ML083540178
| ML083540178 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/13/2008 |
| From: | Exelon Generation Co |
| To: | Fish T H Operations Branch I |
| Hansell S | |
| Shared Package | |
| ML081060602 | List: |
| References | |
| 50-277/09-301, 50-278/09-301, TAC U01776 | |
| Download: ML083540178 (205) | |
Text
Peach Bottom Initial Reactor Operator NRC Examination December 2008 1. Unit 2 experienced a LOCA inside the primary containment. T-112 "Emergency Blowdown" is in progress with all 5 ADS Safety Relief Valves open. Current conditions are as follows: 0 0 0 RPV level is - 170 inches RPV pressure is 390 psig Drywell pressure is 10 psig All ECCS pumps are running Based on the above conditions, MO-'2- 10-25B "RHR Inboard Discharge Valve is -___ (1) and AO-2-10-46B "Testable Check Valve" is (2) . B. (1) closed (2) open C. (1) open (2) closed D. (1) closed (2) closed Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 Level ______ of Knowledge HIGH Choice Time Allowance (minutes)
____ -~ ~-___ ~ Difficulty - -____-._____
__ N/A Correct: ~ c .~ A ___ B Distractors:
ID ____________________~~__~
Basis or Justification MO-25B automatically opens at 450 psig. With RPV pressure at 390 psig, which is above the shut off head (-300 psig) of the RHR pumps, the testable check valve is closed. With RPV pressure at 390 psig, which is above the shut off head (-300 psig) of the RHR pumps, the testable check valve is closed. --______~
_ __ _- -____---__-________________._____~_~_. - -- MO-25B would automatically open at RPV pressure 450 psig; AO 10- 46B would be closed with RPV pressure at 390 psig. MO-25B would automatically open at RPV pressure 450 psig. _- ~ ________~____.____.
Source: Reference(s):
Learning Objective: KIA System Source Documentation New Exam Item Modified Bank Item ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
~__~_~.__
____ _- ___ - ____ - SO 10.7.8-2; T-101 Bases
-~ ~ ~ PLOT-5010-5a 203000 - LPCl Injection RO I SRO 2.7 12.9 -~ _ - -_ KIA Statement K5.01 - Knowledge of the operational implications of testable check valve operation as it applies to LPCl Iniection.
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 2. Unit 2 is in Day 20 of a refueling outage. The Unit is in Mode 5 with fuel rnoves in progress. Shutdown Cooling was in service: using the 2D RHR pump when it tripped on overcurrent.
No other RHR pumps will be available to support shutdown cooling for another 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Based on these conditions, which alternate decay heat removal system(s) need to be placed in service to adequately handle the current decay heat load? Refer to the portion of ON-125 "Loss or Unavailability of Shutdown Cooling", Attachment 1 "Alternate Decay Heat Removal Systems" on the NEXT PAGE. A. RWCU System ONLY B. Fuel Pool Cooling System with 2 heat exchangers ONLY C. Alternate Shutdown Cooling D. RWCU System and Fuel Pool Cooling System with 3 heat exchangers Peach Bottom Initial Reactor Operator NRC Elxamination December 2008 ON-125 PROCEDURE REV. 8 Page 10 of 17 Svs t em Attaghment 1 ALTERNATE DECAY HEAT REMOVAL SYSTEMS - Heal; Limitations
-- Remaval cambili ty RWCU Fuel Pool Cooling 4.4 MW fonts NRWXE 1.1 MW 1. Unit in MODE 5 il HXj 2. Reactor cavity 2.2 MW 3. Fuel Pool Gat-es flooded (2 HX) removed 3.3 MW f3 KX) Alternate Shutdown Cooling 20.5 MW Will inject low- in accordance with per RHR quality water into RPV Hx A0 10.12-2 (3)
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 SRO -d --_______
Level of Knowlee Difficuk __ ____ Time ~____~___
Allowance (minutes)
__.__ ___~__ - Psychometrics HIGH N/A Choice Correct: Distractors:
D A B C ~_______ - ~ ___-~-_~ - _ Basis or Justification Based on the Decay Heat Curve, Unit 2 has approximately 7 MWth of Decay Heat 20 days after shutdown.
To adequately handle removal of that decay heat the RWClJ System (4.4 MWth) and the Fuel Pool Cooling System (3.3 MWth wl all 3 heat exchangers in service) must be placed in service (total of 7.7 MWth decay heat removal capability). Multiple alternate decay heat removal systems may be used together to count as one alternate decav heat removal method. ________~_
~- -_ The RWCU System alone has only 4.4 MWth of decay heat removal capability. This is not enough to properly handle the 7 MWth decay heat. The Fuel Pool Cooling System with only 2 heat exchangers in service has only a total of 2.2 MWth of decay heat removal capability. This is not enough to properly handle the 7 MWth decay heat. _______-~___-______-____~-- - .~ Since no RHR pumps are available for operation, use of A0 10.12-2 is prohibited.
Source: Reference( s): Learning Objective:
~ - __~_.___ __.____ -. KIA Statement Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
______~ _. -_ -_- -~ -~_ _- - ON-I 25 PLOT-PBIG-1550-28b
______~___~__~
~ ~-~- ~ _- - - 203000 - RHR/LPCI Injection Mode
~ Importance:
RO / SRO L ~- -~ 3.8 14.2 G2.4.9 - Knowledge of low power/shutdown implications in accident (e.g. LOCA or loss of RHR) ____~~_~ __ ~~ ~ ~- ~ - - ~ ~- ~- _~- -- --~_ _--_________-
_~~_____________ -
Peach Bottom Initial Reactor Operator NRC Elxamination December 2008 3. Unit 2 is in a forced outage with the ifollowing conditions present:
0 0 0 The 'A' Loop of SDC is in service using the 2C RHR pump. RPV level inadvertently lowers to -3 inches. Reactor pressure is 25 psig and stable.
How will the following RHR System components respond to this transient?
0 0 2CRHRpump MO-2-10-25A - "A Loop RHR Inboard Discharge Valve" MO 10-1 7 and MO 10-1 8 - "RHR shutdown Cooling Suction Valves" A. MO-2-10-25A will close ONLY.
2C RHR pump will continue to run. B. MO-2-10-17 and MO-2-10-18 will close ONLY 2C RHR pump will trip.
C. MO-2-10-17 MO-2-10-18 and MO-2-10-25A will remain open. 2C RHR pump will continue to rum. D. MO-2-10-17 MO-2-10-18 gj MO-2-10-25A will close.
2C RHR pump will trip.
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 B IC Correct: Source : _______- __ Reference( -_____ s) : Learning Objective:
D 0 New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
-~_ .~-_______~
____ ~ ~_ ~- -~ - - GP-8.B COL PLOT-5010-40 Distractors:
-___~ _ - ~ Basis or Justification MO-17 and 18 will close on the PClS Group II isolation signal of c +I" RPV level. MO-25A (RHR Injection valve) will also close on PClS Group II signal of +I" RPV level with MO-17 open and MO-18 open and RPV pressure < 70 psig. The C RHR pump will trip when either MO-17 or R40-18 indicate not full open (loss of suction path).
~- __ _~____________ MO-17 and MO-18 (RLHR suction isolation valves) will also close on PGlS Group I1 isolation signal of c +I" RPV level. The C RHR pump will trip when either MO-17 or MO-18 indicate not full open (loss of suction path).
MO-25A (RHR Injection valve) will also close on PClS Group II signal of +I" RPV level with MO-17 open and MO-18 open and RPV pressure 8: 70 psig. MO-17 and MO-18 wrll close on PClS Group II isolation signal of +I" RPV level. MO-25A (RHR Injection valve) will also close on PClS Group II signal of +I" RPV level with MO-17 open and MO-18 open and RPL pressure < 70 psig. The C RHR pump will trip when either MO-17 or MO- 18 indicate not full open. -____ ~~ -~ -~ __ ._~__.___~_________~_
~- -~ ______-_____~~__~-
~ -~- ~ _~ Importance:
RO I SRO 3.2 I 3.2 __~_ - KIA System 205000 - Shutdown Cooling Shutdown ____ Cooling System including
~_- pump trips. _~ ~__ _~ ~~_~ ~_ ~- ~~_ - ~-
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 4. The following conditions exist at Unit 2: 0 0 0 The HPCI System was manually initiated for level control with the arming collar (23A-S 105) left in the "ARMED" position.
The HPCI turbine subsequently tripped due to the reactor water level exceeding
+46 inches. RPV level stabilized at
+52 inches. Which one of the following describes HPCI pump/system response to the above conditions?
A. HPCI will re-initiate if Drywell Pressure exceeds 2 psig. B. HPCI will re-initiate if the Manual Initiation pushbutton is depressed again. C. HPCI will re-initiate when the Hi(gh Reactor Level trip signal is reset. D. HPCI will ONLY re-initiate when the Lo-Lo Reactor Level setpoint is reached.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowle&e HIGH _______ .- Question # 4 RO Choice -_-__ --I -- .-. Correct: C SRO NIA -_ _ -_ Time Allowance (minutes)
Difficulty
______ _______ __ __ ____~ ---______
Distractors: REQUIRED MATERIALS:
r - NONE ___~ ~ The HPCl turbine trips auto reset and DO NOT seal-in. If the trip condition clears AND an initiation signal is present, THEN the HPCl turbine (pump) will cycle on and off as the high RPV level auto resets itself. With the arming collar left in the "ARMED" position the system initiation signal remains present.
If drywell pressure reaches 2 psig it will have no additional effect on HIXI. With the arming collar left in the "ARMED" position the system initiation signal remains present. Pushing the Manual Initiation pushbutton again will have no additional effect on HPCI. ~._______~_~._~
_____~ With the arming collar left in the "ARMED" position the system initiation signal remains present. If RPV level were to lower to the Lo-Lo setpornt (-48 inches) it would have no additional effect on HPCI. - __~.___~ ~___~______~_~
__ ~ ._ Source Documentation New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam Other Exam Bank
~ .~ ~~ SO 23.7.C-2 __._ Source: Reference(s):
Learning Objective:
--____.___ Notes and Comments:
______-____--_
~
Peach Bottom Initial Reactor Operator NRC E:xamination December 2008 5. An ADS Blowdown has occurred following a LOCA with the following conditions:
0 0 0 0 0 0 ADS Valve Control Switches remain in AUTO. Reactor pressure is 200 psig and lowering slowly. All Core Spray and RHR pumps were initially injecting. "D" Core Spray pump has since tripped.
All RHR pumps were secured when level was restored above Level is being restored using A, 11, and C Core Spray pumps.
100 inches. An additional Core Spray pump needs to be shutdown to control level recovery. Which one of the following statements accurately describes the response of the ADS system to shutting down a Core Spray pump? A. ADS Blowdown will stop if the "A" Core Spray pump is shutdown.
B. ADS Blowdown will stop if the "B" Core Spray pump is shutdown.
C. ADS Blowdown will stop if the "C" Core Spray pump is shutdown.
D. An ADS logic seal-in prevents inadvertent blowdown termination by shutdown of any Core Spray pump.
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 Difficulty
_______ -_____ _-p_____ Level of Knowledge
__ HIGH Choice Time Allowance (minutes)
~ - pp NIA Correct: --_____~ Reference(s):
Objective:
-~________~-
Learning ~._____.-___
Distractors:
0 Modified Bank Item ILT Exam Bank 0 Other Exam Bank -p _-_________~__p-~
_p_ - ~ - _____________-_____-__~___
M-14-52, ARC 227 E-4 ~ __ - ~ -- PLOT-5001 G-5 ~ ~- __- ~ ______~ C KIA System - -___-_______
A B Importance: RO I SRO 3.8 13.9 ~- -p_~__~ - 209001 - LPCS - -_____p- D Basis or Justification ADS logic requires "A or B" AND "C or D" Core Spray pumps to initiate and maintain a blowdown. When the C Core Spray pump is shut down, the logic is no longer satisfied and the blowdown will be terminated.
p____-______
.___ .~ ADS logic requires "A or B" AND "C or D" Core Spray pumps to initiate and maintain a blowdown When the A Core Spray pump is shut down, the logic is still satisfied due to the B AND D Core Spray pumps remaining in operation.
ADS logic requires "A or B" AND 'IC or D" Core Spray pumps to initiate!
and maintain a blowdown When the B Core Spray pump is shut down, the logic is still satisfied due to the A AND D Core Spray pumps remaining in operation.
-- ~ -~ - ADS logic has seal-ins for high drywell pressure and low reactor level, but not for required low pressure injection systems.
Peach Bottom Initial Reactor Operator NRC E:xamination December 2008 6. The Standby Liquid Control System explosive (squib) valves are powered from which of the following sources?
A. 24 / 48 VDC Distribution B. The respective pump Motor Control Center C. 120 VAC Uninterruptible Power Distribution D. Safety Related 125 VDC Distribution Peach Bottom Initial Reactor Operator NRC Examination December 2008 Correct: B Distractors:
~ - D A _. C ________._~
__ Basis or Justification
___~~____
~__~____ -~ ~ The MCC of the SLCS pump selected for injection provides the 480VAC power for both squib valves.
FUNDAMENTAL The MCC of the SLCS pump selected for injection provides the power for both squib valves from Emergency 480 VAC power, not from 24/48 Volt DC distribution.
The MCC of the SLCS pump selected for injection provides the power for both squib valves from Emergency 480 VAC power, not from 120 VAC: Uninterruptible Power. The MCC of the SLCS pump selected for injection provides the power for both squib valves from Emergency 480 VAC power, not from 125 VDC. __ ~ ~~ ~~ ~- ~. N/A I I I I --I Psychometrics -I 1 Level of bme Allowance SRO ~- - ~ ~~ Source: Referencew
____~ Learning 0 bject ive:
KIA System Source Documentation New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
~_ ~__ _______~_~~____ - ARC 21 1 H-3; M-1-S-46 PLOT-501 1-2b ~_____~______~_______ 21 1000 - Standby Liquid Conltrol System 1 Importance:
RO / SRO 3.1 13.2 -- KIA Statement K2.02 - Knowledge of the electrical power __~_________
supplies to the explosive valves. _-~ -~ - ~~ REQUIRED MATERIALS:
~ __- Notes and - Comments:
-~
Peach Bottom Initial Reactor Operator NRC Examination
- 7. December 2008 Which one of the following describes the RPS design feature that assures a scram goes to completion once it is initiated?
A. Any automatic scram signal also trips the manual scram logic. B. Scram reset logic is NOT enabled for 10 seconds following a full scram. C. Backup Scram valve logic can NOT be reset for at least 40 seconds after a full scram. D. Mode Switch to Shutdown scram is NOT bypassed for 2 seconds following a full scram.
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 DifficuN Level of Knowledge FUNDAMENTAL P Correct: SRO N/A _-_____ Time Allowance (minute9 ~-__ ~ ~ -~ __ Distractors: A . Source: -_ 0 bjective:
___-~-- __-______
~_____ ~ _~ Basis or Justification
-____ ___ RPS is designed to ensure that the scram goes to completion once initiated, a scram caninot be reset for at least 10 seconds. ~~ -~ ~__ ___________
-~ ~~ ~_____ 0 New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
_______-____~
~ M-1-S-54, UFSAR Section 7.2 PLOT-5060F-4g The auto scram logic does not trip the manual scram logic K/A Statement K4.08 - Knowledge of RPS design feature(s) and/or interlocks which provide for complete control rod insertion following scram Notes -- ~- and Comments:
_____ ~ _ ~ ___~__--__~_____~~~_~
~- ~ - ~- _- ~ _ __ ~ ~ -~ -~ -_ -~~ - -__ - -__ MATERIALS:
~ _ Backup scram valves are not reset until the RPS logic is reset. The 48 second time frame is ,actually associated with the Alternate Rod Insertion (ARI) logic, not RPS. Mode switch reset after 2 seconds just allows the scram to be reset wrth the Mode Switch still in shutdown where it prevents control rod withdrawai
__~________
__ ~ ~ __ I system 1 212000 - Reactor Protection System ! Importance:
RO/SRO Peach Bottom Initial Reactor Operator NRC Examination December 2008 8. A reactor startup is in progress on Unit
- 3. Power is on Range 2 of the WRNMs when a loss of power to the 'A' WRNM chassis occurs. Under these conditions, the loss of power will cause A. a WRNM Triphop alarm to be generated ONLY B. an RPS full scram signal to be generated C. an RMCS rod block AND an RPS half scram to be generated D. the WRNM chassis input to swap to its alternate power supply Peach Bottom Initial Reactor Operator NRC Examination December 2008 Time Allowance (minutes)
________-.-___---
Diff icuk __~____ Question # 8 RO . - - .- -. . . - SRO N/A Choice Correct: Distractors:
C A B D _ __ -~ ____ ~_____ Basis or Justification Loss of +24 VDC power to any WRNM will generate a rod block to RMCS and an RPS input (I/2 scram signal) if in "Refuel" or "Startup/Hot Standby" Mode. The candidate has to correlate that being on Range 2 of the WRNM relates to being in the "Startup/Hot Standby" Mode and not in the "Run" Mode where the WRhlM Triphop would be bypassed. A WRNM Trip/lnop alarm will be generated on a loss of power, however, it is not the only condition.
It also generates a rod block to RMCS and is an RPS input (1/2 scram signal only). _~____ _~____ ~ ~ _______________~~
~___~_ ~ Loss of power to any WRNM will generate a rod block to RMCS and is an RPS input (112 scram signal only). A full scram is not possible with loss of power to just one WRINM. WRNM does not have an alternate power supply, like the QLVPS that the APRMs have. - _-__ ______~_________~_
____- - _- - Source: Reference(s):
Learning Objective:
WA System Source Documentation New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank ~_~_________~_~~_
~ -_ - - ARC 210 G-3 __~~____ __ ~ _- - PLOT-5060C-4a WA Statement K1.02 - Knowledge of the physical connections and/or cause-effect relationships between IRM System and Reactor Manual Control. REQUIRED MATERIALS:
___ __ __ - ~- -~~ ~- ~ ~~ ~ ~~~ ~ - Notes and Comments:
______.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 9. Which one of the following will result if the recirculation calculated total drive flow signal to APRM '1' fails to zero whilc at 100% power? A. Half Scram ONLY B. Rod Block ONLY C. Rod Block AND Half Scram D. Full Scram ONLY Peach Bottom Initial Reactor Operator NRC Examination December 2008 Difficulty Time Allowance (minutes)
Correct: SRO ~.t Distractors:
B ___- A C ~. -_ ___t_~__ Basis or Justification
__~______~___~_~~__
-~ '1' APRM Hi-Hi trip would cause trip input to voters and Hi alarm would cause block 0.65W
+ 54%. One APRM cannot catuse 112 scrams due to 214 voter (trip logic) operation.
-_ ~________~____~__________ ~ ~ _~___~_________________.~__~__.___ - ~ One APRM cannot cause 112 scrams due to 214 voter (trip logic) operation.
Requires any 2 APRM trips to initiate full scram.
Source Documentation Source: New Exam Item Modified Bank Item 0 Previous NRC Exam Other Exam Bank KIA System rlmportance:
RO I SRO KIA Statement K6.07 - Knowledge of the effect that a loss or malfunction of the flow converter
/ comparator network will have on the REQUIRED _- MATERIALS:
____ Notes and
-_ Comments: - ~ _ ~~ ______t______~_~__~-
~ ~ - __t -_ _ ~ - _ _____ ~~ -t Peach Bottom Initial Reactor Operator NRC Examination December 2008 10. Both RCIC and HPCI initiated on Unit 3 low-low RPV water level.
Current plant conditions are as follows: 0 0 0 0 0 Reactor water level is +I 8 inches and stable Reactor pressure is 1040 psig and rising slowly Drywell pressure is 0.8 psig and stable RCIC is in the CST to CST mode at 600 gpm with the flow controller in AUTO HPCI is injecting to the reactor at 1000 gpm with the flow controller in AUTO The PRO reports Torus level is 15' 8" and rising slowly Based on the above conditions which statement below describes (1) RCIC system response, if any, and (2) the appropriate procedure to respond to the condition? A. (1) RCIC will trip on low suction pressure.
(2) Perform SO 13.7.A-3 "Recovery From RCIC System Isolation or Turbine Trip". B. (1) RCIC speed will rise until the overspeed trip occurs.
(2) Perform SO 13.7.A-3 "Recovery From RCIC System Isolation or Turbine Trip". C. (1) RCIC will remain in the CST to CST mode of operation.
(2) Continue to operate the system using RRC 13.1-3 "RCIC System Operation During A Plant Event".
D. (1) RCIC Torus suction valves (MO-3-13-039 and MO-3-13-041) will auto open.
(2) Continue to operate the system using RRC 13.1-3 "RCIC System Operation During A Plant Event".
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 Level __ of Knowledge HIGH ~_____ Choice Correct: 77- Difficulty Time Allowance (minutes) - _ __ ._ N/A Distractors:
Source: ___________
____ Referen!!*
Learning Objective:
A New Exam Item Modified Bank Item [XI ILT Exam Bank
__ Previous NRC Exam 0 Other Exam Bank
.- ARC 221 c-4 ~ -_-_~~~____~_______-_____
~ PLOT-5013-1 c ____~__ .___ - ______ ____-_.______
~ _- ~- Basis or Justification On high Torus level > 15' 6, HPCl suction from CST closes and Torus suction valves open. This swap also causes MO-24 return to CST to auto close thereby removing the RCIC system flow path back to CST. RClC flow controller will attempt to maintain flow at 600 gpm and increase turbine speed until it trips at 125% of rated speed. RCIC suction pressurle will not be affected by MO-24 closure. No sucP.ion valves will reposition.
~_________-_~_______~_.____~_____
~ _. ~- ~ RClC will not remain in CST-to-CST mode. System will trip on mechanical overspeed as flow controller will increase speed to maintain system flow as MO-24 closes. RClC Torus suction vialves do not have an auto open function.
Realigning RClC suction to Torus must be done manually.
~~ - __ ~- __ ~. - - KIA System 1 217000 - RClC 1 Importance:
RO / SRO 1 3.4 13.3 ____ ~ __ KIA Statement A2.03 -Ability to (a) predict the impacts of valve closures on RClC and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or REQUIRED MATERIALS: - operations.
___~ ~~ - Notes and Comments:
~- -~ ~ ~ ~- ~~ ~~- ~~ - -~ -_ --~ ~~____ --
Peach Bottom Initial Reactor Operator NRC Examination December 2008 1 1. Unit 2 is at 100% power when the folllowing occur: 0 0 0 Annunciator 227 C-5 BLOWDOWN VALVES POWER MONITOR alarms. The GREEN indicating light for the 'A' SRV is NOT lit. The GREEN indicating lights for
&LJ other SRVs are lit. Investigation identifies blown fuses associated with the 'A' SRV solenoid.
Based on these conditions, the 'A' SRV (1) open on automatic ADS initiation.
and (2)- open on manual SRV operation.
A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) will NOT (2) will NOT Peach Bottom Initial Reactor Operator NRC Examination December 2008 Correct: D Distractors:
A IB C ~______ _- __~________
__________ - ~- ~ -~ __~ Basis or Justification
_-____ _ The given conditions indicate a loss of 125 VDC to the 'A' ADS SRV solenoid.
In order for this to occur, both the normal and alternate power supplies to the SRV slolenoid must be unavailable. This being the case, both ADS actuation as well as manual operation of the 'A' SRV is defeated. Both ADS and manual operation of the 'A' SRV are defeated. - ~____ _______ ~~ _~ __ ~ - _-_ Both ADS and manual operation of the 'A' SRV are defeated. Both ADS and manual operation of the 'A' SRV are defeated.
________~
~~ Psychiometrics I Level of Knowled e k-,,,,i--
____ Diff iculk I Time Allowance (minutes)
__ - _~_______~__~~___
t SRO N/A ~. ~- Source: Reference( __~_ s) : Learning 0 bject ive: WA System Source Documentation 0 New Exam Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank Modified Bank Item _______ _~__~______~_______~_________
~ ____________~
-~ __ ._ - M-1-S-52, Sheet 3 PLOT-5001 G-6e --~-____~
_~ ~ - 21 8000 - ADS Importance:
RO / SRO 3.7 13.6 WA Statement REQUIRED MATERIALS:
Notes -. and Comments:
-_ A3.07 - AbiliQ - -_______ to monitor of the ADS including
_- lights and
~~ alarms. -~~ ~- __~ ______~ ~ __ -~ ~-_ ____ -~ -~ ~ ~ ~ ~ -______
Peach Bottom Initial Reactor Operator NRC Examination December 2008 12. A Group 1 PCIS isolation is being reset using procedure GP-SA "PCIS Isolation - Group 1". All Group1 isolation signals are cleared. All Group 1 valve control switches are in the target position required for logic reset. Inboard PCIS Reset Switch 16A-S32 is HELD in the "GRP I" position.
What is the status of the following PCIS reset circuit relays for these conditions? Refer to the portions of electrical schematic drawing M-1 -S-23 Sheet 9 on the NEXT TWO PAGES. Relay 16A-K9 is (1) and relay 16A-K11 is (2) . -- A. (1) energized (2) energized B. (1) de-energized (2) energized C. (1) energized (2) de-energized D. (1) de-energized (2) de-energized Peach Bottom Initial Reactor Operator NRC Examination M-1-S--23 Sheet 9 POV, 60HZ INST BUS SH 3, FIG 2 December 2008 Peach Bottom Initial Reactor Operator NRC Examination December 2008 M-1-S-23 Sheet 9 GROUP 1 pN!yq GROUP 31/m GE TYPE CR2940 SPRING RETURN TO "N" GROUP 3 r---- -7 I I 13004 I I I ' I I&/'; NOIRM 1002 -I, 3004 GROUP TI/m r------ I 1002 1 16A-S32 1 16A-533 Peach Bottom Initial Reactor Operator NRC Esxamination December 2008 Source: ___ Reference(s):
Learning 0 bj ect ive: Choice _.________
Correct: New Exam Item [7 Modified Bank Item ILT Exam Bank M-1-S-23, Sheet 9; GP-8.A PLOT-5007G-5a Previous NRC Exam 0 Other Exam Bank -~--____-~-____________~
-~ - _ . ~___ _~_____~______
____~___ - ~~~~ Distractors:
C ___ A B D Basis or Justification
._____.~_~~___~ - Placing Inboard PClS Reset Switch 16A-S32 in the "GRP I" position will close contacts 1-2 anld complete the circuit to energize 16A-K9.
Relay 16A-K11 does not get energized until switch 16A-S32 is placed in the "GRP 11/111" position.
Relay 16A-K1 I does not get energized until switch 16A-S32 is placed in the "GRP ll/lll" positian.
Placing Inboard PClS Reset Switch 16A-S32 in the "GRP I" position will close contacts 1-2 and complete the circuit to energize 16A-K9. Relay 16A-K11 does not get energized until switch 16A-S32 is placed in the "GRP I1/1 I I" position. Placing Inboard PClS Reset Switch 16A-S32 in the "GRP I" position will close contacts 1-2 and complete the circuit to energize 16A-K9.
_____~ __~_____~______~___
__ ~~ -_ ___ _______ ___~___________
~ SRO N/A I Time Allowance (minutes)
~ ~- ~__ .______ K/A System
~ I 223002 - PClS Importance: RO
/ SRO 2.6 / 2.8 WA Statement AI .04 - Ability to predict and/or monitor changes in parameters associated with operating the Primary Containment Isolation System controls includg ______ individualstem
___ relay ______~~~
status. REQUIRED MATERIALS:
-7G -~ ~~ _.__ _~_ _________~_____
~--__ ~~- ~~ _ -- Notes and Comments:
Peach Bottom Initial Reactor Operator NRC Examination December 2008 13. Unit 2 drywell pressure is 3 psig. 0 0 0 Annunciator 214 D-1 Group I1 / 111 INBOARD ISOL RELAYS NOT RESET is in alarm Annunciator 214 E-1 Group I1 / III OUTBOARD ISOL RELAYS NOT RESET is in alarm NO other PCIS signals have been generated Which one of the following components has FAILED to respond to these conditions?
A. 'B' SBGT Fan - red light ON, greien light OFF B. RWCU Outlet Isolation Valve (MO-68) - red light ON, green light OFF C. DW Instrument N2 Supply Valve (AO-2969A) - red light ON, green light OFF D. SBGTS 'A' Filter Train Inlet Valve (AO-00475-1) - red light ON, green light OFF Peach Bottom Initial Reactor Operator NRC Examination December 2008 Reference( s) : Learning Objective: . ________-_-
Distractors:
t - I - A B D Basis or Justification
_~ - This is an incorrect response. Red light LIT and green light OFF indicate the valve is open. AO-2969A should close on a GRP II signal. This is a correct response. Red light LIT and green light OFF indicate the fan is running. B SBGT fan (also A SBGT fan) will start and both filter trains should align in response to the GRP Ill signal. This is a correct response. Red light LIT and green light OFF indicate the valve is open. RWCUI does not isolate on high DW pressure.
