners are designed to remove hydrogen in containment
| |
| following a LOCA due to generation from the radiolytic decomposition of water, but not due to
| |
| release from the PRT.
| |
| 4 d. The Hectric Hydrogen Recombiners are designed to F C ~ O Vhydrogen
| |
| ~ in containment following a
| |
| 1,OCA due to generation from the zire-water reaction, radiolytic decomposition of water, and
| |
| comosion of met& in containment.
| |
| DIPPICULTY ANALYSIS:
| |
| COMPREHENSIVE I ANALYSIS KNOWLEDGE f RECALL
| |
| DIFFICULTY Ff.A'rING: 3
| |
| EXPLANATION: Knowledge of Tech Spec basis for hydrogen recombiners
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 12
| |
| EOP-EPP-00 1, Loss of AC Power to 1.4-SA and IB-SR Buses, is being performed.
| |
| Concurrent to the loss of power, a small break LOCA occurred.
| |
| The crew has completed the following actions when off-site power is restored to 6.9 KV
| |
| Bus 1.4-SA:
| |
| e Sequencers have been de-energized
| |
| 6 Safeguards pumps autostarts have been disabled
| |
| 6 RCP seals have been isolated
| |
| 6 MSIVs arid FWIVs have been closed
| |
| e Depressurization of SGs to I80 psig has commenced
| |
| Which of the following actions is the FIRST to be taken following the restoration of off-
| |
| site power?
| |
| a. Start an ESWpump
| |
| b. Start a CSIP
| |
| c. Stabilize SG pressures
| |
| ci. Initiate SI
| |
| AK§WEW:
| |
| c. Stabilize SG pressures
| |
| Post Validation Revision
| |
| | |
| IIarris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 12 TIEWGROUP lil
| |
| KAMPBKTANCE: RO 4.3 SRO
| |
| 10CFR55 CONTENT: 41(b) 10 4Xb)
| |
| MA: 000055EAi.07
| |
| Ability to operate and monitor the following as they apply to a Station Blackout: Restoration of power
| |
| from offsite
| |
| OBJECTIVE: EO?-3.7-5
| |
| Given a title of a continuous action step from a foldout and a list of plant conditions, DETERMINE if
| |
| implementation is required
| |
| DEVELOPMENT REFERENCES: EOP-EPP-00 1
| |
| REFERENCES SUPPLIED TO AIPP1,IC:ANT: None
| |
| QUESTION SOURCE: NEW GHGNIFICAWLY MODIFIED
| |
| BANK NZMBER FOR S I 6 CANTLY MODIFIED / DIRECT:
| |
| NHC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER @d):
| |
| a. Plausible since if the power source was an EDG instead of offsite power, it would he important to
| |
| provide cooling flow to the EDG.
| |
| b. Plausible since a small break LOCA exists and RCS inventory is being lost, but the first action is to
| |
| stabilize SG pressure.
| |
| d e. Upon restoration of power to at least one bus, the first action taken is to stabilize S G pressures
| |
| d. Plausible since a small break ILKX exists and RCS inventory is being iost, but the first action is to
| |
| stabilize SG pressure.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS JXNBWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge of required actions when power is restored following a loss of all
| |
| AC power
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 13
| |
| While performing an Operating Procedure, the Reactor Operator comes to a step which
| |
| states:
| |
| Request Chemistry to sample the RHT for boron concentration.
| |
| The Reactor Operator believes the step is NOT essential to achieving the purpose for
| |
| which the procedure is being used and that the omission ofthe step does NOT violate the
| |
| precautions and limitations of the Operating Procedure.
| |
| Which of the following is the MINHMUM requirement(s) that must be met to allow
| |
| marking the step NjA?
| |
| a. Step must be initialed by the Reactor Operator prior to perfornlance
| |
| b. m Step must be initialed by the Reactor Operator prior to performance
| |
| e A written explanation ofwhy the step is N!A must be prowded in the
| |
| Comments section of the procedure
| |
| c. m Step must be initialed by the SCO prior to performance
| |
| d. m Step must be initialed by the SCO prior to performance
| |
| m A written explanation of why the step is N!A must be provided in the
| |
| Comments section of the procedure
| |
| ANSWEW:
| |
| d. Step must be initialed by the SCO prior to performance
| |
| A written explanation ofwhy the step is N/A must be provided in the
| |
| Comments section of the procedure
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 13 TIE19/GM OU P 3
| |
| KA IMPORTANCE: RO 3.9 SRO
| |
| 10CFR55 CONTENT: 41(b) 10 43W
| |
| KA;: 2.1.23
| |
| Ability to perfom specific system and integrated plant procedures during all modes of plant operation
| |
| BBJEC1'IVE: PP-2.0-2
| |
| L)ISCI!SS the requirements in FRO-NGGC-0200 concerning the following:
| |
| e Procedure user's responsibilities
| |
| DEVELOPMENT REFERENCES: FRO-NCiGC-0200
| |
| REFERENCES SUPPI,I(EDTO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM BISTORW None
| |
| DHSTRACI'OR JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| a. Plausible since the RO discovered the cause for marking the step N!A, but a supervisor must initial the
| |
| step prior to performance and a written explanation must be provided in the Comments section.
| |
| b. Plausible since a written explanation must be provided in the Comments section, but a supervisor must
| |
| initial the step prior to performance.
| |
| e. Plausible since a supervisor niust initial the step prior to perforniance, but a written explanation must
| |
| be provided in the Comments section.
| |
| 4 d. The step is initialed by the responsible supervisor prior to performance and a written explanation is
| |
| provided in the Comments section.
| |
| ICUI,TY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULI'Y RATING: 2
| |
| EXPLANATION: Knowledge of use of N.4 during procedure usage
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUES'I'IQN: 14
| |
| Given the following conditions:
| |
| e Spent resin is being sluiced from the Cation Demineralizer to a Spent Resin Storage
| |
| Tank.
| |
| e The operator reports that it appears that a pipe in the overhead of's hallway is plugged
| |
| with resin.
| |
| o HP reports the results of a radiation survey as follows:
| |
| e 2500 m r h on contact with pipe
| |
| e 1200 m r h @ 18 inches from the pipe
| |
| o 5 m r h r at floor level below the pipe
| |
| Which one ofthe following describes the required radiological postings?
| |
| a. NO pcrstings are required because a ladder is required to access the pipe area
| |
| b. Very High Radiation Area with red flashing light
| |
| c. High Radiation Area with a red flashing light
| |
| d. High Radiation Area, but NO red flashing light required
| |
| ANSWER:
| |
| c. High Radiation Area with a red flashing light
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 14 TIEWGRBUP: 3
| |
| KA IMPORTANCE: 130 2.5 SRO
| |
| 10CFR55 CONTENT: 41(b) 12 43BW
| |
| KA: 2.3.2
| |
| Knowledge of facility A L A M program
| |
| OBJECTIVE: W-35-13
| |
| Define the following terms as defined in IOCFRZO:
| |
| d. High radiation area
| |
| DEVELOPMENT REFERENCES: AI'-504
| |
| TS 6.12.2
| |
| REFERENCES SUPP1,IE:D TO APPLICANT: None
| |
| QUESTION SOURCE: SIGNHP1CA"FLY MODIFIED DIRECT
| |
| CANTLY MODIFIED I IIIKECT: 0 20657
| |
| NRC EXAM HISTORY: Harris NRC 2002
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER J'd):
| |
| m. Plausible since normal access to the area requires a ladder. hut it is accessible so a red warning light is
| |
| required.
| |
| b. Plausible since a common error is defining the difference between a very high radiation area and a
| |
| high radiation area, but this is a high radiation area requiring a red warning light.
| |
| 4 c. Accessible areas where rad levels exceed 1000 mWhr are required to be locked. Where it is not
| |
| practicai to lock the area, a red warning light shall he in place to warn personnel.
| |
| (8. Plausible since this is defined as a high radiation area, but a red warning light is also required.
| |
| ICULTY ANALYSIS:
| |
| COklPREIIENSWE / ANALYSIS KNOWLEDGE I RIXhLI,
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION: Knowledge of radiological posting requirements
| |
| Post Validatioii Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 15
| |
| The Containment Fan Coil Units (AIi-37 / 38 / 39) provide area cooling to ...
| |
| a. the reactor vessel supports and reactor coolant leg nozzles.
| |
| b. the clearance between the reactor vessel and primary shield wall.
| |
| c. the reactor coolant pumps.
| |
| d. contairment for norma1 operation and accident conditions.
| |
| ANSWER:
| |
| c. the reactor coolant pumps.
| |
| Post Validation Revision
| |
| | |
| Harris hXC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NLWBER: 15 TIEWGROUIP: 211
| |
| KA IMPORTANCE: RO 3.2 SRO
| |
| 10CFR55 CONTENT: .bI(b) 9 43(h)
| |
| H(A: 02262. I .28
| |
| Knowledge o f t h e purpose and function of major system components and controls. (Containment
| |
| Cooling)
| |
| OWJECTBVE: CCS-3.0-A1
| |
| Sl'ATE the purpose of the following five Containment Cooling Subsystems
| |
| e Containment Fan Coil [Jnits
| |
| DEVELOPMENT REFERENCES: SD-169
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY M0DIFIE.D DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CCS-AI 010
| |
| NRC EXAM HISTORY: None
| |
| DIS'TIPAC'FOR JUS'TIFICACTION (CORRECT ANSWER .I'd):
| |
| a. Plausible since containment cooling components provide cooling to these components, but it is
| |
| provided by the reactor supports cooling system not the containment fan coil units.
| |
| h. Plausible since containment cooling components provide cooling to these components, but it is
| |
| provided by the primary shield cooling units not the containment fan coil units.
| |
| 4 c. Air is drawn from containment space, through the cooling coils, to the fan suction. Cooling air from
| |
| the fan coil unit is directed to the reactor coolant pump subcompartments.
| |
| d. Plausible since containment cooling components provide cooling to these components, but it is
| |
| provided by the containment fan coolers not the containment fan coil units.
| |
| DIFFICULTY ANALYSIS:
| |
| COl\gPREHENSA'E / AKALYSIS KNOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION: Knowledge of purpose of Containment Fan Coil Units
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 16
| |
| Given the folIowing conditions:
| |
| e Following a Reactor Trip and Safety Injection due to a leak in the PRZ steam space,
| |
| the Critical Safety Function Status Trees (CSFST) are being monitored.
| |
| e The CSFST for RCS Inventory first checks PRZ level and then checks the Reactor
| |
| Vessel Level Indicating System (RVLIS).
| |
| If PW. level is indicating greater than 92%, why is a check of RVI,IS then perform'?
| |
| a. Determine if the cause of the high PRZ level is excessive RCS inventory or
| |
| voiding in the Reactor Vessel head
| |
| b. Determine if SI termination criteria i s met to allow reducing the excessive RCS
| |
| inventory
| |
| c. Determine if Adverse Containment conditions have caused erroneous indications
| |
| ofthe PRZ level instmnients
| |
| d. Determine if the cause of the high PRZ level is excessive RCS inventory or
| |
| expansion due to an RCS heatup
| |
| ANSM'KR:
| |
| a. Determine if the cause of the high PRZ level is excessive RCS inventory or
| |
| voiding in the Reactor Vessel head
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 16 TIEWGROCP: 1:1
| |
| KAIMPORTANCE: RQ 3.2 SRO
| |
| POCFR.55 CONTENT: 41(b) 7 43m
| |
| KA: 000008(;2.1.28
| |
| Knowledge of the purpose and function of major system components and controls. (Pressurizer Vapor
| |
| Space Accident)
| |
| QBJECTIVE: ICCM-3.0-1
| |
| LIST the two major functions of the Inadequate Core Cooling Monitor (ICXM)
| |
| DEVELOPMENT REFERENCES: FOP Rackground for Inventory Status Tree
| |
| LPEOP-3.12
| |
| REFERENCES SUPPLIED TQ APPLICANT: None
| |
| QUESTIQN SOZJRCE: NEW SIGNIFICANTLY MODIFIED
| |
| OR SIGNIFICANTLY MODIFIED / DIRECT:
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| d a. Once a determination has been made that P E lwei is full, RVLIS is then used to confirm whether the
| |
| cause ofthe full PRZ is excessive inventory or voiding in the head rcgion.
| |
| h. Plausible since RVLIS is used throughout the EOP network to determine if SI termination criteria has
| |
| been met, but in this instance it is used to determine the cause of the high PRZ level.
| |
| c. Plausible since a steam space break in the PRZ will affect the level indications, but RVLIS is used to
| |
| determine the cause ofthe PRZ high level condition.
| |
| d. Plausible since RVEIS is part of the Inadequate Core Cooling Monitoring System and a heat up of the
| |
| RCS will cause expansion ofthe RCS, but but RVIM is used to determine the cause of thc PW. high
| |
| level condition.
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| I)IFFHCUI,TY RATING: Knowledge of the purpose of monitoring RVLIS during accident conditions
| |
| EXPLANATION:
| |
| Post Validation Revision
| |
| | |
| IrillTis NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 17
| |
| Given the following conditions:
| |
| e The plant is shutdown for work on Reactor Coolant Pump seals.
| |
| e The Reactor Vessel Head is still installed.
| |
| e The running Residual Heat Removal (RIIR) pump trips and the crew is unable to start
| |
| the standby RHR pump.
| |
| e Time to reach core boiling is determined to be 26 minutes.
| |
| e Time to reach core boil-off is determined to be 53 minutes.
| |
| Of the following two (2) methods of RCS makeup, which of the following is the
| |
| PREFERRED method of makeup and why is it preferred over the other method?
| |
| a. Gravity feed from the RWST to the RCS is prefened over starting a CSIP since
| |
| starting a CSIP under these conditions would violate Technical Specifications
| |
| b. Gravity feed from the RWST to the RCS is preferred over starting a CSIP since
| |
| Reactor Makeup to the CSIP may be insufficient to makeup for core boil-off
| |
| c. Starting a CSIP is preferred over gravity feed from the RWST since gravity feed
| |
| flow may he insufficient to makeup for core boil-off even if the RCS is
| |
| depressurized
| |
| d. Starting a CSlP is prefemd over gravity feed from the RWST since the RCS may
| |
| be pressurized and prohibit gravity flow
| |
| ANSWER:
| |
| d. Starting a CSIP is preferred over gravity feed fiom the R W S l since the RCS may
| |
| be pressurized and prohibit gravity flow
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Resctor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 17 TIEWGROUP: l/I
| |
| KA IMPORTANCE: RO 3.1 SRO
| |
| IQCFR55CONTENT: 4l(b) 8/10 43th)
| |
| K k 000025AK3.01
| |
| Knowledge of the reasons for the following responses as they apply to the Loss ofResidua1 Heat
| |
| Removai System: Shift to alternate flowpath
| |
| OBJECTIVE: AOP-3.20-3
| |
| Given a set of entry conditions and a copy of .4OP-020, IETEKMhE the appropriate response
| |
| DEVELOPMENT REFERENCES: AOP-020
| |
| AOP-020-BU
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNHFICANI'LY BIODIETED DIRECT
| |
| BANK NUh5RE.R FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DISTR4CTOR .RJSTIFHCACI'ION (CORRECT ANSWER +(I):
| |
| a. Piausible since TS requires that a C S P he made inoperabie before these plant conditions are
| |
| established, but GP-008 requires that at least one CSHP be lhctional under these conditiuns.
| |
| b. Plausible since the CSII' can provide more flow than Reactor Makeup is capable of providing, hut the
| |
| suction source for the CSIP would be the RWST.
| |
| c. Plausible since starting a C S P is preferred to gravity feed, but only because the KCS may be
| |
| pressurized. If the RCS is depressurized, gravity k e d will provide adequate flow.
| |
| d d. If the RCS is pressurized, gravity flou. may be insufficient to provide adequate makeup to the RCS.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE I IRECALL
| |
| mmcucri~RATING: 3
| |
| EXPLANATION: Analysis of plant conditions to determine appropriate response and reason for
| |
| response
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 18
| |
| Given the f~ololowingconditions:
| |
| e Containment temperature is 96 "F.
| |
| e Containment Fan Coolers (AH-1 2 / 3 / 4) are operating in the Normal Cooling
| |
| i\/lQde.
| |
| e i\ loss of offsite power occurs and thc plant responds as expected.
| |
| 'The Containment Fan Coolers should he aligned with one ( i ) fan associated with each
| |
| fan cooier operating in ...
| |
| a. high speed and discharging to the concrete airshaft
| |
| b. high speed itrid discharging to the post-accident discharge duct
| |
| c. low speed and discharging to the concrete airshaft
| |
| d. low speed and discharging to the post-accident discharge duct
| |
| ANSWER:
| |
| a. high speed and discharging to the concrete airshaft
| |
| Post Validation Revision
| |
| | |
| Harris NKC Written Emmination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QIJESTIONNUMBER: 18 TIEWGROUP Ill
| |
| KA 1-MPORTANCE: WO 2.7 SRO
| |
| 10CFR55 CONTENT: 41(b) 9 43w
| |
| KA: 000056AA2.09
| |
| Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Operational
| |
| status of reactor building cooling unit
| |
| OBJECTIVE: CCS-3.0-R4
| |
| PKEDICT the response(sj of the Containment Cooling Subsystems to the foliowing signals.
| |
| * 1,OSP
| |
| DEVELOPMENT REFERENCES: SI)-169
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CCS-R4 001
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFHCACTION (CORRECT ANSWER dd):
| |
| 4 a. Orie fan per unit will start on high speed and discharge to the concrete airshaft.
| |
| b. Plausible since one fan per unit will start on high speed, but the discharge is to the concrete airshaf?
| |
| not the post-accident discharge duct.
| |
| e. Plausible since this fan response is the response to a LOCA start signal and they do discharge to the
| |
| concrete airshaft, but the fans operate in high speed following a loss of offqite power.
| |
| d. Plausible since this is the response to a LOCA stari signal, hut the fans operate in high speed and they
| |
| discharge to the concrete airshaft following a loss of offsite power.
| |
| IClJLTY ANALYSIS:
| |
| COMPREHENSnE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge ofthe response ofthe containment fan cooler fans to a loss of
| |
| offsite power
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| cpmsrIorv: 19
| |
| \;\ihich ofthe following can all be used to confirm that an inoperable / stuck rod is to be
| |
| considered misalibaecl?
| |
| a. a Delta-H indication
| |
| Power range channels
| |
| 6 Reactor vessel level indication
| |
| a Core AT
| |
| b. = Delta-I indication
| |
| a Power range channels
| |
| a QPTR calculation
| |
| a Core outlet thermocouples
| |
| c. e Core AT
| |
| e Power range channels
| |
| a QPTR calculation
| |
| 6 Core outlet thermocouples
| |
| d. a Delta-I indication
| |
| a Power range channels
| |
| a Reactor vessel level indication
| |
| 6 Core outlet thennocouples
| |
| ANSWER:
| |
| b. a Delta-I indication
| |
| e Power range channels
| |
| a QPTR calculation
| |
| a Core outlet thermocouples
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION N I M U E R 19 TIEW/GROUP: 112
| |
| KAIMPORTANCE: I90 2.5 sa0
| |
| 10CFR55 CONTENT: 41(b) 6/10 43(b)
| |
| KA: OOOOOSAK2.02
| |
| Knowledge of the interrelations between the Inoperable i Stuck Control Rod and the following: Breakers,
| |
| relays, disconnects, and control room switches
| |
| OBJECTIVE: ROP-3.1-3
| |
| LIST the indications of a misaligned rod specified in AOP-001, Attachment I, Indications of Miwligned
| |
| Rod
| |
| DEVEI ,OPMENT REFERENCES: AOP-OO 1
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW S m m c A i v r L Y MODIFIED DIRECT
| |
| BAWK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AOP-3.1-R3 002
| |
| AOP-3.1-R3 003
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JWXIFICACTION (COMECT ANSWER d'd):
| |
| a. Plausible since delta-I and QPTR are used to determine a misaligned rod, hut RVkIS and core A'I' are
| |
| not used.
| |
| d b. Delta-I, QPTR, power range channels, and core outlet thermocouples are all used to determine a
| |
| miwdigned rod.
| |
| c. Plausible since QPTR, power range channels, and core uutlet thermocouples are all used to determine
| |
| a misaligned rod, hut core AT is not used.
| |
| d. Plausible since delta-I, power range channels, arid core outlet thermocouples are all used to determine
| |
| a misaligned rod, hut KVLIS is not used.
| |
| DIFFICULTY ANALYSIS:
| |
| DIFFICULTY RATING: 3
| |
| EXPIANATION: Knowledge of the indications of a misaligned rod per procedure
| |
| Post Velidation Revision
| |
| | |
| IIarris NRC Written Examination
| |
| Keactor Operatcr
| |
| QUESTION: 20
| |
| Given the following conditions:
| |
| * Following an accident, EOP-FRP-C. 1 "Response to Inadequate Core Cooling," is
| |
| ~
| |
| being performed.
| |
| e ERFIS is inoperable.
| |
| e Plant parameters are as follows:
| |
| e ICCM highest TC = 672" F
| |
| a RCS WR Itmjxrature (highest) = 688" F
| |
| * RCS pressure PT-440 = 1535 psig
| |
| e RCS pressure PT-402 = 1635 psig
| |
| a CNMT pressure PT-95 1 = 4.5 psig
| |
| What value of superheat should be reported?
| |
| a. 63'F
| |
| b. 71 '1:
| |
| c. 49'F
| |
| d. 87 'F
| |
| ANSWER:
| |
| a. 63 OF
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Datu Sheets
| |
| QUESTION NWfBER: 20 TIIERIGROUP: i /2
| |
| KAIMFQKTPLNCE: RO 4.6 SRO
| |
| P(DCPR55CONTENT: 41(b) 5 43@1
| |
| IBA: 000074EA2.01
| |
| Ability to determine or interpret the following as they apply to a Inadequate Core Cooling: Subcooling
| |
| margin
| |
| OBJECTIVE: EOP-3.19-4
| |
| Given a set of conditions during EOP implementation, IiETERiMINE the correct response or required
| |
| action based upon the EOP User's Guide general information
| |
| e Determining an KCS subcooling value
| |
| DEVELOPMENT KEFEKENCES: EOP-IJsers Guide
| |
| KEFEFENCES SUPP1,IED TO AFT'LHCANT:
| |
| ~
| |
| Steam Tables
| |
| QUESTION SOUKCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| RANK KUMIEK FOR SIGNIFICAN'rLY MODIFIED 1DIRECT: EOP-3.19-R4 003
| |
| NRC EXAM IIISTORY: None
| |
| DISTRACTOR JUSTIFICAQTHON(CORRECT ANSWER \I7@:
| |
| 4 a. When ERFIS is not available, the highest ICCM temperature should be used. If EWIS is not
| |
| available and adverse containment conditions exist, PT-402 should be used for pressure. Saturation
| |
| temperature for 1635 psig is 609 OF, so the amount of superheat is 63 "E' (672-609).
| |
| b. I'lausible since the superheat determined using the %CCMtemperature and saturation for the lowest
| |
| RCS pressure of 1535 psig (not used because of adverse containment conditions) is 71 "F (672-601).
| |
| E. Plausible since the superheat determined using the hot leg temperature (not used if ICCM is available)
| |
| arid saturation for the PT-402 pressure of I635 psig is 79 "E' (588-609).
| |
| d. Plausible since the superheat detemiincd using the hot leg temperature (not used if ICCM is available)
| |
| and saturation for the lowest KCS pressure of 1535 psis (not used because of adverse containment
| |
| conditions) is 87 "F (688-601).