___ -~ ~ -~ __~_~_~___
__~-~ __ - This is a correct response. Red light LIT and green light OFF indicate the valve is open. SBGTS 'A Filter Train Inlet and Outlet Valves should be oDen on auto initiation of SBGT. Psychometrics I Level of Knowledge hGH Difficulty
____ -- I Allowance (minutes)
SRO -~~ NIA Source : ____-_____
KIA Statement Source Documentation 0 New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
~_~______~
_______~-__________
---~ -~- GP 8.B PLOT-5007G-3x 223002 - PClS - I-tance: RO / SRO 3.4 13.4 -~
Peach Bottom Initial Reactor Operator NRC E:xamination December 2008 14. Unit 2 was initially at 100% power. A complete loss of Instrument Air occurred.
NO operator actions have been ta.ken.
Which one of the following correctly describes the pneumatic supply to the Safety Relief Valves (SRVs) for these conditions?
A. The CAD tank is supplying nitrogen to the SRVs.
B. ADS accumulators are the only pneumatic supply available to the SRVs. C. The Instrument Nitrogen system i.s supplying the SRVs via the Instrument Nitrogen header. D. The Backup Instrument Nitrogen to ADS System is supplying the SRVs.
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 N/A Choice Source: Correct: I B 0 New Exam Item Modified Bank Item 0 Previous NRC Exam 0 Other Exam Bank Distractors:
--~A -~ ~-~___ -~ Reference(s):
Learning Objective:
WA System ____________~___ - - _- _______ - - Basis or Justification
_____~_~ __~__ ___~_ ~ -~ ____ ~ With no instrument air, the AO-2969 valves fail closed, which means i IO pneumatic supply to !3RVs except ADS accumulators. The CAD tank supply has to be manually valved in. The question stern specifically states there have been NO operator actions taken. ____~_~ ___~_~_~_______~_~_~
~ -~ ~~ ~ -_~ ~ ~ _~_ ~ ON-119; SE-11 Bases, Step 13-4 ~- -- PLOT-5036-3a
_~ _____~ ~_ ~ - 239002 - SRVS Tlmportance:
RO / SRO 3.1 / 3.3 ~___ ~- ~ ~ __~ With no instrument air, the AO-2969 valves fail closed, which means no pneumatic supply to SRVs via the Instrument Nitrogen header.
__ __ __._~_______~
~ The Backup lnstrumeint Nitrogen to ADS System must be manually valved in-service from the main control room. The question stem specifically states there have been NO operator actions taken. ~ ~~ __ ~~ - -
Peach Bottom Initial Reactor Operator NRC Examination December 2008 15. Unit 2 reactor startup is in progress.
0 RPV pressure is 450 psig with 3 bypass valves open. The 2C RFPT is being placed in service using SO 6C. 1 .A-2 "C Reactor Feedwater Pump Startup With Vessel Level Control Established Through AO- 8091". MSC SELECT is lit for the 2C RFPT on Panel 20C005A.
2C RFPT is on turning gear.
0 0 In accordance with procedure SO 6C. 1 .A-2, pressing and releasing the RFPT "AUTO START" pushbutton at this time will raise RFPT speed to the (1) and -.--(2)-.-
be aborted by pushing any speed "LOWER" or "RAISE" pushbutton.
A. (1) minimum governor control speed of approximately 400 to 600 rpm (2) cannot B. (1) Low Speed Stop setting of approximately 2600 to 2900 rpm (2) can C. (1) minimum governor control speed of approximately 400 to 600 rpm (2) can D. (1) Low Speed Stop setting of approximately 2600 to 2900 rpm (2) cannot Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 Time Allowance (minutes)
________ Difficulty
____~____~_~~ - Choice SRO ~- Correct: ~ B Source: -. TA Distractors: New Exam Item Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam Other Exam Bank
-- ___~_ _ IC Reference(s):
Learning 0 bjective:
ID - __ ____ -_ SO 6C. 1 .A-2 PLOT-5006-4a
~~ __ -_______ ____..~_ Basis or Justification When the "AUTO START" pushbutton is depressed then the RFPT wi I ramp to the Low Speed Stop (LSS) setting of 2600 to 2900 rpm. The auto start can be aborted by pushing any speed LOWER or RAISE pushbutton and the turbine speed1 will be controlled at the speed the turbine is at when the button was pushed. ___-_______ - - _ ~____-_ - ____ _______ - Speed range is not correct for pushing the "AUTO START pushbutton. Minimum governor control speed is correct for depressing the "SLOW' or "FAST RAISE" pushbiuttons.
Speed range is not ccrrect for pushing the "AUTO START pushbutton. Minimum governor control speed is correct for depressing the "SLOW or "FAST RAISE" pushbuttons.
_____________~_____.~_
__~ ~ _____ - The auto start can be aborted by pushing any speed LOWER or RAISE pushbutton and the turbine speed will be controlled at the speed the turbine is at when the button was pushed. _~_____.~
___ FA System 259002 - Reactor Water Level Control KIA Statement Importance:
RO / SRO 3.8 13.6 ~.. I A4.01 -Ability to manually operate andlor monitor in the control room: All individual component REQUIRED MATERIALS: controllers in
__ the manual mode. - _ ~ _- ~ __ ~ __~-~ ______ ____ - - ___ _~ _ _ ~~ -~ - - __ _- ~- -- - ~~ ~ _ Notes _~ and Comments:
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 16. A transient resulted in RPV level lowering to
-50 inches on Unit 3. What is the status of the Standby Gas Treatment (SBGT) System for this condition?
A. SBGT fans B and C initiated with filter train B ONLY aligned for flow.
B. SBGT fans A, B, and C initiated ,with filter trains A and B aligned for flow. C. SBGT fans A and C initiated ONLY with filter trains A and B aligned for flow.
D. SBGT fans B and C initiated ONLY with filter trains A and B aligned for flow.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 SRO 4 ----____ Level of Knowledge . ._________
Difficulty Time Allowance (minutes)-
--__~.__- - FUNDAMENTAL NIA Psychometrics Answer Key i- -. __. .~ ' Question # 16 RO . - - - .. . Choice ._____~____
Correct: Distractors:
D A B C ___________~________
~- Basis or Justification For Unit 3, the B&C SBGT fans will auto start and the 'A' and 'B' filter trains will align for service on RPV level 1". __-___ ______~____~____________
~ ____ The 'A' SBGT filter train will also align for system flow. 'A' and 'B' filter train align for service with an initiation from either unit.
The 'A SBGT fan will not auto start in response to a Unit 3 event, only Unit 2. The 'A' SBGT fan will not auto start in response to a Unit 3 event, only Unit
- 2. -________
_~-~____~___-~- - -_ _ Source: ____~ Reference(s):
Learning Objective:
WA System ___..~_____
Source Documentation New Exam item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank Modified Bank Item _ ___ _______ ~-~ ~ __ _~______~_~_______
~ __ SO 9A.1 .C PLOT-5009A-4a Importance:
RO I SRO 3.7 13.8 -~ ___ __ 261 000 - SGTS __ K/A Statement K4.01 - Knowledge of SBGT design fFature(s) and/or interlocks which provide for automatic systern initiation. REQUIRED MATERIALS: Notes and Comments:
____________~
-~ ~- - __ ~~ __ - ~ - ~_________
__~_____~__
Peach Bottom Initial Reactor Operator NRC E:xamination December 2008 17. A loss of all off-site power (LOOP) occurred.
0 0 0 0 Unit 2 and Unit 3 were manually scrammed.
Unit 2 RPV water level is -35 inches and rising slowly.
Unit 3 RPV water level is -20 inches and rising slowly. Emergency Diesel Generator output breaker E-12 failed to close.
Based on the above conditions, which one of the following identifies the status of the SBGT fans? A. 'A' fan is running 'B' fan is running 'C' fan is running B. 'A' fan is running 'B' fan is off 'C' fan is running C. 'A' fan is off 'B' fan is running
'C' fan is off D. 'A' fan is off 'B' fan is running 'C' fan is running Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge Difficulty
_ ____ ~~ Time Allowance
~- (minutes)
-~ Correct: SRO N/A -~ ~ ~ ~_ ID ___ Distractors:
L- IC - ~__ ~__ -~____ - Basis or Justification All three fans started on 4 inch low RPV level condition.
The 'A SBGT fan is powered by the E-I EDG from the E-12 4kV bus through E-124-T-E3 and would be off sine E-12 bus is de-energized. The
'6' SBGT fan is powered by the E-2 EDG from E-22 bus through E-224-T-B and would be unaffected.
The 'C' SBGT fan is powered by the E-3 EDG from E-33 bus through E-334-R-B arid would be unaffected.
_~~-____ ~ ~ ___~ The 'A' SBGT fan would NOT be running since it is powered from E-12 bus through load center E-124-T-B and the E-12 bus is de-energized.
The 'A' SBGT fan would NOT be running since it is powered from E-12 bus through load center E-124-T-B and the E-12 bus is de-energized.
'6' and 'C' fans would be running since they are powered from E-22 and E-33 busses respectively.
'C' fan would be running since it is powered from E-33 bus.
________.~_~__________~~_~_
~ -~ ~ ~ -- Source Documentation New Exam Item 0 Modified Bank Item Previous NRC Exam 0 Other Exam Bank
__________ - --_ _~_____ ~___ P LOT-5009A-2b
-~ 3.0 / 3.1 - __- __~__ __ -_ ________~
...________-_____.
_ ~- ~- _-_____- ~ ~- _-___ ~~ Source: Learning Objective:
K/A System K/A Statement K6.03 - Knowledge of the effect that a loss or malfunction of the Emergency Diesel Generator System REQUIRED MATERIALS:
Notes and Comments: - _- ~~ ______~~~
have on __ the Standby Gas treatment %stem.
-________________
~- ~
Peach Bottom Initial Reactor Operator NRC Examination December 2008 18. Unit 2 is operating at 100% power. 0 0 The PRO manually taps down the: 2 Startup Transformer 00x003 by placing the Load Tap Changer (LTC) control switch to "LOWER. After releasing the LTC control switch the LTC continues to LOWER for another 15 seconds before stopping. The voltage on the normal offsite feeder for the E-12 bus degrades to 89% of rated voltage and remains steady at that value. The PRO checks the status of the E- 12 Bus after 2 minutes have elapsed.
In accordance with ARC 001 11-1 "E12 Bus Undervoltage", the PRO would find the E-12 Bus energized from the -( l)-.- and the crew will need to reset an
-(2)- isolation.
A. (1) alternate offsite feed (2) Outboard Group I1 B. (1) E-1 Diesel Generator (2) Outboard Group I1 C. (1) alternate offsite feed (2) Inboard Group I1 D. (1) E-1 Diesel Generator (2) Inboard Group I1 Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge Difficulty
__ ______ FUNDAMENTAL
______. Choice SRO NIA Time Allowance
~- (minutes)
~~~ Correct: - ARC-001 D-1; SO -~~ 54.7.A PLOT-5054-6b Distractors:
C A B D __~~______ ~~ ~ ~ Basis or Justification
~~_____________
~~ ~- The off-site feeder breaker (E-212 or E-312) will trip if supply voltage degrades to 99.8%
for nominally 61 seconds with NO LOCA signal present. The E-12 bus will be supplied via the alternate feeder breaker (fast transfer will occur). The E-124 load center supply breaker opens on the load shed and results in a loss of 20Y033 panel and a subsequent Inboard Group II isolation due to the power loss ___~~ of PClS relays. While the E-12 bus transfers to its alternate feed, an outboard Group II isolation does not occur.
~ ___________ E-12 transfers after 6'1 seconds (127E relay);
E-I D/G does not start.
Also, an outboard Group II iisolation does not occur. E-I 2 transfers after 61 seconds (1 27E relay). I ~ I Source: 0 New Exam Item Previous NRC Exam Modified Bank Item ILT Exam Bank - 0 Other Exam Bank Reference(s):
__. Learning Objective:
1 262001 - AC Electrical Distribution Fmportance:
RO / SRO 2.9 I 3.1 ~IL~.~~ ~ WA Statement AI .03 - Ability to predict andlor monitor changes in bus voltage associated with operating the A.C Electrical Distribution controls.
~~ REQUIRED MATERIALS: Notes and Comments:
.. . ~~~ ~_ ~~ __._~.~._______.~..___~
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 19. Both Units are at 100% power. The 4KV System is in a normal 1 ine-up, except for the E-3 12 breaker, which is racked out for elevator mechanism preventative maintenance. The PRO inadvertently opens the E-212 breaker, de-energizing the E-12 bus.
Based on these conditions, the E-1 EDG -(l)- automatically start. Entry into Technical Specification 3.8 "Electric,al Power Systems" is (2) A. (1) will (2) required due to the E-12 bus being INOPERABLE B. (1) will (2) NOT required due to the E-12 bus remaining OPERABLE C. (1)willm (2) required due to the E-12 bus hleing INOPERABLE D. (1) will NOT (2) NOT required due to the E- 12 bus remaining OPERABLE Peach Bottom Initial Reactor Operator NRC Examination December 2008 -~ -__ Level of Knowledge HIGH Question # 19 RO .- ______ Difficulty
~_____~___________
Time Allowance (minutes)
~ _~ N/A Choice Correct: Distractors:
C A B - _________~
~ ~- Basis or Justification The E-212 and the E-312 breakers are the normal and alternate feeder breakers for the E-I2 bus. With E-312 breaker racked out and E-212 breaker taken to OPEN (both breakers "green-flagged"), EDG auto-start is defeated.
Tech Spec (3.8.7) entry is required since the E-I2 4kV bus is considered inoperable when deenergized.
The E-212 and the E-312 breakers are the normal and alternate feeder breakers for the E-I2 bus. With E-312 breaker racked out and E-212 breaker taken to OPEN (both breakers "green-flagged"), EDG auto-start is defeated.
__ __ ______~_ ~~ - . ____ ___ ~ --__ - ~ _ -- -~ The E-212 and the E-312 breakers are the normal and alternate feeder breakers for the E-I2 bus. With E-312 breaker racked out and E-21 2 breaker taken to OPEN (both breakers "green-flagged"), EDG auto-start is defeated. Tech Spec (3.8.7) entry is required since the E-I2 4kV bus is considered inoperable!
when deenergized.
Tech Spec (3.8.7) entiry is required since the E-I2 4kV bus is considered inoperable when deeniergized.
___-___._____~_
__ Source Documentation Source: --_____ Reference@
-~ Learning Objective:
WA System c] New Exam Item Modified Bank Item 0 Previous NRC Exam 0 Other Exam Bank ILT Exam Bank
__ ~____ PLOT-5054, SO 52A.1 ~__ B; Tech ~ Spec 3.8.7 PLOT-5054-04b 262001 - A.C. Electrical Distrilbution Importance:
RO / SRO ~ 3.1 14.2 WA Statement G2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on REQUIRED MATERIALS:
the status of limiting Notes and Comments:
._~.._____
___.~__
Peach Bottom Initial Reactor Operator NRC Examination December 2008 20. The following conditions are present on Unit 2: 0 0 The 20Y050 panel was aligned to the bypass (alternate) source using the Static Switch "Load to Bypass" pushbutton.
Subsequently, the alternate supply breaker on E-1 24-R-C is inadvertently opened.
For these conditions, the Static Switc,h (1) automatically transfer the 20Y050 panel back to the Static Inverter output and 20Y050 panel power will be (2) A. (1) will (2) temporarily interrupted due to Static Switch operation B. (1) will (2) maintained during Static Switch operation C. (1) will NOT (2) lost until the Static Switch "Load to Inverter" pushbutton is operated D. (1) will NOT (2) lost until the Manual Bypass/lsolation Switch (MB/IS) is placed in the "BYPASS" position Peach Bottom Initial Reactor Operator NRC Examination December 2008 Source: ~ _~________.
Reference(s):
Objective:
~~_______
__ Learning Choice 7- _____~ Correct: IC 0 New Exam Item 0 Modified Bank Item ILT Exam Bank _~_ 0 Previous NRC Exam Other Exam Bank ARC 220 F-5; PLOT-5058 PLOT-5058-5c
___. ~~-~____~.____~
__ ~~ ~~ ~__ __ ~~ a- ________ Distractors:
IB i Basis or Justification
______-______
~. IF manually transferred it will not automatically transfer back. - Static Switch only auto transfers to the Inverter if it first "auto" transferired to the alternate.
Static Switch only autio transfers to the inverter if it first "auto" transferred to the alternate.
-- ~ -~ __ This will transfer to alternate which is de-energized.
Time __ Allowance (minutes)
-_ SRO __. 7 -- N/A Importance:
RO I SRO 2.7 12.9 I ~ ~ A -~ -~ -~ - 262002 - UPS (ACIDC) KIA System KIA Statement K6.01 - Knowledge of the effect that a loss or malfunction of AC electrical power will have on the UPS ___ IAC/DC)_ ______~ __ ______ ~ - ~_~-- REQUIRED MATERIALS:
__ -- Notes and Comments:
__ ~~ -~~ - -~ ________~
-~
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 21. Unit 3 was at 100% power when the following conditions occur:
0 0 0 0 0 220 F-5 Inverter Trouble" alarms 210 C-1 "SDV Hi Water Level Trip" alarms 21 1 B-1 "A Channel Reactor Auto Scram" alarms 21 1 C-1 "B Channel Reactor Auto Scram" alarms A loss of Control Rod Position Indication on the Full Core Display Based on the above conditions, reactor power is monitored from (1) and if it's less than 4%, procedure
--( 2)- is entered.
A. (1) 20C05 WRNM Operator Displays (2) T- 1 0 1 "RPV Control" B. (1) 20C05 APRM Operator Displays (2) T- 1 00 "Scram" C. (1) Safety Parameter Display System (SPDS)
(2) T- 1 0 1 "RPV Control" D. (1) 20C036 WRNM Chassis (2) T- 100 "Scram1' Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge HIGH .-______ _- Choice Correct: _~_______-
Difficulty
__ Time ~- Allowance (minutes)
_____ __ N/A Distractors:
i--- - Source: Reference(s): - -_ Basis or Justification
_ --~-______
_____ _- - __ ___ With a loss of uninterruptible power, the WRNM indicators on the bacrc panel (20C036) are operable.
With power less than 4%, T-100 is entered 20C05 operator displays fails as is. If power is less than 4%, even with an ATWS, T-100 entered, not T-101.
20C05 APRM Operator Displays fail as is on a loss of uninterruptible power. SPDS is not allowed to be used to determine reactor power for decision making out of the TRIP procedures. T-101 is the wrong TRIP to enter if power was less than 4%. -____--~-
_____~____
~ - _ -__ _________________~
.-~~_______
__ ______ __~_________
_____ - -_ - _ -- -_ 0 New Exam Item 0 Modified Bank Item 0 Previous NRC Exam 0 Other Exam Bank 1x1 ILT Exam Bank
-____~-__ - ~- T-100; ON-I 12-3; ARC 220 F-5 -______~____-~~_____---__---____
Learning Objective:
PLOT-1560-11 KIA System ~ 262002 - UPS (AC/DC) -- ___ Importance:
RO I SRO 2.6 12.8 ~____ A2.01 -Ability to (a) predict the impact of undervoltage on the Uninterruptible Power Supply (ACIDC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions o-erations.
____~~__~
-~ - - - REQUIRED MATERIALS: Notes and Comments: - ~_ ________.~
~ - - ~~ ~ -- __ ~- _____ -
Peach Bottom Initial Reactor Operator NRC Examination December 2008 22. An electrical fault has resulted in the loss of the Unit 3 Div. I Station 125 VDC distribution panel 30D23 (3PPC). A few minutes later, a LOCA signal :provides an automatic start signal to the Emergency Diesel Generators. Which one of the following is the diesel response resulting from the loss of DC power? A. E- 1 diesel will NOT start automa tically. B. E-1 diesel will start but NOT field flash.
C. E-3 diesel will NOT start automatically.
D. E-3 diesel will start but NOT field flash.
Peach Bottom Initial Reactor Operator NRC E>xamination December 2008 Choice Correct: -_________
Distractors:
C A B D Basis or Justification - ~____~~___
~ __.~_ ~ Div I 125 VDC Panel 3PPC supplies the E-3 EDG 125VDC logic and solenoid power.
E-3 will not start. The E-1 EDG will auto start. E-1 control power is from 20D21 panel and not the 30D23 panel. The E-I EDG will auto start and be able to flash its field as designed.
E-1 control power is from 20D21 panel and not the 30D23 panel. _____________
~ - ~ -______~_~-._~~__________~_~ ~~ ~~ E-3 EDG cannot start since it has lost all control power.
Source: Reference(s):
Learning 0 bject ive
- WA System Source Documentation 0 New Exam Item 0 Previous NRC Exam Modified Bank Item E ILT Exam Bank 0 Other Exam Bank
~- ____ ___~_____~~
E-27, Sheet I, E-26 Sheet 2, PLOT-5052
___________-_____________.____~_-____--__-~-
~ ~~-~ PLOT-5052-2~
263000 - D.C. Electrical Distribution Importance:
RO / SRO 3.4 13.8 WA Statement K3 01 - Knowledge of the effect that a loss or malfunction of the D.C. Electrical Distribution will have on Emergency Generators. - -~ ~ -~ ~~ ~___~-~ ~ REQUIRED __ -~__ MATERIALSF[= ~~ ~~ Notes and Comments:
~-
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 23. Given the following:
0 0 0 0 0 0 0 Unit 2 was at full power when a loss of all off-site power occurred. Diesel Generator E-1 failed to start. RPV level is -10 inches and steady. Reactor pressure is 950 psig. CRD ACCUMULATOR LO PRI?SS/HI LEVEL (21 1 E-2) is in alarm.
2A DC POWER PANEL LO VOLTAGE (209 C-3) is in alarm.
2A DC Bus voltage at Panel 20C021 (CSR) is 90 VDC.
Which one of the following actions is required to be performed for these conditions?
A. Determine plant impact of low DC Bus voltage in accordance with SE-13 "Loss of a 125 or 250 VDC Safety Related Bus".
B. Restart the 2A CRD Pump in accordance with SO 3.1 .-2 "CRD Hydraulic System Startup with the System Filled and Vented".
C. Place the alternate 2A battery charger in service in accordance with SO 57B. 1-2 'I 125/250 Volt Station Battery Charger Operations".
D. Transfer the 2A battery charger power source from E-124-T-B to E-134-T-B in accordance with A0 57B.6-2 "Transfer of 125V Battery Charger 2AD003 to Alternate Power and Return to Normal".
Peach Bottom Initial Reactor Operator NRC Examination December 2008 - Choice - ___~____~____
Correct: A _-~___ __~. Distractors:
B ~____ _-____ _______-.____~
~ ___ ~ -~~ Basis or Justification
_____~_~ -.__.____
This is an SE-13 entry condition
... the referenced alarm and voltage or1 a safety-related 125 VDC distribution panel less than 107.45 VDC requires entry into SE-13. Cannot start the 2A CRD Pump due to no power available to the E-12 bus. ~___-_____~~_~.__-__-~-~ ~ ~ __-- SRO '-1 Psychometrics
-_____ Level of Knowledge Difficulty Time __- Allowance (minutes) ~ ~ ~~ HIGH N/A -____-- ___-___ ~ ~ Source Documentation Source : New Exam Item 0 Previous NRC Exam I Modified Bank Item ILT Exam Bank Other Exam Bank
--_______.____
____~_,________
Reference( s): Learning PLOT-1555-1 ARC 209 C-3; SE-13;
-- A0 57B.6-2 __. ~- ________-~
_____ Objective:
I system 1 263000 - D.C. Electrical Distribution I Importance:
RO I SRO GI .2.23 - Ability to perform specific system and integrated plant procedure during all modes of -- operation. REQUIRED MATERIALS:
___ - - ~-___ ~~ __ ~~ ~- ~ - ___ ~~ --____~~ _____~ ___--~-~ - ~ ~-~ ~~ ~-- t Notes and Comments:
Peach Bottom Initial Reactor Operator NRC E;xamination December 2008 24. The E-4 Diesel Generator is paralleled to the 2 Emergency Auxiliary Transformer (OAX004) and Bus E42 and loaded as follows: 0 1000 kW 0 200KVARS Assuming the load on the system was constant, how would real load and reactive load be expected to respond if the E-4 Diesel Generator Voltage Control switch was momentarily placed in the
'RAISE' position?
Real Load (KW)
Reactive Load (KVARS) Remain approximately the same A. Raise B. Lower Raise C. Remain approximately the same Raise D. Remain approximately the same Lower Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 KIA Statement K5.05 - Knowledge of the operational implications of paralleling A.C. power sources as they apply the Emergency Diesel Generators.
~ ~- ~~ -- _- . _~ ~- -______ REQUIRED MATERIALS:
LONE Correct: constant grid load, real load would remain constant. Increasing the excitation of the generator would cause its reactive load to increase. will not be affected by changes in generator excitation.
_____~ _______ __ With a constant grid load, real load would remain constant. Increasing the excitation of the generator would cause its reactive load to increase.
Distractors:
1 B I With a constant grid load, real load would remain constant.
_______~____
_- Increasing the excitation of the generator would cause its reactive load to increase.
I Psvchometrics I ~____ 1 FUNDAMENTAL Difficulty
__ Time Allowance (minutes)
SRO NIA ___ Source: _____~ Learning KIA System Source Documentation 0 New Exam Item 0 Modified Bank Item ILT Exam Bank SO 52A.1 .B, Section 4 2 0 Previous NRC Exam 0 Other Exam Bank
___~~ ~___ -_ ____~ ___ ~ ~~ ~ - - PLOT-5052-5b 264000 - EDGs Importance:
RO I SRO ~ I 3.4 13.4 ~___________
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 25. Unit 2 is experiencing a loss of Instrument Air transient.
Per procedure ON-1 19 "Loss of Instrument Air", as Instrument Air pressure LOWERS from its normal value, the MAXIMUM pressure when the Backup Instrument Air Compressor 2DK00 1 will automatically start is (1) and the MAXIMUM pressure when the Backup Air Control Valve (AO-80250D) will automatically open is -(2)--. A. (1) 90 psig (2) 90 psig B. (1) 90 psig (2) 100 psig C. (1) 100 psig (2) 90 psig D. (1) 1OOpsig (2) 100 psig Peach Bottom Initial Reactor Operator NRC Examination December 2008 Question # 25 RO Choice Correct : - C A B D _ ~ ~________p___~
-~ -~~ __ Basis or Justification
_______~ ___p_________________________--
~- The Backup Instrument Air Compressor 2DK001 will automatically start when both the 'A' and
'8' Instrument Air receiver pressures drop to 100 psig. AO-80250D will open when both the 'A' and
'8' Instrument Air receiver pressures drop to 90 psig. The Backup Instrument Air Compressor 2DK001 will automatically start at Instrument Air Receiver pressure of 100 psig, not 90 psig. The Backup Instrument Air Compressor 2DK001 will automatically start when both the 'A' and 'B' Instrument Air receiver pressures drop to 100 psig. AO-80250D will open when both the 'A' and 'B' Instrument Air receiver pressures drop to 90 psig. ~ - ______ __p_______~______~.____~
___~ . ~__ AO-80250D will open when both the
'A and 'B' Instrument Air receiver pressures drop to 90 psig as sensed by PS-2481 A and B. Psychometrics I Difficulty
.~ __ ! Time Allowance (minutes)
SRO NIA I Source Documentation 0 New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
_-_~________
____ ____ ~~- -~ _~ ON-I 19 ______~ ~- ~ _~- Source: _____ -p~ ____ Learning Objective:
I system 300000 - Instrument Air I Importance:
RO I SRO K4.01 - Knowledge of Instrument Air System design feature(s) and /or interlocks which provide for the _______ manuallautomatic transfers of control. REQUIRED MATERIALS:
Notes and Comments: - __ -~ ~- ~ __--_ ~ _____~_____~__~_ - -- _______~__
__~ .~ ~ - -~ ~__ -~__
Peach Bottom Initial Reactor Operator NRC E:xamination December 2008 26. Unit 3 is operating at 100% power. 'The Drywell Chilled Water (DWCW) System is aligned as follows: 0 0 0 3A and 3B Drywell Chillers in scwice 3B and 3C DWCW Pumps running 3C Drywell Chiller in STBY 3A DWCW Pump in STBY The #4 Auxiliary Bus becomes de-energized and can NOT be restored.
Assuming no operator action, which one of the follow describes the response of the DWCW System? The 3A Drywell Chiller will remain running and (1) continue supplying DWCW loads. Emergency powered swap-over valves (2) reposition to allow RBCCW to supply DWCW loads.
A. (1) will (2) will NOT B. (1) will NOT (1) will NOT C. (1) will (2) will D. (1) will NOT (2) will Peach Bottom Initial Reactor Operator NRC Examination December 2008 Choice Correct: -- Distractors:
B C -~ ~ Basis or Justification
__~_____ -~- _- The 38 and the 3C Diywell Chiller Units are powered by the #I Aux. Bus. On a loss of the #I Aux. Bus both of these coolers are unavailable AFJD emergency powered nnotor-operated transfer valves and air-operated blocking valves will reposition allowing the RBCCW System to provide automatic back up cooling water flow to the DWCW System in the event of a loss of power to 2 of the 3 chiller units.