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANAL.YSIS KNOWLEDGE /RECALL
| |
| DIFFICUI,TY RATLVG: 3
| |
| EXPLANATION: Knowledge of instruments to use and calculation of subcooling by applying
| |
| steam tables
| |
| Post Validation Revision
| |
| | |
| Harris hXC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 21
| |
| A failure of Containment Cooling causes equilibrium Containment temperature to
| |
| increase from 105 OF to 130 "F.
| |
| Assuming no change in Tave, P M pressure, or Letdown flow rate, how will this effect
| |
| ICs-231, FK-122.1 CHARGING FLOW?
| |
| a. It will throttle open slightly during the course of the temperature change and then
| |
| retuni to its original position
| |
| b. It will throttle closed slightly during the course of thc temperature change and then
| |
| return to its original position
| |
| e. It will throttle open slightly during the course of the temperature change and
| |
| remain in that position
| |
| d. It will throttle closed slightly during the course ofthe temperature change and
| |
| remain in that position
| |
| ANSWER:
| |
| b. It will throttle closed slightly during the course of the temperature change and then
| |
| return to its original position
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 21 TIEWGROUP: 2i 1
| |
| KAIMPORTANCE: RO 3.0 SRO
| |
| lOCFR55 CONTENT: 4l@) 719 43m
| |
| KA: 022K3.02
| |
| Knowledge of the effect that a loss or tnalfunction of the CCS will have on the following: Containment
| |
| instrumentation readings
| |
| O B ~ E c r w E : CVCS-3.0-R3
| |
| DESCRIBE the controls and interlocks of remotely operated CVCS valves, including the following:
| |
| e CVCS controllers, including transfers between automatic and manual control, setpoint determination
| |
| and adjustment, and output control
| |
| DEVELOPMENT REFERENCES: SD-I 00.3
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: SIGNIFICANTLY MODiFIED
| |
| C m m Y MODIFIED DIRECT:
| |
| NRC EXAM IIISTORW None
| |
| DISTWACIOR .JLrSTIFICACTION (CORRECT ANSWER dd):
| |
| 10. Plausible since a cotitainnient temperature increase will affect indicated pressurizer level, but
| |
| indicated level will increase so charging flow would decrease.
| |
| d b. As containment temperature increases, indicated pressurizer level increases due to heating of the
| |
| reference leg. This would result in a smaller Awhich would indicate that pressurizer level is high.
| |
| Charging flow will decrease to lower actual level and then return to its original value to match
| |
| letdown flow.
| |
| 6. Plausible since a containment temperature increase will affect indicated pressurizer level, but
| |
| indicated level will increase so charging flow would decrease.
| |
| d. Plausible since a containment temperature increase will affect indicated pressurizer level and charging
| |
| Row will decrease, but the flow will return to its original value to match letdown flow.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE I ANALYSIS KNO\VEDGE / RECALL
| |
| DIFFICUI,TY RATING: 4
| |
| EXPLANATION: Analyze the effect ofthe temperature change on pressurizer level and then
| |
| determine how this change affects the operation ofFK-122.1
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QIJESTION: 22
| |
| Given the following conditions:
| |
| e The unit is operating at 12% power.
| |
| e The following RCP vibrations are observed:
| |
| INDICATION WCP 6A' RCP 'B' KCP 'CY
| |
| Frame Vibration 3.6 mil and ? at 2.8 mil and stable 4 mil and ? at
| |
| 0.3 mil per hr 0.1 mil per hr
| |
| Shaft Vibration 12 mil md ? at 7 mils and stable 14 mils and ? at
| |
| 0.3 mil per hr 0.6 mils per hour
| |
| Which of the following describes the actions required for this condition?
| |
| a. Stop RCP 'A' and initiate a plant shutdown
| |
| b. Trip the reactor, stop RCP 'A', and go to PATH- 1
| |
| c. Stop RCP 'C' and initiate a plant shutdown
| |
| d. Trip the reactor, stop RCP IC', and g o to PATH-1
| |
| ANSWER:
| |
| a. Stop RCP 'A' and initiate a plant shutdown
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Keactor Operator
| |
| Data Sheets
| |
| QUF,STBON NUMBER 22 TIEWGROUP: 2: 1
| |
| KA IMPORTANCE: RO 2.9 SRO
| |
| 10CFR55 CONTENT: 4P(b) 3/10 43(b)
| |
| IGa. 003A1.01
| |
| Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) asscciated
| |
| with operating ?he RCPS controls including: RCP vibration
| |
| OBJECTIVE: AOP-3.18-3
| |
| Given a set of plant conditions and a copy of AOP-018, DE'FEKMTNE the appropriate response
| |
| DEVELOPMENT REFERENCES: AOP-018
| |
| REEE,NENCES SUPPLIED TO APPLICANT: AOP-01 8, Attachment 1
| |
| QUESTION SOURCE: NEW SIGNIFKCANTEY MODIFIED DIECT
| |
| BANK NUMBER FOR SIGNIFICANTLY RIODIFIED /DIRECT: AOP-3.18 01 9
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOW JIJSTHFICACTION (COKRECl ANSWER d'd):
| |
| 4 a. 'A' RCP vibration has exceeded limits and the pump must be stopped. With the plant in Mode 2, a
| |
| reactor trip is not required, but the plant must he shutdown.
| |
| b. Plausible since these would be the correct actions if the plant was in Mode 1. but the plant is in Mode
| |
| c. Plausibie since these are the correct actions, but 'C' RCP has not reached any trip limits while 'A' RCP
| |
| has.
| |
| d. PlausibIe since these would be the correct actions if the plant was in Mode I, but 'C' RCP has not
| |
| reached any trip limits while 'A' RCP has and the plant is in Mode 2.
| |
| DIFFICUL'IT ANALYSIS:
| |
| COMPWEIIFX'"E /ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Analysis to determine which RCP must be stopped and comparison to power
| |
| level to determine proper action
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 23
| |
| Given the following conditions:
| |
| a The Pressurizer Relief Tank (PRT) is being cooled by recirculation through the
| |
| Reactor Coolant Drain Tank Heat Exchanger per OP-100, Reactor Coolant System,
| |
| and 01-120.08, Radioactive Equipment Drain System.
| |
| e IED-143, RCDT RECIRC ISOIATIOM, loses its air supply.
| |
| 1ED- 143 will fail open, ...
| |
| a. but NOT effect the PRT cooldown because it is already open during this
| |
| evolution.
| |
| b. slowing down the cooling of the PRT due to starting a recirculation of the RCDT.
| |
| c. but NOT affect the PRT cooldown because 1ED- 139 is shut.
| |
| d. causing a lowering level in the PRT as coolant is diverted to the RCDT.
| |
| ANSWER
| |
| d. causing a lowering level in the PRT as coolant is diverted to the RCDT
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Keactor Operator
| |
| Data Sheets
| |
| QUEST1O.Y NUMBEW: 23 TIEWGROUP: 2/1
| |
| K4 IMPORTANCE: RO 2.6 SRO
| |
| B O C F RC~o~m m r : 41(b) 3 @@I
| |
| KA: 007K4.01
| |
| Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following: Quench tank
| |
| cooling
| |
| OBJECTIVE: PZR-3.0-3
| |
| Given il flow diagram of the PRT or associated subsystems and the appropriate procedure, correctly
| |
| ALIGN the P W for filling, draining, recirculation, or cooldown
| |
| DEVEI,OPMENF REFERENCES: C A R 2165-S-1313
| |
| OP-120.08
| |
| REFERENCES SUPPLIED TO APPIJCANT: CAR 21654-1313
| |
| QUEsTHON SOURCE: NEW SIGNIFICAVTLY MODIFIED DIRECT
| |
| RANK NUMBER FOR SIGNIFICANTLY MODIFIEI) I DIRECT: PZR-I33 002
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTHFICACTHON (CORRECT ANSWER +d):
| |
| 8. Plausible since IEIP-143 does fail open. but it is not open during the cooldown evolution
| |
| b. Plausible since 1ED-143 does fail open and will cause the KCDT to recirc, but it is not open during
| |
| the cooldown evolution.
| |
| C. Plausible since IEIP-143 does fail open, but iED-I39 is not closed during the cooldown evolution.
| |
| d d. iED-143 failing open will result in the PRT level decreasing and the KCDT level increasing as water
| |
| is transferred from the IRT to the RCUT.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE I ANALYSIS KNOWLEDGE 1RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Anaiysis of the effect of a valve failure on PRT cooldown, having knowledge of
| |
| valve alignment required
| |
| Post Validation Revision
| |
| | |
| Harris NIPC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 24
| |
| While operating at 100% power, 125 VDC bus DP-1B-SW is isolated due to a fault.
| |
| Which of the following identifies two (2) Technical Specification Action Statements that
| |
| must be entered as result of the bus fault?
| |
| a. 0 3.4.1.2, AFW Modes 1,2, and 3, due to the TDAFW pump being inoperable
| |
| as a result o f a loss ofpower to one (i) of the steam supply valves
| |
| D 3.4. I. I, Reactor Coolant Loops and Coolant Circulation, due to the RCPs
| |
| being inoperable as a result o f a loss of tripping power to the motor breakers
| |
| b. e 3.4.1.1, Reactor Coolant Loops and Coolant Circulation, due to the RCPs
| |
| being inoperable as a result of a loss of tripping power to the motor breakers
| |
| e 3.6.5, RCS Leak Detection, due to RM-3502A being inoperable as a result of
| |
| the sample isolation valves automatically closing
| |
| c. e 3.7.1.2, AFW Modes I, 2 , and 3, due to the TDAFW- pump being inoperable
| |
| as a result of ti loss o f power to one (1) of the steam supply valves
| |
| e 3.8.1 .l, AC Sources Operating, due to the E D 6 being inoperable as a result
| |
| of a loss ofpower to the EDG governor control circuit
| |
| d. 3.8.1. I, AC Sources Operating, due to the EDG being inoperable as a result
| |
| of a loss of power to the EDG governor control circuit
| |
| e 3.6.5, RCS Leak Detection, due to KM-3502A being inoperable as a result o f
| |
| the sample isolation valves automatically closing
| |
| ANSWER
| |
| c. e 3.7.1.2, AFW Modes I, 2 , and 3, due to the TDAFW pump being inoperable
| |
| as a result o f a loss ofpower to one (1) ofthe steam supply valves
| |
| e 3.8.1 .l, AC Sources Operating, due to the EDG being inoperable as a result
| |
| of a loss of power to the EDG governor control circuit
| |
| Post Validation Revision
| |
| | |
| Harris NRC: Written Examination
| |
| Reactor Operator
| |
| nata Sheets
| |
| QUESTION NUMBER: 24 TIWGROUP: lil
| |
| KAIMPORTANCE: RQ 3.4 SRQ
| |
| IOCPR55 CONTENT: 41@) 8 4Xb)
| |
| KA: 064K2.03
| |
| Knowledge of kDG bus power supplies to the following: Control power
| |
| OB3ECTIVE: AOP-3.25-3
| |
| Given plant conditions, DISCUSS the following notes, cautions, and procedural steps as they apply
| |
| The effects of a loss of a DC bus on equipment operability (Le., DG, sequencer, and TL) AFW)
| |
| DEVELOPMENT REFERENCES: AOP-025
| |
| REFERENCES
| |
| ~~~ -~ SUPPLIED TO MPLICANE None
| |
| QBJESTION SOURCE: NEW SIGNIFICANTLY MODIFBED
| |
| RANK NUMBER FOR SIGKIFICANTLY MODIFIED / DIRECT: AOP-3.25-R3 004
| |
| NRC EXAM HISTORY: None
| |
| PIISTRACTOR JUSTIFICACTPON (CORRECT ANSWER .Id):
| |
| a. Plausible since the TDAFW pump is inoperable and the KCPs use TIC power for the tripping coils, but
| |
| tripping the RCPs is not part of the operability requirement.
| |
| b. Plausible since the RCPs use DC power for the tripping coils and the RCS leak detection sample
| |
| valves will isolate on a loss of power, but tripping the RCPs is not pa^ of the operability requirement
| |
| and the sample isolation valves close on a CVIS due to a loss of AC power.
| |
| d e. The TDAFW pump is inoperable due to a loss of power to a steam supply valve and the EDG is
| |
| inoperable due to a loss of power to the governor, as well as to the generator excitation control circuit
| |
| and sequencer.
| |
| d. Plausible since the EDG is inoperable and the RCS leak detection sample valves will isolate on a loss
| |
| of power, hut the sample isolation vaIves close on a CVIS due to a loss of AC power.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECXLL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Analysis of the effect o f a loss ofDC power on equipment operability and
| |
| knowledge of the BSLCOs affected
| |
| Post Validation Revision
| |
| | |
| Harris iWC: Written Examination
| |
| Reactor Operator
| |
| QUESTION: 25
| |
| Given the folIowing indications during a plant s t m p :
| |
| e Power Kange Channel N-41 26.0%
| |
| e Power Range Channel N-42 24.5%
| |
| e Power Range Channel N-43 24.5%
| |
| e Power Range Channel N-44 25.0%
| |
| Loop'A' AT 25.5%
| |
| e Loop'B' AT 25.5%
| |
| e Loop'C' AT 25.5%
| |
| e Turbine Load 24.5%
| |
| Which of the following power levels should he reported as being actual reactor power?
| |
| a. 24.5%
| |
| b. 25.0%
| |
| C. 25.5%
| |
| d. 26.0%
| |
| ANSWER:
| |
| c. 2 5 9 0
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 25 TIEWGROUP 212
| |
| KA IMPORTANCE: RO 3.6 SRO
| |
| 1QCFR55CONTENT: 4L(h) 3 43(W
| |
| MA: 002K5.10
| |
| Knowledge o f the operational implications of the Following concepts as they apply to the RCS:
| |
| e Relationship between reactor power and RCS differential temperature
| |
| OBJECTIVE: NIS-3.0-13
| |
| Discuss the cautions associated with monitoring NI power levels during plant start-up and power
| |
| operations
| |
| DEVELOPMENT REFERENCES: GP-005
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| OR SIGNIFICANTLY MODIFIED / DIRECT: NIS-RIO 004
| |
| NRC EXAM HISTORY None
| |
| DHSTRACTOR JUSTIFICACTION (CORRECT ANSWER '/'(I):
| |
| a. Plausible since this is the lowest given power level and may be considered to be the most
| |
| conservative, but GP-005 provides guidelines for which power level should be considered.
| |
| b. Plausible since this is the average NIS power level, hut the highest as identified by GP-005
| |
| requirements is the average loop AT^
| |
| d c. Until a calorimetric is perfomled at 30% power, true reactor power shall be assumed equal to the
| |
| highest of the following indicators: average Power Range NI value, average percent AT, or Main
| |
| Turbine load
| |
| d. Plausibie since this is the highest given power level and may be considered to be the most
| |
| conservative, but CP-00s provides guidelines for which power level should be Considered.
| |
| ICULTY A N a Y S I S :
| |
| COMPREHENSIVE /ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Calculatiun of average power indications and detemlination of most
| |
| consewativc value
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 26
| |
| AH-82A, NORMAL PURGE SCJPPLY FAN AH-82A, fails to start when the control
| |
| switch is placed in START.
| |
| Which of the following intsrloclcs would prevent the fan from starting?
| |
| a. Normal Purge Inlet and Discharge Valves are open
| |
| b. AW-82A fan inlet damper is closed
| |
| c. Fan inlet air temperature is low
| |
| d. Containment differential pressure is zero
| |
| ANSWER
| |
| d. Containment differential pressure is zero
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 26 TIEWGROUP: 2i2
| |
| KAIMPORTANCE: RO 3.0 SRO
| |
| 10CFR55 CONTENT: 41(b) 9 43m
| |
| KA: 029A1.03
| |
| Ability to predict and/or monitor changes in parmeters to prevent exceeding design limits) associated
| |
| with operating the Containment Purge System controls inchding: Containment pressure, temperature,
| |
| and humidity
| |
| OBJECTIVE: CVS-3.0-I22
| |
| LOCATE the controls and MII.AIN the interlocks associated with the following major components
| |
| NCPMU units, including AH-82 fans
| |
| DEVELOPMENT REFERENCES: OP-168
| |
| REFERENCES SUPPLIED TO MPLBCANT: None
| |
| QIJESrION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
| |
| BANK IVUMEER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
| |
| NRC: EXAM HISTORY None
| |
| DISTR4CTOR JUSTIFICACTIBN (COIPRECT ANSWER dd):
| |
| a. Plausible since the valves are interrocked to close if fan N M 2 A is stopped, but are manually opened
| |
| prior to the start of the fan.
| |
| &. Plausible since the inlet damper is interlocked to open when the fan i s started, Rut are closed when the
| |
| fail is started.
| |
| c. Plausible since a low inlet air temperature will cause an alarm condition: hut will not prevent the fan
| |
| from starting.
| |
| d d. Fan AH-S2A will only start if containment AP is more negative than -0.400 N O .
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREIIENSRE /ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge of interlocks associate with containment purge fans
| |
| Post Validation Revision
| |
| | |
| Harris hRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 27
| |
| Given the following conditions:
| |
| 6 The plant is at the Point of Adding Heat (POAII) when a SG PORV fails open.
| |
| 6 RCS temperature decreases and stabilizes at 548 "F.
| |
| Which of the following predicts the plant response and the operator actions requircd?
| |
| a. Reactor power increases; withdraw control rods and dilute, in a controlled
| |
| manner, to restore RCS temperature to program within 15 minutes
| |
| b. Reactor power increases; trip the reactor if RCS temperature CANNOT be
| |
| restored above 551 O F in a controlled manner within 15 minutes
| |
| c. The reactor becomes subcritical; trip the reactor if criticality CANNOT be
| |
| restored in a controlled manner within 15 minutes
| |
| d. The reactor becomes subcritical; immediately trip the reactor
| |
| ANSWER:
| |
| b. Reactor power increases; trip the reactor if RCS temperature CANNOT be
| |
| restored above 551 *F ia a controlled manncr within IS minutes
| |
| Post Validation Revision
| |
| | |
| Hamis NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 27 TIEWGROUP 211
| |
| KAIMFORTANCE: RO 3.3 SRO
| |
| 1QCFR55CONTENT: 41(b) 6/10 43(b)
| |
| KA: 039.42.05
| |
| '4bility to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (h) based
| |
| on predictions, use procedures to correct. control, or mitigate the consequences of those malfunctions or
| |
| operations: increasing steam demand, its relationship to increases in reactor power
| |
| OBJECTIVE: E-3.10-1
| |
| Apply the philosophies of OMM-001 and PLF-629 regarding safe and conservative decisions that must
| |
| be made by a control room crew
| |
| DEVELOPMENT REFERENCES: OMM-Q01
| |
| IE-LP-3.60 (Salem Event, SOER 94-01)
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QgJESTION SOURCE: NEW SIGNIF1CA"TLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DISTR4CTOR ~S'HFICACTPON(CORRECT AKSWER d'd):
| |
| a. Plausible since reactor power will increase, but temperature is not to be restored using two different
| |
| methods of reactivity control simultaneously and the 15 minute limit is to restore temperature above
| |
| 55 1 O F , not to program.
| |
| 4 b. The first operator action should be to attempt to stop the cause (e&, secure the overfeeding) of the
| |
| transient. 'i emperature may then be recovered by using control rods in a slow and controlled manner.
| |
| Temperature has to be restored to greater than 55 1 "F within 15 minutes due to the requirements of TS
| |
| 3.1.1.4.
| |
| c. Plausible since the 15 minute time limit is associated with restoration, but the reactor does not become
| |
| subcritical.
| |
| d. Plausible since the reactor is to he tripped if it becomes subcritical due to a malfunction per OMM-
| |
| 001, but the reactor does not become subcritical.
| |
| I)IFFI'ICULTY ANALYSIS:
| |
| COMPKEIIENSIVE I ANALYSIS KNOWLEDGE [RECALL
| |
| DIFFICULTY HATING: 3
| |
| EXP1,ANATION: Analyze the plant response to an increase in steam demand and determine
| |
| appropriate actions
| |
| Post Validation Revision
| |
| | |
| Harris N R C : Written Examination
| |
| Reactor Operator
| |
| QUESTION: 28
| |
| The plant is operating at 100% power with the following conditions:
| |
| T d Ambient Temp CT Basin 'Temp
| |
| I500 35 O F 64 "F
| |
| 1900 20 "F 60 OF
| |
| 2300 10 O F 58 "F
| |
| Which of the following describes the correct CT Deicing Gate Valve alignment for these
| |
| conditions?
| |
| 1900 2300
| |
| a. Full Open Full Open
| |
| b. Full Open Half Open
| |
| C. Half Open Full Open
| |
| d. Half Open Half Open
| |
| ANSWER:
| |
| b. Full Open Half Open
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 28 TIEWGROUP: 3
| |
| KA IMPORTANCE: RO 2.8 SRO
| |
| BCCFH55 CONTENT: 41(b) 10 43(b)
| |
| m: 2.1.25
| |
| Ability to obtain and interpret station refexme materials such as graphs, monographs, and tables which
| |
| contain perfonnance data
| |
| OBJECTIVE: CT-R3
| |
| Given OP-141, Attachment 5 , ANALYZE a set ofadverse weather conditions and DESCRIBE the
| |
| operation of the Cooling Tower System to prevent ice damage to the fill material
| |
| DEVEI,OPMENT REFERENCES: OP-141
| |
| REFERENCES SUPPIJED TO APPLICANT: OP-141, Attachment 5
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGXTIFICANT1,YMODIFIED / DIRECT: CT-R3 001
| |
| NRC EXAM HISTORY: Harris NRC 2000
| |
| DISTRACTOR .JUSTIFICACTION (CORRECT ANSWER .Id):
| |
| 8. Plausible since valves should he open at 1900, hut are required to be changed to half open at 2300,
| |
| d lp. At 1500 conditions call for valves to be full open, at 1900 conditions call for no change in position,
| |
| and at 2300 conditions call for change to half open.
| |
| c. Plausible since valves should he changed between 1900 and 2300, but should go from full open to half
| |
| open.
| |
| d. Plausible since valves should be half open at 2300, hut should be full open at 1900 due to no change
| |
| from 1500.
| |
| DIFPICULTY ANALYSIS:
| |
| COMPKEIHENSIVE I ANALYSIS 0 KNOWLEDGE / RECALL
| |
| DIFFICULW RATING: 3
| |
| EXPLANATION: Application of given data to curve to determine required operation of deicing
| |
| valves
| |
| Post Validation Kevision
| |
| | |
| Harris NRC Written Examination
| |
| Rcilctor Operator
| |
| QUESTION: 29
| |
| Which of the following conditions requires processing a Radioactive Gaseous Release
| |
| BATCH permit'?
| |
| a. Manual operation of the Containment Vacuum Relief System
| |
| b. Resetting and starting the Containment Pre-Entry Purge following an automatic
| |
| isolation
| |
| C. Startup of the standby Airborne Radioactive Removal fan (S-I) following a trip of
| |
| the running fan
| |
| d. Swapping the operating Normal Containment Purge Makeup (AH-82) fans frum
| |
| Train A to Train B
| |
| ANSWER:
| |
| b. Resetting and starting the Containment Pre-Entry Purge following an automatic
| |
| isolation
| |
| Post Validation Revision
| |
| | |
| Hmis NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 29 TIEWGROUP 3
| |
| KAIMPORTANCE: RO 2.4 SRO
| |
| 10CFR55 CONTENT: 41(b) 12 43m
| |
| KA: 2.3.11
| |
| Ability to control radiation relases
| |
| OBJECTIVE: CVS-3.0-R4
| |
| EXPLAIN the conditions which require a radioactive release permit prior to operating components
| |
| associated with the Containment Ventilation System
| |
| DEvELOpMENr REFERENCES: OP-168
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODPFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED 1 DIRECT: CVSR4 00 1
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTEPICACTION (CORRECT ANSWER +a):
| |
| 1p. Plausible since the Containment Vacuum Relief System interfaces with Containment atmosphere, but
| |
| it supplies air to Containment and does not require a release permit to operate.
| |
| s' b. For initial start-up of the Containment Pre-Entry Purge (start of an outage) or if purge was secured for
| |
| radiological reasons, a Ratch release pennit is required
| |
| c. Plausible since if this was the initial startup ofthe system a batch release wrouId be required, but once
| |
| the system is in operation for a period oftime only a continuous release permit is required.
| |
| 8. Plausible since if this w'as the initial SVdI'tllp of the system a batch release would be required, hut once
| |
| the system is in operation for a period of time only a c.ontinuous release permit is required.
| |
| DIFFICL%TY ANALYSIS:
| |
| GQMPREIPENSNE I ANALYSIS KNOWLEDGE I KECALI,
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge of the radiological release permits required
| |
| Post Validation Revision
| |
| | |
| Hanis NRC Written Examination
| |
| Reactor Operator
| |
| QUESTPOW: 30
| |
| Which of the following two (2) conditions are both identifial hy EOP-EPP-Oi3, LOCA
| |
| Outside Containment, as being used to identify that the LOCA has been isolated?
| |
| a. e RCS pressure increasing
| |
| E- RAD local room temperalures
| |
| b. RAE3 iocal room temperatures
| |
| E- RAB radiation levels decreasing
| |
| e. e RAR radiation levels decreasing
| |
| * Local observation of the isolation
| |
| d. e RCS pressure increasing
| |
| E- Local observation of the isolation
| |
| ANSWER:
| |
| d. E- RCS pressure increasing
| |
| m 1,ocai observation of the isolation
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NU-MBER: 30 TIERKROUP: 1/1
| |
| KAIMPORTANCE: RO 3.5 SRQ
| |
| 10CFR55 CONTENT: 41(b) 10 43(W
| |
| KA: WE04EK1.2
| |
| Knowledge ofthe operational implications ofthe following concepts as they apply to the (LOCA Outside
| |
| Containment) Normal, abnormal and cmergency operating procedurcs associated with &OCA Outside
| |
| Containment)
| |
| OBJECTIVE: EOP-2.3-R4
| |
| Using appropriate plant procedures and prints, determine the following:
| |
| e Transitions to other EOPs
| |
| DEVELOPMENT REFERENCES: EOP-EPP-0 13
| |
| REFERENCES SUPPIJED TO APPLICANT: None
| |
| QUESTIQN SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DPKECT: EOP-3.3 024
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR SIJSTIFICACTION (CORRECT ANS\IER dd):
| |
| a. Plausible since RCS pressure increasing is one of the indications used, hut pressurizer level may not
| |
| be indicative of actual RCS inventory or the leak being isolated and is not used in IIPP-013.
| |
| b. Plausihle since these may both be indications that might support that the leak is isolated, but
| |
| pressurizer level may not be indicative of actual RCS inventory or the leak being isolated and is not
| |
| used in EPP-013.
| |
| c. Plausible since local observation is one of the indications used, but KAB radiation levels may he
| |
| elevated for some time after isolation and is not used in EPP-013.
| |
| 4 d. EPP-013 determines that the ILEA outside containment is isolated if RCS pressure is increasing and
| |
| if local observation confirms the isolation.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge ofthe conditions required by EPF-013 to determine that a LOCA
| |
| outside containment is isolated
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 31
| |
| Which of the following is the reason for purposely tripping the Reactor Coolant Pumps
| |
| (RCPs) under accident conditions'!
| |
| a, Ensure RCPs are available later in the event if they should be needed in response
| |
| to an inadequate core cooling condition
| |
| b. Prevent RCP runout in the event of a large break LOCA
| |
| c. Prevent excessive depletion of RCS inventory through a small break in the RCS
| |
| d. Prevent damage to RCPs due to pumping a two-phase mixture event
| |
| ANSWER:
| |
| c. prevent excessive depletion of RCS inventory through a srnali break in the RCS
| |
| Post Validation Revisioti
| |
| | |
| Harris NKC Written Examination
| |
| Reactor Opcrator
| |
| Data Sheets
| |
| QUESTION NUMBER: 3 1 TIEWGROUP: 111
| |
| KA IICIPORTANCE: RO 4.2 SRO
| |
| 10CFH55 CONTENT: 4f(b) 3/10 43(b)
| |
| KA: 000009EK3.23
| |
| Knowledge of the reasons for the following responses as the apply to the small break LOCA: RCP
| |
| tiipping requirements
| |
| OBJECTIVE: BD-3.1-1
| |
| Analyze the Reactor Coolant Pump (RCP) trip criteria. This analysis should include, at the minimum, the
| |
| following topics:
| |
| The reason for purposely tripping the KCPs under c e r t h accident conditions
| |
| DE\'E;LOPMEN'F REFERENCES: Generic Issues of ERG Background - Executive Voiume
| |
| LP-BD-3.1
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: BD-3.1 001
| |
| NHC EXAM IIISTORY: None
| |
| DISTRACTOR JUSTIFICACTIBN (CORRECT ANSWER d'd):
| |
| a. Plausible since for most accidents it is desirable to have RCPs available, particularly those cases where
| |
| an inadqnate core cooling condition might exist.
| |
| b. Plausible since little work is required by the RCPs in the event o f a large break L O W , but this would
| |
| result in a lower pump current. not a runout condition.
| |
| d C. Tripping the RCPs during the early stages of a small break L.OCA limits the amount of mass lost out
| |
| the break, thereby increasing the mass available for heat removal in the event the purrips werc not
| |
| tripped but tripped at a later time.
| |
| d. Plausible since RCPs are not designed to pump a two-phase mixture and it would be desirable to
| |
| protect the pumps from damage.
| |
| DIFFICULTY ANALYSIS:
| |
| COMMPREIIENSIVE I ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION Knowledge of the reasons for tripping RCPs during a small break LOCA
| |
| Post Validation Revision
| |
| | |
| Harris MRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 32
| |
| WhiIe operating at 100% power, a failure ofthe Pressurizer Pressure AUTO controller
| |
| (PK-444A) occurs and the Reactor Operator takes manual control ofthe controller.
| |
| While restoring from the failure, in order to maintain PRZ pressure at 2235 psig, the
| |
| Reactor Operator should adjust PK-444A to setpoint of approximately ...
| |
| a. 31%.
| |
| c. 69%.
| |
| d. 89%.
| |
| ANSWER:
| |
| c. 67%.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER. 32 TKEKIGROUP: 1/1
| |
| hXIMPf3XTANCE: RO 2.6 SRO
| |
| 1UCFRSS CONTENT: 41(b) 3/7 43(W
| |
| Kk 000027AK2.03
| |
| Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the
| |
| following: Controllers arid positioners
| |
| OBJECTIVE: PZRPC-3.0-2
| |
| E)ITL.AIN how the pressurizer pressure control setpoint is both determined and adjusted for a desired
| |
| KC:S pressure
| |
| DEVELOPMENT REFERENCES: SD-100.3
| |
| QUESTION SOURCE:
| |
| - -
| |
| REFERENCES SUPPLIED TO APPLICANT:
| |
| u NEW
| |
| AOP-019
| |
| None
| |
| SIGNIFICANTLY MODIFIED u DIRECT
| |
| BANK NUMBER FOX SIGNIFICANTLY MODIFIED / DIRECT: PZRPC-R5001
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOX JUSTIFICACTlON (CORRECT ANSWER d'd):
| |
| a. Plausible ifthe setpoint is calculated by dividing 535 psig (2235-1700) by 1700 p i g (Low end of
| |
| span), with a result of 31.4%.
| |
| b. Plausible since this is the mid-point of the 0-100% scale
| |
| 4 E. PK-444A setpoint is detennined by calculating the percent of span of the contmller. Controller span
| |
| is 800 psig (1700 - 2500). 2235 psig - 1700 psig := 535 psig. 535 psigl800 psig = 66.9%
| |
| d. PIausible if the ratio of 2235 psig to the upper end of the span (2500 psig) is calculated, with a result
| |
| of 89.4%.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE I ANALYSIS KNOWLEDGE I REGALL
| |
| DLFIWUII.TY RATING: 3
| |
| EXPLANATION: Calculation of required setpoint for pressurizer pressure control
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 33
| |
| Which one of the following correctly describes how and why the speed of the Condensate
| |
| Booster Pumps (CBPs) is varied?
| |
| a. Changing the coupling impeller vane pitch to maintain a constant 430 psig feed
| |
| pump suction pressure
| |
| b. Changing the coupling impeller vane pitch to maintain desired flow from the
| |
| CBPs to the feed pumps
| |
| c. Varying the miount of oil to the coupling between the pump and motor to
| |
| inaintain a constant 430 psig at the fed pump suction
| |
| d. Varying the amount of oil to the coupling between the pump and motor to
| |
| maintain a desired flow from the CBPs to the feed pumps
| |
| ANSWER:
| |
| c. Varying the amount of oil to the coupling between the pump and motor to
| |
| maintain a constant 430 psig at the feed pump suction
| |
| Post Validation Revision
| |
| | |
| Ifanis NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 33 TIERIGROUB: 211
| |
| KA IMPORTANCE: RO 3.2 SRO
| |
| 10CFR55 CONTENT: 41(b) 4 4Xb)
| |
| KA: 056G2.1.28
| |
| Knowledge ofthe purpose and function of major system components and controls. (Condensate)
| |
| OBSECTTVE: CFW-3.0-4
| |
| DESCRIBE lhe basic constmetion and operation ofthe following CFW System components /
| |
| subsystems
| |
| o CRP Variable Speed Fluid Coupling (VSPC)
| |
| DEVELOPMENT REFERENCES: SD-134
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| Qt?ESTIOX SOURCE: NEW SIGNIFIC.4NTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIIPECT: CFW-R.? 001
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORXECT ANSWER d'd):
| |
| 8. Plausible since the variable speed coupling maintains 430 psig at the feed pump suction, but it is
| |
| maintained by using oil between the motor and pump coupling.
| |
| b. Plausible since this is a means of providing a variable flow rate, but the CBPs used a variable speed oil
| |
| coupling.
| |
| 4 e. An oil bath between the motor and pump coupling causes the pump tto operate at a variable speed to
| |
| rnahtain a constant 430 psig suction at the feed pump.
| |
| d. Plausible since an oil bath between the motor and pump coupling causes the pump to operate at a
| |
| variable speed, but it is designed to maintain a constant 430 psig suction at the feed pump rather than a
| |
| constant flow rate.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS #IQ\VLEI>GE. /RECALL
| |
| DIFFICULTY HATING: 3
| |
| EXPLANATION: Knowledge of the operation of the CBPs
| |
| Post Validation Revisicn
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 34
| |
| Given the following conditions:
| |
| 0 The plant is operating at 100% power.
| |
| A tube leak has been detected on 'W'SG.
| |
| The Condenser Vacuum Pump Rad Monitor, REM- 1TV-3534, and H-X- 15 curves are
| |
| being monitored every 15 minutes to estimate the leak rate.
| |
| CYPE is operating with NQ motivating air.
| |
| Which of the following readings noted on REM-1TV-3534 is the MINIMUM reading
| |
| that would require a plant shutdown per Technical Specifications?
| |
| a. 5.40 E -7
| |
| b. 6.00 E -7
| |
| c. 1.08 E -6
| |
| d. 1.80 E -6
| |
| ANSWER:
| |
| c. 1.08 E -6
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 34 TIEWGROUP 1/2
| |
| KAIMPORTANCE: KO 3.2 SRO
| |
| 10CFH55 CONTENT: 41(b) 7;10 43(b)
| |
| KA: 000037AA2.10
| |
| Ability to determine and interpret the foliowing as they apply to the Steam Generator Tube Leak Tech-
| |
| Spec limits for RCS leakage
| |
| OBJECTIVE: AOP-3.16
| |
| For a primary-to-secondary leak, DESCRIBE when a power reduction or unit shutdown is required.
| |
| DEVELOPMENT REFERENCES: AOP-0 16 (unknown)
| |
| Curves PI-X-15aihic
| |
| REFERENCES SUPBIdED TO APPLICANT: Curves H-X-ISalbic
| |
| QUESTION SOURCE: NEW SPGNIFICM'TI,Y MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: Hams NRC 2000-80
| |
| NRC EXAM HISTORY: Hams NRC 2000
| |
| DHSTRACTOW .PUSTIFICACTPON (COHWECT ANSWER d'd):
| |
| 1%. Plausible since this exceeds would exceed I'SAL 2 limits if operating on full motivating air (curve H-
| |
| X-ISa), but the incorrect curve is used.
| |
| b. PIausihle since this exceeds would exceed PSAI, 2 limits if operating on intermediate motivating air
| |
| (curve H-X-ISb), but the incorrect curve is used.
| |
| 4 e. Lowest level that would exceed 75 gpd (PSAL 2) which would require a TS shutdown
| |
| d. Plausible since this exceeds the PSAL 3 limit which would require a 'I'S shutdown, but this is not the
| |
| lowest level that would require the shutdown.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSEVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Interpretation of plant data on KCS leakage curve and comparison to procedural
| |
| reyuirenients
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTIQN: 35
| |
| FRP-J.2, Response to Containment Flooding, directs that the containment sump be
| |
| sampled for activity, and then to notify the operations staff of sump level and the sample
| |
| results.
| |
| What action will the operations staff be considering based on this information?
| |
| a. Isolation of the Cold Leg Accumulators
| |
| b. Isolation of the CNMT spray additive tank
| |
| e. Shift to Hot Leg Recirculation
| |
| d. Transfer of sump water to tanks outside containment
| |
| ANSWER
| |
| d. Transfer of sump water to tanks outside containment
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESlXOX KkrhXBEW: 35 TIEWGROUP: I 12
| |
| KAIMPORTANCE: KO 2.4 SRO
| |
| IIICFRR55 CONTENT: 41@) 9/10 43(b)
| |
| Kh: WE15EK1.2
| |
| Knowledge of the operational implications of the following concepts RS they apply to the (Containment
| |
| Flooding) Nonnal, abnormal and emergency operating procedures associated with (Containment
| |
| Flooding)
| |
| OEECTIVE: EOP-3.13-4
| |
| Given the following EOP steps, notes, and cautions, DESCRIRE the associated basis
| |
| 0 Sampling the CNMT sump for activity (J.2)
| |
| DEVELOPMENT REFERENCES: EOP-FW-J.2
| |
| LP-IOP-3.13
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESrION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BAKK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.13 010
| |
| NHC EXAM HISTORY: None
| |
| DPSIRACTOR JUSTIFIICACTION(CORRECT ANSWER dd):
| |
| a. Plausible since if flooding has occurred it is likely that a large RCS leak has also occurred and the
| |
| accumulators have dumped to containment and would no longer be needed, but this sample is to
| |
| detennine whether the water can be transferred.
| |
| b. Plausible since if flooding has occurred it is likely that a large RCS leak has also occurred and the
| |
| spray chemical addition tank may have emptied to containment and would no longer be needed. but
| |
| this sample is to detennine whether the water can be transferred.
| |
| e. Plausible since a shift to hot leg recirc from the sumps will eventually be required in the event of a
| |
| large break LOCA, but this sample is to determine whether the water can be transferred.
| |
| 4 d. The containment sump is sampled to determine if excess water can be transferred to storage tanks
| |
| located outside containment.
| |
| DIFFICULTY AKALYSIS:
| |
| COMPREIIENSIVE / ANALYSIS KNOWLEDGE I RFXALL
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION: Knowledge of purpose for sampling sumps following flooding inside
| |
| containment
| |
| Post Validation Revision
| |
| | |
| Hank NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 36
| |
| Given the following conditions:
| |
| KHR Pump A-SA is tagged out.
| |
| Following a large break LOCA, the crew was performing EOP-EPP-010, Transfcr to
| |
| Cold Leg Recirculation.
| |
| 1SI-301, CONTAINMENT SUMP TO RHR PUMP B-SB, failed to open and the
| |
| crew transitioned to EOP-EPP-012, Loss of Emergency Coolant Recirculation.
| |
| Both Containment Spray Pumps automatically transferred to the Containment Sump.
| |
| Two (2) Containment Fan Chlers are operating.
| |
| Containment pressure is 12 psig and decreasing slowly.