Loss of power is sensed by (an undervoltage device located in the 480V Load Center, upstream of the chiller breakers.
3C Drywell Chiller will be unavailable as it is powered from
- M Aux Bus. Emergency powered motor-operated transfer valves and air-operated blocking valves will reposition allowing the RBCCW System to provide automatic back up cooling water flow. 3A Chiller will be running, however, RBCCW will be supplying loads. Emergency powered motor-operated transfer valves and air-operated blocking valves will open allowing the RBCCW System to provide automatic back up cooling water flow to the DWCW System.
____~ _~ In the event of a loss of power to 2 of the 3 chiller units (3B and the 3C; Drywell Chiller Units) (emergency powered motor-operated transfer vatves and air-operated blocking valves will reposition, preventing the remairing 3A Drywell Chiller frorn supplying DWCW loads. ____~_ ~____________ ~ ~ ~ ~. Difficulty - Time Allowance (minutes)
___. Learning Objective: K/A System K/A Statement Source Documentation 0 New Exam Item Modified Bank Item ILT Exam Bank
[7 Previous NRC Exam 0 Other Exam Bank
-____~______~- ~ ~~ ~ ~- A0 44A. 1-3 PLOT-5044-2 b -__________~-~~
~ _________~
Importance:
RO / SRO 2.9 12.9 400000 - Component Cooling Water
.~ ~~ K2.02 - Knowledge of electrical power supplies to CCW valves.
__ -~ _____~..~~
REQUIRED-MATERIALS:
--E Peach Bottom Initial Reactor Operator NRC Examination December 2008 27. Alternate Rod Insertion (ARI) solenoids are powered from (1) and (2) to actuate. A. (1) 125 VDC (2) de-energize B. (1) 125 VDC (2) energize c. (1) RPS (2) de-energize D. (1) RPS (2) energize Peach Bottom Initial Reactor Operator NRC Examination December 2008 Diff icultE Time -~____ Allowance (minutes)
Choice SRO - Correct: FUNDAMENTAL Distractors:
N/A B A C D Basis or Justification
__ -_____________________.~~ ~ ~~ ~ - ~ ARI valves are powered from DIV 1/11 125 VDC and energize to actuate from manual pushbuttons or upon auto initiation signal of either -48" or 1106 psig RPV pressure.
ARI valves energize to actuate from manual pushbuttons or upon auto initiation signal of either -48" or 1 106 psig RPV pressure.
~ ARI valves are powered from DIV 1/11 125 VDC, NOT RPS, and energize upon initiation signal of either -48" or 1106 psig RPV pressure.
+I" is the RPV level associated with RPS scram and PClS Group I1/111 isolations.
ARI valves are powered from DIV 1/11 125 VDC, NOT RPS, and energize upon initiation signal of either -48l or 1106 psig RPV pressure.
Source : Reference(s):
________.~__
Learning Objective:
KIA System -____-____
WA Statement Source Documentation 0 New Exam Item Modified Bank Item ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
-~ -~ .~ ~ _____ ___ _~___ -~- ____ ~__. E-3030, Sheet
~ 1 PLOT-5003A-5h 201 001 - CRD Hydraulics Importance:
RO / SRO 4.5 / 4.5 - - - --~- - ~ K2.05 -~~ - - Alternate Rod Insertion (ARI)
-- - valve - ~ solenoids;
~~ - -- -- ~ ~~ ~ - - REQUIRED MATERIALS:
Notes and Comments:
___-_ ._~______~ - -~ ~ ~~~ -~ -- ~ __- - --__-___._____-
Peach Bottom Initial Reactor Operator NRC Examination December 2008 28. The Control Rod Select Power Switch supplies power to the (1) . [t is the onlymethod of (2) . A. (1) Rod Select Matrix (2) manually deselecting all control rods B. (1) Rod Select Matrix (2) resetting a control rod drift alarm C. (1) Four Rod Display (2) resetting a control rod drift alstrrn D. (1) Four Rod Display (2) manually deselecting all control rods Peach Bottom Initial Reactor Operator NRC Examination December 2008 Time Allowance (minutes)
~- _~______._.~
Difficulty
____~_ _____~- Level of Knowledge Choice ~--______
Correct: SRO ~ __~ - Distractors:
~_-____~ Reference(sL Learning 0 bjective:
A __ B C D ~_ ~-~ _ _-_ ~-___-___
Basis or Justification The rod select power switch supplies power to the control rod select matrix AND is the ONLY method of manually de-selecting all control rods. Resetting a control rold drift alarm is the function of the rod drift alarm test switch, not the rod select power switch. The four rod display is not powered by the control rod select power. Resetting a control rod drift alarm is the function of the rod drift alarm test switch, not the rod select power switch. ~ .________-__-
_- -~ .___ ..___ -- The four rod display is not powered by the control rod select power.
-______~_______
Source: New Exam Item 0 Previous NRC Exam Modified Bank Item 0 ILT Exam Bank [I] Other Exam Bank
___ ____ 1 system SO 62.1 .A-2 PLOT-5062-5
-. ________.___~~___ 201 002 - RMCS ___.___ L. I KIA Statement 1 Importance:
RO / SRO A4.06 -Ability to manually operate andlor monitor the rod select matrix power switch in the main
-~ - -___~ ~ ____ - _ --~_ ~~ - ~__ ~- - _ ~ Notes and Comments:
~~-____ _______~___
Peach Bottom Initial Reactor Operator NRC Examination December 2008 29. Unit 2 reactor startup is in progress: RPV pressure is 150 psig. Control rod 30-3 1 is at target position 12.
The accumulator for HCU 30-3 1 developed a water-side leak. An Equipment Operator immediately closed HV-2-3A-13 1 13 "Charging Water Riser Isolation Valve" for HCU 30-3 1 to isolate the leak.
HCU 30-3 1 accumulator pressure lowers to IO0 psig as read on PI-13-13 1. One minute after HV-2-3A-13 1 13 was closed, a reactor scram occurred. Control Rod 30-3 1 (1) scraim because (2) . Refer to the portion of P&ID M-357 ;Sheet 1 "Control Rod Drive Hydraulic System" on the NEXT PAGE. A. (1) will (2) reactor pressure is high enough ONLY B. (I) will (2) reactor pressure and accumulator pressure together are high enough C. (1) will NOT (2) reactor pressure and accumulator pressure are too low D. (1) will NOT (2) accumulator pressure ONLY i,s too low Peach Bottom Initial Reactor Operator NRC Examination P&ID M-357 Sheet 1 December 2008 Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowled e Diff icuJy Choice __ __ ___ Correct: !c Time Allowance (minutes) - SRO NIA ti- _-________
Distractors:
Source : Reference(s):
lB New Exam Item Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
___ _________~~__~______~_____~ - ~ -~ M-357; UFSAR, PLOT-5003A
_~ __ ~~ ~ ___~ ~~ ____~~_________________~
_- _ ~- __ Basis or Justification
_____________~____________ - Reactor pressure is oinly 150 psig and accumulator pressure is only 130 psig. The control rod will not scram with pressure below 400 psig due to the net initial upward force to scram the drive being zero. Reactor pressure is oinly 150 psig and accumulator pressure is only 100 psig. The control rod will not scram with pressure below 400 psig due to the net initial upward force to scram the drive being zero. Reactor pressure is only 150 psig and accumulator pressure is only 100 psig. The control rod will not scram with pressure below 400 psig due to the net initial upward force to scram the drive being zero. ~___~~____~______________________
__ - ~- _______--~____
_~ -~ ~ ~ _~ Learning PLOT-5003A-5 0 bj ect ive:
KIA System KIA Statement
___ ~ ___ Importance:
RO / SRO - 3.0 13.0 r ~ __ --_____ _~____________
201 003 - Control Rod and Drive Mechanism
_~__ __ __ --____~__
Both reactor pressure and accumulator pressure are too low in pressure.
The control rod will nolt scram with pressure below 400 psig due to the net initial upward force to scram the drive being zero. ~~_~_~~____
~_______.~-_~
_~ -~
Peach Bottom Initial Reactor Operator NRC E.xamination December 2008 30. A hydraulic ATWS exists on Unit
- 2. As power is reduced, a Rod Worth Minimizer (RWM) insert rod block occurs. Control rods can be inserted by ____ A. placing the Rod Control switch
(.3A-S2) in the "IN" position B. venting the scram air header IAU' T-214 "Isolating and Venting the Scram Air Header" C. placing the Emergency In / Notch Override switch (3A-S3) in the "EMERG ROD IN" position ONLY D. bypassing the RWM AND placin,g the Emergency In / Notch Override switch (3A-S3) in the "EMERG ROD IN" position Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 ~______ Level of Knowledge FUNDAMENTAL
_~____ Distractors:
___. ~- Difficultx
~ Time Allowance (minutes)
__ NIA iC Basis or Justification The RWM must be bypassed in order to allow rods to be inserted in a7y sequence.
Then the Emergency In I Notch Override switch is used to insert control rods. An insert rod block exists. Normal rod movement with the rod control switch is prevented by the insert rod block. ~- -~ ________~__.
~. - ~ - - ______ ~~ Since a hydraulic AlWS exists, this negates the use of T-214 to vent the scram air header. An insert rod block exists. Rod movement with the Emergency In I Notch Override switch is prevented by the insert rod block.
Source: - ._ Reference(s):
__ Learning Objective:
KIA System -~ - KIA Statement Source Documentation New Exam Item 0 Modified Bank Item ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
~-~-~____
-~ T-220; M-1-S-20 Sheets 9, 12 ________~_
______.~_____~___ - -_ PLOT- 1 560-3 201 006 - RWM Importance: RO
/ SRO 3.4 13.5 ~~ K4.04 - Knowledge of RWM System design feature(s) and/or interlocks which provide for system bypass. REQUIRED MATERIALS:
Notes and Comments:
___~_ ~ _ ~- ~ - ___~____ -~ ~_ __ ~_ -~ -~ ___-__-______~
___ ~ ____ ____~ - ~_ -_ ~~ ~ ~- -
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 3 1. A reactor startup is in progress on Unit 2 when the variable leg for Reactor Narrow Range level becomes plugged with dlebris. As the startup continues, predict the response of INDICATED level as RPV pressure rises toward rated AND what procedure, if any, would be used to correct this condition. Assume no change in actLial level. Indicated level on the affected narrow range indicators will (1) . -- (2) would be entered to correct this condition.
A. (1) rise due to the rising pressure on the variable leg (2) OT-1 10 "Reactor High Level"'
B. (1) lower due to the rising pressure on the reference leg (2) OT- 100 "Reactor Low Level C. (1) lower due to the reduced pressure on the reference leg (2) OT- 100 "Reactor Low Level'" D. (1) remain the same as the pressure rise is felt on both the variable and reference legs equally (2) No procedure Peach Bottom Initial Reactor Operator NRC Examination December 2008 ~___ Level ~ of ~- Knowledge Difficulty Choice Correct: _~ Time Allowance (minutes)
___. __ . Distractors:
Source: I ic 0 New Exam Item 0 Previous NRC Exam 0 Other Exam Bank Modified Bank Item
_____p__~--____~__~__-___-~_~_~-__--
Basis or Justification As the reactor vessel pressure reaches rated power conditions, the pressure on the reference leg will rise while the pressure on the variahle leg will remain constant, thereby causing indicated level to lower. OT-I 00 would be entered for an unexpected or unexplained drop in reactor water level. As the reactor vessel pressure reaches rated power conditions, the pressure on the reference leg will rise while the pressure on the variable leg will remain constant, thereby causing indicated level to lower. The pressure on the reference leg will rise while the pressure on the variable leg will remaiin constant, thereby causing indicated level to lower.
As the reactor vessel pressure reaches rated power conditions, pressure on the reference and variable legs will NOT remain the same. The pressure on the reference leg will rise while the pressure on the variable leg will remain constant, tlhereby causing indicated level to lower.
_______~__~
~ ___- -_ ____-____ - -_~-_-___-__--___~---~~~~-p
--___-_--__-~-~-____.~-
~_____--~ ~ ~ ~-__ ~_____ ________________
_ ~ Importance:
1 2.9 13.0 ~ _-_____ __-_______
216000 - Nuclear Boiler Instrumentation IZ WA System K/A Statement A2.02 - Ability to predict the impacts of instrument line plugging on Nuclear Boiler Instrumentation and based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions. Notes and Comments: - - _______ __~____________~____-
_- ~ ~-~___~__~~- REQUIRED MATERIALS: ~ ~ -_ ____p_ ~ __________~~~
__ ~~ -_ - __-_~ _________
Peach Bottom Initial Reactor Operator NRC Examination December 2008 32. Unit 2 is in MODE
- 4. The "B" Loop of RHR is lined up in Torus Cooling using the 2D RHR pump. The "A" Loop of RHR is lined up in Shutdown Cooling using the 2A RHR pump. A fault on the E-42 bus results in annunciator 005-B 1, E-42 BUS DIFFERENTIAL OR OVERCURRENT RELAYS.
Which one of the following correctly describes the effect of this event?
The E4 Diesel Generator auto starts and A. loads the E42 Bus. Shutdown Cooling using the 'A' RHR Loop remains in service. B. loads the E42 Bus. Torus Cooling using the 'B' RHR Loop remains in service. C. does NOT load the E42 Bus. Shutdown Cooling using the 'A' RHR Loop is lost. D. does NOT load the E42 Bus. Torus Cooling using the
'B' RHR Loop is lost.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 NIA Question # 32 RO . .. - - . . . Source : _______ _- Reference(s):
Learning Objective:
Choice New Exam Item 0 Modified Bank Item [XI ILT Exam Bank Previous NRC Exam Other Exam Bank --_________~
~ _- ~~-_~_-_______-_____~-~-~_.
-~ _____ ________ ARC 005 B-I PLOT-501 0-2a Correct: Distractors:
A C ___________________~____ - ~ Basis or Justification
~- ___. -________~-
__ E4 Diesel will auto start on low E-42 bus voltage, but does not load orlto the E-42 bus due to the bus fault condition.
With E-42 bus de-energized the 2D RHR pump has no power and therefore RHR Torus cooling is lost. E4 Diesel output breaker is locked out from closing due to the E-42 bus fault. E4 Diesel is locked out, and the 2D RHR Pump will trip of loss of E-42 bus power. ~- -__ - ____~~~_____
~- _ - ~ 2A RHR Pump is powered from the E12 Bus. Shutdown cooling will not be lost. KIA System 21 9000 - RHR Torus Cooling Mode 1 Importance:
ROISRO 3.1 13.3 -~ KIA Statement
__ REQUIRED - MATERIALS: Notes and Comments:
K2.02 - Knowledge of to the pumps. ~ _ _-___ ~____-_ _____-~ -~ __ ~____~_________-_-__
____- _ - ~_
Peach Bottom Initial Reactor Operator NRC Examination December 2008 33. A large LOCA has caused the following conditions on Unit 2: 0 0 0 Drywell pressure is 12 psig Reactor pressure is 390 psig Drywell sprays are in service IAW T-204 "Initiation of Containment Sprays Using RHR' If drywell pressure continues to lower and no operator actions are taken, drywell spray will remain in service until the spray valves automatically close at psig drywell pressure.
A. <O B. >Oand<l C. >1 and<2 D. >2and<3 Peach Bottom Initial Reactor Operator NRC Examination December 2008 _____~- Level of Knowled e FUNDAMENTAL Difficuk Choice Time Allowance (minutes)
NIA Correct: Distractors:
B A D -~ _______~________
~ _ Basis or Justification
__________
~ _ ________~____
_ If spray operation lowers drywell pressure
I psig) and the LPCl initiation signal is sealed in, the sprriy valves will automatically close. If spray operation lowers drywell pressure
<I psig (before reaching 0 psig), and the LPCl initiation signal is sealed in, the spray valves will automatically close. If spray operation lowers drywell pressure
I psig) and the LPCl initiation signal is sealed in, the spray valves will automatically close. ~ -- -~ _ ______-____~_~~- ~ ~ ____ _ 2 to 3 psig is below a reasonable Drywell pressure control band (3 to 5 psig) that a Reactor Olperator could be given while spraying the Drywell using T-204 "Initiation of Containment Sprays Using RHR". The Examinee could mistake this control band as the pressure in which the spray valves will auto close since it is below the 3 to 5 psig control band. - - ___.___ ~ ~ _ _ Source: Reference(s):
__~. Learning 0 bject ive: KIA System Source Documentation New Exam Item Modified Bank Item 0 ILT Exam Bank Previous NRC Exam Other Exam Bank
.__ ________ _______-_-~
_ ~ ~ T-204; PLOT-5010 PLOT-5010-6i KIA Statement REQUIRED MATERIALS:
--_ ~________
______ _ - Notes and Comments:
__~_~____________~~___
___ _-- ~- -_
Peach Bottom Initial Reactor Operator NRC Examination December 2008 34. A LOCA occurred on Unit
- 2. The following conditions are present:
0 Reactor level is
-10 inches Reactor pressure is 740 psig Drywell pressure is 9 psig and rising slowly Torus Sprays are in service using the 2B RHR pump A loss of off-site power (LOOP) occurs All EDGs have started and energized their respective 4kV busses Based on these conditions, in order to re-establish Torus Sprays, the RHR pump -__ (1) and the Torus Spray flowpath (2) . A. (1) must be manually restarted (2) will automatically realign B. (1) will automatically restart (2) will automatically realign C. (1) must be manually restarted (2) must be manually realigned D. (1) will automatically restart (2) must be manually realigned Peach Bottom Initial Reactor Operator NRC Examination December 2008 Diff icuk .~ Time ___ Allowance (minutes)
Choice ~_ N/A Correct: Distractors:
C A - .- B D Basis or Justification On a LOOP the normal feeder breakers will trip on undervoltage AND all bre-ikers being fed from the 4kV IIUS will trip, including any RHR pumps that were in smice. To re-establish Torus Spray the RHR pump will have to be manually restarted. The flowpath motor operated valves will de-energize on the LOOP, but will not realign themselves automatically.
They will have to be manually closed prior to starting the RHR pump ~ The flowpath motor opeirated valves will de-energize on the LOOP, but will not realign themselves automatically.
They will have to be manually closed prior to starting the RHR pump. On a LOOP the normal lfeeder breakers will trip on undervoltage AND all breakers being fed from the 4kV bus will trip, including any RHR pumps that were in service. To re-establish Torus Spray the RHR pump will have to be manually restarted The flowpath motor operated valves will de-energize on the LOOP, but will not realign themselves autoimatically.
They will have to be manually closed prior to starting the RHR pump. The pump will not restart automatically.
On a LOOP the normal feeder breakers will trip on undervoltage AND all breakers being fed from the 4kV bus will trip including any RHR pumps that were in service. _~______ ~ ~ - -~ -~~ __ _- Reference(s1:
Learning i 0 bject ive:
I Source Documentation New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam c] Other Exam Bank ________~_~_~
-~~ SO 54.7.A; T-204 ___ -~ ~ - PLOT-5010-6a 1 ___ - --~ ~ _~___ .____ 230000 - RHR Torus Spray Mode Tymportance:
RO / SRO .- 3.2 13.3 -~ __ K1.05 - Knowledge of the physical connections and/or cause-effect relationships between RHR/LF'CI: ~~~ ~~~ Torus Spray Mode and A.C electrical. - - ~ - ~_ - _~ kti and Comments:
~ ~ -~ ~-
Peach Bottom Initial Reactor Operator NRC Examination December 2008 35. Unit 2 is in Mode 1 at 100% power. 0 0 There has been a loss of all Fuel Pool Service Water Booster Pumps. Fuel Pool temperature is 89 degrees F. Fuel Pool level is normal. For the above conditions, what is an (alternate source of cooling to the Fuel Pool Cooling heat exchangers in accordance with SO 19.7.A-2 "Loss of Fuel Pool Cooling"?
A. HPSW using A0 32.3-2 "HPSW Injection into the Fuel Pool" B. RBCCW using A0 35.1-2 "RBCCW Backup to Fuel Pool Cooling" C. RHR using A0 10.3-2 "RHR System to Fuel Pool Cross-Connect Operation" D. ESW using AO-33.2 "Emergency Service Water System Manual Startup and Operations" Peach Bottom Initial Reactor Operator NRC Examination December 2008 __ - -~___ ___._~_______
___ Answer Key ____________________~~__
~__ _-_ - Difficulty Level of Knowledge FUNDAMENTAL
~--_____ Choice Time Allowance
_____._~_________
Correct: Source: Reference(s):
__ Learning Objective:
B New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank ~ ~ _ -_ __ SO 19.7.A-2; A0 35.1-2 ____ ___________~_-
__ -~ __ ~ - -~ PLOT-5019-5a Distractors:
A C D __ Basis or Justification - -__~ -~ SO 19.7.A-2 prerequisite 2.1 states that IF a loss of cooling is due to ti loss of Fuel Pool Service Water Booster pumps THEN the RBCCW System shall be available to supply cooling water to the Fuel Pool Cooling heed Exchangers with A0 35.1-2. A0 32.3-2 with HPSVIl should only be used when it is necessary to provide Fuel Pool injection whlen an actual or imminent condition exists for uncovering fuel in the Fuel Pool. _ A0 10.3-2 with RHR will not supply alternate cooling to the Fuel Pool Cooling heat exchangers. The A0 is used when there has been a complete loss of Fuel Pool Cooling or Fuel Pool Cooling is inadequate to maintain Spent Fuel F'ool water temperature. The ESW System provides a safety-related cooling water supply to the Emergency Diesel Generators and ECCS Room Coolers. It does not provide an alternate clooling source for the Fuel Pool Cooling heat exchangers.
________-~
__________._~____~____~
__ ~~ ___ _____________
~~ ~__ ~- _. K/A System 233000 - Fuel Pool Cooling __ __ TGrtance:
RO / SRO 2.5 12.7 K/A Statement K5.01 - Knowledge of the operational implications of heat removal mechanisms as it applies to Fuel --____ Pool Coolin% _______. . __- ~ - Notes and Comments:
-~
Peach Bottom Initial Reactor Operator NRC Examination December 2008 36. Unit 2 is operating normally at 100?4 power when:
0 0 FEEDWATER FIELD INSTRUMENT TROUBLE (201 H-1) goes into alarm. "B" main steam line flow indicator FI-2-06-088B on Panel 20C08A instantaneously fails downscale.
Which of the following describes the response, if any, of the Digital Feedwater Control System (DFCS)? DFCS will A. remain in three element control arid maintain normal RPV water level
(+23") B. remain in three element control and maintain a higher than normal RPV water level C. shift to single element control andl maintain normal RPV water level
(+23") D. shift to single element control and! maintain a lower than normal RPV water level Peach Bottom Initial Reactor Operator NRC Examination SRO _~_______
Level of Knowledge Difficulty
~- -___ Time Allowance (minutes)
~- Psychiometrics HIGH NIA December 2008 Question # 36 RO - -. - I Choice Correct: Distractors:
._____~..
C A B D _ Answer Key Basis or Justification On a loss of 1, 2 or 3 steam flow signals DFCS will automatically tranFfer to single element controll.
As stated in Step 3.2 of OT-100, "If any feed\n;ater flow indication is upscale or any steam line flow indication is downscale, then verify the Feedwater Level Control System is operating in single element control".
Since there was no plant transient (no change in actual feed flow, steam flow or RPV level), DFCS will maintain RPV level as is in single element controll.
~~ ____._______- ~ ~ ~ On a loss of 1, 2 or 3 steam flow signals DFCS will automatically transfer to single element control. On a loss of 1, 2 or 3 steam flow signals, DFCS will automatically transfer to single element control. Since there was no plant transient (no change in actual feed flow, stearn flow or RPV level), DFCS will maintain RPV level as is. Since there was no pliant transient (no change in actual feed flow, steam flow or RPV level), DFCS will maintain RPV level as is in single elemeit ~~ ~- _~-._~____._____
_~ - Sou rce Documentation Source: I 0 New Exam Item Previous NRC Exam - 2007 Modified Bank Item 0 Other Exam Bank ILT Exam Bank - _. - ~- Learning Objective:
__~_________
~________~-
_____ -___ ____.- ~ 1 KIA System 239001 - Main and Reheat Steam SystemTImportance:
RO I SRO KIA Statement K3.15 - Knowledge of the effect that a loss or malfunction of the Main and Reheat Steam System have on the following: Reactor Water Level Control.
-- ~~ - ~- - ~________________~_ REQUIRED MATERIALS:
__ ~_ _~~____________________~
-- ~~ ~__ Notes and Comments:
~~_________~___
~
Peach Bottom Initial Reactor Operator NRC Examination December 2008 37. While performing Main Turbine shell warming in accordance with SO 1B. 1 .A-2 "Main Turbine Startup and Normal Operation" the operator is cautioned to ensure turbine first stage pressure remains below 100 psig. The reason for this caution is to prevent A. rolling the main turbine off the turning gear B. differential expansion between the turbine shell and rotor C. exceeding the setpoint for the power-to-load unbalance (load reject) trip D. exceeding the setpoint for the turbine stop valve and control valve scram bypass Peach Bottom Initial Reactor Operator NRC Examination Source: December 2008 New Exam Item 0 Modified Bank Item 0 Previous NRC Exam 0 Other Exam Bank Choice Correct: f Distractors:
lB i 1 IC I__- L ~_ .-_______-_
-~ -- ~- Basis or Justification
___~_____________~__
__ _ This is stated in the CAUTION for step 4.9.10 of SO 1 B. 1 .A-2, and also in GP-2. Note that even though the scram bypass setpoint would be exceeded if first stage! pressure rose above 138 psig, a scram would rot occur since the TSVITCV low power scram bypass is locked in by procedure (GP-2, Attachment 7). This is the reason why 6 of 10 lift pumps are secured prior to shell warming, as stated in the NOTE for step 4.9.4 of SO 1 B.l .A-2. As stated in the NOTE: for step 4.7 of SO lB.l .A-2, "differential expansion concerns are addressed by the pre-warming direction provided in this procedure."
The power-to-load unbalance trip receives a pressure input signal from the turbine cross-around header (HP turbine exhaust), not the turbine first stage. ~ ~____~____
__ Psychometrics
-____ Level of Knowledge Time Allowance (minutes)
~- FUNDAMENTAL N/A ILT Exam Bank
_________~
_.._ ~ ~ ~~~ GP-2, SO I B.1 .A-2 PLOT-5001 B-Id ____~_~ __ . ~ ~~ RO I SRO 4.6 14.6 ~~~~ ~~~ ----____--___
245000 - Main Turbine Generator and Auxiliaries Systems ~_ ~ . . i - -~ Reference(s):
Learning Objective:
WA System WA Statement
__ REQUIRED MATERIALS: Notes and Comments:
____~___~_
-~ ___.~. __
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 38. Unit 2 is at 80% power during a plant startup. Shortly after placing the third reactor feed pump in service, annunciator 2 10 H-2 "REACTOR HI-LO WATER LEVE L" alarms. The following conditions exist: 0 0 0 0 RPV level is +3 1 inches and rising. Total feed flow is greater than total steam flow. "A" RFP flow is 4.1 MLBM/HR and rising. "B" RFP flow is 3.9 MLBM/HR and lowering. "C" RFP flow is 4.0 MLBM/HR and steady.
Based on the above indications, the -.-( 1)- RFP is operating correctly and the (2) RFP must be taken to mariiual control.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Difficulty Time Allowance (minutes)
~ _____~__-
.- __ ~ .________ Level of -______ Knowledge_
HIGH Answer Key _ ____ SRO ~ N/A Distractors:
1A I I- _______ ~ ~_ __~_ __ ~~ - Basis or Justification If feed flow is > steam1 flow, RPV level will rise. The RFP master level controller will attempt to lower speedlflow for ALL RFPs. Only the "B" RFP is operating correctly. The "C" RFP control is not responding (flow is constant). The "A" RFP controller has failed because flow is rising. "AI RFP flow is rising The Operator should take manual control of the: "A1 RFP due to flow rising contrary to feedwater system control.
The "C" RFP flow should be lowering in response to the high RPV water level. However, "A" RFP flow is actually rising and has an immediate detrimental effect on rising water level and should be the higher priority for manual control "A" RFP flow is rising and is not operating properly in response to a high reactor level.
-~ _~ ~ _______~~
___ _______________
____ ~~__ __ ___ ________ ~~______ ____________
~ -~_ ___ ~ _- Source : Reference@):
Learning Objective: Source Documentation 0 New Exam Item Ix] Modified Bank Item (XI ILT Exam Bank Previous NRC Exam [7 Other Exam Bank - .-____. _~ ~- -~ -_ ~ ~ ~ OT-I10 PLOT-1540-4 Importance:
RO / SRO 3.3 13.3 259001 - Reactor Feedwater System ~ ~ - WA System WA Statement AI .01 -Ability to predict and/or monitor changes in parameters associated with operating the Reactor ~_~________
_~ ~~ ~._____~ Notes and Comments:
_______~ ~ ~ ~_______~ - ~~ ~~
Peach Bottom Initial Reactor Operator NRC Examination December 2008 39. Given the following: Unit 2 is operating at 95% power. A recirculation flow reduction event results in entry into Region 2 of the Power tcl Flow Map.