| |
| W i l e performing EPP-012 the Reactor Operator notes that WWST level is 2%)with
| |
| both CSIPs, both Containment Spray Pumps, and RHR Pump B-SR operating.
| |
| Which ofthe following actions are to he taken?
| |
| a. Stop the RHR pump ONLY
| |
| b. Stop both CSIPs and the RHR pump ONLY
| |
| c. Stop both CSIPs, the RIHR pump, and one Containment Spray pump ONLY
| |
| d. Stop both CSIPs, the RKR pump, and both Containment Spray pumps
| |
| ANSWER:
| |
| b. Stop both CSIPs and the RIlR pump ONLY
| |
| Post Validation Revision
| |
| | |
| Harris NKC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 36 TIER/GROUP 1/1
| |
| KAIMPORTANCE: RO 3.7 SRO
| |
| 10CFRS5 CONTENT: 41(b) 7 43(b)
| |
| MA WEllEKl.1
| |
| Knowledge of the operational implications ofthe following concepts as they apply to the (Loss of
| |
| Emergency Coolant k2circulation) Components, capacity, and function of emergency systems
| |
| OBJECTIVE: EOP-2.3-S2
| |
| Predict how each ofthe following could impact efforts to maintain core cooling during a LOCA
| |
| e Failure of valves to realign for cold-Ieg recirculation
| |
| DEVELOPMENT REFERENCES: EOP-CPP-012
| |
| REFERENCES SBJPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGSIFICANTLY MODIFIED DIRECT
| |
| RANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.3-K5 004
| |
| NRC EXAM HISTORY: None
| |
| DISTRACIQR JUSTIFICACTION (CORRECT ANSWER dd):
| |
| a. Plausible since the KHR pump is still aligned to the RWST and must be stopped, but the CSIPs are
| |
| also aligned to the RWST and must likewise be stopped.
| |
| 4 b. The RIIR pump and the CSPs are still aligned to the RWST and must be stopped when the RWST
| |
| empty alarm is received at 3% level.
| |
| c. Plausible since the RIIR pump and the CSIps must be stopped, but the spray pumps can continue to
| |
| operate since they are no longer aligned to the RWST.
| |
| d. Plausible since the FWR pump and the CSIPs must be stopped, but the spray pumps can continue to
| |
| operate since they are no longer aligned to the RWST.
| |
| PCULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLE.DGEI RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPIANATION: Analyze plant conditions to determine which pumps are taking a suction from
| |
| the RWW to determine the pumps which are to be stopped
| |
| Post Validation Kevision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 37
| |
| Given the following plant conditions:
| |
| e The plant is operating at 100% power.
| |
| e 1CS-7,45 GPM Letdown Orifice A, and lC§-S,hO GPM Letdown Orifice B, are
| |
| closed.
| |
| e 1CS-9,60 GPM Letdown Orifice C, is open.
| |
| e The Reactor Makeup System is setup properly and is in AUTO.
| |
| e VGT level transmitter, LT-112, fails high.
| |
| Assuming NQ operator action, which of the following describes the piant response?
| |
| a. Charging Pump suction is eventually lost as VCT level decreases
| |
| h. ICs-120 (LCV-l15A), Letdown VCT/Hold Up Tank, aligns to the VCT and N O
| |
| automatic makeup will occur
| |
| c. ICs-120 (LCV-1 15A), Letdown VCT/Hold Up Tank, aligns to the HUT and a
| |
| CONTINIJOUS makeup to the VCT will occur
| |
| d. ICs-120 (LCV-I 15A), L,etdown VCT/Hold Up Tank, aligns to the HUT and
| |
| INTERMITTEXT makeups at normal setpoints will occw
| |
| ANSWER:
| |
| d. IC§-120 (LCV-I 15A), Letdown VCT/Hold IJp Tank, aligns to the HUT and
| |
| INTEKMITTENT makeups at normal setpoints will occur
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 37 TIEWGROUP: 211
| |
| KA IMPORTANCE: RO 3.0 SRO
| |
| IOCFWS CONTENT: 41(b) 7 4303)
| |
| KA: 00441.06
| |
| Ability to predict andfor monitor changes in parameters (to prevent exceeding design limits) associated
| |
| with operating the CVCS controls including: VCT level
| |
| OBJECTIVE: CVCS-RS
| |
| PREDICT the response of the CVCS to the following failures
| |
| c. LT-I 12 or LT-115 faiiure (high or low)
| |
| DEVELOPMENT REFERENCES: AOP-003
| |
| REFERENCES SUPPLIED TO APPLICAWT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| RANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Harris NRC 2002-32
| |
| NRC EXAIM HISTORY: Harris NRC 2002
| |
| DISTRACTOR JLJSTIFHCACTION (CORRECT ANSWER dd):
| |
| 5. Plausible since this would occur with no operator action if the high failure were 1.T-I 15 instead of LT-
| |
| I12
| |
| b. Plausible since a low failure of LT-I 15 would result in this response
| |
| e. Plausible since this would occur if letdown were in excess or equal to makeup capability. but letdown
| |
| is less than makeup capability under the given conditions.
| |
| d d. LT-I 12 failing high causes LCV-115A to fully divert to the HUT tank at 60 gptn letdown flow. VCT
| |
| level decreases and automatic makeup raises level at 120 gpm, causing niakeup to stop until level
| |
| drops again.
| |
| DIFFICULTY ANALYSIS:
| |
| COMYREIIENSIVE/ ANALYSIS KNOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPI,ANATION Requires analysis of plant response to failures in CVCS given initial plant
| |
| conditions
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 38
| |
| The plant is operating at 100% power with all e q u i p m t operable and properly aligned.
| |
| Which of the following describes changes to the Component Cooling Water System
| |
| alignment following a Safety Injection signal?
| |
| a. CCW to the Gross Failed Fuel Detector and Primary Sample Panel isolates
| |
| b. Both CCW pumps start and the Non-Essential header isolates
| |
| c. CCW to and from the RCP Motor Coolers isolates
| |
| d. Both CCW pumps start and the Thermal Banier Wx Return isolates
| |
| ANSWER:
| |
| a. CCW to the Gross Failed Fuel Detector and Primary Sample Panel isolates
| |
| Post Validation Revision
| |
| | |
| Iiarris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESlION NUMBER: 38 TIEWGROUP: 211
| |
| KAIMF'QRTANCE: RO 3.6 SRO
| |
| 10CFR55 CONTENT: 4I(b) 4 43w
| |
| KA: 008A3.08
| |
| Ability to monitor automatic operation of the CCWS, including: Automatic actions associated with the
| |
| CXWS that occur as a result ofa safety injection signal
| |
| OBJECI'ZVE: CCWS-3.0-K2
| |
| STATE how the CCWS responds during each ofthe following conditions:
| |
| Safety Injection signal
| |
| DEVELOPMENT REFERENCES: SD-145
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: SIGNIFICANTLY MODIFIED B DIRECT
| |
| cmmy MODIFIED DIRECT: ccws-ru002
| |
| NRC EXAM IIISTORY: None
| |
| DISTHACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| d a. On an SI signal, both the GPFD and sample panel receive isolation signals.
| |
| b. Plausible since the pumps will get a start signal, but only the GFFD and sample panel in the non-
| |
| essential header are isolated.
| |
| c. Plausible since the CCW to RCP isolations ciose on a Phase B signal, but Phase 5 is not generated by
| |
| an SI signal.
| |
| d. Plaiisible since the pumps will get a start signal, but the thermal banier heat exchangers are only
| |
| isolated on a Phase 5 signal.
| |
| ICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPIANATION: Knowledge ofthe response of CCWS to an SI signai
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 39
| |
| Given the following conditions:
| |
| 6 The plant is operating at 23% power.
| |
| e Steam pressure channel PT-475 is selected for control of SG A
| |
| a Stcam pressure transmitter PT-475 fails high.
| |
| Assuming NO operator action, which of the following statements describes the response
| |
| ofthe Steam Generator Water Level Control System (SGWLCS)?
| |
| a. An increase in steam flow from SG A is sensed and responds by increasing
| |
| IFW-140, MN FW A REG BYP FK-449.1, position to increase feed flow to SG
| |
| A and level increases
| |
| b. An increase in s t e m flow fmm SG A is sensed and responds by increasing
| |
| 1FW-133, MAIN FW A REGULATOR FK-478, position to increase feed flow to
| |
| SG A and level increases
| |
| c. A decrease in steam flow from SG A is sensed and responds by decreasing IFW-
| |
| 140, MN FW A REG BYP FK-479.1. position to decrease feed flow to SG A
| |
| and level decreases
| |
| d. A decrease in steam flow from SG A is sensed and responds by decreasing IFW-
| |
| 133, MAIN FW A REGULATOR FK-478, position to decrease feed flow to SG
| |
| A and level decreases
| |
| ANSWER:
| |
| b. An increase in steam flow from SG A is sensed and responds by increasing
| |
| 1FW-133, MAIN FW A REGULATOR FK-448, position to increase feed flow to
| |
| SG A and level increases
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMEiER: 39 TIEWGROUP !
| |
| 26
| |
| KAIMPBRTANCE: RO 3.0 SRO
| |
| 1QCFR55CONTENT: 41(b) 4 430~1
| |
| KA: 059.44.08
| |
| Ability to manually operate and monitor in the control room: Feed regulating valvc controller
| |
| OBJECTIVE: SGWLC-3.0-2
| |
| Given the status of the various S G W K related control switch positions and controllers, PREDICT holv
| |
| a malfunction of the following will effect the SGWIC System:
| |
| 0 S G pressure channels
| |
| DEVE1,OPMENT KEFERENCES: SD-126.02
| |
| REPEREWES SUPPLIED TO APPLICANT: None
| |
| QLJESTION SOURCE: NEW SIGNIHCAKTLY MODIFIED 0 DIRECT
| |
| BANK RUMBER FOR SIGNIFICANTLY MODIFIED / UIRECT: SGWLC-RZ 002
| |
| NRC EXAM HISTORE None
| |
| DISTRACTOR JUSTlFlCACTION (CORRECTANSWER dd):
| |
| a. Plausible since steam pressure failing high causes the steam flow to increase, resulting in SF > FF, but
| |
| the feed reg valve is in operation at this power level.
| |
| d b. Steam pressure failing high causes the steam flow to increase, resulting in SF > FF. The feed reg
| |
| valve. in operation at 15% power, opens to cause FF and level to increase.
| |
| E. Plausible since steam pressure failing causes the steam flow to changee,resulting in a SF - FF
| |
| mismatch, but the feed reg valw will open to increase FF.
| |
| d. Plausible since steam pressure failing causes the steam flow to cliange, resulting in a SF - FF
| |
| mismatch, but the feed reg valve will open to increase FF.
| |
| ICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXP1,ANATIQN: Analyze the effect of the failure on the control system and recognize which
| |
| valve will be controlling at the power level given
| |
| Post Validation Revision
| |
| | |
| Harris hXC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 40
| |
| The plant is operating at 80% power with rod control in automatic and pressurizer
| |
| pressure at 2240 p i g .
| |
| After a rapid power reduction the plant is stabilized at 40% power, when the Reactor
| |
| Operator notes thc followinp conditions:
| |
| e Pressurizer pressure is 2275 psig and slowly decreasing.
| |
| e Pressurizer level is 45% and slowly decreasing.
| |
| a Both pressurizer spray valves indicate mid-position.
| |
| e All pressurizer backup heaters are de-energized.
| |
| These conditions are indicative of.. .
| |
| a. a normal plant response following an outsurge from the pressurizer.
| |
| h. a failure in the Pressurizer Pressure control circuitry, which opened the spray
| |
| valves.
| |
| c. a failure in the Pressurizer Level control circuitry?which failed to energize the
| |
| backup heaters.
| |
| d. a normal plant response following an insurge into the pressurizer.
| |
| ANSWER:
| |
| c. a failure in the Pressurizer Level control circuitry, which failed to energize the
| |
| hackup heaters.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 40 TIEWGROUP: 2!2
| |
| KA IMPORTANCE: RQ 3.1 SRO
| |
| 10CFR55 CONTENT: 41(b) 3!7 43(b)
| |
| KA: 011K6.04
| |
| Knowledgc of the effect of a loss or malfunction on the following will have on the PZR I,CS: Operation
| |
| of PZR level controllers
| |
| OBTECTNE: PZKIC-3.0-5
| |
| EXPLAIN how the system controls pressurizer level, including the input parameters and the components
| |
| that receive output signals
| |
| DEVELOPMENT KEFERENCES: SD-I 00.3
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGKIPICANTLY MODIFIED / DIRECT: PZRLC-R7 00 1
| |
| NAC EXAM HISTORY None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
| |
| 8. Plausible sirice the response is correct, with the exception of the pressurizer heaters not being
| |
| energized, fix an ontsurge from the pressurizer.
| |
| b. Plausible since a downpower shouid result in an insurge which would cause the spray valves to open,
| |
| but the heaters should also he energized.
| |
| d e. A rapid downpower transient will result in an insurge to the pressurizer. This should result in the
| |
| conditions noted, including a high pressurizer level causing the heaters to be energized even during a
| |
| high pressure condition causing the spray valves to be open. The heaters not being energirxd with
| |
| level more than 5% high is indicative o f a level control system failure.
| |
| d. Plausibic since the rcsponse is corrcct, with the exctlption o f :he pressurizer heaters not k i n g
| |
| energized, for an insurge to the pressurizer.
| |
| ICULTY ANALYSIS:
| |
| COMPREIIENSIVE / ANALYSIS KYOWLEDGE /RECALL
| |
| DIFFICULTYRATING: 3
| |
| E.XPLANATIQN: Analysis of the expected plant response and the actual plant response to an
| |
| insurge into the pressurizer
| |
| Post Validation Revision
| |
| | |
| IIanis NRC Written Examination
| |
| Reactor Operator
| |
| Given the following conditions:
| |
| e Feed flow lo the s t e m &eneratorsis being transferred from the Auxiliary Feedwater
| |
| (AFW) System to the Main Feedwater (MFW) System in accordance with OP-134-1,
| |
| Feedwater System.
| |
| e The Motor-Driven AFW Pumps are operating with all Flow Control Valves throttled
| |
| in mid-position.
| |
| e The Turbine Driven AFW Pump is in standby with all Flow Control Valves full open.
| |
| e MFW Pump A is operating with the Feed Reg Bypass Valves throttled slightly
| |
| open.
| |
| AI1 AFW and MFW Isolation Valves are open.
| |
| If a condition occurs which results in a valid AFW Isolation Signal, how will the
| |
| following AFW and MFW valves on SG B respond?
| |
| T D A W FLOW MFW ISOLATION MFW FEED KEG
| |
| CONTROL VALVE VALVE BYPASS VALVE
| |
| a. Remain Open Remain Open Rmain Open
| |
| b. Remain Open Close Close
| |
| C. Close Remain Opcn Remain Open
| |
| d. Close Close CIOSC
| |
| ANSWER:
| |
| d. Close Close Close
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 41 TIEWGROUP: 212
| |
| KALMPORTANCE: RO 4.2 SRO
| |
| 10CFR55 CONTENT: 41(b) 4 43(W
| |
| KA: 035KI.01
| |
| Knowledge of the physical connections andor cause-effect relationships bdween the S/GS and the
| |
| following systems: MFWIAFW systems
| |
| OBJECTIVE: AFS-3.0-Ii2
| |
| IESCRIBE the controls and interlocks of AFW System valves and controllers
| |
| DEVEI,OPMENT REFERENCES: OP-137
| |
| 33-137
| |
| SD-134
| |
| SI)-103
| |
| IPEFEIZENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW fl SIGNIFICANTLY MODIFIED
| |
| OR SIGNIRICAFiTLY MODIFIED / DIRECT:
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOW JUSTIPICACTION (CORRECT ANSWER +d):
| |
| a. Plausible since a AFW isolation signal closes the motor-operated isolation valves so these valves
| |
| could remain open, but they also close.
| |
| b. Plausible since the F W valves wili go closed, but the TDAFW Pump flow control valves also close.
| |
| e. Plausible since the TDAFW Pump flow control valves will go closed, but the FW valves slso close.
| |
| 4 d. AFW isolation occurs on a Main Steam Isolation Signal, which will also close the F W valves,
| |
| concurrent with a high steam line differential pressure signal.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWL.EDGEI RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Analysis of the cause of AFW isolation signal, its effect on AFW, as well as
| |
| other signals generated which cause FWI
| |
| Post Validation Revision
| |
| | |
| Hanis NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 42
| |
| Given the following conditions:
| |
| E A reactor trip with SI has occurred.
| |
| E The immediate action steps, ECCS flow verifications, and AFW flow verifications
| |
| are performed.
| |
| E SG levels are 10% and the required AF\V flow CANNOT be estabiished.
| |
| E FRP-H. 1, Response to Loss of Secondary Heat Sink, is entered.
| |
| m RCS pressure is checked and determined to he less than intact SG pressure.
| |
| Which of the folIowtrig describes the pImt conditions?
| |
| a. A large break LOCA is in progress AND a secondary heat sink is required
| |
| b. A large break LOCA is in progress AND a secondary hcdt sink is NOT required
| |
| c. A small break LOCA is in progress AND a secondary heat sink is required
| |
| d. A small break LOCA is in progress AND a secondary heat sink is NOT required
| |
| ANSWER:
| |
| b. A large break LOCA is in progress AND a secondary heat sink is NOT required
| |
| Post Validation Revisioii
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESFION NUMBER 42 TIEWGROUP: lil
| |
| KAIMPORTANCE: HO 3.9 SRO
| |
| 10CFR55 CONTENT: 41(b) 7 43m
| |
| K k WE05EK2.2
| |
| Knowledge of the interrelations between the (Loss of Secondaly Heat Sink) m d the following: Facilitys
| |
| heat removal systems, including primary coolant, eniergency coolant, the decay heat renioval systenrs,
| |
| and relations between the proper operation of these systems tu the operation of the facility
| |
| OBJECTIVE: EOP-3.11
| |
| Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
| |
| d. Requirements for a heat sink
| |
| DEVELOPMENT REFERENCES: FKP-I. 1
| |
| LP-EOP-3. I 1
| |
| REFERENCES SUPPLIED TQ APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.11 023
| |
| NRC EXAM HISTORY: Harris NRC 2000
| |
| DISTRACTOR JUST%F%CACTION (CORRECT ANSWER dd):
| |
| a. Plausible since a large break LOCA has occurred, but a secondary heat sink is riot required.
| |
| d b. With KCS pressure less than SG pressure a large hreak LOCA has occurred and adequate heat
| |
| removal will occur from SVbreak flow.
| |
| c. Plausible since a I D C A has occurred, but the LOCA is a large break and a secondary heat sink is not
| |
| required.
| |
| d. Plausible since a secondary heat sink is not required, but the LOCA is a large break.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Comparison of heat removal systems and plant conditions to determine
| |
| requirements
| |
| Post Validation Revision
| |
| | |
| Harris hXC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 43
| |
| Given the following conditions:
| |
| e The plant had been operating at 100% for three (3) weeks when a Reactor Trip
| |
| occurred.
| |
| o Six (8) hours following the trip, a reactor startup is planned.
| |
| Which one ofthe following is PROHIBITED at SHNFP as a result of industry wide
| |
| premature criticality events?
| |
| a. A startup rate in excess o f + 0.3 dpm
| |
| b. Delaying the startup until xenon begins to decay
| |
| c. Operators performing the EXSPACK estimated critical conditions (ECC)
| |
| d. A difference of 400 pcm hetween the POWERTRAX and EXSPACK ECCs
| |
| ANSMJEW:
| |
| d. A differenec of 400 pcm hetween the POWERTRAX and EXSPACK ECCs
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTIOX NUMBER: 43 TIERKKOUP: 3
| |
| KAIMPORTANCE: RO 3.7 SRO
| |
| 10CFK55 CONTENT: 41(b) 10 43W
| |
| Kn: 2.2.1
| |
| Ability to perform prc-startup procedures for the facility, including operating those controls associated
| |
| with plant equipment that could affect reactivity
| |
| OBJECTIVE: GP-3.4-6
| |
| SWMAKIZE at least three conditions which have contributed to premature criticality events within the
| |
| industry; also SUR04ARIZE actions taken at SHNPP to prevent similar occurrences
| |
| DEVELOPMENT REFERENCES: GP-004
| |
| REFEWNCXS SUPPLIED TO APPLICANT: None
| |
| QGESTION SOURCE: SIGNIFICANTLY MODIFIED 0 DIRECT
| |
| CANTLY MODIFIED / DIRECT: ( 3 - 3 . 4 01 1
| |
| NKC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| 8. Plausible since cxcessive startup rates can contribute to lack of reactivity control, but limitations are
| |
| placed on startup rate after criticality is achieved.
| |
| b. Plausible since xcnon decay will he adding positive reactivity to the core while the startup is being
| |
| perfomicd, but is accounted for in the time after trip in the ECC.
| |
| c. Plausible since SIMPP required any manual ECC calculations be perfomed by Reactor Engineering,
| |
| but EXSPACK is normally perfomed by Operations.
| |
| 4 d. The threshold for perfomring a reactor startup following a power history of >RO% equilibrium power
| |
| is 250 pcm difference between P O W E R T M and EXSPACK and 500 pcm for transient history and
| |
| steady state below 80%.