Per OT-112 "UnexpectedAJnexplained Change in Core Flow", which one of the following instrumentation responses indicates the reactor core is experiencing thermal hydraulic instability (THI)?
A. Peak-to-peak oscillations on RBh4 are 10% and growing larger.
B. Peak-to-peak oscillations on APRMs are 10% to 12% and their magnitude is growing larger.
C. Oscillations on WRNMs and short period alarms are received on a 10 to 20 second frequency.
D. Steady confirmation counts on the OPRM display with "OPRM PRE-TRIP" alarms.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Difficulty
___~~____ . Time Allowance
_______. (minutes)
Question # 39 RO SRO N/A Choice ~_____~___
Correct: Distractors:
B A ~_ ~ Basis or Justification
__ .___ ___--_______ - __________~____ Core Thermal Hydraulic Instability (THI) may be occurring if any of the following conditions exist: *Steadily increasing confirmation counts or OPRM display with few to no resets.
- Any APRM flux noise signal grows by 2 or more times its initial level.
- APRM flux oscillations rise greater than or equal to 10% (peak to peak).
RBM not referenced its a nuclear monitoring instrument for THI No reference to period indication as a nuclear monitoring instrument for THI. Steadily increasing confirmation counts on OPRM display causing repetitive "OPRM Pre-trip Condition" alarms is an indication of THI. Source Documentation Source: New Exam Item Previous NRC Exam - 2007 0 Modified Bank Item ILT Exam Bank Other Exam Bank ________~
-~ - __ -_ -~ -_ -~ ~~ -_ - - ~ Reference@
Objective:
Learning PLOT-1540-3
_-_ . _- - Importance:
RO / SRO -______ T Forced .____________ Core Flow Circulation
~__ WA System K/A Statement AA2.02 - Ability to determine andlor interpret Neut,ron Monitoring as it applies to Partial or Complete _____ REQUIRED MATERIALS:
Notes and Comments:
I 3.1 13.2 _~_ .~__________.__
295001 - Partial or Complete Loss of -~ Loss ~-_______________~
of Forced Core Circulation.
. _______.._
Peach Bottom Initial Reactor Operator NRC Examination December 2008 40. Given the following conditions on Unit 2: 0 0 0 125/250 Volt Battery Charger 2AD003 is performing an "equalize" charge on its battery. During the charge, AC power to the charger is lost due to a momentary loss of power to the E-12 bus. Power to the charger is restored in 15 seconds when the E-12 bus is reenergized by the diesel generator.
Which of the following describes the status of the 2A Battery when the E- 12 bus is reenergized? The 2A Battery is A. in the "float" charge mode on the original charger B. in the "equalize" charge mode on the original charger C. in the "float charge mode on the standby charger D. in the "equalize" charge mode on the standby charger Peach Bottom Initial Reactor Operator NRC E:xamination December 2008 ________. Level of Knowledge HIGH Question # 40 RO . - . .. . - - Time Allowance
______ (minutes)
__ ~ Difficulty - ~ ._ N/A Choice Source: __ Reference(s):
Learning 0 bj ect ive: - _______ ___ Correct: New Exam Item Modified Bank Item ILT Exam Bank ~ ~ ~ -~ Previous NRC Exam Other Exam Bank SO 578.1-2 ~___~_~~- - ~~ PLOT-5057-6a Distractors:
____~ __~___ REQUIRED MATERIALS:
Notes __ and Comments:
~~___ ~ ~ B ~-~~ -~ ~- -~ ~ __ ~ -~ -_ NONE ~ ~ ~ ~- A C D Answer Key ~ - Basis or Justification From Note 2 in SO 57'B.l-2 "Upon a loss of AC input power, the battery charger will return to the same mode it was in once power is restored. IF the battery charger WBS in the Equalize mode, THEN the timer will pick up where it was interrupted AND time out." The charger will return to the equalize charge mode. ~ ~- -___ ~____~ The standby battery charger is manually placed in service only.. . there is no automatic transfer to the standby charger.
~ The standby battery ciharger is manually placed in service only.. . there is no automatic transfer to the standby charger.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 41. Given the following:
0 Unit 2 is operating at 100% powcr. An electrical transient results in iii blown output fuse on battery charger Prior to the transient, charger 2BD003-1 was supplying the Division I1 250 VDC bus. 2BD003-1. Assuming no operator action, how will the plant respond to this event? The Division I1 250 VDC bus will be supplied by the A. 2B station battery ONLY B. 2D station battery ONLY C. 2B AND 2D station batteries D. 2B station battery AND in-service 2D charger Peach Bottom Initial Reactor Operator NRC Examination December 2008 Choice Correct: Distractors:
D A _. __ B C ~ __ __________._~-_.___--
Basis or Justification When the output brealker for charger 2BD003-1 trips, the charger no longer supplies power to the Division II 250 VDC bus. The bus loads would then be supplied by the 28 battery and the in-service 2D charger ~ - -- ______ _-_______~_______~~.
When the output breaker for charger 2BD003-1 trips, the charger no longer supplies power to the Division II 250 VDC bus. The bus loads would then be supplied by the 28 battery and the 2D charger. The 2B battery will supply the 250 VDC loads along with the 2D charger.
-~ -~ ~~ _____~-__.______-~-.
The 2D charger remains in service supplying the 250 VDC loads along with the 2B battery.
-__--_______ Level of Knowledge Difficulty Time Allowance (minutes)
SRO -. ~- NIA Source: ~. Referencets):
Learning Objective:
KIA System Source Documentation 0 New Exam Item [XI Modified Bank Item 0 Previous NRC Exam 0 Other Exam Bank
__~~_____~~
[XI ILT Exam Bank
~ .___ ~- - ~- _______ ~__-____-
-~ E-26 PLOT-5057-3~
295004 - Partial or Complete Loss of D.C. Power ~ I 3.1 13.1 ~ Importance:
RO I SRO KIA Statement AK2.01 - Knowledge of the interrelations between Partial or Complete Loss of D.C. Power and Battery ___ ~~ - Charger. ~_____ REQUIRED ~ MATERIALS: Notes and Comments:
~~~ . ~ - ~~ ~~ ___~- - -~
Peach Bottom Initial Reactor Operator NRC Examination December 2008 42. Given the following:
0 0 0 A plant startup is in progress on Unit 2 The main turbine is in Shell Warming The "EHC 125V DC LOSS TRIP" annunciator is received and a turbine trip occurs Which one of the following identifies the positions of the Turbine Control Valves (TCV), Intercept Valves (IV), and Intermediate Stop Valves (ISV) for these conditions?
IV 7 ISV - TCV - A. Open Closed Closed B. Closed Closed Open C. Closed Closed Closed D. Open Open Closed Peach Bottom Initial Reactor Operator NRC Examination December 2008 SRO __- FUNDAMENTAL Choice N/A Correct: Source: _..___ __ Distractors: New Exam Item Modified Bank Item 0 Previous NRC Exam 0 Other Exam Bank C A B D _._______
~ - ~- ~ __ _______~._--~
_ -~ Basis or Justification
___ - A loss of 125 VDC results in a turbine trip at less than 1400 RPM, as indicated by the annunciator. Even though the turbine is initially in Shell Warming, all listed turbine valves close on a turbine trip condition.
This is the initial lineuip while in Shell Warming, prior to the turbine trip condition.
~ ~________~__
_____- - -~-_ The lSVs will also clo:je on a turbine trip condition.
All listed turbine valves will close on the trip condition.
________ _-___ ____ Ref ere nce( s) : Learning Objective:
~ KIA System M-302; M-303; M-304; ARC 206 C-4 ~- --_ - ~~ PLOT-5001 B-4a ~- ~ - _ ~__~____ ~ Importance: RO I SRO 3.1 13.1 I - -- ~ 295005 - Main Turbine Generator Trip
__- ~~~ ~ KIA Statement AA2.03 - Ability to determine and/or interpret Turbline Valve Position as it applies to Main Turbine REQUIRED MATERIALS: - ~____ ____ ___ ~ - -____ Notes and Comments: ~~ ~~ -
Peach Bottom Initial Reactor Operator NRC Examination December 2008 43. Given the following:
0 0 0 0 Unit 3 was scrammed from full plower due to a loss of both Recirc pumps Reactor water level is 20 inches and steady All control rods are fully inserted The scram can NOT be reset Which one of the following explains why CRD flow is required to be minimized IAM SO 3.2.A-3 "Control Rod Hydraulic System Shutdown"?
A. To prevent CRD pump runout.
B. To stop input to the scram discharge volume.
C. To prevent damage to CRD mechanism seals.
D. To limit thermal stratification in the RPV bottom head.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 FUNDAMENTAL Choice N/A I Correct: Distractors:
D A B C Basis or Justification
_______________~
_____..___________~~.. As stated in T-A00 Bases, and Step 4.1 of SO 3.2.A-3. CRD Hydraulic Systern design (orifice) prevents CRD pump runout following a scram.
Following a scram, input to the SDV is stopped once it is filled and becomes solid. This is not the reason for minimizing CRD system flow following a scram. __ ~...~~____
__________~_ . ~. I I I --I Psychometrics LLevel of Knowledge - DifficuB ~ &me ~~_______~~
Allowance (minutes)
~ SRO . Source: Reference(s):
Learning Objective:
WA System Source Documentation New Exam Item c] Modified Bank Item T-100 Bases, SO 3.2 A-3 0 Previous NRC Exam 0 Other Exam Bank 0 ~ ILT _ Exam Bank . - ~ - -_~ ~_ ~~ _____ _~__~____
~ --~ PLOT-PBIG-2100-TI 00-9 295006 - Scram . RO / SRO 3.5 / 3.6 WA Statement AA1.06 ________ -Ability to operate ___. and/or System as __~____. it applies to a Scram. ________._
~~ - ---_________
-~ ___ - ~~~ - REQUIRED MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator NRC Examination December 2008 44. Which one of the following statements describes the reason for disabling control room controls IAW SE-10 "Plant Shutdown from the Alternative Shutdown Panels" after abandoning the control room'? A. To maintain High Pressure Coolant Injection (HPCI) System automatic operation.
B. To ensure interlocks associated with operation of safe shutdown equipment are defeated.
C. To ensure fire-induced circuit faults will NOT prevent operation of safe shutdown equipment.
D. To prevent simultaneous operation from the control room and the Alternative Shutdown Panels.
, Question # 44 RO -. .. -. . . Choice .-- Level of Knowledge FUNDAMENTAL
~ Correct: ___.~- Difficulty Time Allowance (minutes) - - - - N/A 1 Distractors:
Source : ________ C [XI New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
_________~__~ - --_ - - - - A B D _______- ._ __._ -- ~ ~ Basis or Justification
___~_ - - From the UFSAR: "trainsfer/isolation switches provide electric circuit isolation between alternative shutdown circuits and circuits that could De affected by a fire in orie of the four areas of concern (control room, cable spreading room, computer room, emergency shutdown panel area)." 'rom 1 OCFRSO, Appendix F?: "The safe shutdown equipment and systems for each fire area shall be known to be isolated from associated non-safely circuits in the fire area so that hot shorts, open circuits, or shorts to ground in the associated circuits will not prevent operation of the safe shutdown equipment . " HPCI system automatic system operation (trips, isolations and automatic start) is defeated when control is transferred to the Alternative Shutdown Interlocks are defeated when operation is transferred to the Alternative Shutdown Panels, but this is not the design basis reason for disabling Control Room controls. This could be (and is) accomplished procedurally; it is not the design basis reason for disabling Control Room controls. - - ~ ____~ ~ - __ -~ Panel. _ -- - _. ________________~.
Objective: KIA System
~ ~ - SE-10; UFSAR FPP, Ch. 5 PLOT-1555-9
___~_________~_
._ __- 295016 - Control Room Abandonment
-- Lpp- t KIA Statement AK3.03 - Knowledge of the reasons for the followiiig responses as they apply to CONTROL ROOFA - -~_ - ~ __ -- ____~_ ~ _~_______~
~_- - Notes and Comments:
Peach Bottom Initial Reactor Operator NRC Examination December 2008 45. Unit 3 was operating at full power when the following transient occurred:
0 0 0 "A" RBCCW pump tripped due to an overcurrent condition. "B" RBCCW pump started automatically but at a reduced discharge pressure. RBCCW system temperatures are rising steadily. Under these conditions, ON-113
LOSS of RBCCW' directs the RWCU pumps tripped and the system isolated. According to the ON-113 Bases, the reason for these actions is to: A. Isolate a likely primary-to-secondary leak in the RBCCW heat exchangers.
B. Allow more time to diagnose and correct the cause of the RBCCW problem. C. Prevent RWCU pump cavitation due to high reactor water inlet temperatures.
D. Reduce the required RBCCW system flow rate thereby preventing pump runout.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge
___~_ ____- Oifficuw Time Allowance (minutes)
FUNDAMENTAL Choice SRO NIA ~ Correct: Source : - Reference(s):
Objective:
Learning Distractors: New Exam Item Modified Bank Item ILT Exam Bank Previous NRC Exam - 2005 0 Other Exam Bank
-- -~ ~- _- --- OT-I 13 Bases PLOT-1550-03 B A C - D Basis or Justification
____~_~______~--~~~-
__-.~ - Removing RWCU frorn service removes a significant heat load which will greatly slow the heatup of the RBCCW system, thereby providing more? time to correct the problem.
A primary-to-secondaiy leak in the RBCCW heat exchanger is plausib e, but is not what the isolation of the system is based on according to Ohl-I 13 Bases. ______~_~~__.~_ . ~- ._~____ - ~ Although the RWCU pump inlet temperatures could rise, these steps are not based on preventiing this condition. Although isolating RWCU will reduce the required heat input to the system, it does not reduce RBCCW system flow rate because the isolation occurs on the Reactor Water side of the system. _______ ~___- ________-__
___________~
~ Importance:
RO I SRO 2.9 13.2 __ - ___p -~ 295018 - Partial or Total Loss of CCW -- 1 KIA System KIA Statement AK3.01 - Knowledge of the reasons for securing individual components as it applies to Partial or Total ______ -________ REQUIRED MATERIALS:
~~~ -- -- Notes and Comments:
__ __
Peach Bottom Initial Reactor Operator NRC Examination December 2008 46. Given the following:
Unit 2 was initially operating at 100% power A complete loss of Instrument Air occurred T-261 "Placing The Backup Instrument Nitrogen Supply From CAD Tank In Service" was implemented due to loss of both Instrument Nitrogen compressors Based on these conditions, which Main Steam Isolation Valves (MSIVs), if any, have a long-term pneumatic supply?
A. Inboard ONLY B. Outboard ONLY C. BOTH the inboard AND outboard D. NEITHER the inboard NOR outboard Peach Bottom Initial Reactor Operator NRC Examination December 2008 B Choice ~_________
Correct: The outboard MSlVs are supplied by Instrument Air Distractors:
D Basis or Justification
-~___ ______-~__
~ - - -_- inboard MSlVs are supplied with Instrument N2 from both the 'A' ,and '8' Instrument N2 headers; the CAD tank (T-261) backs up the 'B' Instrument N2 header. instrument Air supplies the outboard MSIVs. Therefore, there is a long-term pneumatic source to the inboard MSIV!; but not the outboard MSl\/s. The inboard MSlVs are supplied by Instrument N2.
Level of Knowledge Diff iculk- __ -_____ Ti me Allowa nce&i n -~ u tes) FUNDAMENTAL C 1 The outboard MSlVs are supplied by instrument Air SRO N/A ~~ ~- __ Source: ________~__ - Reference(s):
Learning Objective:
c] New Exam Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank Modified Bank item _____. _~- ~- ~ __- ON-I 19; M-333; M-351; T-26'1 ~- -_ --- ~- PLOT-5001A-1 k, -1 w _- . -~ __ ~ ~~~~ Importance:
RO / SRO - - - -- 3.4 13.4 T- - -~ WA System WA Statement AK2.05 - Knowledge of the interrelationship between Partial or Complete Loss of Instrument Air and REQUIRED MATERIALS: 29501 9 - Partial or Complete Loss of Instrument Air -~ ___- ~- ~ -- ~ _______ Main Steam System. _ Notes _______~~_.__ and Comments:
~ ~- ~ ~___ ___~- ___~-- - ~ ~
Peach Bottom Initial Reactor Operator NRC Examination December 2008 47. Unit 2 has just entered Mode 4 in preparation for a refueling outage. A failure of MO-2-10-0 1 8 "Shutdowin Cooling Inboard Isolation" results in a loss of Shutdown Cooling and necessitates establishing Alternate Shutdown Cooling using SRV's and Torus Cooling, per A0 10.12-2 "Alternate Shutdown Cooling."
Conditions have been established as follows: 0 0 0 0 0 0 Torus Cooling is in service using the '2A' RHR pump The '2E' and '2H' SRV control switches are in the OPEN position Injection to the RPV has been established with the '2D' RHR pump RPV level is 150 inches on Refuel Range (LI-86) RPV pressure is 30 psig Torus pressure is 0 psig Which one of the following describes the action required in order to establish Alternate Decay Heat removal'?
-- (1) RPV injection until RPV (2) A. (1) Lower (2) level is < 100 inches B. (1) Raise (2) level is > 200 inches C. (1) Lower (2) pressure is < 25 psig D. (1) Raise (2) pressure is
> 50 psig Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level _______ of Knowledge HIGH Correct: ~ Difficulty
__ Time - Allowance (minutes)
-~ -~ NIA Distractors:
D A B C ~- ___ - - _- Basis or Justification
____ _______________~
~_ - A 50 psig DIP is required across the SRV's in order for them to open Zind establish alternate heat removal. -- _- ~ ______~________._~__~__~~_
~~ While 108 inches is an important milestone in OT-I IO, Reactor High Level, when trying to prevent filling the Main Steam lines, filling them is required for this procedure in order to establish flow through the SRV(s).
RPV level at 150 inches is high enough to fill the Main Steam lines.
At 30 psig RPV pressure, LI-86 is reasonably accurate.
In order to establish allternate heat removal the SRV's must open, which requires a 50 psig DIF' between the RPV and Torus. The given conditions require raising reactor pressure, not lowering. Source Documentation Source: 0 New Exam Item 0 Modified Bank Item ILT Exam Bank
[XI Previous NRC Exam (LGS - 2006) 0 Other Exam Bank
~~ _~ ~ - ~ ______~_~~__~_
__ _~ __ - ~- Learning PLOT-5001A-5f; PLOT-5007-If Objective:
Importance:
RO I SRO L 3.7 13.7 - -_ KIA System KIA Statement AA1.04 -Ability to operate and or monitor Alternaie Heat Removal Methods as they apply to Loss of Shutdown Cooling.
295021 - Loss of Shutdown Cooling -- - ~ ~_ ~ ~. -_ -~~ _- REQUIRED ~ Notes and Comments: - __ ~_____~
Peach Bottom Initial Reactor Operator NRC Examination December 2008 48. Unit 2 is in a refueling outage.
Fuel is being loaded into the reactor pressure vessel.
The fuel being loaded is NOT directly adjacent to any WRNM. Which one of the following conditions requires entry into ON-1 24 "Fuel Floor and Fuel Handling Problems"?
WRNM (1) (2) between Core Component Transfer Authorization Sheet (CCTAS) steps during fuel handling.
A. (1) period (2) doubles B. (1)period (2) doubles two times C. (1) count rate (2) doubles D. (1) count rate (2) doubles two times Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge Difficulty
~ ______. Answer Key _. Time Allowance (minutes)
_~____._____.___________
~ Choice Correct: Distractors:
D A B C -_____ _____ _______ ~ _- ~ Basis or Justification
__~_~~_____
____________~ ON-124 entry is required if/when WRNM count rate "doubles two time i between CCTAS steps" during fuel handling.
ON-I 24 directs operato3 actions to ensure inadvertent criticality does not occur during fuel handling activities.
Not an ON-124 entry condition.
_________~_~_~_~__~._~ - __ Not an ON-I 24 entry condition.
Not an ON-124 entry condition.
Source : Reference(s):
Learning Objective:
WA System ._______~- Source Documentation New Exam Item 0 Previous NRC Exam
-~~ - Importance:
RO / SRO 4.2 / 4.1 -~ 295023 - Refueling Accidents WA Statement
__-~ G2.4.31 - -~ Knowledge of annunciator alarms, indic,ations, or response __ procedures.
-~ REQUIRED - MATERIALS:
NONE ~- Notes and Comments:
~- _. - -~ ~- _-__ - - ~-
Peach Bottom Initial Reactor Operator NRC Examination December 2008 49. Given the following conditions:
A large break LOCA has occurred Drywell pressure reached a maxirnum of 22 psig 0 Torus sprays have NOT been placed in service 0 Drywell sprays are in service 0 Drywell pressure is 18 psig and lowering While spraying the drywell under these conditions IAW T- 102 "Primary Containment Control", the Torus-to-Drywell Vacuum Breakers will (1) and the Reactor Building-to-Torus Vacuum Breakers will -- (2) . A. (1) cycle OPEN (2) cycle OPEN B. (1) cycle OPEN (2) remain CLOSED C. (1) remain CLOSED (2) cycle OPEN D. (1) remain CLOSED (2) remain CLOSED Peach Bottom Initial Reactor Operator NRC Evamination December 2008 Importance:
RO / SRO 3.4 / 3.4 _~ WA System 1 WA Statement 295024 - High Drywell Pressure
__________~
EA1 .I6 - Ability to vacuum breakers. - -~ - Notes and Comments:
.____~___
-~ _ _- -_ - bE&KD MATERIALS:
~ ~. ~ Choice _______ Correct: Distractors:
-. B A C ____ D ~-~______
_ Basis or Justification
~~______ - __- - _____ - __ As drywell pressure continues to lower, the torus-to-drywell vacuum breakers will open to maintain torus-to-drywell dlp < 0.5 psid (Tech Spec value). Since torus pressure remains well above the opening setpoint for the reactor building-to-torus vacuum breakers of 0.75 psid (Tech Spec value), these vacuum breakers will not open. The reactor building-to-torus vacuum breakers remain closed provided torus pressure remains > 0.75 psid above reactor building pressure.
Spraying the drywell IAW T-102 will ensure this happens. ~ ~ _. _ The torus-to-drywell viacuum breakers will open to maintain torus-to-drywell d/p < 0.5 psid (Tech Spec value). The reactor building-to-torus vacuum breakers remain closed provided torus pressure remains
> 0.75 psid above reactor building pressure.
Spraying the drywell IAW T-I 02 will ensure this happens. The torus-to-drywell vacuum breakers will open to maintain torus-to-drywell d/p < 0.5 psid (Tech Spec value). -_ _ ~ _~___ ~~ - - Psychometrics -I Level ~____ of Knowledge
_~ . I HIGH Time - Allowance (minutes) - ~. SRO . N/A Source Documentation Source: 0 New Exam Item Previous NRC Exam r Ix] Modified Bank Item 0 Other Exam Bank Reference(s)
- Learning 0 bject ive: _______ - Ix] ILT Exam Bank Tech Spec 3.6.1.5 / 3.6.1.6 _ - PLOT-5007-5a Peach Bottom Initial Reactor Operator NRC Examination December 2008 50. Given the following conditions:
0 Unit 2 was initially operating at 100% power when a Group I isolation occurred. Reactor pressure peaked at 1300 psig, at which time the reactor scrammed on higl drywell pressure.
All Safety Relief Valves (SRVs) and Safety Valves (SVs) cycled to relieve pressure during the event.
Based on these conditions, the safety function of RPS was (1) and the safety function of the SRVs and SVs was (2) during this transient.
A. (1) OPERABLE (2) OPERABLE B. (1) OPERABLE (2) NOT OPERABLE C. (1) NOT OPERABLE (2) OPERABLE D. (1) NOT OPERABLE (2) NOT OPERABLE Peach Bottom Initial Reactor Operator NRC Examination December 2008 Question # 50 RO Choice Correct: Distractors:
C A B D __________-~_~___~.
_ - ~- ~ Basis or Justification
~- ~ __- . ---- The safety function of RPS was not operable; the safety function of the SRVs and SVs is operable.
The safety function of RPS was not operable since a scram did not ocsur on MSlV closure, high RPV pressure, or high neutron flux.
The safety function of RPS was not operable since a scram did not occur on MSlV closure, high RPV pressure, or high neutron flux. The safety function of the SRVs and SVs is operable since, collectively, they operated to limit RCBP pressure to below the Safety Limit (1325 psig).
The safety function of the SRVs and SVs was operable.
_________________~____
-_____._____
~ _______ _~___ - ~ __~-___ _______.-
~~- L Psychometrics I Level of Knowledge HIGH Difficulty I-- ~ Time Allowance (minutes)
__ SRO NIA Source: Reference(s):
Learning Objective:
KIA System ____ KIA Statement Source Documentation New Exam Item 0 Modified Bank Item ILT Exam Bank Previous NRC Exam 0 Other Exam Bank ________________
~~ _~ Tech Spec 2.1.2, 3.4.3 and Bases
~-__- - -_ ~ - ~~ PLOT-5001 A-7; PLOT-5060F-7 REQUIRED MATERIALS: - Notes and Comments:
__ _- ____________
.. ..
Peach Bottom Initial Reactor Operator NRC Examination December 2008 5 1. A transient occurred on Unit 3 that resulted in the following plant parameters:
0 Reactor pressure:
900 psig 0 Drywell pressure:
18 psig 0 0 Torus pressure:
16 psig 0 0 Torus level:
15 feet Drywell temperature:
235 degrees F Torus temperature:
145 degrees E: Which one of the following conditions will cause the marain to the Heat Capacity Temperature Limit (HCTL) to be reduced? A. RPV pressure lowers B. Torus level lowers C. Torus pressure rises D. Drywell temperature rises Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowleae Difficulty - Time Allowance (minutes)
_ Answer Key ~ ~- i-- SRO ~ ~ IB Correct: Source: -~ -~ Reference(s):
Objective:
Learning Distractors:
1 A New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
___~___ __~__~______~_~____~__
~ - TRIPEAMP _____~~ Curves, Table _______.__ and Limits - Bases ~~ ~~ PLOT-PBIG-2100-TI 02-4, -523 Basis or Justification
__ ~~___ ~____ Lowering Torus level will cause the HCTL to be more restrictive.
Lowering RPV pressure will cause the HCTL to be less restrictive.
-~ _______._ - Torus pressure has no effect on HCTL. Drywell temperature has no effect on HCTL.
Importance:
RO / SRO __ ___ 1 3.5 13.7 1 _____~~___
295026 - Suppression Pool High Water Temperature WA System WA Statement EK2.06 - Knowledge of the interrelations between Suppression Pool High Water Temperature and Suppression Pool Level. ~ REQUIRED MATERIALS: Notes and Comments:
~- ~ ~ ___~ ~~ - - __~ _ ~ ~~ ~~ -~ ~- __ ~- - ~- ~-~ _~___________
Peach Bottom Initial Reactor Operator NRC Examination December 2008 52. Technical Specification LCO 3.6.1.4 "Drywell Air Temperature" must be entered when drywell average air temperature exceeds
-( 1)- degrees F and is applicable in mode(s) -(2)-. A. (1) 145 (2) 1 ONLY B. (1) 145 (2) 1,2 and 3 c. (1) 200 (2) 1 ONLY D. (1) 200 (2) 1,2 and 3 Peach Bottom Initial Reactor Operator NRC Examination December 2008 C Choice TS 3.6.1.4 entry is required when Drywell average air temperature reaches 145 degrees F and is applicable in modes 1, 2 and 3. Correct: D Distractors:
TS 3.6.1.4 entry is reqluired when Drywell average air temperature reaches 145 degrees F. B Level - of Knowledge . FUNDAMENTAL A SRO NIA _-________
Difficulty - -~ Time -~ Allowance (minutes) . ~ Basis or Justification TS 3.6.1.4 entry is required when drywell average air temperature 145 degrees F, and is applicable in modes 1, 2 and 3. T-I 02 is also entered at 145 degrees F Drywelll temperature.
TS 3.6.1.4 entry is required when drywell average air temperature 145 degrees F, and is applicable in modes 1, 2 and 3. T-IO2 requires a GP-4 scram and RPV depressurization at 200 degrees F, but this is well beyond the Tech Spec LCO for Drywell temperature.