| |
| ICIJLTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSL9 KNOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge of the administrative requirements prior to criticality being
| |
| achieved
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 44
| |
| While reviewing the MCB annunciators prior to relieving the off-going shift, YOU note
| |
| that an annunciator has a RED bar attached to it.
| |
| This indicates that the annunciator is in alarm due to ...
| |
| a. the aimn being defeated.
| |
| b. the associated system being tested.
| |
| c. the alarm window itself being inoperable with a Work Request to repair it written.
| |
| $. the associated system being under clearance.
| |
| ANSWER:
| |
| d. the associated system being under clearance.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 44 TIEWGROUID: 3
| |
| KAIMPORTANCE: HO 3.3 sa0
| |
| iOCFR55 CONTENT: 41(b) 5 43(b)
| |
| KA: 2.4.31
| |
| Knowledge of annunciators alarms and indications, and use of the response instructions
| |
| OBJECTWE: PP-2.0-R3
| |
| DISCUSS the requirements in OMM-001/AP-002/hP-100 concerning the following:
| |
| k. MCI3 annunciators
| |
| DEVELOPMENT REFERENCES: OMM-00 1
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: 0N E W SIGXIFICANTLY MODIFIED c]DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MQDIFIED /DIRECT PP-3.0-R3 003
| |
| NRC EXAM HISTORY: None
| |
| DISTHACI'OR JUSTIFICACTKON (CORRECT ANSWER d'd):
| |
| a. Plausible since a color is used for coding this condition, but the abann being defeated is
| |
| indicated by black color coding.
| |
| b. Plausible since a color is used for coding this condition, but the associated system is being
| |
| tested is indicated by pink color coding.
| |
| E. Plausible since a color is used for coding this condition, the alarm window itself being
| |
| inoperable with a Work Request to repair it written is indicated by yellow color coding.
| |
| 4 d. Red color coding indicates that the associated system is under ctearance
| |
| DIFFICULTY ANALYSIS:
| |
| 0 COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFPICUI,TY RATING: 2
| |
| EXPLANATION: Knowledge of the color coding for alarm conditions
| |
| Post Validation Revision
| |
| | |
| IIarris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTIQN: 4.5
| |
| Given the following conditions:
| |
| e A Reactor Trip occurred fiom 100% power.
| |
| e The plant stabilized at 5.57 OF for several minutes.
| |
| e Shortly thereafter, a Safety Injection signal actuated.
| |
| Which of the following describes the effect of this sequence on the Main Feedwater
| |
| System?
| |
| a. e A f t a the Reactor Trip occurred, the SGs could be fed using the Feedwater Reg
| |
| Bypass Valves
| |
| e After the SI occurred, the SGs could be fed using the Feedwater Reg Bypass
| |
| Valves
| |
| b. e After the Reactor Trip occurred, the SGs could be fed using the Main
| |
| Feedwater Reg Valves or the Feedwater Reg Bypass Valves
| |
| After the SI occurred, Main Feedwater could NOT be used to feed the SGs
| |
| c. e After the Reactor Trip occurred. the SGs could he fed using the Feedwater Reg
| |
| Bypass Valves
| |
| e After the SI occurred, Main Feedwater could NOT be used to feed the SGs
| |
| d. e After the Reactor Trip occurrd, the SGs could he fed using the Main
| |
| Feedwater Reg Valves or the Feedwater Reg Bypass Valves
| |
| m After the SI occurred, the SGs could he fed using the Feedwater Reg Bypass
| |
| Valves
| |
| ANSWER:
| |
| c. e After the Reactor Trip occurred, the SGs could he fed using the Feedwater Reg
| |
| Bypass Valves
| |
| e After the SI occurred, Main Feedwater could NOT be used to feed the §Cis
| |
| Past Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 45 TIEWGHOUP: 2i I
| |
| KclIlMPORTANCE: RO 3.2 SRO
| |
| 1QCFR55CONTENT: 41(b) 4 43@)
| |
| KA: 059K4.19
| |
| Knowledge of MPW design feature($)and/or interlock(s) which provide for the following.:Automatic
| |
| feedwater isohtion of MFW
| |
| ORECCTIVE: AFW-3.0-A6
| |
| EXPLAIN the response of major CFW System valves to the following signalskonditions
| |
| hPain Peedwater Isolation Sigrxal (MFES)
| |
| Reactor trip (P-4) coincident with low Tavg(< 564OF)
| |
| DEVELOPMENT REFERENCES: SD-153
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QKESTION SOURCE: NEW' c]SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR s m w I c A l r l T r , Y MODIFIED / DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER +d):
| |
| a. Plausible since on a reactor trip with low Tave (564 "F), the SGs can still be fed with the bypass
| |
| valves, but on an SI or high-high SG level MEW can no longer supply the SGs.
| |
| b. Plausible since the SGs can no longer be fed using MFW on an SL but on a reactor trip only the
| |
| bypass valves can he used to feed the SGs.
| |
| 4 e. On a reactor trip with low Tave (564 OF), the SGs can still he fed with the bypass valves, but on an SI
| |
| or high-high SG level MFW can no longer supply the Scis.
| |
| d. 1'PausibIe since on a reactor trip with low 'l'ave (564 "F). the SGs can still be fed with the bypass
| |
| valves, hut not the main feed reg valves, and on an SI or high-high SG level MFW can no longer
| |
| supply the SGs.
| |
| DIPFICLrLI'Y AKALYSIS:
| |
| COiMPRHIENSIVE / ANALYSIS c]KNOWLEDGE / RECALL
| |
| DIFFICULTY KATING: 3
| |
| EXPLANATION: Comprehension that on a reactor trip where the plant stabilizes at no-load
| |
| temperature, the P-4 with Low 'rave signal allows feeding with the hypass and
| |
| SI isolates all MFW
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 46
| |
| Which of the following describes the design of Phase A and a Phase B Containment
| |
| Isolation signals?
| |
| a. a Phase A O m limits radioactive releases following a LOCA
| |
| a YRase B O x limits radioactive releases following a LOCA or secondary
| |
| system break inside Containment
| |
| h. a Phase A limits radioactive releases A X minimizes Containanent
| |
| overpressurization following a LOCA
| |
| a Phase 3 limits radioactive releases Amminimizes Contitinment
| |
| overpressurization following a LOCA or secondary system break inside
| |
| Containmcnt
| |
| c. a Phase A O m l i m i t s radioactive releases following a LOCA
| |
| a Phase B limits radioactive releases following a LOCA A x p r e v e n t s an
| |
| excessive RCS cooldown following a secondary system break inside
| |
| Containment
| |
| d. 8 Phase A limits radioactive releases Amminimizes Containment
| |
| overpressurization following a LOCA
| |
| a Phase J3 limits radioactive releases following a LOCA Amprevents an
| |
| excessive RCS cooidown following a secondary system break inside
| |
| Containment
| |
| ANSWER:
| |
| a. a Phase A ONI,Ylimits radioactive releases following a LOCA
| |
| a Phase B O x l i m i t s radioactive releases following a LOCA or secondary
| |
| system break inside Containment
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER. 46 TIENGROUP: 1/1
| |
| KAIMPORTANCE: RO 3.5 SRO
| |
| 10CFR55 CONTENT: 41(b) 9 43@)
| |
| Kk 00001 1EK3.00
| |
| Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: Actuation
| |
| of Phase A and B during LOCA initiation
| |
| OBJECTIVE: CIS-3.0-1
| |
| STATE the purpose ofthe Containment Isolation System
| |
| DEVELOPMENT REFERENCES: SI)-1 14
| |
| REFERENCES SZJPPLIEDTO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: CIS 006
| |
| CIS 009
| |
| NRC EXAM HISTORY None
| |
| DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER d'd):
| |
| 4 a. Phase A serves to limit the release of radioactive materials to atmosphere following a LOCA. Phase 13
| |
| acts to limit radioactive releases by actuating on a kOCA or a steam or fecdwater line break inside
| |
| containment.
| |
| b. PlausibIe since both Phase A and Phase B act to limit the release of radioactive materials to
| |
| atmosphere, but overpressurization is limited by spray actuation, main steam line isolation, and feed
| |
| water isolation.
| |
| c. Plausible since both Phase A and Phase B act to limit the release of radioactive materials to
| |
| atmosphere, but overpressurization and RCS cooldowns are limited by spray actuation, main steam
| |
| line isolation, and feed water isolation.
| |
| d. Plausible since both Phase A and Phase H act to limit the release of radioactive materials to
| |
| atmosphere, but ovepressurimtion and RCS cooldowns are limited by spray actuation, main steam
| |
| line isolation, and feed water isolation.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION: Knowledge ofpurpose of Phase A and Phase B signals
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examinaticn
| |
| Reactor Operator
| |
| QUESTION: 47
| |
| An entry into FRP-S.1, Response to Nuclear Power GeneratiodATWS, has been made
| |
| from PATH- 1. The following conditions currently exist:
| |
| e The reactor trip breakers are closed.
| |
| e Rods are being inserted manually.
| |
| e Control Bank D is at 12 steps.
| |
| e Power Range Instruments are all indicating 8%.
| |
| e Intermediate Range SUR is NEGATIVE
| |
| Which of the following conditions must be met in FRP-S.1 allow a return to PATH-I?
| |
| a. One of the reactor trip breakers must be opened
| |
| b. Both ofthe reactor trip breakers must be opened
| |
| c. Power Range indication must be reduced below 5%
| |
| d. Control Bank A must be inserted fully
| |
| ANSw ER :
| |
| c. Power Range indication must be reduced below 5%
| |
| Post Validation Revision
| |
| | |
| Ilarris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NCMBER: 47 TIEWGROUP 1/1
| |
| KAIMPORTANCE: RO 4.4 SRB
| |
| IOCFR55 GOhTENT: 4l(h) 7/10 43(b)
| |
| KA: 000029EA2.0 1
| |
| Ability to determine or interpret the following as they apply to a ATWS: Reactor nuclear instrumentation
| |
| OIBJECTWE: EOP-3.1-3
| |
| DEMONSTRATE the below-assumed operator knowledge from the SHhTP Step Deviation Documents
| |
| and WOG ERGS that support performance of EQP actions:
| |
| a. Verification of reactor trip
| |
| DEVELOPMENT REFERENCES: EOP-FW-S. 1
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QhiE.STION SOURCE: REW SIGNIFICANTLY MODIFiED DIRECT
| |
| BANK hWMBER FOR SIGNIFl[CANTI,Y MODIFIED f DIRECT: EOP-3.15-135002
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (COBRECT ANSWER .\l'd):
| |
| a. Plausible since this would cause the reactor to be tripped, but it is not required to be done to exit PRP-
| |
| s.1.
| |
| b. Plausible since this would cause the reactor to be tripped, but it is not required to be done to exit FRP-
| |
| s.1.
| |
| d c. Exiting FW-S.1 requires that PR N E be less than 5% and IR NIS startup rate be negative. Reactor
| |
| trip breaker position is not a condition for exiting the procedure, although actions are taken to open the
| |
| breakers.
| |
| d. Plausible since this would cause the reactor to be adequately shutdown, but it is not required to be
| |
| done to exit FRP-S.I.
| |
| D w F I c m x Y ANALYSIS:
| |
| 0 COMPREHENSIVE / ANALYSIS ICUOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowled&eof the procedural requirements to exit FRP-S. 1
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 48
| |
| Given the following conditions:
| |
| m A plant cooldown is being performed.
| |
| e All Steam Generators (S6s) are currently at approximately 50 psig.
| |
| Auxiliary Feed Water (AFW) Pump A-SA is being used to feed the SGs.
| |
| e The supply breaker on 120 VAC IDP-IA-SI for 1AF-i9, AUX FW MOTOR PMP
| |
| A-SA DISCHARGE \'&V, trips open.
| |
| Which of the following describes the effect of this loss of power on the operation of
| |
| AFW Pump A-SA?
| |
| 3. Operates at shutoffhead
| |
| b. Operates on minimum recirculation flow
| |
| c. Operates on maximum recirculation flow
| |
| d. Operates at runout conditions
| |
| ANSWER:
| |
| d. Operates at runout conditions
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Opentor
| |
| Data Sheets
| |
| QUESTION NUMBER: 48 TIEWGROUP: 2/ 1
| |
| MAIMPORTANCE: RO 2.5 SRQ
| |
| 10CFR55 CONTENT: 41(b) 4 4309
| |
| Kk 061K6.01
| |
| Knowledge of the effect of a loss or malfunction of the following will have on the AFW con~ponents:
| |
| Controliers and positioners
| |
| OBJECTIVE: AFS-3.OR5
| |
| DFXKIBE how the AFW system is impacted by a loss of l2Ovac unintermptible power supplies (SI, S I ,
| |
| SIII, SIV)
| |
| DEVELOPMENT REFERENCES: SD-137
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOUIPCE: NEW SIGNIFICANTLY k1ODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AFS-A3 001
| |
| AFS-A3 007
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR m w I F I c A 6 : T I o N (CORRECT ANSWER +@:
| |
| a. Plausible since p o w a is lost to the discharge valve, but the vr~lvcfails open causing flow to increase.
| |
| b. Plausible since power is lost to the discharge valve, but the valve fails open causing flow to increase.
| |
| e. Plausible since the valve fails open and flow increases, but the pump does not mn on recirculation
| |
| flow.
| |
| ./ (8. The loss of power causes AFW Pump A-SA to reach rnnout conditions due to IAF-19 failing open
| |
| and having the SGs at such a low prcssure.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSRE / ANALYSIS 0 KNOWLEDGE /RECALL
| |
| DIr,-FICUI,TY RATING 3
| |
| EXPIANATION: Analysis of the effect o f a failure of the PCV after determining the fail position
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 49
| |
| Given the following conditions:
| |
| The plant is in Mode 5.
| |
| a ALB-008-1-4, RWMU STORAGE TANK MINIMUhUHIGH LEVEL, a l m s .
| |
| m WWMU tank level is decreasing with NO VCT makeup in progress.
| |
| Which one of the following procedures would be the most appropriate to implement?
| |
| a. AOP-003, Maifunction of Reactor Makeup Control
| |
| b. AOP-08, Accidental Reiease of Liquid Waste
| |
| c. AOP-016, Excessive Primary Plant Leakage
| |
| d. AOP-020, Loss of Reactor Coolant Inventory / RHR While Shutdown
| |
| ANSWER:
| |
| b. AOP-008, Accidental Release of Liquid Waste
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 49 TIEWGROUP: 1/2
| |
| KAIMPBRTANCE: RO 4.0 SBO
| |
| 1QCFR55CONTENT: 41(b) 10/13 43(b)
| |
| KA: 00005962.4.4
| |
| Ability to recognize abnormal indications for system operating parameters which are entry-level
| |
| conditions for emergency and abnormal operating procedures. (Accidental Liquid Radwaste Release)
| |
| OBJECTWE: AOP-3.8
| |
| 1I)ENTIFY symptoms that require entry into AOP-008, Accidental Release of Liquid Waste
| |
| DEVELOPMENT REFERENCES: AOP-008 (unknown)
| |
| QUESTION SOURCE:
| |
| -
| |
| REFERENCES SUPPLIED TO APPLICANT
| |
| u NEW u
| |
| I--.I
| |
| None
| |
| SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SPGNIFI'ICANT1,YMODIFIED i DIRECT: AOP-3.8 001
| |
| NRC EXAM HISTORY: Hams NKC 2000
| |
| DISTRACTOR SUSTIFICACTION (CORRECT ANSWE.R d'd):
| |
| 8 . Plausible since KMUW tank supplies makeup to VCT, but AOP-003 addresses conditions regarding
| |
| valve / transmitter failures, not loss oftank source.
| |
| d b. Entry conditions have been met for AOP-008.
| |
| e. Plausible since F&lIJW tank supplies makeup to RCS and candidate may imply that loss of supply
| |
| results in a loss ofprimary inventow, but conditions are met for entry into AOP-008.
| |
| d. Plausible since RMUW tank supplies makeup to RCS and candidate may imply that loss of supply
| |
| results in a loss of primary inventory with plant shutdown, but conditions are met for entry into AQP-
| |
| 008.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE i ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION: Knowledge of entry requirements ftx accidental liquid release
| |
| Post Validation Revision
| |
| | |
| Hamis NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 50
| |
| Which of the following actions would be most effective in responding to a Pressurized
| |
| Thermal Shock condition in accordance with EOP-FRP-P. 1, Response to Pressurized
| |
| Thennal Shock?
| |
| a. Close the block valve for any open PRZ PORV
| |
| b. Start a RCP once SI has been terminated
| |
| c. Direct an operator to locally isolate any stuck open SG PORV
| |
| d. Direct an operator to locally open any failed closed BIT outlet valve
| |
| ANSWER:
| |
| c. Direct an operator to locally isolate any stuck open SG PQRV
| |
| Post Validation Revision
| |
| | |
| Harris NR(: Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 50 TIEFUGROUP I12
| |
| KA MPORTANCE: RO 2.9 SRQ
| |
| 10CFR55 CONTENT: 4l@) 10 43W
| |
| #A: WE08G2.1.30
| |
| Ability to locate and operate components, including locai controls. (Pressurized Thermal Shock)
| |
| OBJECTWE: EOP-3.14-1
| |
| DESCRIBE the purpose ofthe following EOPs including the type of event for which they were designed
| |
| and the major actions performed
| |
| a FRP-P. 1, Response to Imminent Pressurized Thermal Shock
| |
| - -
| |
| DEVELOPMENT REFERENCES: EOP-FRP-P.I
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED u DJHECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC E?XV HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION ( C Q W C T ANSWER d'd):
| |
| a. Plausible since closing the block valve for a stuck open PW, PORV is an action taken in FW-S.1,
| |
| though it is performed to maintain RCS inventory and will cause pressure to increase which would
| |
| cause the severity of a PTS event to worsen.
| |
| b. Plausible since an RCP is started in VRP-S.1 to cause mixing of any SI water with the RCS, but only if
| |
| SI cannot be terminated.
| |
| d c. A stuck open SG PORV would contribute to the cooldown associated with a PTS event. Locally
| |
| isolating the SG PORV would stop any cooldown caused by the SG PORV.
| |
| d. Plausible since the BIT out6et valves are supposed to bz opened during an SI condition, but the I3IT is
| |
| isolated in FRP-S.1 and local action is taken to close the vaive in the event it cannot be closed
| |
| remotely.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Analysis of plant conditions during a PTS event to determine the most
| |
| appropriate course of action
| |
| Post Validation Revision
| |
| | |
| Hams NFK Written Examination
| |
| Reactor Operator
| |
| QUESTION: 51
| |
| Given the following conditions:
| |
| e An operator has been sent to rack out a 480 VAC breaker by the SCO.
| |
| e Inadvertently, the incorrect cubicle is opened and the control power fuses are
| |
| removed from the wrong breaker.
| |
| Which of the following describes how the breaker is affected by the removal of the
| |
| control power fuses?
| |
| a. All Main Control Board indications will he lost for the breaker and if the breaker
| |
| is closed, it will trip and CANNOT be closed until control power is restored
| |
| b. All Main Control Board indications will be lost for the breaker and if the breaker
| |
| is open>it can only be closed mechanically locally
| |
| c. Main Control Board indication will still be available for the breaker, but ifthe
| |
| breaker is closed, it will trip and CANNOT be closed until control power is
| |
| restored
| |
| d. Main Control Board indication will still be available for the breaker, but if the
| |
| breaker is open, it can only be closed mechanically locally
| |
| ANSWER:
| |
| h. All Main Control Board indications will be lost for the breaker ,and ifthe breaker
| |
| is open, it can only be closed rncchanically lociilly
| |
| Post Validation Revision
| |
| | |
| Hairis NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NIMBER: 5 1 TIEWGROIOP: 2; I
| |
| KAIMPQRTANCE: RO 2.6 SRO
| |
| IOCFR55 CONTENT: 41(b) 7 43(W
| |
| KA: 062A4.04
| |
| Ability to manually operate and/or monitor in the controi room: 1,ocal operation of breakers
| |
| OBJECTIVE: 480V-3.O-RI
| |
| State the function of breaker control power and discuss the effects o f a loss of breaker control power
| |
| DEVELOPMENT REFERENCES: OP-156.02
| |
| 48OV-LP-3.0
| |
| REFERENCES SUPP1,IED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
| |
| RANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 48OV-RI 001
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
| |
| a. Plausible since MCB indication will be lost, but the breaker will not trip open on the loss of control
| |
| power.
| |
| 4 h. A loss of control power will cause MCB indication to go out and cause the remote operation ofthe
| |
| breaker to be defeated.
| |
| e. Plausible since a loss of control power to causes a loss of the ability to operate the breaker, but the
| |
| breaker will not trip and MCB indication will he lost.
| |
| d. Plausible since the breaker can only be operated locally, but the loss of control power will result in a
| |
| loss of MCB indication.
| |
| DIFFICWLTY ANALYSIS:
| |
| 0 COMPREHENSIVE / ANALYSES KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowlcdge ofthe effect of a 108s of control power to a 48OV breaker
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 52
| |
| Which ofthe following situations would likely result in an inadvertent dilution event
| |
| during Mode I operation and, after the crew has adjusted core reactivity to compensate
| |
| for the change in boron concentration, which procedure would be used to address the
| |
| cause of the event?
| |
| a. = RCP thermal barrier heat exchanger leak
| |
| = AOP-OI 6, Excessive Primary Plant Leakage
| |
| b. o The boric acid pump trips during an automatic makeup
| |
| AOP-004, Malfunction of Reactor Makeup
| |
| e. o A mixed bed demineralizer that was last in service three weeks ago is
| |
| mistakenly placed in service at the end-of-cycle
| |
| = AOP-033, Cheniistry Out of Tolerance
| |
| d. o A tube leak in the Seal Water heat exchanger
| |
| m AOP-014, Loss of Component Cooling Water
| |
| ANSWER:
| |
| d. = A tube leak in the Seal Water heat exchanger
| |
| 0 AOP-014, Loss of Component Cooling Water
| |
| Post Validation Kevision
| |
| | |
| Harris NRC: Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NGMBER 52 TIEWGROUP 216
| |
| KAIMPORTANCE: RO 4.2 SRO
| |
| 10CFR55 CONTENT: 41(b) 6/10 43(b)
| |
| KA: 004A2.06
| |
| Ability to ( a ) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based
| |
| on those predictions, use procedures to correct, control, or mitigate the consequences of those
| |
| malfunctions or operations: Inadvertent boratioddilution
| |
| OBJECTIVE: E-3.12-3
| |
| Identify systems whose operation may alter RCS boron concentration and discuss how operation of these
| |
| systems may affect boron concentration
| |
| DE\EI;OPMEWT REFERENCES: SOER 94-2
| |
| AOP-0 14
| |
| AOP-14-BB)
| |
| REFERENCES SUPPLIED TO APPLPCANT: None
| |
| QUESTION SOURCE: SIGNIFICANTLY MODIFIED DIRECT
| |
| CANTLY MODIFIED / DIRECT: IE-3.12-R3001
| |
| NRC E.XAI\I HISTORY: None
| |
| DISTUCTOR JUSTIFICACTION (CORRECT ANSWER $d):
| |
| a. Plausible since the thermal barrier interfaces with a non-borated system (CCW), but leahge would be
| |
| out of the RCS to CCW and would not affect RCS boron concentration.
| |
| b. Plausible since boric acid is required for the proper blended flow, but an automatic makeup would be
| |
| terminated automatically in the event of a boric acid pump trip.
| |
| c. Piausible since boron concentration will change in CVCS, but this would result in an inadvertent
| |
| bordtion rather than a dilution.
| |
| d d. A seal water HX leak will result in CVCS being diluted by CCW. This failure is to be addressed by
| |
| AOP-014.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREIIENSIVE / ANALYSIS 0 KNOW1,EDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXP1,ANATION: Analyze the effect of each failure on KCS boron concentration and determine
| |
| the required procedure to address the failure
| |
| Post Validation Revision
| |
| | |
| IIturis NRC Written Exaniination
| |
| Reactor Operator
| |
| QUESTION: 53
| |
| While establishing a bubble in the PRZ per GP-002, Normal Plant Heatup From Cold
| |
| Solid to Hot Subcritical MODE 5 to MODE 3, letdown pressure control valve ICs-38
| |
| (PK-145.1), Low Pressure Letdown Pressure Controller, opens.
| |
| Which of the following describes why PK-145.1 opens?
| |
| a. Thermal expansion of liquid in the pressurizer
| |
| b. Change in CCW heat load
| |
| c. Spray valves are shut while drawing a bubble
| |
| d. Switchover of letdown to orifices from RHR-CVCS cross-connect
| |
| ANSWER:
| |
| a. Thermal expansion of liquid in the pressurizer
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Exaniiriation
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 53 TIEWGROUQ: 2: 1
| |
| KAIMPORTANCE: HO 2.9 SRO
| |
| 10CFR55 CONTENT: 41(b) 7 43W
| |
| KA: 010K1.06
| |
| Knowiedge of the physical connections andor cause-effect relationships between the PZR PCS and the
| |
| following systems: CVCS
| |
| OBJECTIVE: CiP-3.2-2
| |
| DISCUSS drawing a bubble in the pressurizer, including
| |
| b. The parameters used to determine when the bubble has bee11 drawn
| |
| DEVELOPMENT REFERENCES: GP-002
| |
| 1.P-(3-3.2
| |
| REFERENCES SUPPLIED TO APPLICANT
| |
| ~~ None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED mmcr
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EWQ 1081 1
| |
| NHC EXAM HISTORY: Hams NWC 2002
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
| |
| d a. Thermal expansion of the liquid doe to the heaters being energized results in a pressure increase in the
| |
| RCS. PK-145.l opens to maintain letdown pressure, resulting in increased letdown flow.
| |
| b. Pballsible since the letdown heat exchanger is cooled by CCW, but temperature has little effect on the
| |
| response ofPK-145.1.
| |
| e. lkdusible since the spray valves are shut while a bubble is being drawn, but PK-145.1opens to
| |
| maintain letdown pressure, not RCS prcssure.
| |
| d. Plausible since KHR letdown may be placed in service at low temperature and pressure conditions, but
| |
| is not in service while drawing a bubble.
| |
| DIFFICCJLTY ANALYSIS:
| |
| ~~
| |
| COI\IPHEIIENSNE / AKALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY MATING: 3
| |
| EXPLANAFION: Comprehension ofthe eEects of drawing a bubble on CVCS components
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 54
| |
| 125 VDC battery 1A-SA is rated for 1170 amp-hours at a 4-hour discharge rate.
| |
| I f DC load shedding is performed such that the loading on the battery is reduced from
| |
| 292 amps to 146 amps, how long should the battery be available to supply the remaining
| |
| loads?
| |
| b. More than 4 hours, hut less than 8 hours
| |
| c. 8 hours
| |
| d. More than 8 hours
| |
| ANSWER:
| |
| d. More than 8 hours
| |
| Post Validation Revision
| |
| | |
| Hamis NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 54 TIEWGROUP: 2/1
| |
| KAIMPORTANCE: RO 2.5 SRO
| |
| lOCFR§§ CONrENT: 41(b) 8 43w
| |
| Kik 063A1.01
| |
| Ability to predict and/or monitor changes in parameters associated with operating the 1 X electrical
| |
| systeni controls including: Battery capacity as it is affected by discharge rate
| |
| OBJECTIVE: DCP-3.0-A3
| |
| STATE the fiinction and EXPLAIN the basic operation of the following major components of the DC
| |
| Power System:
| |
| e Batteries
| |
| DEVELOPMENT REFERENCES: EOP-EPP-00 1
| |
| ADEL-LP-2.6
| |
| DCP-LP-3.0'
| |
| MEFERENCES SUPPLIED TO AF'PLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| BHSTHACTOR JGSTIFICACTION (CORRECT ANSWER d'd):
| |
| 8. Plausible since the battery is rated for 4 hours, but at a discharge rate of approximately 293 amps per
| |
| hour and decreasing the discharge rate would increase the capacity.
| |
| h. Plausihle since the discharge rate has been decreased which would extend the capacity of the battery
| |
| for a period of time, but the time would be more than doubled.
| |
| C. Plausible since the discharge rate has been halved, so it would appear that the capacity would be
| |
| doubled, but it is a n o n - h e x relationship.
| |
| d d. Reducing the discharge rate on a battery increases the battery capacity in a non-linear function such
| |
| that decreasing the discharge rate by half, increases the capacity by more than double.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KVOWLEDGE /RECALL
| |
| m F m x z r y RATING: 4
| |
| EXPLANATION: Calculation of the nominal discharge rate of a battery and comprehension of the
| |
| effect of reducing discharge rate on battery capacity
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 55
| |
| Given the following conditions:
| |
| 0 The plant has experienced a -rrge Break Loss of Coolant Accident during a reactor
| |
| startup.
| |
| All equipment functioned as designed and the crew has reached the point in PATH-1
| |
| where monitoring Critical Safety Function Status Trees is required.
| |
| Which one of the following statements describes the IMMEDIATE result that voiding in
| |
| the downcomber region would have on the Source Range instrumentation and procedure
| |
| used to mitigate these plant conditions?
| |
| a. 0 The displacement of downcomber water would increasc the neutron leakage
| |
| and result in a higher source range counl rate.
| |
| m The crew should continue in PATH-1 rather than transition to EOP-FRP-S.2,
| |
| Resporise to Loss of Core Shutdown.
| |
| b. o A decrease in downcomber water density would reduce fission and result in a
| |
| lower source range count rate.
| |
| 0 The crew should transition to EOP-FRP-S.2, Response to Loss of Core
| |
| Shutdown, rather than continue in PATH- I .
| |
| c. 0 The displacement of boron from the downcomber region would increase
| |
| fission and result in a higher source range count rate.
| |
| 0 The crew should continue in PATI-1-1 rather than transition to EOP-FRP-S.2,
| |
| Response to Loss of Core Shutdown.
| |
| d. 0 A decrease in downcomber water density would reduce fission and result in a
| |
| lower source range count rate.
| |
| 0 The crew should continue in PATH-1 rather than transition to EOP-FRP-S.2,
| |
| Response to Loss of Core Shutdown.
| |
| ANSWER:
| |
| a. The displacement of water would increase the neutron leakage and result in a
| |
| higher sonrce range count rate.
| |
| o The crew should continue in PATH-1 rather than transition to EOP-FRP-S.2,
| |
| Response to Loss of Core Shutdown.
| |
| Post Validation Revision
| |
| | |
| Hams NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NCJMBER: 55 TIEWGROUP 2l2
| |
| &%IMPORTANCE: RO 3.3 SRO
| |
| ICCFRSS CONTENT: 41(h) 2 43(b)
| |
| MA: 015A2.05
| |
| Ability to (a) predict the impacts of the following malfunctions or operations on the NE; and (b based on
| |
| those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions
| |
| or operations: Core void formation
| |
| OBJECTIVE: BD-3.10-7
| |
| Explain the NIS response to different void fractions in the core and downcomer region
| |
| DEVELOPMENT REFERENCES: IIO-BD-3.10
| |
| REFERENCES SUPPLIED TO APPLICANR None
| |
| QUESTION SOURCE: 17 NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGXWICANTLY MODIFIED / DIRECT: HNPO 20608
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOH JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| 4 a. Downcornber voiding results in higher source range indication due to increased leakage. The crew
| |
| should continue in PATH-I rather than transfer to FRP-S.2 since entry conditions to F a - S . 2 are a
| |
| Yellow path condition.
| |
| b. Plausible since a severe decrease in core water density would result in less moderation and a lower
| |
| power Icvel, but downcornher density has little effect on core reactivity.
| |
| E. Plausible since displacing core boron would result in a higher power level, but downcomber density
| |
| has little effect on core reactivity.
| |
| d. Plausible since a severe decrease in core water density would result in less moderation and a lower
| |
| power level, but downcomber density has little effect on core reactivity.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE i RECALL
| |
| DIFFICULTY MATING: 3
| |
| EXPLANATION: Analysis ofthe effects of core voiding on SK indication and knowledge of the
| |
| procedure hierarchy during the pcrfonnance of the EOPs
| |
| Post Validation Revision
| |
| | |
| Hamis NRC Written Examination
| |
| Reactor Operdtor
| |
| QUESTION: 56
| |
| Given the following conditions:
| |
| e A transition has just been made to FRP-S. 1, Response to Nuclear Power Generation
| |
| /ATWS, from PATH-I.
| |
| e The Reactor Operator is manually inserting control rods.
| |
| e Ail Turbine Throttle Valve (TV) and Turbine Governor Valve (GV) indications show
| |
| the RED light OFF and the GREEK light ON, with the exception of TV-3 and (37-2
| |
| which have both the RED fight and GREEN light ON.
| |
| o Turbine speed is decreasing, and is currently 1680 rpm.
| |
| e The Main Steam Isolation Valve (MSIV) Bypass valves are closed.
| |
| Which of the following actions should be taken next?
| |
| a. Verify all AFW pumps running
| |
| b. Manually trip the Turbine from the MCB
| |
| c. Pface both Turbine DEH pumps in PULL-TO-LOCK
| |
| d. Shut all MSiVs
| |
| ANSWER
| |
| b. Manually trip the Turbine from the MCR
| |
| Post Validation Revision
| |
| | |
| Hams NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NBIILIBER: 56 TIER/GROt.JP 22
| |
| KAIMPORTZTANCE: RO 2.8 SHO
| |
| 10CFR55 CONTENT: 4l(b) 10 4309
| |
| KA: 045A4.06
| |
| Ability to manually operate andor monitor in the control room: Turbine stop valves
| |
| OBJECTIVE: FOP-3.15-4
| |
| Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
| |
| e Order of preference for turbine trip steps from the MCR
| |
| DEVELOPMENT REFERENCES: EOP-PKP-S. I
| |
| REFERENCES SKJFPLKED TO APPLICANT None
| |
| QCESTION SOURCE: NEW SIGNHPICANTLY MODIFIED DIRECT
| |
| BANK hWMBER FOR SIGNWICANTLY MODIFIED / DIRECT EOP-3.15-RZ 001
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR SLTSTIFPCACTION (CORRECT ANSWER dd):
| |
| a. Plausible since GV-2 and TV-3 are associated with opposite steam chests and it may be assumed that
| |
| as long as the GVs are closed for 1 steam chest a i d the TVs are closed for the other steam chest with
| |
| turbine speed decreasing, and starting AFW is the next step in the procedure, however the turbine
| |
| should not be considered to be tripped.
| |
| 4 b. Verification of a turhine trip requires either all 4 TVs be closed or all 4 GVs be closed. If one set of
| |
| these valves are not all closed, then the RNO directs manually tripping the turbine from the MCB.
| |
| c. Plausible since the turhine should not be considered to he tripped based on indications, and this is an
| |
| RNO action, bnt should not be performed until a manual trip from the MCB is attempted.
| |
| d. Plausible since the turbine should not be considered to he tripped based on indications, and this is an
| |
| RNO action, but should not be perfnnned until a manual trip from the MCB is attempted.
| |
| DIFFICULTY ANALYSIS:
| |
| 0 COMPREHENSIVE / AP4ALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge of the required indications for a turbine trip and the priority for
| |
| tripping the turbine if a trip cannot be verified
| |
| Post Validation Revision
| |
| | |
| IIarris NTC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 57
| |
| Given the following conditions:
| |
| e The Main Control Room has been evacuated and control transferred to the Auxiliary
| |
| Control Panel (ACP).
| |
| AOP-004, Remote Shutdown, is being perfonned when a loss of offsite power
| |
| coincident with a Safety Injection signal occur.
| |
| Which of the following describes the response of the plant?
| |
| a. The Emergency Diesel Generators automatically start and the sequencers load the
| |
| ED@ due to the undervoltage signal
| |
| b. The Emergency Diesel Generators automatically start and the sequencers load the
| |
| EDGs due to the safety injection signal
| |
| e. The Emergency Diesel Generators automatically start, but must he manually
| |
| loaded with the required loads
| |
| d. The Emergency Diesel Generators must be manually started and manually loaded
| |
| with the required loads
| |
| ANSWER:
| |
| a. The Emergency Diesel Generators automatically start and the sequencers lndd the
| |
| EDGs due to the ~ n d e r v o h g esignal
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 57 TIENGROUP 211
| |
| KAIMPORT.4NCE: RO 3.6 SRO
| |
| 10CFR55 CONTENT: 41(b) 8 Wb)
| |
| KA: 064A3.07
| |
| Ability to monitor automatic operation of the EDiG system, including: Load sequencing
| |
| OBIECTTVE: AOP-3.4-R5
| |
| DISCUSS how a transfer to the auxiliary control panel would affect the following inputs to the ESF
| |
| sequencers
| |
| * Safety injection signal
| |
| Safety bus undervoltage signal
| |
| DEVEIBPMENT REFERENCES: AOP-004
| |
| AOP-004BD
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY I\.ZQDIIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: AOP-3.4-R6 001
| |
| NRC EXAM HISTORY: None
| |
| DISTEPnC'I'QR JUSTIFICACTION (CORREC'I' Ah'SWER d'd):
| |
| 4 a. The EDGs should automatically start OR the UV condition and the UV signal will still cause the
| |
| sequencer to operate. Only the SIAS input to the sequencer is defeated upon transfer to the ACP.
| |
| b. Plausible since the EDG will automatically start, hut loading will be based upon the Ut' signal.
| |
| c. Plausible since the EDG will automatically start, but loading will he based upon the LW signal.
| |
| d. l'lausihle since many automatic functions are defeated when control is transferred to the ACP. but the
| |
| EDG will automatically start and loading will be based upon the UV signal.
| |
| DIFFICULTY ANALYSIS:
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATIQN: Analysis ofthe effect o f a transfer to the ACP on the EDG and sequencer
| |
| operation
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 58
| |
| Given the following conditions:
| |
| An I&C technician reports that both of the Control Room Normal Outside Air Intake
| |
| Isolation radiation monitors bave failed detectors.
| |
| e It will take somewhere between four (4) and eight (8) hours to replace the detectors.
| |
| m i c h of the following states the action which must be taken within one (1) hour, in
| |
| accordance with Technical Specification 3.3.3. l ?
| |
| a. Establish operation of the Control Room Emergency Filtration System in the
| |
| Recirculation Mode of Operation
| |
| b. Initiate the preplanncd altcmate method of radiation monitoring
| |
| C. Return the monitors to service, or be in Hot Standby within the next six ( 6 )hours
| |
| d. Pcrforni a surveillance test on the Control Room Emergency Filtration System, or
| |
| be in Hot Standby within the next six ( 6 )hours
| |
| ANSWER
| |
| a. Establish operation of the Control Room Emergency Filtration System in the
| |
| KecircuIation Mode of Operation
| |
| Post Validation Revision
| |
| | |
| Harris NRC W-&en Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 58 TIEWGROUP: 3
| |
| KAIMPORTAIVCE: RO 2.6 SRO
| |
| 1 0 ~ ~ COXTENT:
| |
| ~ 5 5 4i(b) ioiii 43(b)
| |
| KA: 2.2.24
| |
| Ability to analyze the affect ofmaintenance activities on LCO status
| |
| OBJECTIVE: KMS-11
| |
| DEMONSTRATE knowledge of the Technical Specifications associated with the Radiation Monitoring
| |
| System:
| |
| a. RECOGNIZE the LCO limits associated with action statements of one hour or less
| |
| DEVELOPMENI REFERENCES: TS Table 3.3-6
| |
| REFERENCES SUPPLIED TO APPLICAIVT: None
| |
| QUESTION SOURCE: NEW c]SIGNIFICANTIiY MOI9IFIED DIRECT
| |
| BANK NUMBER FOH SIGNIFICANTLY MODIFIED / DIRECT: INIO I R442
| |
| NRC EXAM HISTORY: Hanis NRC 2002
| |
| 1)KSTRACTOR JlJSTIFICACTION (CORRECT ANSWER dd):
| |
| 4 a. With no outside air intakes available, maintain operation of the Control Room Emergency Filtration
| |
| System in the RecircuIaticn Mode of Operation.
| |
| b. Plausible since this would permit monitoring of the Control Room environment, but is not directed
| |
| c. Plausible since this is a typical action requirement in Tech Specs for inoperable equipment, but does
| |
| not apply to this condition.
| |
| d. Plausible since this is a typical action requirement time limit in Tech Specs for inoperable equipment,
| |
| but does not apply to this condition.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 4
| |
| EXPLANATION: Knowledge of 1 hour actions required by Technical Specifications
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 59
| |
| Given the following conditions:
| |
| 8 A LOCA has occurred inside Containment.
| |
| 8 Containment pressure is 5.5 psig.
| |
| 8 RCS Wide Range Pressure indications are:
| |
| (BLACK BEZELED INSTRUMENTS)
| |
| PI440 = 1060 psig
| |
| PI-441 = 1040 p ~ i g
| |
| (YELLOW BEZELED INSTRUMENTS)
| |
| PI-402 = 980 psig
| |
| PI-403 = 980 psig
| |
| PI-402A 700 psig
| |
| RCS pressure should be reported as ..,
| |
| a. 700 psig.
| |
| b. 980psig.
| |
| C. 1040 psig.
| |
| d. 10GOpsig.
| |
| ANSWER:
| |
| b. 980psig.
| |
| Post Validation Revision
| |
| | |
| Ilarris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 59 TIEWGROUP 3
| |
| KAIMPORTANCE: RO 3.5 SRO
| |
| 10CFR55 CONTENT: 41(b) 6 43W
| |
| KA: 2.4.3
| |
| Ability to identi@ post-accident instrumentation
| |
| OWECTn7E: EOP-3.19
| |
| DESCRIBE Control Room usage of EPPs, foldouts, and FWs as it relates to the following:
| |
| g. Use of RCS wide-range pressure indication
| |
| DE\EI,BPBIENT REFERENCES: EOP Users Guide
| |
| IlEFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICAhTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.19 034
| |
| NRC EXAM HISTORY: Harris NRC 2005
| |
| DISTRACTOR JIIJSTIFICACTION(CORRECT ANSWER dd):
| |
| a. Plausible since yellow bezeled instniments are qualified for post-accident monitoring. The lowest
| |
| qualitid instrunlent following an accident should be used unless the narrow range instrument PI-
| |
| 402A is on scale with KCS pressure helow 700 psig.