~___ ~ ~- ~~ ~ ~- ~ ~ ~~ _____________~_____
Source: Reference(s):
Learning Objective:
_____.~_~.
KIA System Source Documentation New Exam Item 0 Modified Bank Item 17 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank _~~-~___________~ ~~ ~ ~ _____________~__~___
Tech Spec 3.6.1.4 ~_ - ~ ~~ PLOT-5007-8 Importance:
RO I SRO 4.0 14.7 ~_ ._ -~ ~~ I -~ - - 295028 - High Drywell Temperature KIA Statement
~- G2.2.22 - Knowledge of limiting conditions for operations and safety limits. ~~... REQUIRED MATERIALS:
I NONE - ____---_____
~~
Peach Bottom Initial Reactor Operator NRC Examination December 2008 53. Why does T-102 "Primary Containment Control" require an Emergency Blowdown to be performed when Torus water level CANNOT be maintained above 10.5 feet? In order to A. ensure the HPCI vortex limit is NOT exceeded B. minimize the driving force of the primary system breach C. depressurize the reactor before thie SRV tailpipes are uncovered D. depressurize the reactor before thle downcomer vents are uncovered Peach Bottom Initial Reactor Operator NRC Examination December 2008 __ Level of Knowledge
__ Choice SRO 1 Time Allowance (minutes)
_.___._______
_____ Correct: Distractors:
I Basis or Justification T-102 Bases identifies a level of 10.5 feet as the level of the downcomers, and directs an Emergency Blowdown The HPCl vortex limit is below 9.5 feet, which is the level at which HPCI must be secured in order to prevent uncovering the HPCl exhaust line A Torus break is not a primary system breach.
__ _______~_~_____~
~- ~_ T-102 Bases identifies a level of 10.5 feet as the level of the downcorners, and directs an Emergency Blowdown, whereas the SRV tailpipes are at a level of 7 feet. ____~__~ Source: Reference(s):
Learning Objective:
WA System Source Documentation 0 New Exam Item 0 Previous NRC Exam Modified Bank Item ILT Exam Bank __ __ 0 Other Exam Bank
-___ T-102 and Bases; SAMP-2 Bases
______~.___~_______________
_~ PLOT-1560-9 Importance:
RO / SRO _- 3.8 14.1 295030 - Low Suppression Pool Water Level -~ WA Statement EK3.01 - Knowledge of the reasons for Emergency Depressurization as it applies to Low Suppression Pool Water Level. REQUIRED - MATERIALS:
Notes ~~- and Comments:
~~ _~ -~ - _ ~- +E ~ ~ ~.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 54. Given the following:
0 0 0 0 Unit 2 was shutdown 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> ag,o. The 2D RHR pump is lined up in shutdown cooling.
RPV pressure is 30 psig. RPV level is 25 inches. The following conditions occur: 0 0 RPV level lowers to
-10 inches. All appropriate PCIS isolations occur.
In this instance, raising RPV level to :> +50 inches is necessary to A. satisfy Technical Specification requirements of forced RPV coolant circulation B. reset the Group I1 isolation in order to return shutdown cooling to service ONLY C. promote natural circulation and help prevent stagnation of coolant in the core D. restore decay heat removal capabilities to prevent a heatup of the reactor coolant system Peach Bottom Initial Reactor Operator NRC Examination December 2008 Difficub Time __ Allowance (minutes)
__________
_____~___ . ... Question # 54 RO - -. - - .. SRO __..~ _ - - Choice FUNDAMENTAL Correct: NIA Distractors:
Source: ~ Reference@):
Objective:
Learning C 0 New Exam Item 0 Modified Bank Item ILT Exam Bank c] Previous NRC Exam 0 Other Exam Bank
__..___-___________-_
~ -~ ________~
_- ~ -~~ ~_ ON-125 and Bases _______-_
-~- ~ __ _ PLOT-1550-3 A __ _~ _ __ Basis or Justification As stated in ON-125 Bases, "raising RPV level to > +50" promotes natural circulation and helps prevent stagnation of coolant in the core." Raising RPV level to :* +50" only promotes natural circulation. To satisfy Tech Spec requirements for forced circulation a Recirc or RHR pump must be in service. __ _ __ ______-___
-~~ ~_ ____- ~ ~~ ~ ~ -~ - -~ ~~ While restoring level to > +I inch would allow the rest of Group II isolation, level need only be raised above the Group II setpoint of 1 inch, not to > +5Ot. RPV level > +50 inches will promote natural circulation, but it is not a form of adequate decay heat removal.
KIA System 295031 - Reactor Low Water Level EKI .02 - Knowledge of the operational implications of Natural Circulation as it applies to Reactor Low REQUIRED MATERIALS: Notes and Comments:
_~ Water -~ ____~~____________
Level. _~ ~- ~~~ ___ _~ ~_ _ ~- --~~ ~__~ ~ ~ -_ - __ __~__-~_______
~ ___________________~
Peach Bottom Initial Reactor Operator NRC Eyramination December 2008 55. Unit 2 is operating at 100% power when a Turbine Trip occurs. Two minutes later the following conditions are observed:
0 Reactor power is 0% 0 0 0 0 Rod 26-1 1 is at position "48" Rod 50-3 1 is at position "48" Rod 22-5 1 is at position "48" All other control rods are at "00" Subsequently, the URO begins to inst:rt rods. Which one of the following describes control rod positions that meet the criteria for termination of an ATWS? A. Rods 26-1 1,50-3 1, and 22-5 1 are at position "04". B. Rods 26-1 1 and 50-3 1 are at position "02", AND rod 22-5 1 is at position "04". C. Rods 26-1 1 and 50-3 1 are at position "OO", AND rod 22-5 1 is at position "48". D. Rod 26-1 1 is at position "OO", AND rods 50-3 1 and 22-5 1 are at position "48".
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Choice .- -~ Correct: .__~____-_
Distractors: Level of Knowledge Difficuk C SRO NIA Time Allowance (minutes)
~______~ A __ B D - Basis or Justification
~___-_-~____~____________
-~ -~ Meets the Tech Spec definition of shutdown margin with the single strongest rod fully withdrawn and all other rods fully inserted. Does not meet the Tech Spec SDM criteria.
_________._~
-_~ _________~-_
___________________
-~ ~ Criteria stipulate that if one rod is fully withdrawn or less, all other rods must be fully inserted.
Having one out and all others to or beyond "02" does not meet the definition. Does not meet the Tech Spec SDM criteria.
~ - - Source: Reference(s):
______ Learning 0 bjective:
KIA System Source Documentation 0 New Exam Item 0 Modified Bank Item Previous NRC Exam (LGS - 2006) 0 Other Exam Bank a ILT Exam Bank -__ ~- _____-__ ~- T-101 (Note 24); Tech Spec dlefinition of Shutdown Margin _ _ ~ PLOT-PBIG-2100-TI 01 -4, -6 295037 - SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Importance:
RO I SRO 4.2 14.3 ~- ~~ KIA Statement EA 2.05 -Ability to determine andlor interpret Control Rod Position as it applies to Scram Condition
~_____- Present and Reactor Power - Above ________~~-___~__
APRM Downscale or Unknown. - REQUIRED ._ -__ MATERIALS:
-~ ~- Notes and __ Comments:
____._____~_
~ - -
Peach Bottom Initial Reactor Operator NRC Examination December 2008 56. T- 104 "Radioactivity Release" requires an emergency blowdown when certain conditions are reached. What is the reason for performing an emergency blowdown as it applies to T- 104? A. Establish or maintain adequate cme cooling. B. Prevent exceeding the General Emergency offsite release limit.
C. Preserve the pressure-suppression function of the primary containment.
D. Reduce the discharge of reactor coolant from un-isolated primary system breaks.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge FUNDAMENTAL I Choice Time Allowance (minutes) 1 -- Difficu4 - ~_.____-~__~_
~_ ~- N/A Correct: Source: Reference(s):
Objective:
___ Learning -.~.___ .~ Distractors:
17 New Exam Item Modified Bank Item ILT Exam Bank
--~ 0 Previous NRC Exam Other Exam Bank T-I 04 Bases _____________________-________~_
-- -__- ~. ~ PLOT-PBIG-2100-TI 04-5 D A B C Basis or Justification Per T-104 Bases, "An emergency depressurization is required when the offsite radioactive release rate cannot be maintain below the GE level and the primary system causing the radioactive release has not been isola':ed. This emergency blowdown may be performed prior to reaching the GE threshold since an earlier blowdown could reduce onsite and offsite dcses. However, an earlier emergency blowdown may not be advisable in all events." This is not a reason for performing an emergency blowdown for T-104 Performing an emergency blowdown per T-I 04 does not prevent exceeding the GE offsite release limit. This is not a reason for performing an emergency blowdown for T-104 WA System ~ 295038 - High Off-Site Release Rate I Importance:
ROISRO 3.6 13.9 . ~- WA Statement EK3.04 - Knowledge of the reasons for Emergency Depressurization as it applies to High Offsite -~ - -~ Release ___________
Rate. ~~ _ ~ REQUIRED __ MATERIALS:
~- ~ - ~ ~ ~- _____~~_________~__~-
__ ~ Notes and Comments:
_______~_~____
Peach Bottom Initial Reactor Operator NRC Examination December 2008 57. The Plant Reactor Operator (PRO) has just received a fire alarm from the Turbine Building.
The Fire Brigade has been dispatched. In accordance with FF-01 "Fire Brigade", the PRO is required to call for OFFSITE fire fighting support A. immediately if the fire spreads into two or more T-300 fire areas B. immediately if plant safe shutdown systems or ECCS are in jeopardy C. after 15 minutes if the Incident Commander reports the fire is NOT extinguished D. after 20 minutes if the Incident Commander reports the fire is NOT under control Peach Bottom Initial Reactor Operator NRC Examination December 2008 .~ Level ____ of Knowledge Diff icuk Time Allowance (minutes)
FUNDAMENTAL Question .- . - # . 57 - . RO _- .. Choice ~ ~ - ~ NIA Correct: Source : -_ Distractors:
0 New Exam Item
[7 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank ~_____-~--
D Ref ere nce( s) : A FF-01 Notes
~~ -- B Learning Objective:
C PSEG-0214L-03 Basis or Justification
_~-________~--__--~_-
Correct, per FF-01. - The size of the fire is not defined by FF-01. -~____ ~. ~-- - This is a requirement From ON-1 14 to scram the reactor. ~~ ~- This is associated with the time limit for performing EAL classifications.
Importance:
RO I SRO 2.9 13.1 1 ~___~___ ~~ _--- 600000 - Plant Fire On Site 1 WA System WA Statement AKI .02 - Knowledge of the operational implications of the following concepts as they apply to Plar REQUIRED MATERIALS:
___~____~_
~_______~__~~
____-______--
~ ~- Fire On-Site:
~__ Fire __ ~ -~ ~~ -- - _ _____-_~ - _________~~_~ - Notes and __ Comments:
Peach Bottom Initial Reactor Operator NRC EIramination December 2008 58. Unit 2 is operating with the following conditions:
0 0 0 Hydrogen pressure:
60 psig Main Generator volts:
21.9 KV Main Generator MW:
1100 MWe Main Generator VARS:
120 MVARS A grid disturbance results in steadily rising grid voltage.
The Main Generator voltage regulator responds as designed by attempting to lower Main Generator terminal voltage. With no operator action, this could result in Refer to the PBAPS "Generator Capability Curve" (Figure 1 of A0 50.7-2) PROVIDED SEPARATELY.
A. exceeding the Generator Under Excitation Limit B. overheating the Main Generator stator windings C. a Generator Lockout due to reverse power relay trip D. a Generator Lockout due to field over-excitation relay trip Peach Bottom Initial Reactor Operator NRC Examination December 2008 SRO Psychiometrics
_______ Level of Knowledge
.~ Difficulty
.~ Time Allowance (minutes)
~ ___ _~ - HIGH NIA ______ - Answer Key Choice Correct: Distractors:
A B C D ~____~~~~_______--~-~~~ ~ ~ _____-~-~ - __ Basis or Justification
_~_____ _______ With no operator action, if grid voltage rises above generator output voltage, generator VARS will lower.
If the voltage mismatch is large enough, VARS could llower to the point where the generator is operating in the OUT (LEAD) area of the Generator Capability Curve, and possibly to the point at which the Under Excitation Limit (UEL) is exceeded. This could happen if generator voltage was higher than grid voltage to the point where picking up additional VARS would result in exceeding the capability curve (B-C) for 60 psig hydrogen pressure.
-_______~~
~- ~ -~ A reverse power trip occurs when real load (MW) is reduced to the point where the grid supplies the generator.
The given conditions would not result in lowering MW, especially to the point of reverse power. Field over-excitation results from too high field current, which can be caused by the voltage regulator raising generator output voltage (VARS) too high. ________~~_________________~~~
~- ~- - ~ ___~_~_______.____~____~~_
-- Source : Reference( s): - Learning Objective:
t WA System Source Documentation 0 New Exam Item Modified Bank Item 0 Previous NRC Exam c] Other Exam Bank 700000 - Generator Voltage and Electric Importance: RO
/ SRO 3.3 13.4 ______ ~ Grid Disturbances
-~ WA Statement AKI .03 - Knowledge of the operational implications of under-excitation as it applies to Generator Voltage and Electric Grid Disturbances REQUIRED MATERIALS:
Ture I of A0 !iO.7-2 -~______ ~-
Peach Bottom Initial Reactor Operator NRC Examination December 2008 59. During a high reactor pressure transicmt on Unit 2, the Plant Reactor Operator notes the following Safety Relief Valve (SRV) indications:
0 0 0 11 SRV white lights are illuminated.
Three SRV red lights are illuminitted. All other SRV green lights are illuminated.
No safety valve white lights are illuminated. What was the minimum peak reactor pressure during this transient and what is the approximate current reactor pressure? Minimum Peak Pressure Approximate Current Pressure A. 1135 psig 1135 psig B. 1145 psig 1145 psig C. 1155 psig 1135 psig D. 1260psig 1145 psig Peach Bottom Initial Reactor Operator NRC Examination December 2008 Source : - _ Reference(s):
__ Learning Objective:
Choice -~ ~_ ~ ~ New Exam Item Ix] Modified Bank Item Ix] ILT Exam Bank _- 0 Previous NRC Exam 0 Other Exam Bank
-~ __ Tech Spec 3.4.3 __ PLOT-5001 A-4d Correct: Distractors:
C A B D Basis or Justification SRV setpoints range form 1135 psig to 1155 psig. If all 11 white memory lights are lit, then pressure reached 1155 psig. With three SRVs still open (red lights lit), pressure is at lowest range value of 1135 psig. If I135 psig was the peak pressure only 4 SRV's would have the white memory lights lit.
_. - _ - - - __ ~____~~_~
_________
-.. If 1145 psig was the peak pressure only 8 SRVs would have the white memory lights lit.
1260 psig is the setpcaint for safety valve (not SRV) actuation.
Psychometrics
>I _ ~ Level of Knowledge HIGH Difficulty I 1 Time Allowance
_-_________ (minutes)
SRO N/A WA System 295007 - High Reactor Pressure
~ Importance:
RO/SRO 3.9 14.1 __ ~ I WA Statement AA1.04 - Ability to REQUIRED MATERIALS:
~ -- ~~_ __ __ _-_~~ ~ __ -~-~_________-~
________~
Notes and Comments:
~ ___________ -
Peach Bottom Initial Reactor Operator NRC Examination December 2008 60. A plant startup is in progress on Unit 3. The crew is conducting HPCI surveillance testing in accordance with ST-0-023-301-3 "HPCI Pump, Valve, Flow and Unit Cooler Functional and In-service Test."
The reason for placing Torus cooling in service during this surveillance test is to prevent exceeding the
-_____ A. HPCI booster pump NPSH limit B. HPCl cooling water temperature limit C. Torus heat capacity temperature limit D. Torus temperature limit of Technical Specifications Peach Bottom Initial Reactor Operator NRC Examination December 2008 Question # 60 RO Choice Correct: Distractors:
D A B C _~- -_ -~___ -~ ____ _~- ~~ Basis or Justification Per Tech Spec 3.6.2.1, suppression pool temperature shall be less than or equal to 105 degrees F when at or above 1% power (on any WRNM channel) and testing that adds heat to the suppression pool is being performed.
Reactor plower has to be above the POAH to perform HPCI testing. Per ST-0-023-301-3, Note 1 of Step 6.4, 'I.. .torus cooling is placed in service to ensure the 105 degree F torus temperature limit is not reached."
--~.___ _-_____ --___ ~ _ ____ - The HPCl booster purnp NPSH limit is well above the maximum allowed torus temperature during testing (1 05 degrees F). The HPCl cooling water limit is well above the maximum allowed torus temperature during testing (1 05 degrees F). C is incorrect - the HCTL is well above the maximum allowed torus temperature during testing (1 05 degrees F). Psychiometrics 1 Level of Knowled e tFUNEZGE&--
Difficulty ~~ ~~ Time Allowance (minutes)
! SRO NIA _.___ Source: Reference( _____ s) : Learning Objective:
KIA System KIA Statement
-___ Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
._____~__
~- ~_-~___.____
-~ ST-0-023-301-3; Tech Spec:3.6.2.1
_____ _ - - PLOT-5007-1 k, -8 295013 - High Suppression F'ool Water [ Importance:
_ 3.6 13.8 AK3.01 - Knowledge of the reasons for Suppression Pool cooling operation as it applies to High SuEression Pool Water REQUIRED MATERIALS:
~~ ~- -~ _____ ______ ~~ ~ - -_~ -~_- - ~ _ __ ~- _ -~ ___~___~_ - Notes and Comments:
~_ ~ ___~____~_
Peach Bottom Initial Reactor Operator NRC Examination December 2008 61. The following conditions exist on Unit 2: Reactor power is 100% 0 0 An inadvertent Group II/III isolatiion resulted in loss of chilled water to the drywell TI-80 146 "Drywell Bulk Average Temperature Indicator
is presently inoperable While performing RT-O-40C-530-2 "Drywell Temperature Monitoring" the crew determined that out of service TI-2501 Point 136 reads 132 degrees F TI-250 1 Zone 1 temperature points (1 19 through 124) are Which of the following procedures must be entered? Refer to RT-0-4OC-530-2 "Drywell Temperature Monitoring" PROVIDED SEPARATELY.
ON- 120 "High Drywell Temperat&
T- 102 "Primary Containment Control" A. Yes No B. No Yes C. Yes Yes D. No No Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge - HIGH Choice ~. ____ SRo N/A J DifficuQ Time ---__ Allowance (minutes)
Correct: Source: -.____ -~ Reference@):
___ Learning Objective:
WA System WA Statement
~ ___.~_ ________.~__
A lB Distractors:
C D ~ ______ ~~ _- -_~ Basis or Justification Per ON-I20 Bases (alnd RT Notes), if Tl-80146 is unavailable then TI-2501 points should be used to determine Drywell Bulk Average Temperature.
Further, if all points in a zone are unavailable, the calculated temperatJre is invalid. In this case, the operator is directed to use TI-2501 Point 136 (or TR-804) and add 10 dlegrees F to determine Drywell Bulk Average Temperature. For the given conditions (142 degrees F), entry into OW-120 is required, but T-102 entry is not. ON-120 entry is required if Drywell temperature is above 140 degrees F; T-102 entry is not required.
__ ~ _ --_ _ __ T-102 entry is not required until Drywell temperature exceeds 145 degrees F. ON-120 entry is required if Drywell temperature is above 140 degrees F I Psvchiometrics I Source Documentation 0 New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam Other Exam Bank
_.__________-_
~ -~ __ - - -_ _-_ ON-I 20 Bases; RT-0-40C-530-2
_- _- PLOT-PBIG-1550-3 295020 - Inadvertent Containment Isolation 1 Importance:
RO / SRO AA2.02 - Ability to determine and/or interpret Drywell temperature as it applies to Inadvertent
_~_ _ ~ _ _~ ~ _~ __ __ _ ~_- ~ ~_ ~~ ~~- - _ ~_ ~-
Peach Bottom Initial Reactor Operator NRC Examination December 2008 62. A startup is being performed on Unit 3. The following conditions and trends are observed: Reactor pressure is 850 psig. At time 11:25 AM 'B CRD PUMP TRIP' (31 1 G-1) annunciator is received.
At time 1 1 :30 AM the 'A' CRD pump is started but immediately trips. At time 1 1 :33 AM, several accumulator trouble lights illuminate on the Full Core Display for withdrawn control rods.
At time 11:37 AM, CRD Charging Header pressure drops below 940 psig. In accordance with ON-107 "Loss of CRD Regulating Function", based on the above conditions you are required to
-___ (I) at time (2) . A. (1) perform a Manual Scram (2)11:37 AM B. (1) perform a Manual Scram (2) 11:57 AM C. (1) insert control rods in reverse order of the approved sequence (2) 11:37 AM D. (1) insert control rods in reverse olrder of the approved sequence (2) 11:57 AM Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge Difficulty Time Allowance (minutes)
~ --________
~______.._
-- HIGH Correct: ~ -~ NIA ~- ___ Distractors:
I Source Documentation Source: New Exam Item Previous NRC Exam Modified Bank Item 0 Other Exam Bank Ix] ILT Exam Bank ____~ ___ __ ON-107; Tech Spec 3.1.5 ~______ ~ -~~ _ Reference(s):
Learning PLOT-? 550-1 2a Objective:
~ __________
-~ A B I---- ic D Basis or Justification Per ON-107, if reactor pressure is e 900 psig, charging header pressure is 940 psig, and all control rods with accumulator trouble indicators lit .are not fully inserted, a scram is required immediately.. .there is not a 20-minute a I Iowan ce. 20 minutes to restore charging header pressure once the condition of both multiple accumulator lrouble alarms and low (e 940 psig) CRD charging header pressure is only allowed when reactor pressure is
> 900 psig. ~- __ ~ - - __________~-~___~
~ -~ The time is correct but the action (GP-9) is incorrect.
The time and action (GP-9) are both incorrect. - KIA System t __~_________
295022 - Loss of CRD Pumps
__~ - Importance: RO I SRO r 4.2 14.2 KIA Statement G2.4.47 -Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
~ Notes and Comments:
_ ________~
-~ -~ MATERIALS:
~ ______~ ~____ - __- _. ~________
_____
Peach Bottom Initial Reactor Operator NRC Examination December 2008 63. Unit 3 is operating at 100% power when the following conditions occur:
0 0 0 HIGH AREA TEMP (3 10 5-3) goes into alarm. Point #2 on TRS-3-13-139 (RCIC room) is 120 degrees F and rising. other points on TRS-3-13-1.39 are below the alarm level.
T- 103 "Secondary Containment Control" has been entered. A steam leak in the RCIC room has been confirmed. What are the operational implications of RCIC room temperature rising above 135 degrees F? Refer to T- 103 Table SC/T on the W,XT PAGE. 1. RCIC will automatically isolate. 2. RCIC will be required to be manually isolated.
- 4. An emergency blowdown will be required if RCIC CANNOT be isolated.
A. 1 and 3 B. 1 and4 C. 2and3 D. 2 and4 Peach Bottom Initial Reactor Operator NRC Examination NIT 2 230 December 2008 - UNIT: 190 - T-103 "Secondarv (Containment Control" TABLE SC/T-3 TEMPERATURE-ALARM AND ACTION LEVELS ALMM LEVEL (OF) ~~ ACT ION LEVEL (OF) AREA STATUS TORUS ROOM 1 35 li5 PT 8.9.14.15.20, OR 24 110 110 135 1 so PT 2 PT 3 RCIC ROOM OR HPCI ROOM 110 110 135 135 135 135 135 135 __I- -- PT 17 PT 29 A RHR ROOM OR C RHR ROOM B RHR ROOM OR 0 RHR ROOM 110 110 PT 23 PT 6 A CS ROOM OR C CS ROOM 110 110 ~~ TI-2(3)501 PT 151 TI-2( 31501 PT 152 - B CS ROOM OR 0 CS ROOM STEAM TUNNEL TI -2( 3)SOl PT 153 TI-2( 3)501 PT 154 110 110 175 PT 1 OR 16 A ISOL VALVE ROOM (SOUTH) 165 PT 12 B ISOL VALVE ROOM (NORTH) 165 130 PT 18 OR 21 PT 30 ISOL VALVE PIT 165' EL 140 1 so RWCU REGEN HX ROOM OR A NON REGEN HX ROOM OR 160 130 130 115 1 os PT 11 PT 28 PT 5 PT 10 OR 27 PT 4 NO ACT1 ON LEVEL -- 13s -___. B NON REGEN HX ROOM OR A OR B RWCU FLTR OEMIN ROOM OR RWCU BACKWASH VALVE ROOM GENERAL AREA 165' EL (MAY AFFEGT RPV LEVEL INST) 105 PT 22 Peach Bottom Initial Reactor Operator NRC Examination SRO Psychometrics HIGH NIA Source Documentation
-4 Time Allowance (minutes)
__ Level __ of Knowledge Difficulty
_ -~ ____-___ ____-____
__ Source : New Exam Item Previous NRC Exam Choice _.._______
Correct: ~ ___ Referencets):
Objective:
Learning Distractors:
Modified Bank Item ILT Exam Bank 0 Other Exam Bank
________-
-- ~-- ~- - _~~_ ~- -~ -- -- _- ~-_____- ARC 31 0 J-3; T-I 03 and Bases PLOT-1560-3 C Basis or Justification
_____ ~ _-__ ___ Both #2 and #3 are correct. T-I 03, step SCm-4 will direct isolating RCIC (isolate all systems discharging into the area.. .). T-103, step SCC-8 will direct a reactor scram if a primary system is discharging into the Reactor Building and any parameter exceeds an action level.
- I is incorrect. RClC will not automatically isolate until RClC area temperatures (RCIC room, pipe chase) reach 200 degrees F (Tech Spec value). _______--_____~_-____-~ - #I is incorrect. RClC will not automatically isolate until RClC area temperatures (RCIC room, pipe chase) reach 200 degrees F (Tech Spec value). # 4 is incorrect.
T-I 12 "Emergency Blowdown" is not required until the same parameter exceeds an action level in more than one area (and if the "primary system breach" has not been isolated).
- 4 is incorrect T-I I2 "Emergency Blowdown" is not required until the same parameter exceeds an action level in more than one area (and il the "primary system breach" has not been isolated).
__ ___ _- ~- -~._____~
__-____ _- ~_-~~_______-_-~
____-__--
Peach Bottom Initial Reactor Operator NRC Examination December 2008 64. A small steam leak has been reported. in the Unit 2 Reactor Building.
The following conditions are present:
0 0 There are NO ARMS in alarm Reactor Zone Vent Exhaust is reading above normal but NOT in alarm 2 VENT EXH STACK RAD MONITOR HVTROUBLE A (21 8 B-5) is alarming 2 VENT EXH STACK RAD MONITOR HUTROUBLE B (21 8 C-5) is alarming 2 VENT EXH STACK RAD MONITOR HI-HI A (21 8 B-4) is alarming Based on the above conditions, which one of the following actions is required?
A. Enter ON-104 "Vent Stack High
]Radiation" T-103 "Secondary Containment Control".
B. Enter ON-104 "Vent Stack High Radiation" T- 104 "Radioactivity Release Control". C. Enter T- 1 03 "Secondary Containment Control" ONLY. D. Enter T-104 "Radioactivity Relea,se Control" ONLY.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Source: Choice Correct: Distractors:
B A C D Basis or Justification ON-I 04 and T-I 04 miust both be entered (and executed concurrently)
____~__~~
____. ~ -~~~ The given conditions do not require entry into T-103.
The given conditions do not require entry into T-I 03. ON-I04 entry is also required SRO -'i NIA Time _____ Allowance (minutes)
Reference(s):
Learning 0 bject ive:
KIA System I KIA Statement Source Documentation New Exam Item 0 Modified Bank Item ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
-~~ __~__________
-~ - ~ ARC-218 B-4; ON-I 04; T-104 PLOT-2104-1
-~~ Importance:
RO / SRO - 3.8 14.5 295034 - Secondary Containment Ventilation High Radiation
~ _- __~_ ~-~
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 65. Unit 2 is operating at 100% power when the following occur:
0 C RHR PUMP ROOM FLOOD (224 C-5) annunciator is received The crew enters T- 103 "Secondary Containment Control" An Equipment Operator reports a leak from the 2C RHR heat exchanger Which one of the following will also occur as a result of these conditions?
A. Flood alarms in the other (A, B, or D) RHR pump rooms due to backflow through the floor drain system. B. Increased influent into the Waste Collector Tank due to additional input from the 2C RHR Room Sump.
C. Increased pump out of the Reactor Building Equipment Drain Sump due to overflow from the 2C RHR Room Sump. D. Increased pump out of the Reactor Building Floor Drain Sump due to additional input from the 2C RHR Room Sump.
Peach Bottom Initial Reactor Operator NRC E;xamination December 2008 SRO Choice -.-~___-___
Correct: Distractors: , A Source: - ___-_ Reference(s):
-_______ Objective:
WA System Learning B New Exam Item 0 Modified Bank Item ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
~ _- _ ~ ~____ ARC 224 C-5; T-103 Bases; M-368, M-518 - ~ ~_ ~- - - _ PLOT-5020-1 C, -1 d ~ _~_ - Importance: RO
/ SRO I--- 3.1 / 3.2 295036 - Secondary containment High Sump/Area Water Level ,___ . __- ~- ~~ ~ C Basis or Justification
-~ The RHR Room Sumps are pumped directly to the RB Floor Drain SLmp, which will cause the sump pumps to operate more frequently. Per Note in ARC 224 C-5, "floor drains are plugged during power opeiation to prevent flooding in more than one room. Per T-103 Bases, "only orie ECCS room flood alarm should be received for an ECCS suction piping rupture." One must irifer the same is true for an RHR heat exchanger leak. The RHR Room Sumps are pumped to the RB Floor Drain Sump, which is pumped to the Floor Chain Collector Tank, not the Waste Collector Tank. ___~-~---__-_____-___~---~
~ ~-_____-- - _______~ __~~__-~_~~__~-__ - - _~ --___ - ___~______~__ - ~ The RHR Room Sumps are pumped directly to the RB Floor Drain Sump, which overflows to the RB Equipment Drain Sump. The RHR Room Sumps do not overflow to the RB Equipment Drain Sump. --__ _-- K2.01 - Knowledge of the interrelations between Secondary Containment High SumpIArea Water Level --_____ REQUIRED MATERIALS:
-- Notes __ and Comments:
-_ --____ _ ~_ ~ ~~- ~_ ~ ~- ~- ~-- ~- -~ --- ~ ~~ - _
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 66. In accordance with OP-PB- 108- 10 1 - 100 1 "Simple Quick ActdTransient Acts", whicl \ one of the following tasks can be performed by Reactor Operators during a transient, as directed by the CRS, without immediate procedure references?
A. Manual initiation of Standby Liquid Control.
B. Manual driving of control rods during an ATWS. C. Manual adjustment of HPCURCIC speed to control injection.
D. Manual initiation of HPClI via individual component manipulation.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level of Knowledge FUNDAMENTAL Time Allowance (minutes)
~- Distractors:
1 A SRO N/A -- t Source : ____ ~___ Reference(s):
Learning Objective:
KIA System ____ ~ ID 0 New Exam Item 0 Modified Bank Item ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
_________-_____~~~-~
-- -~ OP-PB-108-101-1001
__- PLOT-1529-1 j __- -~ -- _____~___________~_-__--~
~- Importance:
RO / SRO 1 3.8 14.2 N/A ~ _~--- OP-PB-108-101-1001, step 3.1.8, considers adjusting process controllers in automatic to maintain a process parameter a "Transient Act".
Initiating SBLC is not listed under "Transient Acts" (tasks that can be performed by reactor operators during a transient as directed by the C, RS without immediate procedure reference). Therefore, the Rapid Response Card is required.
____~-~ ~- ~ ~ - - This is not a "Transient Act"; T-200 procedure is required.
This is not a "Transient Act"; Rapid Response Card is required.
____~-~~-___.~__~._
-. Source Documentation Peach Bottom Initial Reactor Operator NRC Examination December 2008 67. The ultimate responsibility for maintaining and fulfilling NRC license requirements and reporting changes to a license rests with the A. Site Operations Director B. Site License Coordinator C. Individual License Holder D. Regulatory Assurance Manager Peach Bottom Initial Reactor Operator NRC Examination December 2008 FUNDAMENTAL
_- NIA I Choice Source: Correct: I c 0 New Exam Item Modified Bank Item 0 Previous NRC Exam Other Exam Bank - LORT _____.____
___- I Distractors:
_____-_ ~ Basis or Justification
_____ ____-- -__ ___ Correct, as stated in paragraph 3.1 of OP-AA-105-101: "The ultimate responsibility for maintaining and fulfilling license requirements and reporting changes to licenses rests with the individual Licensee.
Incorrect per OP-AA-'I 05-101. __ ______~____.___ Incorrect per OP-AA-'105-101.
Incorrect per OP-AA-'105-101 i Psychometrics Level of Knowledgg Difficulty - Time Allowance (minutes)
SRO ___________.____
~ - -_ 0 ILT Exam Bank _~ ~ Reference( s): ____ ~ ~- _~ ~ __ ____ Learning PLOT-I 525-1, -3 Objective:
WA System NIA 3.3 13.8 _- ~. G2.1.4 - Knowledge of individual licensed operator responsibilities related to shift staffing, such as _ -_ REQUIRED MATERIALS:
~ -~ -__ ~ _ _ ~ ~______~___
~ Notes _________
and Comments:
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 68. When reporting critical parameters to the CRS, OP-PB- 10 1-1 1 1 - 100 1 "Strategies for Successful Transient Mitigation" requires the RO to report parameter, A. value and trend B. value and threshold C. threshold and trend D. value and milestone Peach Bottom Initial Reactor Operator NRC Examination December 2008 Psychtometrics SRO FUN DAM ENTAL NIA ___~ Level of Knowledge Difficuw _ Time ______-__
Allowance (minutes)
~ Answer Key ~. Choice ______-____
Correct: -________
Distractors:
A B C -. D ~-____________~-_______~~
Basis or Justification Per OP-PB-101-1 1 1-11 001 "Strategies for Successful Transient Mitigation', "critical parameter reports should include parameter, value and trend. Per OP-PB-101-1 I 1-1 001 "Strategies for Successful Transient Mitigation", the CRS (not the RO) "provides direction for frequency of reports on critical para meters including ,milestones and thresholds.
" Per OP-PB-101-111-1001 "Strategies for Successful Transient Mitigation", the CRS (not the RO) "provides direction for frequency of reports on critical parameters including ,milestones and thresholds." Per OP-PB-101-111-1001 "Strategies for Successful Transient Mitigation", the CRS (not the RO) "provides direction for frequency of reports on critical parameters including milestones and thresholds."
~___________-____
__ --- ____ ______ _ _- ________._____---
-~__- __ ____________________~-.~~
__ _ -_ __.___ _-_____________
-~___ Source : .--~______
Reference( s) : - Objective:
KIA System KIA Statement
~________ Source Documentation 0 New Exam Item Modified Bank Item ILT Exam Bank Previous NRC Exam 0 Other Exam Bank .~______ .~ - ~ ____ _- ~ ~- ~___~_ OP-PB-101-111-1001 PLOT-DBIG-1529-1 b
___________~_
_-__- ____-_______~-_
~_ - _ __ __ verbal reports. ~ ~ _ _____~~ _ ~~~ ~ ~ - Notes and Comments:
Peach Bottom Initial Reactor Operator NRC Examination December 2008 69. Unit 2 and Unit 3 are at 100% power. Using Print E- 1 "Single Line Station Diagram" PROVIDED SEPARATELY, determine the automatic system response to a 50 14 Line fault.
A. The 215 breaker trips ONL,Y, then attempts to reclose.
B. The 205 and 2 15 breakers will BOTH trip, then the 205 breaker attempts to reclose. C. The 2 15 breaker trips ONLY, then the Unit 2 Main Generator will lockout.
D. The 205 breaker trips ONLY and its associated motor operated disconnects will open.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 Correct: ! B _- ~ - ~ ~- ~~~ Basis or Justification Breakers 205 and 215 will open on the fault and only the 205 breaker will attempt reclosure. The 21 5 breaker is a Unit output breaker and does not have a reclosure feature.
The 215 breaker is a IJnit output breaker and does not have a reclosure feature. Also, for a fault on the 5014 Line the 205 breaker would trip open as well. __ - ____ ___ ~_ ____~ -_______-
-~ ~_____ ~~ ~~ ~ ~ The Unit 2 Main Generator will not lockout. Output breaker 225 will remain closed. Breakers 205 and 215 will open on the fault. The 205 and 215 breaker motor operated disconnects are manually operated ONLY.
--_____ Difficulty - Time Allowance
_._______~____~ (minutes)
SRO NIA Source Documentation 1 Source: 1 New Exam Item Previous NRC Exam Modified Bank Item 0 Other Exam Bank
~---__-~_.~~______ ~~ ~ - ~ ~- __ -~ - ~___________
Reference(s):
__ Learning Objective:
KIA System importance: RO I SRO 3.9 14.3 ~-~ ~__ KIA Statement G2.2.15 - Ability to determine the expected plant configuration using design and configuration con documentation, such as drawin s, line-u s, tag-outs, etc. REQUIRED MATERIALS:
p'[FhpSheet
_____~ 1, Station Single Line Diagram Notes and Comments:
-__ _._______~___.___~___
~ __~~ ~
Peach Bottom Initial Reactor Operator NRC Examination MEASURED VALUE (Set) 87.3 December 2008 RETEST ALEZRT LIMITING VALUE R?UiGE VALUE INITIAL (Sec a s: Sec ) SAT UNSAT Ref 4 .3 4 Ref 4.3.3 (Set;) Ref 4.3.5 .c 56.5 86.5 r ?6,5 I 70. Given the following:
MO-2-23-057 "HPCI Torus Suction Outboard" was declared INOPERABLE for preventative maintenance. Following maintenance, operators performed a stroke test on MO-2-23-057 using a partial ST-0-023-301 -2 "HPCI Pump, Valve, Flow and Unit Coolers Functional and In-service Test". The stroke test data is shown below: Based on the guidance in ST-0-023-301-2 NOM-P-11 .I "Operability", MO-2-23-057
.- A. must remain INOPERABLE B. must be retested to determine operability C. is OPERABLE pending Engineering review D. operability status is indeterminate Peach Bottom Initial Reactor Operator NRC Examination December 2008 ______- Difficulty Time Allowance (minutes)
Choice Correct: ~~ SRO ~- NIA Distractors:
A B C D __ Basis or Justification
_ . __ Per ST-0-023-301-2 Limitation 4.3.2, any valve that exceeds its limitirig stroke time criteria shall be immediately declared inoperable.
-_ ~__ ~- This applies to valves that stroke-tested in the alert range initially and then again during a re-test This applies to valves that stroke-tested in the alert range initially and then again during a re-test.
Per NOM-P-11 .I there is no "indeterminate" status. The component is either operable or inoperable.
~ __________
.___ ___ -___ ____ Source : Reference(s):
Learning Objective:
KIA System Source Documentation New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
-____ ____________~
~ ST-0-023-301-2; NOM-P-11.1 PLOT-DBIG-1529-1 c --_ - ~ ___ ______ - - -~~ NIA Importance:
RO I SRO 2.9 14.1 _ ~ ______ _ KIA Statement operability requirements.
___,_____ ~ ~ ~- - - _- REQUIRED _____ MATERIALS:
~~ _ - - __~ ~~- ~__ ~- _- _- -_ Notes and Comments:
Peach Bottom Initial Reactor Operator NRC Examination
- 71. December 2008 An operator is needed to perform a filter alignment in an area where the general area radiation level is 60 mR/hr. 0 0 The job will take 30 minutes for one operator, but can be completed in 20 minute:t with two operators. Area radiation levels could be reduced to 30 mWhr with shielding.
It will take 20 minutes for one individual to install shielding.
To ensure the job dose is maintained "AS Low As Reasonably Achievable", -- (1) operator(s) should perfirm the lineup (2) shielding. Consider total personnel dose only.
A. (1)one (2) with B. (1) one (2) without c. (1) two (2) with D. (1) two (2) without Peach Bottom Initial Reactor Operator NRC Examination December 2008 Choice Correct: Distractors:
_.~______
Basis or Justification
__ _. 1 individual to perform lineup without shielding is a total exposure of 30 mR/hr (30 minutes in a 60 mFUhr field). 1 individual to install shielding is 20 mWhr (20 minutes in a 60 mR/hr Yield), plus one individual to perform lineup is 15 mR/hr (30 minutes in a 30 mR/hr 1 individual to install shielding is 20 mR/hr (20 minutes in a 60 mR/hr field), plus two individuals to perform lineup is 20 mR/hr (20 minutes in a 30 mR/hr field, times 2), ,For a total exposure of 40 mR/hr. ~- for a total exposure of 35 mR/hr. ________~~._____
__________ - 2 individuals to perform lineup without shielding is a total exposure of 40 mR/hr (20 minutes in a 60 mR/hr field, times 2). I I I I 4 Psychometrics Level of Knowledge 1 ~. Difficulty
~___ 1 Time Allowance (minutes) 1 SRO Source: Learning 0 bj ect ive: WA System WA Statement
______.___
_~_____.______ Source Documentation 0 New Exam Item Modified Bank Item ILT Exam Bank 0 Previous NRC Exam Other Exam Bank
_____,~~~_____
_- ~ - - ____~~________~_.~~~-~
~-_ RP-AA-400 PLOT-I 770-2, -3 N/A 1 Importance:
RO I SRO G2.3.12 - Knowledge of the radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, ___ -- ~ -~~ ~ __~__________~__~
-_ Notes and Comments:
~- ~ __ ~ _-_
Peach Bottom Initial Reactor Operator NRC Examination December 2008 72. Unit 2 is operating at 100% power when a steam leak occurs in the Reactor Building.
0 The Reactor Building exhaust duct radiation monitors reach the PCIS Group 111 setpoint. All systems operate as expected EXCEPT that all SBGT filter inlet dampers fail to open. Which one of the following would result from this event with no further operator action? A. Higher release rates through the Main Stack due to fission products not being adequately filtered.
B. An unfiltered ground-level radioactive release due to the Reactor Building not being maintained at negative pressure.
C. Higher release rates through the I Jnit 2 Vent Stack due to forced flow from the Reactor Building.
D. A monitored ground-level radioactive release due to the Reactor Building not being maintained at negative pressure.
Peach Bottom Initial Reactor Operator NRC E,xamination December 2008 Level of Knowledge
___ HIGH __ -___- __ Choice NIA Difficulty
___ __ Time Allowance (minutes) - --___-~ Correct: 1 B lA Distractors:
Basis or Justification
--_____---________
________._____
~- The Group Ill PClS isolation will trip and isolate Reactor Building ventilation.
The failed filter inlet dampers will prevent SBGT from maintaining Reactor Building negative pressure. This will result in an unmonitored and unfiltered ground-level release.
SBGT would not be exhausting Reactor Building air to the Main Stack ______- - -~ - - Reactor Building ventillation dampers close on a PClS Group Ill isolation and isolate the Reactor Building from the Vent Stack. The release would not be through a monitored path.
I Psvchometrics I Source: Reference(s):
Learning Objective:
KIA System _________
____.~____ Source Documentation New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank - -_ __~_~~_~______
~ - -_~ - T-I 03 Bases (step SCC-2) ~ ~--~ ~ -_ PLOT-5009A-3a N/A Importance:
RO / SRO 3.4 13.8 KIA Statement G2.3.14 - Knowledge of radiation or contaminatiori hazards that may arise during normal, abnormg emergency conditions or activities.
_~___.__ -____-___
__ __ REQUIRED _________~
MATERIALS:
__ ENE ___~_ Notes and Comments:
Peach Bottom Initial Reactor Operator NRC Examination December 2008 73. An ATWS is in progress on Unit 2. 'The CRS directed inhibiting the Automatic Depressurization System (ADS) per ' r- 1 17 "Level/Power Control".
This is done to prevent . A. core damage due to large irregular neutron flux oscillations B. exceeding 1 10 degrees F Torus temperature before boron is injected C. potential loss of, or inaccuracies in, RPV level instrumentation D. substantial fuel damage due to a large reactor power excursion Peach Bottom Initial Reactor Operator NRC E:xamination December 2008 Choice -_______ Correct: D Distractors: A B -____-~_-_ - - C -- ~ _ _- ~- __ ~ Basis or Justification The ADS safety function is inhibited to give priority to other systems ( e., provide additional time for SLC, RPS, etc. to perform their safety funclions).
From T-I 17 Bases: ADS initiation would complicate efforts to maintaii RPV level within required level ranges.
FURTHER, rapid and uncontrolled injection of large volumes of relatively cold, un-borated water from lorn pressure injection systems may occur. With the reactor either critical 3r shutdown on boron alone, the positive reactivity addition due to boron dilution and temperature reduction may result in a reactor power excursion large enough to cause substantial fuel damage. ADS is inhibited to prevent this from happening.
ADS initiation would riot cause large irregular neutron flux oscillations, ..it would cause a rapid reduction in reactor power due to voids. During an ATWS Torus temperature may exceed 110 degrees F before boron injection anyway due to SRV operation.. .this is not the reason for inhibiting ADS. Depressurization due to ADS initiation must also be accompanied by elevated Drywell temperature for this to occur ... this is not the reason fa- inhibiting ADS. ~ -~~___.-__~.______-_-_---_-_______~~-__~~--
____ -___~ ____ ~____ ~~_________~___
-_ -_ _- - _ ____.--__-~
__ ~_- Psychometrics I Difficulty - _____~_ Level of Knowledge FUNDAMENTAL 1 Time Allowance (minute9 ~________-_
__ Source Documentation Source : New Exam Item Previous NRC Exam Modified Bank Item ILT Exam Bank - __ Other Exam Bank - ____---- - __ Reference(s):
Learning Objective: KIA System PLOT-2 1 1 7-5a __________
___ IN1 A G2.4.22 - Knowledge of the bases for prioritizing safety functions during abnormallemergency Terations.
___~ _ ___._~~-_
_-_ -- - __ __ _ --_ --________~__
REQUIRED MATERIALS:
~- ~ ~____ --- Notes and Comments:
~____~ - --IF ~- -_ - -_ ~~ ~ - -~ _- -_- -
Peach Bottom Initial Reactor Operator NRC Elxamination December 2008 74. The Control Room had to be abandoned while HPCI was injecting into the RPV. An operator was stationed at the Unit 2 HPCI Alternative Shutdown panel and placed the HPCT TransfedIsolation switches in "EMERGENCY".
AFTER the transfer was completed, the following conditions are observed at the Alternative Shutdown Panel: Reactor level (LI-2-2-3-112) is 0 inches and rising Reactor pressure is 350 psig and steady CST LEVEL LOW (233 A-1) annunciates Based on these conditions, HPCI will A. trip due to low suction pressure B. re-align its suction path to the Toi-us C. remain running with suction from the CST D. trip due to placing the HPCI Transfer/Isolation switches in "EMERGENCY" Peach Bottom Initial Reactor Operator NRC Examination December 2008 Level ____~ of Knowwe __ HIGH Choice _______ Difficulty Time Allowance (minutes)
NIA Correct: Distractors:
C A B D ~ Basis or Justification HPCl suction swap to the Torus is defeated following transfer of contrd to the Alternative Shutdown Panel. Therefore, the HPCl suction path wil remain aligned to the CST. HPCl auto trips are defeated following transfer of control to the Alternative Shutdown Panel. HPCl interlocks are defeated following transfer of control to the Alternative Shutdown Panel. HPCl does not trip when control is transferred to the Alternative Shutd3wn Panel. --________ - _____-_~-~ ~ ~ __-~__~___________~_-_
__ -- _________~_~~-__________._~---
__- _~~-~______________~~___________.~
Source: ____~.~..-
~~~ Reference(s):
Learning 0 bjective:
KIA System Source Documentation 0 New Exam Item 0 ILT Exam Bank 0 Previous NRC Exam Modified Bank Item Other Exam Bank - LORT _- _______ _--__________~
______ -~ -~ -~_ ~ SE-10 Attachment 9; ARC-233 A-I ; M-I -S-36 PLOT-5023-4r
-~__ -- ______________
__ __- -~ - ~~ NIA ___- Importance:
RO I SRO 4.2 14.1 KIA Statement G2.4.34 - Knowledge of the RO tasks performed outside the main control room during an emergency and the resultant operational effects. REQUIRED MATERIALS: - TONE _____ ________-
~- ____- --- ~-- __~~________.
~ ~_______~ Notes and Comments:
Peach Bottom Initial Reactor Operator NRC Examination December 2008 75. Unit 2 is operating at 80% power when several Control Room annunciators are received, including the following:
A CONDENSATE PUMP BRK 'TRIP (203-E2) REACTOR HI-LO WATER LEVEL (2 10-H2) GENERATOR PROTECTION CIRCUIT ENERGIZED (206-L 1) The alarms have been validated. Prioritize these alarms and determine the required operator action.
A. Perform GP-4 "Manual Scram".
B. Perform GP-9-2 "Fast Reactor Power Reduction".
C. Insert ALL GP-9-2 Appendix 1 control rods ONLY. D. Verify A and B Recirc Punips runback to 45%.
Peach Bottom Initial Reactor Operator NRC Examination December 2008 A B Choice Basis or Justification
~- -~ ______ -~ A valid GENERATOR PROTECTION CIRCUIT ENERGIZED annunciator indicates a loss of Stator Cooling.
OT-I 13 directs a manual scram IAW GP-4 if a valid loss of Stator Cooling exists and generator load is greater than 7760 amps (-23%0 reactor power). This action is directed by OT-I 00 for a low reactor water level condition, based on availability of makeup capability. This action would be appropriate if it weren't for the loss of Stator Cooling condition.
_ _ - _______ ._____ __ Correct: D Distractors: The given conditions indicate a trip of the 'A' Condensate pump, which would result in a Recirc runback to 45% if Feedwater flow was > 85%. With the Unit at 80% power, Feedwater flow is < 85%; therefore, a Recirc runback will not occur when the Condensate pump trips.
_ Allowance (minutes)
-- . I Psvchometrics I Level of Knowledge HIGH Difficulty SRO N/A Source : I-- Objective:
Source Documentation 0 New Exam Item Modified Bank Item ILT Exam Bank Previous NRC Exam - 2007 Other Exam Bank ___ ~________________~~____~_
-~ ~ -~ GP-9-2; GP-4; OT-100;
__~_- OT-I 12; OT-I 13 - -~ PLOT-1540-3 NIA 1 Importance:
RO 1 SRO I 4.1 14.3 interpret the smnificance
~- ~ of each annunciator or alarm. __ ~ _________~_____ - __ __ ~~
Peach Bottom Initial Senior Reactor Operator NRC Examination December 200:1 1. Restoration of ST-0-0 1 1 -30 1-2 "Standby Liquid Control Pump Functional Test for IST" has been completed.
HV 1 1 - 15 "SBLC Discharge Header to RPV Outboard Isolation Valve" was inadvertently left in the CLOSED po,sition. Refer to the portion of P&ID M-358, Sheet 1, on the NEXT PAGE. Subsequently, Unit 2 is manually scriunmed due to loss of both CRD pumps. An ATWS occurs. 0 0 The CRS directs SBLC placed in-service The URO places SBLC in service:
using RRC 1 1.1-2 "SBLC System Initiation During a Plant Event" and reports the following:
o Both squib valves fired o SBLC pump dischargc:
pressure is 1400 psig o RWCU is isolated Predict the impact of these conditions on the SBLC System to identify the correct procedural response.
Refer to the portion of T- 10 1 "RPV Control" provided AFTER THE NEXT PAGE. Perform A. T- 10 1 "RPV Control" step RC/Q-16 B. T-210 "CRD System SBLC Injection" C. T-211 "CRD System Non-enriched Boric Acid and Borax Injection" D. T-212 "RWCU System SBLC Injection" Peach Bottom Initial Senior Reactor Operator NRC Examination December 200h ~-__.__~ ~_-~_____~__--_________--~_-~---_~._
~ r-- SRO Psychiometrics
~ -_ Level of Knowledge Difficulty
___~_.___
Time Allowance (minutes)
_~_ ___ HIGH 1 OCFR55.43( b) 5 Choice Correct: Source: __~______
Reference&
Objective:
Learning D New Exam Item Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank __~ ~- - ________.~____________
-_ ____ P&ID M-358, Sheet
_~____ 1; ST-O-011-301-2 PLOT-5011-4h A B C Answer Key ___ ~-_____ ~~~____-_____
~___ __ _~~~______~~~_
~ ___ _ Basis or Justification
_- ~- ~- Based on the given conditions, SBLC is not injecting into the RPV: 1490 psig pump discharge Ipressure indicates the SBLC pump discharge relief valve is lifting (due to the closed header isolation valve). T-210 and T 21 1 cannot be performed without at least one CRD system pump available.
Therefore, T-212 is the only option available, which can be implemented even though _~ RWCU .~ is isolated.
Execution of T-101 step RC/Q-16 is based on SBLC injecting into the RPV. SBLC will not inject into the RPV with header isolation valve HV-2-11-15 closed. The candidate must know that T-210 cannot be performed without at least one CRD system pump available.
In other words, use of T-210 requires CRD system piping arid an available CRD pump. The candidate must know that T-211 cannot be performed without at least one CRD system pump available.
In other words, use of T-211 requires CRD system piping arid an available CRD pump.
_~__~___________~
~- ~ -~ _______~~___
______________~____~
~ ______~____~-~ - ~________
~___-_--_______~_- - ~ SRO 3.2 WA System WA Statement A2.07 - Ability to (a) predict the impacts of valve closures on the Standby Liquid Control System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of 21 1000 - Standby Liquid Control
~~ ~- ____. ~ __ . ___..______.._____ . _~ _.
Peach Bottom Initial Senior Reactor Operator NRC Examination - Reactor Water Level (inches) 2. Drvwell Torus Temp Pressure (psi& Jderrrees F) An ATWS condition exists on Unit 2.. +23 30 . December 2008 4 87 6 92 7 96 9 99 .- The Unit RO and the PRO have been reporting the following data to the Unit Supervisor:
0 2 3 4 6 24 20 12 3 5 1 -. The HPCI system is being used for RPV level control.
Based on the above set of data. what
- step of T-117 "Level /Power Control" must be performed next?
Refer to T- 1 17 PROVIDED SEPAR/iTELY.
Step A. LQ-11 B. LQ-13 C. LQ-18 D. LQ-31 Peach Bottom Initial Senior Reactor Operator
'NRC Examination December 2008 1 OCFR55.43(13)(5)
Answer Key Source: ________ __ Reference(s):
_____ Learning Objective:
I Choice New Exam Item 0 Modified Bank Item 0 ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
__ --_ _ ________~~____-~._~- - ~- T-117 PLOT-PBIG-2117-3 1 Correct: REQUIRED MATERIALS:
Distractors:
T-I 17 "PowerlLlevel Control" --~~p- .- - -~ - __ ~- B A C ~ D ___-_____________~~_~._~_~_~-~
Basis or Justification With reactor power initially
>~YO AND Torus Temperature
<I IO degrees F, T-I 17 step LQ-13 will require performance of T-240 Attachment 1 Figure 1 to lower RPV level < -60 inches. If RPV level goes above
-60 inches AND reactor power goes
>4% during performance of subsequent T-I 17 steDs, checkhe-check step LQ-15 directs re-entry at step LQ-6 which will reqire re-performance of step LQ-13 to lower RPV level < -60 inches.
Step LQ-11 would only be performed if Torus temperature was > 110 degrees F. The highest Torus temperature was 97 degrees F. With RPV level above -60 inches and reactor power
> 4%, checkhe-check step LQ-I 5 directs re-entry at step LQ-6 which will require re-performance of step LQ-13 to lower RPV level < -60 inches. Step LQ-18 would only be performed if RPV level was not intentionally lowered to control reactor power. With RPV level above -60 inches and reactor power > 4%, checkhe-check step LQ-15 directs re-entry at step LQ-6 which will require re-performance of step LQ-13 to lower RPV level performed NEXT. _p-___ _- -~ _~~____.______
~- __ -___-- _ ___________~~- - --_-_______~
__~__~___~
__-____________________~~_______~_~__-
__ -60 inches. Step LQ-31 would not be ______ __________
______. -. - Source Documentation WA System 21 5005 - APRM/LPRM 1 Importance:
SRO Notes - and Comments:
-. -
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2005; 3. Unit 2 had a stuck open SRV and wa:s shutdown using GP-4 "Manual Scram." Subsequent damage to the SRV tailpipe resulted in a steam leak into the Drywell. The following conditions exist: Reactor Water Level is -20 inches and rising rapidly WV pressure is 240 psig and lowering Drywell pressure is 9 psig and rising Torus pressure is 6 psig and rising Torus temperature is 182 degrees F and rising 2B Core Spray pump is injecting at 3600 gpm RCIC is injecting at 600 gpm CST level is 4.5 feet Predict the impact of these conditions on the RCIC System to determine the required procedural direction. Refer to the NEXT PAGE for the applicable portion of T-102 "Primary Containment Control" the RCIC NPSH Curve.
A. Continue to inject with RCIC using RRC 13.1-2 "RCIC System Operation During a Plant Event."
B. Lower RCIC system flow to ensure adequate NPSH using RRC 1 3.1-2 "RCIC System Operation During a Plant Event."
C. Remove RCIC from service using RRC 13.1-2 "RCIC System Operation During a Plant Event" due to high
'Torus temperature and adequate core cooling. D. Defeat the auto open signal to the Torus suction valves using A0 13.3-2 "RCIC Torus Suction Valve, Defeat of Auto Open Signal" and realign RCIC suction to the Torus.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 ______ ____- ~-_____-~_____________________ - 1 OCFR55.43( 3)(5) Choice Source: Reference(s):
Learning Objective:
KIA System __ Correct: New Exam Item 0 Modified Bank Item 0 ILT Exam Bank
-- 0 Previous NRC Exam 0 Other Exam Bank T-102 and Bases -~ ~-~ PLOT-5011-4d
_______-_
___~ _~ __ __- ____- 217000 - Reactor Core Isolation Cooling -Importance:
SRO Distractors:
C ___ A -.__ B D RClC suction automatically swaps to the Torus when CST level IS < 5.25 fee-. T- 102 CAUTION #5 baser; states that Torus temperatures above 180 degrees = can cause high bearing oil temperature, which can lead to possible bearing damsge and control valve operational issues.