| |
| 4 h. Yellow bezeled instruments are qualified for post-accident monitoring. The lowest qualified
| |
| instrument following an accident should be used unless the narrow range instrument PI-402A is on
| |
| scale with RCS pressure below 700 p i g .
| |
| c. Plausible since this is the lowest black bezeled instrument, but yellow bezeled instruments should be
| |
| used due to post-accident conditions.
| |
| d. Plausible since this is the highest black bezeled instrument, but yellow bezeied instruments should be
| |
| used due to post-accident conditions.
| |
| DIFFICCZTY tAL\SIs:
| |
| COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE / RECALL
| |
| DIFFICULTY HATING: 3
| |
| EXPLANATION: Analysis of plant conditions to determine required actions during adverse
| |
| containment
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 60
| |
| Which of the following is the MINIMUM required operable components to meet
| |
| Technical Specification 3.8.2.1, DC Sources - Operating, Limiting Condition for
| |
| Operation?
| |
| a. 0 IA-SA 125-V emergency batterybank
| |
| 0 1B-SB 125-V emcrgency battery bank
| |
| b. 0 1A-SA 125-V emergency battery bank
| |
| 1B-SB 125-V emcrgency batterybank
| |
| 1A-SA 125-V full capacity charger
| |
| c. IA-SA 125-V emergency batterybank
| |
| 1R-SB 125-V emergency battery bank
| |
| e 1A-SB 125-V full capacity charger
| |
| e 1B-SA L25-V full capacity charger
| |
| d. c IA-SA 125-V emergency battery bank
| |
| 1B-SB 125-V emergency battery bank
| |
| IA-SA 125-V full capacity charger
| |
| * 1A-SB 125-V full capacity charger
| |
| B IB-SA 125-V full capacity charger
| |
| a IB-SB 125-V full capacity charger
| |
| ANSWER:
| |
| c. e iA-SA 125-V emergency battery bank
| |
| e i B-SB 125-V emergency battcry bank
| |
| 1A-SB 125-V full capacity charger
| |
| 0 1B-SA 125-V full capacity charger
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 60 TIEWGROUP: 211
| |
| IC4 IMPORTANCE: RO 3.2 SRO
| |
| IOCFR55 CONTENT: 41(b) 10 43(W
| |
| KA: 00005862.1.33
| |
| Ability to recognize indications for system operating parameters which are entry-level conditions for
| |
| tcciuiical specifications. (Loss of DC Power)
| |
| OBJECTIVE: DCP-3.0-RI
| |
| Given the name of a component in the DC power system, state whether or not that component is
| |
| Technical Specification related
| |
| DEVE1,OPMENT REFERENCES: TS 3.8.2.1
| |
| REPERENCFS SUPPLIED TO APPLICANT: None
| |
| QUESTIQN SOURCE: 0 NE.W SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NLJIRERFOR s ~ ~ x w ~ C m r ~ Y / DIRECT:
| |
| MODIFIED DC-KI no1
| |
| NRC EXAM HISTORY: None
| |
| DISTRAC'FOR JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| a. Plausible since both battery banks are listed, but each train also requires a full capacity charger.
| |
| b. Plausible since both battery hanks and a single full capacity charger are listed, hut each train also
| |
| requires its own full capacity charger.
| |
| ./ E. Minimum requirements are 125 VDC Emergency Battery Bank IA-SA and either fir11 capacity
| |
| charger IA-SA or IH-SA AND 125 VDC Emergency Battery Bank ID-SB and either full capacity
| |
| charger IA-SB or 1U-SB.
| |
| d. Plausible since this would meet the requirements of the 'IS,but this includes all components and is not
| |
| the minimum required.
| |
| ICULTY ANALYSIS:
| |
| COMPREHESSIVE i mALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPL.NATION: Knowledge ofthe TS requirements for Dc power
| |
| Past Validation Revision
| |
| | |
| Harris MKC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 61
| |
| Given the following conditions:
| |
| e The plant is operating at 100% power when ALB-OlO-l-lB, RCP A UPPER OIL
| |
| RSVR LOW-LEVEL, alarm is received.
| |
| e The operator checks the computer points for GD AOP-018 and finds RCP A motor
| |
| thrust-bearing temperature at 195F and RCP A tipper radial bearing at 185F with
| |
| both slowly increasing.
| |
| Which of the following actions are required?
| |
| a. Stop RCP Aand initiate a rapid plant shutdown in accordance with AOP-038,
| |
| Rapid Downpower
| |
| b. Manually trip the reactor and go to PATH-I, stopping RCP Aas time permits
| |
| c. Continue monitoring RCP A temperatures, tripping the reactor and entering
| |
| PATH-I if RCP A temperatures exceed 300F
| |
| d. Stop RCP A, manually trip the reactor and go to PATH-I
| |
| ANSWER:
| |
| b. Manually trip the reactor and go to PATH-1, stopping RCP Aas time permits
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 61 TIERGROUP: 1/1
| |
| KAIMPORTANCE: RO 3.4 SHO
| |
| 1QCFH55CONTENT: 41(b) 3/10 43(b)
| |
| KA: 000015iI7AA2.08
| |
| Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions
| |
| (Loss of RC Flow): When to secnre RCPs on high bearing temperature
| |
| OBJECTIVE: AOP-3.18-3
| |
| Given a set of plant conditions and a copy of AOP-018, DETERMINE the appropriate response
| |
| DEVELOPMENT REFERENCES: AOP-018
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: MEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AOP-3.18 019
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR ~ S T I F I C A C T I O N(CORRECT ANSWER +d):
| |
| a. Plausible since the RCP is to be stopped, but must be stopped immediately which requires that the
| |
| reactor be tripped.
| |
| d b. RCP motor temperatures require the punip be stopped. With power above 48%, the reactor must be
| |
| tripped prior to tripping the RCP.
| |
| c. Plausible since this is a trip setpoint for stator winding temperature:, but the pump must be tripped
| |
| immediately based on the given temperatures.
| |
| d. Plausible since these are the correct actions, but the reactor should be tripped first and the pump
| |
| stopped when time permits.
| |
| ICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION: Knowledge of RCP motor temperature tripping requirements
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 62
| |
| Given the following conditions:
| |
| e Path-2 is being performed due to an SGTR.
| |
| e The Main Stcam Isolation Valves (MSIVs) on the intact SGs are open.
| |
| e The MSIV on the ruptured SG is closed.
| |
| a The Condenser is available for Steam Dump operation.
| |
| e A cooldown to 485 OF from 557 OF at the maximum rate is required.
| |
| Which one of the following describes the method to accomplish this cooldown?
| |
| a. Fully open the Stcam Dumps as fast as possible
| |
| b. Fully open the Steam Dumps as fast as possible without causing main steam line
| |
| isoiation
| |
| c. Fully open the intact SG PORVs as fast as possible
| |
| d. Fully open the intact SG PORVs as fast as possible without causing a main s t e m
| |
| line isolation
| |
| ANSWER:
| |
| b. Fully open the Steam Dumps as fast as possible without causing rnain steam line
| |
| isolation
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 62 TIEWGROUP: 1/1
| |
| KAIMPQRTANCE: RO 4.3 SRO
| |
| 10CPR55 CONTENT: 41(b) 10 43m
| |
| KA: 000038EA1.36
| |
| Ability to operate arid monitor the following as they apply io a SGTR Cooldown of RCS to specified
| |
| temperature
| |
| OBJECTWE: EOP-3.19-R4
| |
| Given a set of conditions during EOP implementation, DETERMINE the correct response or required
| |
| action based upon the EOP Users Guide general information
| |
| e Dumping steam at maximum rate
| |
| -u u
| |
| DEVELOPMENT REFERENCES: EOP Users Guide
| |
| REFERENCES SZJPPLIEDTO MPPI,ICANT: None
| |
| n
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT:
| |
| NRC EXAM HISTORY None
| |
| DIS1RACTOR JUSTIFICACTION (CORRECT ANSWER dd):
| |
| a. Plausible since the maximum cooldown rate can be achieved using maximum steam dump flow. but
| |
| causing too great a rate ofpressure drop will result in the MSIVs going closed which is undesirable.
| |
| 4 b. During a SGTR cooldown an attempt should be made to maximize s t e m dump demand without
| |
| causing SC; pressure to decrease fast enough to cause a main steam line isolation.
| |
| c. Plausible since this action would be taken if the MSIVs on the intact SGBwere already closed, but
| |
| with the M S N s open it is desirable to use steam dumps.
| |
| d. Plausible since causing the MSIVs to close is not desirable, but steam dumps should be used if
| |
| available.
| |
| ICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS
| |
| DIFFICULTY IWTING: 3
| |
| a KNOWLEDGE / RECALL
| |
| EXPLANATION: Knowledge of the EOP Users Guide requirement for performing a maximum
| |
| rate cooldown
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 63
| |
| Given the followiiig conditions:
| |
| e While transferring resin, it is noted that RM-1WR-4644A, SPENT RESIN PUMP 1-
| |
| 4A, radiation monitor is indicating 10 mR&.
| |
| e The monitor is physically located 20 feet away from a suspected clog in the pipe
| |
| which is the source ofthe monitor indication.
| |
| What is the radiation level 5 feet from the pipe? (ASSUME THE CLOG IN THE PIPE IS
| |
| A POINT SQIJRCE)
| |
| a. 20mRem/hr
| |
| b. 4QmRem/hr
| |
| c. 80 mRem/hr
| |
| d. 160mRem/hr
| |
| ANSWER.
| |
| d. 160rnRemhr
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 63 TIEWGRO[?E 211
| |
| KAIMPORTANCE: RO 2.5 SRO
| |
| 10CFR55 CONTENT: 41(b) 11/12 43(b)
| |
| KA: 073K5.02
| |
| Knowledge of the operational implications as they apply to concepts as they apply to the P F N system:
| |
| Radiation intensity changes with source distance
| |
| OBJECTIVE: RP-3.5-21
| |
| Calculate dose rates at different distanccs from point sources and line sources
| |
| DEVELOPMENT REFERENCES: Rk-LP-3.5
| |
| REFERENCES SUI'PIJED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT:
| |
| NRC EXAM IIISTORY: None
| |
| DIS1'RACTOR JUSTIFICACTION (CORRECT AWSWER d'd):
| |
| a. Piausible if the square root of the distances is taken, instead of squared as they should be (IOmR/hr x
| |
| 2 0 " ~ft = 20 mR//hr x 5'" ft).
| |
| b. Plausible if the distances are not squared as they should be (iOmR/hr x 20 fi = 40 mRhr x 5 ft).
| |
| e. Plausible if a mathematical error is made (value selected as a distmcter due to the progression of other
| |
| numbers in distracters).
| |
| 4 d. Using the formula 11d1*=12d;, the intensity ofthe source at 5 feet is calculated to be 160 mRem/hr.
| |
| ICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Calculation of distance using inverse square for radiation
| |
| Post Validation Revision
| |
| | |
| Marris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 64
| |
| Given the following conditions:
| |
| 0 The Control Room has been evacuated due to a fire.
| |
| 0 AOP-004, Remote Shutdown, is being performed.
| |
| 0 The crew is detcrmining the amount of boric acid required to be added to the RCS.
| |
| Which of the following describes the reason for adding boric acid during the performance
| |
| of Section 3.1, Remote Shutdown Due to Fire, of AOP-004?
| |
| a. Ensure adequate shutdown margin is maintained for the first 12 ~ O U F Sfollowing
| |
| the plant trip
| |
| b. Ensure adequate shutdown margin is maintained in the event that access to the
| |
| Control Room is prevented until the core has reached xenon-free conditions
| |
| C. Ensure adequate shutdown margin is maintained in the event that a cooldown to
| |
| Cold Shutdown conditions is required
| |
| d. Ensure adequate shutdown margin is maintained in the event that pressurizer is
| |
| required to be raised to 90% ievel
| |
| ANSWER:
| |
| c. Ensure adequate shutdown margin is maintained in the event that a cooldown to
| |
| Cold Shutdown conditions is required
| |
| Post Validation Rei.,ision
| |
| | |
| Harris NRC: Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NLMBEIP: 64 TIEWGROUP: 1/2
| |
| KAIMPORTANCE: RO 3.3 SRO
| |
| 10CFRSI CONTENT: 41(b) 6 43(W
| |
| KA: 000068AK3.13
| |
| Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation:
| |
| Perfcnning a shutdown margin celculation, inciuding boron needed and boration time
| |
| OBJECTIVE:
| |
| Given a set of plant conditions and a copy of AOP-004, Remote Shutdown, DETERMINE the
| |
| appropriate course of action
| |
| DEVELOPMENT REFERENCES: AOP-004-BI)
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK IVUI\IBERFOR SIGNIPICAVTLY IMODIFIED/ mmcr: N ~ W
| |
| NRC EXAM HISTORY None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER .Jd):
| |
| a. Plausible since shutdown margin is changing for the first 12 hours following a plant trip, hut shutdown
| |
| margin is increasing.
| |
| la. Plausible since shutdown margin will decrease as the core approaches xenon-fke conditions, but the
| |
| boration is performed in the event of a cooldown is required.
| |
| d c. A boration is only perfonned in the event that a cooldown is required to be performed during the
| |
| perforntance of AOP-004.
| |
| d. Plausibie since AOP-004 allows increasing PRZ level to 90% level, but this is done to accommodate
| |
| performance of a boration.
| |
| DIFFICULTY M A L Y S 6 :
| |
| COMPREHENSIVE / ANM,YSIS KNOWLEDGE / RECALL
| |
| DIFFICIJLTY RAlRvG: 3
| |
| EXPLANATION: Knowledge ofthe reason for performing a horation whiIe operating the plant
| |
| from the shutdown panel
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 65
| |
| Given the following conditions:
| |
| e The unit is operating at 50% power.
| |
| e LT-460, Channel 111Pressurizer Level, has failed and all associate, istables are in
| |
| the tripped condition.
| |
| e Power is subsequently lost to UPS Bus IDP-IA-SI.
| |
| Which train(§) of Reactor Protection will actuate, if any?
| |
| a. Neither train
| |
| b. Train SA ONLY
| |
| c. Train SB ONLY
| |
| d. Both trains
| |
| ANSWER:
| |
| d. Bothtrains
| |
| Post Validation Revisioii
| |
| | |
| Harris NKC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 65 TIEWGRQUP: 2: I
| |
| KAIMPQRTANCE: RQ 3.3 SWQ
| |
| 1QCFR55CONTENT: 41(b) 7 43w
| |
| KA: 012K2.01
| |
| Knowledge of bus power supplies to the following: KIPS channels, components, and intercoimections
| |
| OBJECTWE: AOP-3.24-2
| |
| RECOGNIZE automatic actions that are associated with loss of an instrument bus or loss ofNNS UPS
| |
| I)E\ELOPMENT REFERENCES: AOP-024
| |
| SD-103
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICNCLY MODIFIED DIRECT
| |
| BAKK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Iarris NRC 2000-29
| |
| NRC EXAM HISTORY: Harris NRC 2000
| |
| DPSTRAC71OR JUSTIFICACTIQN (CORRECT ANSWER dd):
| |
| a. Plausible since some ESF features arc energized to actuate, hut RPS features are all de-energized to
| |
| actuate
| |
| b. Plausible since an ESF actuation u ~ ~not l doccur on hath trains if required since slave relays require
| |
| power to actuate, but KPS is de-energized to actuate.
| |
| c. Plausible since an ESF actuation would not occur on both trains if required since slave relays require
| |
| power to actuate. hut KPS is de-energized to actuate.
| |
| d d. Bus 1A-SI suppries Channel H pressurizer level and a loss of this supply wiIl result in 2 channels being
| |
| tripped. Each channel inputs both trains of W S , so both trains of R19S will actuate.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Analysis of abnormal conditions regarding Ioss of power to determine response
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 66
| |
| Given thc following conditions:
| |
| o An earthquake has caused damage to the Main Reservoir dam
| |
| Main and Auxiliary Reservoir levels are both currently 240 feet and stable.
| |
| o AOP-022, Loss of SenriceWater, is being performed for a IAM of Ultimate Heat
| |
| Sink.
| |
| e Emergency Sewice Water (ESW) pumps have been aligned to the Main Reservoir.
| |
| e One ( I ) Noma1 Service Water (NSW) pump is operating.
| |
| Which of the following pumps are required to be operating to provide water to the SSE
| |
| Fire Protection Header once the ESW header is aligned to the fire protection header?
| |
| a. ONLY an ESW pump
| |
| b. An ESW pump AND an ESW Booster pump
| |
| c. ONLY a second NSW pump
| |
| d. A second NSW pump AND an ESW Booster pump
| |
| ANSWER:
| |
| b. An ESW pump AND an ESW Booster pump
| |
| Post Validation Revision
| |
| | |
| Iiarris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 66 TIEWGROUP: 21 1
| |
| KA IMPORTANCE: RO 2.5 SRO
| |
| 10CFR55 CONTENT: 41(b) 4 43w
| |
| KA: 076K1.15
| |
| Knowledge of the physical connections and/or cause-effect relationships between the S W S and the
| |
| following systems: FPS
| |
| OBJECTIVE: FP-3.0-3
| |
| STATE the sources of fire water available to the plant including automatic actuation signals
| |
| DEVELOPMENT REFERENCES: AOP-022
| |
| OH-I39
| |
| REFERENCES SUPPLIED TO ABPIXANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED
| |
| OW SIGNIFICANTLY MODIFIED I DIRECT:
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
| |
| a. Plausible since an ESW pump is started, but an ESW Booster pump is also required.
| |
| d b. An ESW pump, aligned to the Main. Reservoir, is started, along with an ESW Booster pump to supply
| |
| thc SSE fire protection header.
| |
| c. Plausible since the first NSW pump is not required to be tripped provided cooling tower basin level is
| |
| adequate and NSW supplies the ESW header (which can supply the fire protection header), but an
| |
| ESW pump is required.
| |
| d. Plausible since an ESW Rooster pump is required to supply the fire header, hut an ESW pump is
| |
| required to supply the booster pump.
| |
| DIFFICULTY ANALYSIS:
| |
| COlcIPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLATATION: Knowledge of the system alignments avaiiable to supply the fire header
| |
| Post Validation Revision
| |
| | |
| Hmis NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 67
| |
| Given the following conditions:
| |
| e The plant is being cooled down to 140°F for maintenance which will NOT require the
| |
| RCS be opened.
| |
| e The crew is in the process of placing the first Residual Heat Removal (RRR) train in
| |
| service for RCS cooling.
| |
| o Current boron concentrations are as follows:
| |
| RHR (train to be placed in service) boron 1021 ppm
| |
| Required Shutdown Margin boron 1200 pprn
| |
| o RCS boron 1341 ppm
| |
| Cold Shutdown boron 1750 ppm
| |
| o Refueling boron 2261 ppm
| |
| Before the RHR train can be placed in service for RCS cooling, WHR boron
| |
| concentration must be increased by a MINIMUM o f . ..
| |
| a. 179ppm.
| |
| b. 320ppm.
| |
| c. 729ppm.
| |
| d. 124Qppm.
| |
| ANSWER:
| |
| a. 179ppm.
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 67 TIER/GR0UID: 21 I
| |
| KAIMPORTANCE: RO 3.2 SRO
| |
| 1OCFR55 CONTENT: 41(b) 8 43W
| |
| K& 005K5.09
| |
| Knowledge of the operational implications of the following concepts as they apply the RHKS: Dilution
| |
| .and boration considerations
| |
| OBJECTIVE: RHRS-2.0-12
| |
| APPLY precautions and limitations of OP-111,KGNS to Hypothetical System Configurations
| |
| DEVE1,OPMENT IWFERENCES: OP-llI
| |
| REFERENCES SUPPLIED TO APPLICANT:
| |
| ~ ~ ~~~ None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DIS'I'RACTOH JUSTIFICACTION (CORRECT ANSWER tl'd):
| |
| d a. RHK boron must be greater than or equal to the required SDM or the required reheling concentration.
| |
| The boron concentration requirements will be dependent on the intended use of the KHR System.
| |
| Using the RHR system for cooldown purposes requires that the boron concentration be greater than or
| |
| equal to the required shutdown margin.
| |
| b. Plausible since this is the difference between RHR and RCS boron concentration, but only the
| |
| required SDM boron is needed.
| |
| e. Plausible since this is the difference between RHR and Cold Shutdown boron concentration, but only
| |
| the required SDM boron is needed.
| |
| d. Plausible since this is the difference between RIIR and rcfuelirig boron concentration, and refueling
| |
| conditions occur at 140°F.but only the required SDM boron is needed.
| |
| DIhTICUI,TY ANALYSIS:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| DnmcucrYw m G : 3
| |
| EXPLANATION: Application of actual versus required boron concentration - must determine
| |
| minimum limiting requirement
| |
| Post Validation Revision
| |
| | |
| Harris WRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 68
| |
| Given the foilowing conditions:
| |
| A liquid waste discharge from a Treated Laundry and Hot Shower (TL&HS) Tank is
| |
| in progress.
| |
| REM-1WL-3540, Treated Laundry and Hot Shower Tank Pump Discharge Monitor,
| |
| goes into high alarm.
| |
| Which of the following terminates the discharge'?
| |
| a. The running TL&HS Tank Pump will automaticalEy trip.
| |
| b. An operator must take manual action to shut the TL&HS Tank Pump Discharge
| |
| Isolation Valve.
| |
| c. The running TL&HS Tank Pump Recirc Valve will automatically open.
| |
| d. The TL&HS Tank Pump Discharge Isolation Valve will automatically close.
| |
| ANSWER:
| |
| d. The TI.&WS Tank Pump Discharge Isolation Valve will automatically close.
| |
| Post Validation Revision
| |
| | |
| Hareis NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 68 TIEFUGROUP: 2i2
| |
| KA IMPORTANCE: RO 3.6 SRO
| |
| IOCFR55 CONTENT: 41(b) 7/13 43@)
| |
| KA: 068A3.02
| |
| Ability to monitor automatic operation ofthe Liquid Radwaste System including: Automatic isolation
| |
| OBJECTWE: LWPS-LP-3.0-7
| |
| UESCRE3E the automatic protection features associated with discharges to the environment from the
| |
| LWPS
| |
| DEVELOPMENT REFERENCES: AOP-005
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED D I ~ ~
| |
| HANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Kh4S-A6 005
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIPICACTION (CORRECT ANSWER .I'd):
| |
| Plausible since the pump will stop the discharge, but there is no auto trip due to high rad.
| |
| h. Plausible since manual isolation will stop the discharge, but an auto isolation will not require operator
| |
| action.
| |
| c. Plausible since placing the tank in recire will stop discharge, but only because of the isolation valve,
| |
| as the recirc valve does not have an auto function.
| |
| ./ d. On a high rad level as sensed by REM 3540, the discharge isolation valve will automatically close,
| |
| terminating any release in progress.
| |
| DIFFICULTY ANhLYsrs:
| |
| COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 2
| |
| EXPLANATION: Knowledge of liquid radwaste design and operation
| |
| Post Validation Revision
| |
| | |
| Harris NKC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 69
| |
| Assuming NO operator actions, which ofthe following describes the effect of a loss of
| |
| instrument air on Volume Control Tank (VCT) level?
| |
| a. VCT level decreases due to maximum charging and letdown isolation valves
| |
| closing
| |
| b. VCT level decreases due to maximum charging and letdown being diverted to the
| |
| Hold Up Tank
| |
| c. VCT levcl increases due to minimum charging and the letdown pressure control
| |
| valve failing open
| |
| d. VC'T level increases due to minimum charging and the letdown orificc isolation
| |
| valves failing open
| |
| ANSWER:
| |
| a. VCT level decreases due to maximum charging and letdown isolation valves
| |
| closing
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examinatian
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBEM: 69 TIER/GROUP: 21 1
| |
| KAIMPORTANCE: RO 3.4 SRO
| |
| 10CFR55 CONTENT: 41(h) 4 43(W
| |
| KA: 048K3.02
| |
| Knowledge of the effect that a loss or malfunction of the KAS will have on the following: Systems having
| |
| pneumatic valves and controls
| |
| OBJECTIVE: AOP-3.17-4
| |
| Given a set of entry conditions, and a copy of AOP-017, DETEKMINE the appropriate response.
| |
| DEVELOPMENT REFERENCES: AOP-0 17
| |
| REFERENCES SUPPLIED TO APPLICANT
| |
| QUESTION SOURCE: NEW a None
| |
| SIGNTFXCtU'TILY MODIFIED
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CVCS-R3 008
| |
| 0 DIRECT
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUS'THFICACTION (CORRECT ANSWER d'd):
| |
| d a. Charging flow control fails open and letdown isolation valves fiail closed on a loss of air, so VCT level
| |
| will decrease.
| |
| b. I'lausibie since VCT level will decrease, but it will be due to letdown isolating, not diverting water to
| |
| the hold up tank.
| |
| C. Plausible since the letdown pressure control vahe faiis open on a loss of air, but the letdown isolation
| |
| valves fail closed, isolating ietdown.
| |
| d. Plausible since the charging Bow control valve and the letdown orifice valve all fail on a loss of air,
| |
| hut fail in the opposite direction as that which would cause this response.
| |
| DPFFICUI,TY ANALYSIS:
| |
| COMPREIIENSWE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIPFICULTY RATING: 3
| |
| EXPI.ANATION: Analyze the response of GVCS aAer determining the fail position of various
| |
| CVCS vaives on a loss of IA
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 70
| |
| Given the following conditions:
| |
| o Following a plant trip, EOP-EPP-004, Reactor Trip Response, is being performed.
| |
| o The crew is verifying Natural Circulation conditions as a result of a loss of power to
| |
| all RCPs.
| |
| m Five (5) core exit thermocouples are failed.
| |
| How do the failed core exit therniocouples affect indications used to verify Natural
| |
| Circulation?
| |
| a. * The Core Exit Temperature indications will be HIGHER than actual
| |
| o RCS Subcooling will indicate MORE subcooling than actual
| |
| b. e The Core Exit Temperature indications will be HIGHER than actual
| |
| RCS Subcooling will indicate LESS subcooling than actual
| |
| c. Core Exit Temperature indications will indicate LOWER than actual
| |
| o RCS Subcooling will indicate MORE subcooling than actual
| |
| d. e Core Exit Temperature indications will indicate the SAME as actual
| |
| e RCS Subcooling will indicate the SAME subcooling as actual
| |
| ANSWER:
| |
| d. Core Exit Temperature indications will indicate the SAME as actual
| |
| o RCS Subcooling will indicate the SAME subcooling as actual
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QCESTION NUMBER: 70 TIEWGHOUP: 2l2
| |
| KAIMPORTANCE: RO 3.5 SRO
| |
| IOCPR55 CONTENT: 41(b) 5 4300
| |
| KA: 017K3.01
| |
| Knowledge of the effect that a loss or malfunction of the ITM system will have on the following: Natural
| |
| circulation indications
| |
| OWECTNE: ICX3Vf-4.0-R6
| |
| DESCRIBE how the plant's subcooling monitor infonnation is processed
| |
| DEVELOPMENT REFERENCES: SD-106
| |
| ICCM-LP-3.0
| |
| mFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESlION SOURCE: NEW SIGNIFICANTLY RIODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT New
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR SUSTIFICACTION (CORRECT ANSWER d'd):
| |
| a. Plausible since the thermocouples are failed, but a failed thermocouple indicates 507: (low) and not
| |
| high.
| |
| b. Plausible since the thermocouples are failed, but a failed thermocouple indicates 50'F (low) and not
| |
| high.
| |
| e. Plausible since the failed theirnocouples indicate 50°F (low), but the ICCM indication uses the highest
| |
| thennocouples and not the lowest.
| |
| d d. The failed thermocouples will not be used to process the indication by the ICCM, so them will be no
| |
| change on core exit temperatures and subcooling margin.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREIPENSNE I ANALYSIS KNOWLEDGE I RFCAI,L
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Analyze the effect of failed thermocouples on temperatures and subcooling
| |
| margin
| |
| Post Validarioti Revision
| |
| | |
| Hwris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 71
| |
| Which of the following EOP network procedures, containing NO Immediate Operator
| |
| Actions, may be directly entered and the conditions under which it may be entered?
| |
| a. EOP-FRP-1.1, Response to Pressurizer High Level, when it is desirable to
| |
| restore pressurizer level following a malfunction of the Pressurizer Level Control
| |
| System and NO accident is in progress
| |
| b. EOP-FRP-I. I, Response to Pressurizer High Level, when it is desirable to
| |
| restore pressurizer level following an inadvertent Safety Injection actuation with
| |
| the plant in Mode 3
| |
| C. EPP-005, Natural Circulation Cooldown. when it is desirable to cuoidown with
| |
| RCPs unavailable and NO accident is in progess
| |
| d. EPP-005, Natural Circulation Cooldown, when it is desirable to cooldown with
| |
| RCPs unavailable due to a loss of offsite power with the plant in Mode 3
| |
| ANSWER:
| |
| c. EPP-005, Natural Circulation Cooldown, when it is desirable to cooldown with
| |
| RCPs unavailable and NO accident is in progress
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 41 TIEWGROFTP 3
| |
| KAIMPORlANCE: RO 4.3 SHO
| |
| 10CFH55 CONTENT: 41(h) 10 43m
| |
| KA: 2.4.1
| |
| Knowledge of EOP entry conditions and immediate action steps
| |
| OBJECTIVE: EOF-3.19-1
| |
| DESCRIBE Control Room usage of the EOP network as it relates to the following
| |
| e Entry into EOP network
| |
| DEVFLOPMENT REFERENCES: EOP Users Guide (page 13)
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW , SIGNIFICANTLY MODKFFIED DIRECT
| |
| BANK NLJMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Mew
| |
| NRC E.XhM " T O R Y : None
| |
| DISTRACTOR JUSTIFKCACTION (CORRECT ANSWER d'd):
| |
| 8. Plausible since FKP-1.1 is used to lower PFZ leve1 during the perfomlance of the EOP network, but is
| |
| entered only by operator judgement when a yellow path condition is encountered.
| |
| b. Plausible since FW-1.1 is used to lower PKZ level during the perfomlance of the EOP network, but is
| |
| entered only by operdtor judgement when a yellow path condition is encountered.
| |
| 4 c. EPP-005, "Natural Circulation Cooldown," may be entered whenever a natural circulation cooldown is
| |
| required and an accident is not in progress.
| |
| d, Plausible since EFP-005 may be entered whenever a natura1 circulation cooldown is required, but no
| |
| accident can be in progress.
| |
| DIFFICUI.TY ANALYSIS:
| |
| 0 COMPREFIENSIVE I ANALYSIS KNO\VLEDGE /RECALL
| |
| DIFFICUL7L1' RATENG: 2
| |
| EXPLAK4'rION: Knowledge of EOPs which can be entered directly
| |
| Post Validation Revision
| |
| | |
| Hmis NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 72
| |
| Which of the following is a reason that containment pressure greater than 45 psig is
| |
| considered an extreme challenge to the containment critical safety function?
| |
| a. Containment structural failure is imminent
| |
| b. Containment leakage could bc in excess of design basis lcakage
| |
| c. Hydrogen recomhiner efficiency is significantly reduced at the pressure
| |
| d. Containment temperature is high enough to prevent adequate core cooling
| |
| ANSWER
| |
| b. Containment leakage could be in excess of design basis leakage
| |
| Post Validation Revision
| |
| | |
| Harris NIPC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUEsTION NUMBER 92 mwcxtom 2/1
| |
| KA IMPORTANCE: NO 3.1 SRO
| |
| IQCFRSCONTENT: 41(b) 9/10 43(b)
| |
| KA: 10362.4.6
| |
| Knowledge of symptom based EOP mitigation strategies. (Containment)
| |
| OBJECTIVE: EOP-3. 13-4
| |
| Given the following EQP steps, notes, and cautions, DESCMBE the associated basis
| |
| CSF decision points
| |
| DEVELOPMENT REFERENCES: EOP-LP-3.13
| |
| REFERENCES SUPPLIED TO APPLICANT:
| |
| QUESTION SOURCE: NEW n None
| |
| SIGNIFICANTLY MODIFIED
| |
| RANK NIJMMBER FOR SIGNIFICANTLY MODIFIED / DIRECT:
| |
| DIRECT
| |
| EOP-3.13-194 001
| |
| NRC EXAM IIISTORY None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
| |
| a. Plausible since this is above the postulated pressure following a large break LOCA or steam break, but
| |
| containment failure is not expected to occur until several times this value.
| |
| d b. 45 psig is above the pressure that design containment leakage rates assumed in off-site radiological
| |
| analysis.
| |
| e. Plausible since the recombiners are located in containment and are conceivably affected by the high
| |
| presmrc, but the 45 psig is selected based an exceeding design leakage rates.
| |
| d. Plausible since core cooling in the event of a LOCA is based upon transfemng heat to the injection
| |
| water which is then collectcd in containment for recirc, but the 45 psig is selected based on exceeding
| |
| design leakage rates.
| |
| DIFFICULTY ANALYSIS:
| |
| COMPREIIENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Knowledge ofthe basis far CSFST decision points for containment pressure
| |
| Post Validation Revision
| |
| | |
| IIarris MRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 73
| |
| Which of the following would require that Independent Verifrcation be performed in
| |
| accordance with OPS-NGGC- 1303, "Independent Verification'?"
| |
| a. During Mode 5, a valve in the Containment Spray system is being repositioned for
| |
| testing and the OP lineup will be completed prior to Mode 4 entry
| |
| b. During Mode 1, a valve in the Main Steam system is being placed under clearance
| |
| and the valve is only accessible with a manlift
| |
| C. During Mode 4, a valve in CVCS inside containment is being positioned for
| |
| draining and the valve is located in an area where the temperature is I34OF
| |
| d. During Mode 3, a valve in CVCS is being placed mder clearance and the valve is
| |
| located in a radiation field of 175 mRcmihr with an estimated verification time of
| |
| 6 minutes
| |
| ANSWER:
| |
| h. During Mode 1, a valve in the Main Steam system is being placed under clearance
| |
| and the valve is only accessible with a manlift
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER 73 TIENGROUP: 3
| |
| KA IMPORTANCE: KO 3.6 SWO
| |
| 1OCFR55 CONTENT: 41(b) 10/12 43(b)
| |
| KA: 2.2.13
| |
| Knowledge of tagging and clearance procedures
| |
| OBJECTIVE: PI-3.1 I-H8
| |
| DISCUSS the forlowing i t e m concerning independent verification
| |
| a. When it is used
| |
| e. When it may be waived for A L M A
| |
| DEVELOPMENT REFERENCES: OPS-NGGC-I 303
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SIGN1FICANTL.Y MODIFIED DIRECT
| |
| RANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: PP-3. i 1-R8 003
| |
| NRC EXAM HISTORY: Harris NRC 2002
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
| |
| a. Plausible since the valve is easily accessible, but rV is not required since plant conditions do not
| |
| require the system to be operable and spray valve positions are covered by a lineup.
| |
| 4 b. IV requirements may be waived if radiation exposures : 10 &em would result, if area temperatures
| |
| in excess of 1200F exist, or ifplant conditions do not require the equipment to be operable and a varve
| |
| lineup controls the position.
| |
| c. Plausible since the valve is k i n g repositioned for draining, but area temperature conditions pennit
| |
| waiving the Ri requirements.
| |
| d. Plausible since the valve is being repositioned for a clearance, but projected radiation exposures
| |
| permit waiving the Hv requirements (17.5 mR > 10 a).
| |
| DIFFICULTY mmysrs:
| |
| COMPREIIENSIVE / ANALYSIS KNOWLEDGE /RECALL
| |
| DrFFIcxrcrY RATING: 3
| |
| EXPLANATION: Application of plant conditions to implementation of administrative procedural
| |
| requirements
| |
| Post Validation Revision
| |
| | |
| IIarris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 74
| |
| Given the following conditions:
| |
| Following an accident, EOP-EPP-015, Uncontrolled Depressurization of All Steam
| |
| Generators, is being performed.
| |
| Due to the excessive cooldown rate, the operators have reduced AFW flow to all
| |
| stcam generators (SG) to mininium as they continue attempts to isolate the SGs.
| |
| Which of the following describes the expected plant response to the AFW flow reduction
| |
| and what actions are to be taken as SG pressures decrease?
| |
| a. RCS hol leg temperatures will eventually begin to increase and the crew will then
| |
| transition to EOP-EPP-008, Safety Injection Termination
| |
| b. RCS hot leg temperatures will eventually begin to increase and the crew will then
| |
| be required to increase AFW flow to maintain RCS hol leg temperatures stable or
| |
| decreasing
| |
| c. The SGs will eventually approach dry conditions and no longer be required and
| |
| the crew will then transition to EQP-EPP-008, Safety Irijection Termination
| |
| d. Tke SGs will eventually approach dry conditions and the crew will then be
| |
| required to isolate AFW flow to all SGs
| |
| ANSWER:
| |
| b. RCS hot leg temperatures will eventually begin to increase and the crew will then
| |
| be required to increax AFW flow to maintain RCS hot leg temperatures stable or
| |
| decreasing
| |
| Post Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Data Sheets
| |
| QUESTION NUMBER: 74 TIEWGROUP: 1/1
| |
| KAIMPORTANCE: RO 3.4 SRO
| |
| 10CFR55 CONTENT: 41(b) 4/10 43@)
| |
| I<A: WEl2EK2.1
| |
| Knowledge of the interrelations between the (Uncontrolled Depressurization of all Steam Generators) and
| |
| the following: Components, and functions of control and safety systems, including instmmentation,
| |
| signals, interlocks, failure modes, and automatic and manuai features
| |
| OBJECTIVE: EOP-3.9-4
| |
| Given actions taken in these emergency procedures, PREDICT the plant response to these actions
| |
| DEVELOPMENT REFERENCES: EOP-EPP-0 15
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUESTION SOURCE: NEW SlGNIFICAh'TLFLY MODIFIED DIRECT
| |
| BANK NLIWBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
| |
| NRC EXAM HISTORY: None
| |
| DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
| |
| a. Plausible since hot leg temperatures will eventuaIly increase, but the comct action is to stabilize
| |
| temperature by increasing AE'W flaw and adjusting steaming rate, if possible.
| |
| d b. As SCi pressure and steam flow decrease, RCS hot leg temperatures will eventually stabilize and may
| |
| increase. Adjusting feed flow and steam dump will control RCS hot leg temperatures.
| |
| e. Plausible since the minimum APW flow will result in the SGs nearing dryout conditions, but as hot
| |
| leg temperature begins to increase the correct action is to stabilize temperature by increasing AFW
| |
| flow and adjusting steaming rate, if possibie.
| |
| d. Piausible since the niininium AFW flow will result in the SGs nearing dryout conditions, but as hot
| |
| leg tcniperature begins to increase the correct action is to stabilize temperature by increasing AFW
| |
| flow and adjusting steaming rate, if possible.
| |
| ICULTY ANALYSIS:
| |
| COI\IPREII%NSIVE/ ANALYSIS KNOWLEDGE /RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Analyze SO response to decreasing pressure and reduced AFW flow and
| |
| determine correct response
| |
| I'oat Validation Revision
| |
| | |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| QUESTION: 75
| |
| Which ofthe following conditions would permit securing Containment Spray per EOP-
| |
| PATH-1 Guide?
| |
| a. a Actuation caused by a LOCA
| |
| 4 Time since LOCA occurred is 6 hours
| |
| Containment pressure is 9 p i g
| |
| b. a Actuation caused by il LOCA
| |
| o Time since LOCA occurred is 3 hours
| |
| a Containment pressure is 5 psig
| |
| c. a Actuation caused by a Steam Line Break
| |
| e Time since Steam Line Break occurred is 3 hours
| |
| 4 Containment pressure is 5 psig
| |
| d. 4 Actuation caused by a Steam Line Break
| |
| a Time since Steam Line Break occurred is 6 hours
| |
| 0 Containment pressure is 9 psig
| |
| ANSWER
| |
| c. 4 Actuation caused by a Steam Line Brcak
| |
| 4 Time since Steam Line Break occurred is 3 hours
| |
| * Containment pressure is 5 psig
| |
| Post Validation Revision
| |
| | |
| -.,. ~ ...,.,...,.............
| |
| Harris NRC Written Examination
| |
| Reactor Operator
| |
| Rata Sheets
| |
| QUESTION NUMBER 75 'HEWGROUP: 21I
| |
| KAIMPORTANCE RQ 3.2 SRO
| |
| 10CFIW CONTENT: 41(b) 7/9 4Yb)
| |
| KA: 026A2.08
| |
| Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based
| |
| on those predictions, use procedures to correct, contro1,or mitigate the consequences ofthose
| |
| malhnctions or operations: Safe securing of containment spray (when it can be done)
| |
| OJUECTIVE: EOP-3.1-7
| |
| Given the following EOP steps, notes, and cautions, DESCRIBE the associated b a i s
| |
| d. Criteria for securing of ChMT spray
| |
| DEVELOPMENT REVERENCES: EOP Guide 1
| |
| REFERENCES SUPPLIED TO APPLICANT: None
| |
| QUES'GI~NSOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
| |
| BANK NUMBER FOR SIGNIFICANTLY MODIFIED / 1)IRECT: EOP-3. I-R5 006
| |
| NRC EXAM HISTORY: Hams m C 2002
| |
| DISTRACTOR JCJSTIFICACTION (CORRECT ANSWER J'd):
| |
| a. Plausible since the minimum time requirement of4 hours of spray has been met, but pressure must be
| |
| reduced below 8 psig prior to resetting.
| |
| b. Plausible since pressure is below 8 psig, hut must meet the time requirements prior to resetting.
| |
| 4 c. Containment pressure must be below 8 p i g to ensure it can automatically reactuate if pressure goes
| |
| back above 10 p i g , and in the event of a LOCA the system must operate for > 4 hours, but in a steam
| |
| break early termination is desired.
| |
| d. Plausible since the minimum time requirement of 4 hours of spray (for 6,MK:A) has been met, but
| |
| pressure must be reduced below 8 psig prior to resetting.
| |
| DIFFICULTY AllALYSIs:
| |
| c]COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
| |
| DIFFICULTY RATING: 3
| |
| EXPLANATION: Kaowledge of the bases for EQP steps
| |
| Post Validation Revision
| |
| }} | | }} |