Since RClC operation is NOT required for adequate core cooling, IRClC should be shutdown using RRC 13.1-2 "RCIC System Operation During a Plant Event" (this should be used to secure the system since it was used to put the system in service following the transiek RClC System operation is not required for adequate core cooling. Based on T- 102, Caution #5, Torus temperature above 180 degrees F can lead to possible bearing damage and control valve operational issues.
Per T-I 02 bases, the equipment (RCIC) "may still be used if required to maintain adequate core cooling." Adequate core cooling can be maintained by 2B Core Spray. For the given Torus watler temperature and pressure, per T-102 sheet 3, RCIC NPSH is adequate.
__ ~~ ~- __ __~ - -~ ~_ -- .. _______ This choice suggests transferring RClC suction from the Torus (due to high Torus temperature) back to the CST. However, use of A0 13.3-2 is only used when RCIC logic bus 'A' power cannot be restored, which is not the case here. RClC suction cannot be transferred back to the CST due to the low CST level (< 5.25 Lp-- .___~____~______-___--____
______.__~__
__ KIA Statement A2.08 - Ability to (a) predict the impacts of loss of lube oil cooling on the RClC System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those ~- ~~ _- -~~ abnormal conditions or operations.
~ __~ _-___-____-~_~
-~ ~ - ~ ~~~ -~ -~ --______~
~- -~ -~ ~~~- ~- -~
Peach Bottom Initial Senior Reactor Operator NRC Examination December 200t1 4. Given the following:
0 0 Unit 2 is operating at 100% power. Core alterations are in progress 011 Unit 3. The 'A' SBGT filter train is determined to be inoperable. Refer to Technical Specification 3.6.4.3 PROVIDED SEPARATELY to determine how these conditions apply to (1) Unit 2 and (2) Unit 3. A. (1) Unit 2 operation may continue for up to 7 days. (2) Unit 3 core alterations may continue for up to 7 days. B. (1) Unit 2 operation may continue:
for up to 7 days. (2) Unit 3 core alterations must be immediately suspended.
C. (1) Unit 2 must be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. (2) Unit 3 core alterations may continue for up to 7 days. D. (1) Unit 2 must be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. (2) Unit 3 core alterations must bc immediately suspended.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Choice Correct: Distractors:
A Basis or Justification
.~ ___________________
-~ - -_ Per TS 3.6.4.3 bases a filter train is part of a SBGT subsystem.
For TS 3.6.4.3 Condition A, with one SBGT subsystem inoperable, it must be restored to operable status within 7 days. If not, Unit 2 must be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. For Unit 3, after 7 days either the operable SBGT subsystem must be immediately placed in service or core alterations must be immediately suspended.
Therefore, Unit 2 operation may continue for up to 7 days and Unit 3 core alterations may continue for up to 7 days. B Unit 3 core alterations may continue for up to 7 days -_ ____ - Unit 2 operation may icontinue for up to 7 days. Unit 2 operation and 1Jnit 3 core alterations may continue for up to 7 days. _~~_______________~______~
~ :A I Psvctiometrics I ______~ ~ Difficulty - I Time Allowance 1 OCFR55.43bM2)
Source : 0 bj ect ive: ~.___ Source Documentation New Exam Item 0 Modified Bank Item 0 Previous NRC Exam Other Exam Bank ILT Exam Bank
~-~ ~ _________.~~
__-~- ~ -- TS 3.6.4.3 and Bases (pre-AST revisio3 PLOT-5009A-8 261 000 - Standby Gas Treatrnent System Importance: SRO 4.7 WA Statement
-_ -_______~
____~_~.~___
_-_________~-
Peach Bottom Initial Senior Reactor Operator VRC Examination December 2008 5. Unit 2 was operating at 100% power when the following conditions occurred:
0 0 0 0 2A-2C BATTERIES GROUND (220 F-1) is received in the control room A0 57B.12-2 "125/250 VDC 2A and C Station Battery Ground Investigation" is in progress.
The ground detection system (GIS-80896A) indicates a ground value of 13,500 ohms. The ground has been isolated to Dliv. I 250 VDC bus 20DO12. Which of the following describes (1) ,what systems/components are impacted by this ground and (2) the correct procedural response to this condition?
A. (1) RCIC DC motor operated valves.
(2) Enter SE-13 "Loss of a 125 or 250 VDC Safety Related Bus." B. (1) HPCI DC motor operated valves.
(2) Enter SE-13 "Loss of a 125 or 250 VDC Safety Related Bus."
C. (1) RCIC DC motor operated valves.
(2) Continue with ground isolation IAW A0 57B. 12-2.
D. (1) HPCI DC motor operated valves.
(2) Continue with ground isolation IAW A0 57B. 12-2.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 1 OCFR55.43(b)(5 Choice Source: Reference(s):
Learning Objective:
Correct: New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam Other Exam Bank
~_____~__ - -~ ARC 220 F-1; A0 57B.12-2; SE-13
__ -~ __ PLOT-5057-1 d C DistractoFi-I.-- B D _____ _____ Basis or Justification
_~__________
~_________~
__ RClC DC motor operated valves ARE powered from Div.
I 20D012. 7 he correct response is to continue with ground isolation IAW A0 57B.12-2, which directs further actions to isolate the ground (Le., circuit, load, et.=.). RClC DC motor operated valves ARE powered from Div.
I 20D012, btit there are no SE-13 entry requirements based on DC system grounds HPCl DC motor operated valves are powered from Div.
II 250 VDC bus 20DO11. In addition, there are no SE-13 entry requirements based on DC system grounds.
HPCl DC motor operated valves are powered from Div. II 250 VDC bus 20D011. -~ ____ ~____ ~___________
__~ ________ ___ ~____-__~_____
__ __~____~_________,__~~
~~ Source Documentation Importance:
SRO I 3.2 KIA System WA Statement A2.01 -Ability to (a) predict the impact of grounds on D.C. Electrical Distribution and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal
______ conditions or operations.
263000 - D.C. Electrical Distribution ~_~_____~-~--
_~ ~-~__________~ ~ ~ REQUIRED MATERIALS=
iNoNE Notes and Comments:
I -
Peach Bottom Initial Senior Reactor Operator
'NRC Examination December 2008 6. Unit 2 is in MODE 1 at 100% power. 0 0 The URO notices the RED light for the 2A Recirc MG Set Drive Motor Breaker is NOT lit. Investigation reveals a jailed trip coil in the breaker. Refer to Technical Specification 3.3.4.. 1 PROVIDED SEPARATELY to determine which one of the following is the MOST LIMITING applicable action(s) for these conditions.
A. Restore ATWS-RPT trip capability within 14 days. B. Restore ATWS-RPT trip capability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
C. Remove the associated recirculation pump from service within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> OR Be in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. D. Restore ATWS-RPT trip capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Otherwise, remove the associated recirculation pump from service within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> OR Be in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Peach Bottom Initial Senior Reactor Operator
'NRC Examination December 2008 Time Allowance (minutes)
Question # 6 SRO - - Choice SRO IOCFR55.43(b)(2
~~ Correct: D - B C -____- ___ Basis or Justification
________ ~______~~~
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion time is associated with the Required Action of Condition C. Per TS 33.4.1 Bases, Condition C applies to multiple, inoperable, un-tripped channels within both Functions not maintaining ATWS-RPT trip capability, as is the case with an inoperable Drive Motx breaker. Condition D applies if the I-hour completion time of Condition C is not met. Per TS 3.3.4.1 Bases, Condition A does not apply if the inoperable channel is the result of an inoperable breaker. Condition D must be entered and its Required Actions taken. __-____~_____________~_~ 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is the completion time with only one function of ATWS-RPT trip capability not maintained.
With a failed trip coil, the breaker will not automatically trip, which results in both functions of trip capability being lost. This is the same as choice
'D' except the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> completion time associated with the Required Action of Condition C is not included.
~________~~____________ - Source: Objective:
Statement Source Documentation 0 New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
-~ -- - Tech Spec 3.3.4.1 and Bases
~ -~ ~ PLOT-5002-8 202001 - Recirculation 1 Importance:
SRO -~ ._ ~ ~ Notes and Comments:
_~ ~ ~-
Peach Bottom Initial Senior Reactor Operator NRC Examination December 200:: 7. Unit 2 Traversing In-Core Probe (TIP) System operation is in progress for an LPRM calibration. A subsequent Feedwater transient results in the following conditions:
The Reactor was manually scrammed. RPV level lowered to
-10 inches and is now at +20 inches. The PRO reports the in-service TIP detector is driving into the core and the ball valve RED light is lit.
Based on the above conditions, (1) assess TIP operation and (2) determine the required action.
A. (1) RPV level did NOT reach the TIP isolation setpoint.
(2) Direct the Reactor Engineer to continue the TIP trace IAW RE-35-2 "TIP System Operation."
B. (1) KPV level has recovered above the TIP isolation setpoint.
(2) Direct the Reactor Engineer to continue the TIP trace IAW RE-35-2 "TIP System Operation."
C. (1) The TIP ball valve failed to isolate automatically. (2) Direct the PRO to manually withdraw the TIP detector and close the ball valve IAW SO 7F.7-A-2 "TIP System Isolation in Event of Containment Isolation."
D. (1) The TIP ball valve failed to isolate automatically.
(2) Direct the PRO to manually fire the shear valve IAW SO 7F.7-A-2 "TIP System Isolation in Event of Containment Isolation."
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 - .____~ Level of Knowledge HIGH Choice c Correct: Difficulty - ~- Time Allowance (minutes)
____ 1 OCFR55.43( 3 (5) -~ Distractors:
Source: Reference(s):
Objective:
Learning L 1 0 New Exam Item [XI Modified Bank Item ILT Exam Bank ~- Previous NRC Exam 0 Other Exam Bank GP-8.B; SO 7F.7.A-2 PLOT-5007F-6 b
___~__~_~_~_~________~~
____~___~________._________
__~ ___ Basis or Justification The TIP detector should automatically withdraw to shield and the ball ivalve should automatically close at l-inch RPV level. For a TIP failure to istilate, GP-8.B directs manual isolation IAW SO 7F.7-2, which directs manually withdrawing the TIP detector to shield and closing the ball valve. The TIP detector should automatically withdraw to shield and the ball valve should automatically close at l-inch RPV level. ~~______ ~ ~___ ____ TIP operations would not be allowed to continue following an isolation fai I u re. The shear valve is only fired if the detector can not be retracted and then only if directed by the Shift Manager.
~- Importance:
SRO 3.7 ! .______~ .______- - -~ KIA System KIA Statement A2.07 -Ability to (a) predict the impact of failure to retract during accident conditions on the TIP System and (b) based on this prediction, use procedures to correct, control, or mitigate the consequences of this abnormal condition oroperaition. REQUIRED MATERIALS:
_______ Notes and - Comments:
215001 - Traversing ln-core Probe
~____ __- ~- ~_~______
__ __ ~_~_~_________~_~~________
Peach Bottom Initial Senior Reactor Operator NRC Examination December 200;3 8. Unit 2 is operating at 100% power when a pneumatic supply line failure causes outboard MSIV AO-86D to rapidly close.
Which one of the following describes (1) the plant impact, if any, and (2) what procedural actions must be taken by the CRS? A. (1) An automatic reactor scram ill occur due to a subsequent Group I isolation.
(2) Enter and execute T-101 bbRF"V Control."
B. (1) An automatic reactor scram u7ill occur due to high neutron flux. (2) Enter and execute T-101 "RP'V Control."
C. (1) An automatic reactor scram will NOT occur. (2) Reduce power IAW GP-5 "Power Operations."
D. (1) An automatic reactor scram will NOT occur. (2) Operation may continue at 100% power IAW GP-5 "Power Operations."
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Level of Knowledge HIGH Correct: Distractors:
SRO lOCFR55.43(!~)(5 Difficub Time Allowance (minutes)
__ . ____~._ -._______-- ~ ~ A- Source: - Reference(s):
Learning Objective: - - __ -__ ____ . Basis or Justification Per Chapter 14 of the UFSAR (T&A analysis), rapid closure of a single MSIV at 100% power will result in a high neutron flux scram. A concurrent high reactor pressure condition will require entry into T-101. The three un-isolated steam lines will pass 100% steam flow without exceeding the high steam flow isolation setpoint
(-140% of rated). ~ _- __________
0 New Exam Item [XI Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
_________~~-
___ ~-__________~~~~ - ~ GP-5; UFSAR PLOT-5001 A-6b Per GP-5, Table 1, the reactor can operate up to 75% power with 1 MSIV closed. Closure of a single MSIV at 100% power will result in a reactoa- scram. Per GP-5, Table 1, the reactor can operate up to 75% power with 1 MSIV closed. __-____ ~. _. ~. .. ~ __ WA System I 239001 - Main and SRO WA Statement 4.2 ._ .. A2.03 -Ability to (a) predict the impact of MSIV closure on the Main and Reheat Steam System and (b) based on this prediction, use procedures to correct, control, or mitigate the consequences of this - - -~ ~~ ~ -~ __ __-- ~ ~ -_ -_______ __ Notes and Comments:
__~-
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 9. Unit 2 was operating at 100% power ,with the OPRM System inoperable when the '2B' Recirc pump tripped. The following conditions currently exist:
0 0 A loop flow (FI-2-2-3-092B) is 4.5 Mlbdhr B loop flow (FI-2-2-3-092A) is 5; Mlbmhr Indicated Core Flow (FR-2-2-3-095 black pen) is 50 Mlbm/hr APRMs are oscillating between 50 and 54% in 4-5 second random intervals Assess these conditions and identify the correct procedural action.
Refer to A0 60A. 1-2 "PBAPS Backup Stability Solution Power Flow Operation Map" PROVIDED SEPARATELY.
The plant is operating in (1)-.. . The required action is to (2) A. (1) Region 1 (2) scram the reactor and enter T-11 00 "Scram" due to being in Region 1 B. (1) Region 2 (2) insert all GP-9-2 control rods per GP-9-2 "Fast Reactor Power Reduction" due to indications of Thermal Hydraulic Instability C. (1) Region 2 (2) immediately exit Region 2 by raising '2A' Recirc pump speed using SO 2A. 1 .D-2 "Operation of the Recirc Pump Speed Control System" D. (1) the normal operating region (2) perform the follow-up actions of OT- 1 12 "Unexpected/Unexplained Change in Core Flow" Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 - Level of Knowledge - HIGH Answer Key - Difficulty - Time _._ __- Allowance (minutes)
~- - Choice REQUIRED MATERIALS:
Correct: A0 60A.1-2 "PE)APS Backup Stability Solution Power Flow Ooeration Mady Distractors:
A B C D ~ Basis or Justification
___ _- ___~___~ The calculation of core flow 50-2(5)
= 40 Mlbm/hr
/ 102.5 Mlbm/hr = 39% (alternatively, 40 Mlbm/hr can be found on the upper 'x' axis). Plottins 40 Mlbm/hr vs. 50-54% power shows the reactor is operating in Region 1 Per A0 60A.1-2, step 4.5, the correct action for operating in Region 1 of the Power Flow _- Map ~ is to perform an immediate scram.
The reactor is operating in Region 1. If a core flow calculation error is made, the candidate could believe the reactor is operating in Region 2 The indications provided do not meet the criteria for THI, but inserting GP- 9-2 rods would be a correct action if operating in Region 2. Per A0 6ClA.I- 2, a reactor scram is rlequired anytime THI is occurring while this A0 is in effect (Le., OPRMs _____ inoperable).
The reactor is operating in Region 1. If a core flow calculation error is made, the candidate could believe the reactor is operating in Region 2 Raising recirc pump speed would be a correct action if operating in Region If a core flow calculation error is made, the candidate could believe the reactor is operating just inside the normal region -_______________
2 without indications
____~_________________
of THI. ~~ ___ Source: Reference(sk Learning 0 bj ect ive
- KIA System
_~___~__ KIA Statement Source Documentation New Exam Item 0 Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam Other Exam Bank . -~ __ - OT-I 12: A0 60A.1-2 PLOT-PBIG-1540-3.
-4 295001 - Partial or Complete Loss of Forced Core Flow Circulation
~ Importance:
s3~g0 . .- AA2.01 - Knowledge of the operational implicatioris of the following concepts as they apply to Partial or
_ __ - ~_ __ Complete Loss of Forced Core Flow Circulation: Power/Flow Map.
_- _._
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 09:OO FB 78 27 1 \\\\\\\\\\\\\\\\
\\\\\\\\\\\\\\\\
09: 15 FB 75 210 09:30 FB 73 189 09:45 FB 76 108 10:oo FB 74 146 - 10. A fire in the Cable Spreading Room required evacuation of the Main Control Room. The following conditions exist on Unit 2 during implementation of SE-IO "Plant Shutdown From Alternative Shutdown Panels":
0 HPCI has been tripped and isolated. Alternate shutdown cooling has been established via the 'A' SRV. Refer to the SE-10 Attachment 11 Cooldown Log below AND the applicable portion of SE- 10 Sheet 2 on the NEXT PAG*i to determine the status of the cooldown rate relative to the SE-10 hourly limit.
In addition, why does SE-10 direct maintaining the
'A' SRV tailpipe temperature above 100 degrees F? Unit 2 SE- 110 Attachment 1 1 Date log started Today Sheet 1 Cooldown Log The cooldown rate is
-( 1 )-- the SE- 10 hourly limit.
The 'A' SRV tailpipe temperature must be maintained above 100 degrees F in order to prevent (2) . A. (1) above (2) excessive stress on the SRV tailpipe B. (I) below (2) excessive stress on the SRV tailpipe C. (1) above (2) exceeding the minimum reactor vessel temperature limits of Tech Spec 3.4.9 D. (1) below (2) exceeding the minimum reactor vessel temperature limits of Tech Spec 3.4.9 Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 ~ Time Allowance (minutes)
Level of Knowledge Diff icuIty ~pp HIGH Choice SRO IOCFR55.43(h 5 . - Correct: Source : __ Reference(s):
Learning Objective:
Distractors:
New Exam Item 0 Modified Bank Item 0 ILT Exam Bank Previous NRC Exam 0 Other Exam Bank
~pp______p~
___~-- -~~~~~p ~ ~~~~ - ~- ~ ___ SE-10 and Bases pp PLOT-I 555-7, -9 C Basis or Justification pp~~-~~-~--~--~--~
___~ The SE-10 cooldown rate limit is 80 degrees F/hour. The calculated cooldown rate from the Cooldown log is 85 degrees F. Per SE-10 Bases, 'A' SRV tailpipe temperature must be maintained above 100 degrees Fa to ensure compliance with Tech Spec Figure 3.4.9-2 (Temperature/Pressure Limits for Non-Nuclear Heatup and Cooldown Following a Shutdown).
A Part 1 is correct; part 2 is incorrect. Per SE-10 Bases, 'A' SRV tailpipe temperature must be rnaintained above 100 degrees F to ensure Both parts are incorrect. The calculated cooldown rate is above the SE-10 limit of 80 degrees F. Per SE-10 Bases, 'A' SRV tailpipe temperature wust be maintained above 'I 00 degrees F to ensure compliance with Tech Spec Figure 3.4.9-2. Part 1 is incorrect; part 2 is correct. The calculated cooldown rate is above the SE-IO limit of 80 degrees F. ~. compllance
__p~ with Tech - Spec Figure 3.4.9-2.
~~ ___ B p- __p~__--~____~
_.._ ~_____ ___-- D Importance:
SRO - _- I 3.5 295016 - Control Room Abanidonment
.. -p~p ~ WA System WA Statement AA2.06 -Ability to determine and/or interpret cooldown rate as it applies to Control Room Abandonment. REQUIRED MATERIALS:
Calculator p~ ~~~ ~~__ Notes and Comments:
~ - ~-p ~- pp~~- ~~ __~p_~~- - - ~~- __~p~_-
Peach Bottom Initial Senior Reactor Operator VRC Examination December 2008 1 1. The Unit 2 RBCCW pumps have been tripped due to a leak in the system.
The following conditions currently exist: Reactor power has been lowered IAW GP-9-2 "Fast Reactor Power Reduction" Initial Recirculation pump '2A' and '2B' parameters on TR-2-2-2-03 1 are as follows: Motor bearing temperaturcs:
Pt 1 - 185°F Pt 2 - 174°F Pt 3 - 190°F Pt 4 - 186°F Pt 13 - Pt 14 - Pt 15 - Pt 16- 89°F 82°F 87°F 85°F Seal cavity temperatures: Pt 8 - 188°F Pt 9 - 192°F Pt 20 - 202°F Pt 21 - 188°F 0 Following additional speed reduction IAW GP-9-2, Recirculation pump '2A' and '2B' parameters on TR-2-2-2-03 1 are as follows: Motor bearing temperaturq: Pt 1 - 189°F Pt 2 - 170°F Pt 3 - 195°F Pt 4 - 190°F Pt 13 - 191°F Pt 14 - 192°F Pt 15 - 190°F Pt 16 - 185°F Seal cavity temperatm:
Pt 8 - 178°F Pt 9 - 172°F Pt 20 - 182°F Pt 21 - 178°F Evaluate these parameters against the portion of ON- 1 13 "Loss of RBCCW' provided on the NEXT PAGE and choose the correct course of action. A. Perform GP-4 "Manual Reactor Scram", trip both recirculation pumps and enter T- 100 "Scram." B. Trip '2A' recirculation pump ONLY. Enter OT- 1 12 "UnexpectedAJnexplained Change in Core Flow" and execute concurrently with ON-113.
C. Continue to reduce '2B' recirculation pump speed ONLY, using SO 2A.1 .D-2 "Operation of the Recirc Pump Speed Control System."
D. Trip '2A' recirculation pump ANI) continue to reduce
'2B' recirculation pump speed. Enter OT- 1 12 "UnexpectetUUnexplained Change in Core Flow" and execute concurrently with ON-1 1 3.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2006 Level of Knowledge - HIGH Choice ________-__
Correct: Difficuh Time Allowance
-~- (minutes) - ~ __ ~~__ -~ 1 OCFR55.43@)(5)
Distractors:
Source : A [XI New Exam Item Modified Bank Item 0 Previous NRC Exam 0 Other Exam Bank B C D _-~____ ~ _____~__-
__._ Basis or Justification Following recirc pump speed reduction IAW GP-9-2, one motor bearin J temperature on the 2A pump is >I94 degrees F, which requires trippirg the 2A pump. In addition, the 2B pump seal temperature is above 180 deqrees F. Since both pumps lhave a temperature point above the level for removing the pump from service, a GP-4 scram is required, followed by tripping both pumps and entering T-100 per step 2.6 of ON-I 13. ON-I 13 gives direction to remove the 2B pump from service (single loop) based on seal temperature. However, since the 2A pump must be tripped due to high motor beairing temperature, both pumps must be removed from service, which first requires a reactor scram.
_______-___
~ ______._~___
~_ ~ ~ The 2A pump must be tripped based on motor bearing temperatures.
Based on the conditions given, GP-9-2 is already in progress and recirs pump speed has already been lowered. Based on the direction in ON-I 13, the 2B recirc pump must be removed from service.
~~___ 0 ILT Exam Bank _- ~- ~- ON-I 13 and Bases
~~ - Learning PLOT-PBIG-1550-3, -18a, -1 81b Objective:
WA System WA Statement
__~_______~ - ~- - 29501 8 - Partial or Complete Loss of T-lmpDrtance:
SRO Component Cooling Water
_______ 3.4 ~ - - __- ____-_. ~- ____ M2.01 - Ability to determine and/or interpret component temperatures as they apply to Partial or __ .~ -- ~~ -- - ~- ~ ~________ REQUIRED MATERIALS: - -~ Notes and Comments:
I Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 12. An ATWS is in progress on Unit 2. RPV water level was intentionally lowered per T- 1 17 "LeveVPower Control."
The following conditions currently exist:
Reactor power is 6% 1 SRV is stuck open RPV level is -200 inches and rising EHC is controlling RPV pressure iat 950 psig Torus temperature is 175 degrees IF and rising RHR loop 'A' is in Torus cooling; 'B' loop is unavailable Torus pressure is 6 psig and slow1:y rising Torus level is 15 feet and stable HPCI is injecting at 5000 gpm Which one of the following describes the required action and the reason for taking the action? Refer to the portions of T-102 "Primary Containment Control" AND T-117 "LeveVPower Control" provided on the NEXT TWO PAGES.
A. Reduce RPV pressure to less than 900 psig in order to maintain on the safe side of T/L-1 "SRV Tail Pipe Limit." B. Perform Emergency Blowdown per T-112 due to inability to maintain RPV level above -195 inches.
C. Reduce RPV pressure to less than 900 psig in order to maintain on the safe side of T/T- 1 "Heat Capacity Temperature Limit."
D. Perform Emergency Blowdown per T-112 due to being on the unsafe side of T/T- 1 "Heat Capacity Temperature Limit."
Peach Bottom Initial Senior Reactor Operator IVRC Examination December 2008 Level of Knowledge Difficu4 - HIGH Choice Time Allowance (minutes)
___ 1 OCFR55.43Q,)(5)
Correct: Source : Reference(s):
Learning 0 bjective:
K/A System ~~~~ __ Distractors:
0 New Exam Item [7 ILT Exam Bank -~ Previous NRC Exam (LGS - 2006) Modified Bank Item 0 Other Exam Bank
~- ~~ . _____ T-102 and Bases PLOT-PBIG-2102-5a 295026 - Suppression Pool High Water TizaK-: SRO ~. ~~ ______--_
Temperature 4.2 ___ __~_____ ~. C A B D Basis or Justification
-_ _-_____ ~ Torus temperature is -3 degrees F from HCTL and rising. If Torus temperature cannot be maintained on the safe side of HCTL, T-I 02 TI r-8 directs maintaining RF'V pressure on the safe side of HCTL. ______~~_______~________.~._~.-~_____ - Torus level is high but 1.6 feet away from T/L-1 limit and level is stable. Reducing pressure for the purposes of maintaining this curve is not warranted.
While RPV Level is below -195 inches, it is only 5 inches below band and is rising due to HPCl injection.
The criterion for T-I 17 LQ-20 is whether or not level can be restored and maintained above -195 inches, which it can. Therefore, T-I 12 is nolt warranted under these conditions.
Operation is on the SAFE side of the HCTL curve.
Peach Bottom Initial Senior Reactor Operator lVRC Examination December 2008 13. Unit 2 was operating at full power when a small break Loss of Coolant Accident (LOCA) occurred. The following conditions currently exist:
Torus level is 17 feet and rising.
Torus pressure is 9.8 psig and rising. Drywell temperature indicated 165 degrees F before TI-80146 "Drywell Bulk Average Temperature Indicator" failed.
Based on T- 102 "Primary Containment Control" NOTE
- 27 below, the crew attempted to perform a manual calculation of Drywell Bulk Average Temperature using RT-0-40C-530-2 "Drywell Temperature Monitoring" but the calculation was invalid.
WUSE TI-80146(90146)
IS OUT OF SERVICE, @ RT-0-40C-530 TO DETERMINE OW BULK AVG Evaluate these conditions to determinle the appropriate action related to spraying the Drywell. A. Do NOT spray the Drywell since the safe side of the DWSIL curve cannot be verified per RT-0-40C-530-2.
B. Do NOT spray the Drywell since
'Torus level may rise above the limit of T-102 "Primary Containment Control" for spraying the Drywell.
C. Spray the Drywell per T-I 02 after verifying the safe side of the DWSIL curve using TI-2501, Point 136 plus 10 degrees F. D. Spray the Drywell per T- 102 after verifying the safe side of the DWSIL Curve using the hottest temperature indicated on TI-250 1, Points 1 19-1 27.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Choice Correct: Distractors:
A B C D -~___ - _~ Basis or Justification
~~______~~____________~______
__ RT-0-40C-530-2 precaution 4.2.2 states that if the calculation of Drynell Bulk Average Temperature is invalid, the safe side of the DWSIL curvv cannot be verified. DO NOT SPRAY THE DRYWELL. Per T-102, the Torus level limit for spraying the Drywell is 18 feet.
If Drywell sprays are required and all other conditions are met, Torus level at 17 feet and rising would not prevent spraying the Drywell. TI-2501, Point 136 (plus 10 degrees F) can be used to calculate approximate drywell temperature for entering ON-120 or T-I 02, but not for spraying the drywell.
____ ~ __-~____________
~~________
___ ~ Using the hottest temperature from TI-2501 points 119-127 is an acceptable method of determining when to initiate RPV blowdown, but it is not acceptable for use on the DWSIL curve. - -~- ._____.~_~_..~___-_______~
I Psvchometrics I Level of Knowledge HIGH ___ Difficub Time Allowance (minutes)
~ - - c Source Documentation Source: 0 New Exam Item Modified Bank Item 0 Previous NRC Exam Other Exam Bank
.___ .. ~~ PLOT-I 560-1 I __ Learning Objective:
WA System ~~-~___ !='%-High Drywell Temperature K/A Statement REQUIRED MATERIALS:
Notes and Comments:
~- of EOP warnings, cautions and notes. ~ ~- -____~ -~____- -~ - - ~- -
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 14. Unit 2 was operating at 100% power !when a Group I Isolation occurred.
The following conditions currently exist:
0 Reactor power is 18%
0 RPV pressure is 1100 psig Instrument Nitrogen has NOT been restored Torus water level is 13.5 feet and lowering Torus temperature is 170 degrees F and steady Which one of the following actions must be taken for these conditions?
Refer to the portions of T-101 "RPV Control" AND T- 102 "Primary Containment Control" provided on the NEXT THREE PAGES.
A. Use the Bypass Jack to lower RPV pressure IAW T-101 "RPV Control."
B. Use manual SRV operation to lower RPV pressure IAW T-101 "RPV Control."
C. Perform T-112 "Emergency Blowdown" BEFORE Torus level drops below 10.5 feet. D. Perform T-112
Emergency Blowdown" BEFORE Torus level drops below 12.5 feet.
Peach Bottom Initial Senior Reactor Operator IVRC Examination December 2008 Level of Knowledge HIGH An swe r Key SRO 1 OCFR55.43(13)(5)
Difficulty - ~ Time Allowance (minutes)
~ __ ~- Choice Source : ~~ ____________
Reference(s):
Learning 0 bjective:
Correct: 0 New Exam Item [XI Modified Bank Item ILT Exam Bank 0 Previous NRC Exam Other Exam Bank
~- ______ ~- - ~ T-101 and Bases; T-102 and IBases _______. ~ PLOT-1560-3 Distractors:
A B C Basis or Justification
__________
__________~~
-~ At a Torus level of 12 5 feet, for 170 degrees F Torus temperature, HC TL would be exceeded. An emergency blowdown is required then by T-1132, step T/T-9 and T/T-IO.
Although lowering RPV pressure is directed by T-I 02, step T/T-6, bypass valves are not availabile due to the Group I isolation.
Although lowering RPV pressure is directed by T-102, step T/T-6, SRVs are not available without Instrument N2 (accumulators are reserved for ADS). - ~ ~- -~ _______ __ - ~- T-I 01, step RC/P-13, directs use of SRVs ONLY when the accumulators are NOT the only source of pneumatic supply. This choice looks attractive since this is when T-102, steps TIL-7
& 8 direct a blowdown based on Torus level. However, based on the HCTL curve, a blowdown would be required before Torus level drops below 12.5.
-~ Importance: SRO 4.2 I- Level I ~- - _____________
~~ 295030 - Low Suppression Pool Water - _____ _- KIA System WA Statement G2.4.47 -Ability to diagnose and recognize trends; in an accurate and timely manner utilizing the -ropriate control room reference material
__ __~____ REQUIRED MATERIALS:
__ ~___ ~~ Notes and -~____~- Comments: - ~- - -~-
Peach Bottom Initial Senior Reactor Operator
'NRC Examination December 2006 15. Unit 2 was operating at 1 OOYO power when a Reactor Scram was attempted due to Main Turbine vibration. The following conditions exist: 0 Main Turbine is tripped 0 0 0 0 0 An ATWS is in progress RPV level is steady at - 150 inches and is being controlled in accordance with T-240-2 "Termination and Prevention of Injection into the RPV" RPV pressure is being controlled with Bypass Valves and SRV's within a band of 950- 1050 psig Reactor Building Steam Tunnel temperatures are 173 degrees F and rising T-221 "MSIV Bypass" is in progress For the above conditions, which one of the following must be performed?
A. GP-8.B "PCIS Isolation - Groups I1 and 111" to reset isolations.
B. T-227-2 "Defeating RWCXJ Isolation Interlock" to restore RWCU.
C. T-222-2 "Secondary Containment Ventilation Bypass" to restore Reactor Building ventilation.
D. A0 40B.1-2 "Raising MSL Tunnel PCIS Group 1 Hi Temp Trip Setpoint" to prevent a Group 1 isolation on high steam tunnel temperature.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Difficufi
_-_______
Level of Knowledge HIGH Question # 15 SRO Time ~- Allowance (minutes) - ~ Choice Correct: -______ Distractors:
C __ A B D __-.________
_~___________
-~ Basis or Justification
_________________
-~___ -~ .~ -_ T-221-2 directs the operator to perform T-222-2 to prevent isolation of the MSlVs due to high steam tunnel temperature that could occur without normal Secondary Containment ventilation.
GP-8.B cannot be used to reset the isolations with RPV level c 1 inch _~~-___ ___ --___-______~~-
___--- ___ T-227-2 is for pressure control; it is not required for the given conditions.
-~-______._________~~~.~
___ - Although performing this A0 would bypass a PClS Group I isolation, it is not required for the given conditions.
I I I Psychiometrics
_ Source: __- ___- Reference( s): Learning 0 bj ect ive: KIA System
~__ KIA Statement Source Documentation 0 New Exam Item 0 Modified Bank Item 0 Previous NRC Exam 0 Other Exam Bank ILT Exam Bank
-~ T-I 17; T-221-2; T-222-2 PLOT-I 560-1 3 _____ -~ _- .__ __ Importance: SRO 4.2 ~ ________~
_____~______~___
295037 - Scram Condition Present and Reactor Power Above APRM Downscale
________.
EA2.07 - Ability to determine and/or interpret the following as it relates to Scram Condition Present and Reactor Power Above APRM Downscale:
containment conditions/isolations.
REQUIRED Notes and
~______ Comments:
-________________~~_~__~_~__-_
-_ ~ -~__~ ___--~
Peach Bottom Initial Senior Reactor Operator NRC Examination December 200:- 16. Given the following conditions:
0 0 Unit 2 was initially operating at 100% power. An EHC System malfunction resulted in a reactor pressure transient. Reactor pressure peaked at 1340 pig. In accordance with Technical Specifications, a Safety Limit Violation (1) . Referring to LS-AA- 1020 "Reportability Reference Manual, Volume 1 - Table SAF" PROVIDED SEPARATELY, the NRC (2) A. (1) has occurred (2) must be notified within one hour B. (1) has occurred (2) must be notified within four hours C. (1) has NOT occurred (2) does NOT need to be notified D. (1) has NOT occurred (2) should receive a courtesy notijication call within four hours Peach Bottom Initial Senior Reactor Operator NRC Examination December 200% - Level of Knowledge Difficulty
._ Time Allowance (minutes)
_____ HIGH Choice C CorrGT _~____ 1 OCFR55.43(5)
- 2) IA ~ Distractors:
Source : _______- Reference(s): - Objective:
Learning c 0 New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank Tech Spec 2.1.2; LS-AA-1020, -_ SAF 1.16 PLOT-1800-9
_______p_~p Basis or Justification
~__ p__ _~___-_ _~______~-~- - ~ Safety Limit 2.1.2, Reactor Steam Dome pressure has been exceeded (1325 psig). This requires notification to the NRC WITHIN four (4) holm per LS-AA-1020 Exelon Reportability Reference Manual, SAF 1.16.
Wrong notification time.
This requires notification to the NRC WITHIN four (4) hours per LS-AA-I 020 Exelon Reportability Reference Manual, SAF 1.16. A Safety Limit has been violated.
___.p__~ - - -~ ________ K/A System 1 295007 - High Reactor Pressure I Importance:
SRO 4.1 ~- _- ____ K/A Statement G2.4.30 - Knowledge of events related to system operation/status that must be reported to internal R~~",~~D~R~g~~02~~~~~a~i,i~
Reference Manual, Volume I - Table organizations or external a encies, such as State, the NRC, or -~___-___-
the transmission ystem operator.
__ ~- ~__-__p_ ~ -___p__ _p.~. - .~ __________~_~_~_______~.______~
_. - -__ Notes and
_____ Comments:
Peach Bottom Initial Senior Reactor Operator NRC Examination
- 17. Unit 3 is operating at 1 OOYO power wlhen the following occurs:
December 200:; Annunciator 3 17 K-5, REAC BLlDG HI-LO DIFF PRESSURE, alarms. Annunciator 3 17 L-1 REAC BLDG REFUELING AREA HI-LO DIFF PRESS, alarms. Reactor Building dP indicates
+ 1 inch and rising on DPI-30003 - 1. An Equipment Operator reports a steam leak in the area of the Unit 3 HPCI Room.
Attempts to isolate the HPCI steam supply line have failed. HPCI Room area temperature is I32 degrees F and rising on TR-3-13-139 Point #3. RCIC Room area temperature is 1 12 degrees F and rising on TR-3-13-139 Point
- 2. Based on these conditions, T-103 entry is required due to the potential for an unmonitored radiation release via the (1) and the CRS must direct (2) Refer to the portions of T- 103 "Secondary Containment Control" provided on the NEXT TWO PAGES. A. (1) Vent Stack (2) plant shutdown using GP-3 "Normal Plant Shutdown" B. (1) Vent Stack (2) GP-4 "Manual Reactor Scram"' AND depressurize per T-101 "RPV Control" C. (1) Refuel Floor blowout panels (2) plant shutdown using GP-3 Normal Plant Shutdown" D. (1) Refuel Floor blowout panels (2) GP-4 "Manual Reactor Scram" AND depressurize per T-101 "RPV Control" Peach Bottom Initial Senior Reactor Operator WRC Examination December 2008 Level of Knowledge Difficu4 ~ _~__ Time Allowance (minutes) . -~ HIGH ______- SRO 1 OCFR55.43(b (5) ~ Choice Source: __ Reference(s):
______. Learning 1 Objective:
Correct: 0 New Exam Item 0 Modified Bank Item [XI ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
~~___ _____~~___________
_____~ ARC-317 K-5; T-103 and Bases PLOT-2103-1
_____~_ Distractors:
-- 295035 - Secondary containment High Differential Pressure D Importance:
SRO 3.9 A B C ______~- ~ - -~ Basis or Justification T-103 bases states thlat if ARC-317 K-5 is received, action must be taben to prevent a possible breach of Secondary Containment via the Refuel Fioor blowout panels. The panels blowout at +7 inches of water (+.25 psig) T- 103 step SCC-8 requires a reactor scram and depressurization since ;I primary system is discharging into the Reactor Building.
The Reactor Building would isolate if exhaust radiation reached the isolation set point, preventing release via the Vent Stack. Also, GP-3 shutdown is directed by T-I 03 only when there is no primary system discharging into the Reactor Building and when the same parameter exceeds an action level in more than one area. ONLY the HPCl Room is affected.
The Reactor Building would isolate if exhaust radiation reached the isolation set point, preventing release via the Vent Stack. GP-3 shutdown is directed by T-I 03 only when there is no primary system discharging into the Reactor Building and when the same parameter exceeds an action level in more than one area. ONLY the HPCl Room is affected.
~- _~_________________
~- - ________________~~______~
~_____~ ~_ .~______-____~_~.__---~-_ ~~~ ~~ ~ KIA System KIA Statement Containment pressure as it applies to _ - _____~__ __ ________-
--
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 18. Given the following conditions:
0 0 0 0 Unit 2 was initially operating at 100% power.
T- 10 1 "RPV Control" and T- 102 "Primary Containment Control" have been entered due to a Loss of Coolant Accident.
RPV level has been recovered to -165 inches and is steady. Twenty minutes after the PRO pla.ced the "A CAC/CAD Analyzer
AND the "B CAC/CAD Analyzer" in service, he reports the following readings on both analyzers:
o Drywell H2 concentration is 0.6% o Drywell 02 concentration is 0.6% o Torus H2 concentration is 0.4% o Torus 02 concentration is 3.5% Pending determination of off-site release rates, what actions are required for these conditions?
Refer to T- 102 "Primary Containment Control" Sheet 2 PROVIDED SEPARATELY.
A. Vent the Torus per step T/G- 1.4 ONLY.
B. Vent the Drywell per step DW/G-1.4 ONLY.
C. Vent the Torus per step T/G-l.4 A.ND the Drywell per step DW/G-1.4.
D. Torus venting is NOT required; Drywell venting is NOT required.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Time Allowance (minutes)
_~_~~__._._________ Level of Knowledge
___~ HIGH Difficulty - Answer Key - SRO 1 OCFR55.43( 3 (5) _- Choice Correct: Distractors:
B -- A C D ___________
Basis or Justification
~___ _~_____ _~___-~ ___._________
___ Drywell venting using the 2" vents is required per T-102, Sheet 2, ster DW/G-1.4.
_ ______ - Drywell (not Torus) venting is required per T-102, Sheet 2, step DW/G-1.4. Drywell venting is required ONLY per T-I 02, Sheet 2, step DW/G-1.4 using a 2 inch vent line only. Drywell venting is required per T-I 02, Sheet 2, step DWIG-1.4.
Source : Reference(s):
Learning Objective:
WA System __________ Source Documentation 0 New Exam Item 0 Modified Bank Item 0 Previous NRC Exam Other Exam Bank ___. ILT Exam Bank
___ - RRC 7J.1-2; SO 7J.l.B-2 -~ PLOT-1560-3
~- _ 500000 - High Containment Hydrogen Concentration - -~ - Importance:
SRO 4.7 ____ WA Statement REQUIRED MATERIALS:
p~~'Primary
_____ Containment Control" SHEET 2 of 3 Notes and Comments:
~__ -~ ~ ~ G2.4.6 - Knowledge - of EOP miti ation strategies
________-
___ -___- _ ~ -- -~ - ~ ~_____
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2001; 19. The following conditions exist on Unit 3: 0 0 Fuel reload is in progress in the "B" core quadrant The fuel being loaded is NOT directly adjacent to any WRNM A fuel assembly has just been seated and the grapple is still engaged The Reactor Operator recorded thle following data for this step:
-___- REFOI.LE AFTER "A" WRNM 20 cp?; 20 cps and steady "B" WRNM 30 cps 150 cps and steady What action is required by ON-124 "Fuel Floor and Fuel Handling Problems" for these conditions?
A. Immediately evacuate the Refuel Floor area.
B. Continue fuel moves and notify R-eactor Engineering.
C. Raise the grappled fuel assembly above the upper grid.
D. Suspend fuel moves and determine operability of "B" WRNM.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Difficuk Time -~ Allowance (minutes)
-~ Choice SRO 1 OCFR55.43(b)(6)
~- Correct: Distractors:
C A B __ D - Basis or Justification
__ ______~__.-~_-~
-~ "B" WRNM has gone lrom 30 cps to 150 cps, which is >2 doublings. ON-124 requires raising the bundle above the core (top guide) if the bundle is still grappled, which it is in this case. Per ON-I 24, this is required IF count rate continues to rise after the fb9l assembly is raised from the core. ~~_____________~__~ - ___._ Per ON-124, if WRNNI count rate doubles 2 times between CCTAS steps, the first action is to raise the fuel assembly from the core so that it clears the upper grid. Then fuel handling operations would be SUSPENDED, and the appropriate notifications would be made (including Reactor Engineering).
Based on the given conditions, the first action is to raise the fuel asseribly from the core, THEN suspend fuel moves. __ ~~~___________~__--
-~ Source: r Reference( s): Learning t Objective:
WA System Source Documentation New Exam Item Modified Bank Item ILT Exam Bank
~ 0 Previous NRC Exam Other Exam Bank -~ - ~~~ ON-I 24 PLOT-PBIG-1550-3, -27a NIA Importance:
SRO 4.6 _.._______
~___~_._ ~.___ ___ or - limitations associated
__~-- with reactivity management. ~~~ ~~~ - ~ . Notes and Comments:
Peach Bottom Initial Senior Reactor Operator
'NRC Examination December 2008 20. A small steam leak inside the Ilrywell occurred on Unit 2. The reactor was depressurized in accordance with T- 1 12 "Emergency Blowdown
due to being unable to restore and maintain drywell temperature below 28 1 degrees F The following conditions existed at the start of the blowdown:
0 0 Indicated RPV level was
-1 40 inches All high-pressure feed sources were unavailable The following conditions exist at the lcompletion of the blowdown:
RPV pressure is 35 psig Indicated Wide Range level is - 1 10 inches All other RPV level indications are upscale Drywell temperature (TI-2501 ponnts 126 and 127) is 295 degrees F Multiple failures prevented LPCI Core Spray systems from injecting Evaluate the response of RPV level instrumentation to determine the required action. Refer to the portions of T- 102 "Primary Containment Control" AND T-112 "Emergency Blowdown" on the NEXT TWO PAGES. A. Enter T-116 "RPV Flooding."
B. Enter T- 1 1 1 "Level Restoration."
C. Establish Shutdown Cooling per T- 1 12 "Emergency Blowdown."
D. Restore RPV level to between
+5 and +35 inches per T-101 "RPV Control."
Peach Bottom Initial Senior Reactor Operator IVRC Examination December 2008 1 Choice Source : - -- Reference(s):
______ Objective:
Learning Correct: .- New Exam Item Modified Bank Item 0 ILT Exam Bank
.~ 0 Previous NRC Exam Other Exam Bank T-101; T-102; T-I 12; T-I 16 PLOT-1560-1 I ~_~________~
~ - Distractors:
A B C D _-~___ ~ ~______________
-~ Basis or Justification
________-__~
~~_.~__~_____
___ Per T-I 02, Table DW/T-l, wide range level plots on the UNSAFE side of the RPV saturation curve. Since it plots UNSAFE and exhibits an unexplained trend (level increase from -140 inches to -1 10 inches afte completion of the blowdown without any makeup sources), wide range level is unavailable.
Since all other level indication is upscale (unavailable) level is unknown and entry into T-I 16 is directed by T-I 12 step EB-2 (T-101 RC/L-1, T-I 11, etc.).
Level cannot be determined.
In addition, T-I 11 is not entered unless RPV level cannot be maintained above
-1 72 inches. _-_______~
~- Level cannot be determined. - _~__ ~____ ._ Level cannot be determined.
I Psvchiometrics 1 I Level of Knowledae 1 Diff icultv I Time Allowance (minutes) 1 SRO I WA System N/A- Importance:
SRO 4.3 WA Statement
__ 2.1.45 -Ability to identify and interpret diverse
_- .~ indications - - ___~-- to validate the response of another indicator_
_. REQUIRED MATERIALS:
I NONE __ Notes and - Comments:
Peach Bottom Initial Senior Reactor Operator IVRC Examination December 2008 2 1. Given the following:
0 0 Unit 2 is operating at 90% power following intermittent and spurious changes in
- 3 Turbine Control Valve (TCV) position.
To support troubleshooting, I&C requested to perform "power noise monitoring" in EHC Cabinet 20C30.
The work order specifies hooking up a strip chart recorder at new monitoring points recommended by the vendor. The System Manager has determined that these are NOT approved test points on the cards. Refer to Attachment 4 "Risk and Rigor Determination Matrix" of MA-AA-716-004 "Conduct of Troubleshooting" PRO\!IDED SEPARATELY to classify this activity.
This is a (1) Risk troubleshooting activity, classified as Category -(2)-. A. (1) High (2) A B. (1) High (2) c C. (1) Medium (2) A D. (1) Medium (2) c Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Level of Knowledge
__ DifficuQ HIGH Choice Time Allowance (minutes)
__.________~__~
__ ___ IOCFR55.43(bM5)
Correct: REQUIRED MATERIALS:
Distractors:
MA-AA-716-004, Attachment 4, "Risk and Rigor Determinatic A B Per Guideline I .3 of Attachment 4, Risk and Rigor Determination Matrix, EHC is High Risk, and^ activity is Diagnostic hookup for monitoring on
_. . equipment in service without test points - Category A. _~ _- EHC is High Risk but activity without test points is Category A; activity with test points is Category C. TTEHC is High Risk and activity is Category A. - D __ __ EHC is High Risk and activity is Category A. I Source Documentation Source: Reference( __ - s): Learning Objective: KIA System New Exam Item 0 Modified Bank Item 0 ILT Exam Bank Previous NRC Exam Other Exam Bank - LORT ~~ MA-AA-716-004, Attachment a4, "Risk and Rigor Determination Matrix" PLOT-DBIG-1570-6 N/A Importance:
SRO 3.8 . ~~ _. ___
Peach Bottom Initial Senior Reactor Operator
'NRC Examination December 2008 22. Equipment Operators need to enter a llocked high radiation area to manually operate Primary Containment Isolation Valves in order to satisfy a Technical Specification required action. The highest dose rate in the area is 16,000 mWhr. Per RP-PB-460-100 1 "Radiation Protection Controlled Keys", which one of the following describes the type of Locked High Radiation Area and the highest level of authorization required for issuing the key? Type of LHRA ,Highest Authorization Required A. Level 1 Radiation Protection Manager B. Level 1 Plant Manager C. Level 2 Radiation Protection Manager D. Level2 Plant Manager Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Time Allowance (minutes)
__ ________.
FUNDAMENTAL Choice SRO 1 OCFR55.43(bM4)
~___~- Correct: Distractors:
I Basis or Justification Per RP-AA-460-1001, Level 2 LHRA is an area with dose rates
> 15R hr. The RP Manager must provide authorization for this entry.
The level is incorrect. The area is a Level 2 (>15R/hr) which requires authorization from the RP Manager for issuing the key. _~____~~____
~______ __ ~-~__ __ level is incorrect, ,and the Plant Manager's authorization is NOT required.
Source : Reference(s):
Learning Objective: K/A System
______________
Source Documentation New Exam Item 0 Modified Bank Item ILT Exam Bank Previous NRC Exam - 2007 0 Other Exam Bank
__-____~_.__________
____ ~ ~__~~_______ . RP-AA-460; RP-PB-460-1001
~- PLOT-I 770-3 NIA Importance:
SRO KIA Statement G2.3.13 - Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked hi h-radiation areas, allgning filters, __ etc. Notes and Comments:
~ - REQUIRED MATERIALS:
__ - ~__-~__. ---NONE --
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 23. Given the following:
0 0 0 A radiological accident condition has occurred at Peach Bottom 2 VENT EXH STACK RAD MONITOR HUTROUBLE A (218 B-5) alarms 2 VENT EXH STACK RAD MONITOR HUTROUBLE B (21 8 C-5) alarms The PRO reports that the Unit 2 Vent Stack Radiation is reading above the HI alarm level The source of the radiation release is the (1) and the CRS must -- (2) A. (1) Standby Gas Treatment Exhaust (2) enter and execute T- 104 "Radioactivity Release B. (1) Radwaste Building Ventilation Exhaust (2) direct the termination of radwaste processing using the appropriate radwaste procedures C. (1) Recombiner Building Venti1at:ion Exhaust (2) direct the evacuation of all unnecessary recombiner personnel using GP-15 "Local Evacuation" D. (1) PEARL Building Ventilation Exhaust (2) direct restarting ventilation using SO 40AA. 1 .A "Setup and Operation of PEARL Heating and Ventilation" Peach Bottom Initial Senior Reactor Operator IIRC Examination December 200h D __ Answer Kev The PEARL Building Exhaust is to the Unit 3 Vent Stack. Restarting ventilation would be appropriate if the PEARL was the source. Choice Level of Knowledge HIGH Correct: SRO 1 OCFR55.43hM5)
Difficuty
____ Time Allowance (minutes)
~- Distractors:
Basis or Justification
_- ._____ _____ __ _____ Buildinig Vent Exhausts to the Unit 2 Vent Stack.
Radnaste operations must be terminated under these conditions.
_~ is to the Main Stack. T-104 would not be entered mtil radiation reached the IHI HI alarm point.
~ The Recombiner Building Exhaust is to the Unit 3 Vent Stack. Evacuating all unnecessary personnel from the recombiner would be correct if it was -7 the source of the leak Source: Objective:
KIA System -~ Source Documentation 0 New Exam Item Modified Bank Item Previous NRC Exam - 2005 0 Other Exam Bank - ~ - ~~ ILT Exam Bank
_~ ~ ____ ON-I 04 PLOT-PBIG-1 540-9a NIA Importance: SRO 3.8 WA Statement G2.3.14 - Knowledge of radiation or contaminatiori hazards that may arise during normal, abnormal or emergency
_. conditions or activities.
_- ~- - ___________________~__~_ - ~- ~ MATERIALS:
1 NONE 1 Notes and Comments:
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 24. Unit 2 was operating normally when IDrywell temperature and pressure began to rise. The crew entered ON- 120 "High Dry well Temperature" and OT-10 1 "High Drywell Pressure".
The CRS determined that Drywell temperature and pressure CANNOT be maintained below 145 degrees F and 2 psig, respectively.
Which one of the following identifies how ON-120 and OT-1 01 are required to be used in conjunction with T- 102 Primary Containment Control"?
A. Exit ON- 120 AND OT-10 1 ; continue actions per T- 102 ONLY.
B. Exit ON- 120 ONLY; continue act ions per OT- 10 1 AND T- 102 concurrently.
C. Exit OT-101 ONLY; continue actions per ON-120 AND T-102 concurrently.
D. Continue actions per ON-120 AND OT-101 concurrently with T-102.
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2001 ____~ _______~ Answer Kev Level of Knowledge FUNDAMENTAL Choice Difficulty ~ ~~~ -_ Time - Allowance (minutes)
~ Correct: I D Distractors:
A lB ____ ~~ - .~ ___ Basis or Justification
~~___~ ~~~ Per ON-120, "Operatar Action" step 2.1, ON-I 20 must be executed concurrently with T-I 02. Per OT-101, "Follow-up Action" step 3.3, OT 101 must be executed concurrently with T-102.
ON-I20 and OT-I 01 rnust be executed concurrently with T-I 02. ~ -~ ON-I20 must also be executed concurrently with T-102.
OT-I 01 must also be (executed concurrently with T-I 02. I Psvctiometrics I Source : Reference(s):
Learning Objective:
WA System Source Documentation New Exam Item Modified Bank Item 0 ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
~_____ ________~
_____ __ - ON-120; OT-I 01 ; T-I 02 PLOT-2 1 02-3 ~.-______~~___~~
____ ~____ ____ ~-_ -. - _ N/A Importance:
SRO 4.5 REQUIRED MATERIALS:
Notes and Comments: . .._. _ _~ ,-- ~ NONE NOTE: this question is designated as SRO ONLY because: (1) It cannot be answered by knowing immediate operator actions or TRIP entry conditions (must know follow-up actions).
(2) It requires recall of a strategy or action that is written into a plant procedure, including when the strategy or action is taken. (3) It is an SRO job function to determine when ____~___~~__~
Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 25. Both units were operating at 100% power when an earthquake occurred. The following conditions were initially present:
0 0 0 0 The tremor was felt in the Main Control Room SEISMIC TRIG ACTIVATION OR LOSS OF SEISMIC TNST POWER (3 16 A-4) alarmed The crew entered SE-5 "Earthqualke" and validated the event Seismic instrumentation registered 0.045 g Several minutes later, the following conditions existed on Unit 3: 0 0 0 0 0 0 Drywell pressure is 2.1 psig and rising Drywell temperature is 155 degrees F and rising DRYWELL HI PRESS TRIP (310 F-1) alarm is received GROUP II/III INBOARD ISOL. ]=LAYS NOT RESET (314 D-1) alarm is received GROUP IHII OUTBOARD ISOL. RELAYS NOT RESET (314 E-1) alarm is received The PRO reports AO-30464 "Reactor Bldg. Supply Valve" failed to isolate Refer to EP-AA-1007, Table PBAPS 3- 1 "EAL Matrix" PROVIDED SEPARATELY to determine the highest EAL classification required for these conditions?
A. Alert per FA1 B. Alert per HA5 C. Unusual Event per FUI D. Unusual Event per HU5 Peach Bottom Initial Senior Reactor Operator NRC Examination December 2008 Level of Knowledge
-~ HIGH Choice ~._-._____-__-
Correct: Difficui __ __~ Time Allowance (minutes)
Distractors:
-- A B C D ~- Basis or Justification ~ ~ Based on Drywell pressure above 2 psig and rising Drywell temperatwe, the conditions for a loss of the reactor coolant pressure boundary havc been met. This ~~___ results in an Alert classification
~______ IAW FAI. Per HA5, seismic instrumentation must exceed 0.05 g for the Alert levti ~ ~ ~- ___ All of the conditions for a loss of Primary Containment have not been met. Specifically, although ia RB ventilation damper failed to isolate, the line is isolated by a second clamper and therefore no pathway to the environrnent The criteria for declaring a UE per HU5 have been met but this is not the highest classification for the given conditions.
exists. ___ ~ - Source: Reference( s) : Learning Objective: Source Documentation
[XI New Exam Item 0 Modified Bank Item ILT Exam Bank 0 Previous NRC Exam 0 Other Exam Bank
~ __ ___ __ ____~ ~~ ___~ ~ ____ __ __ . EP-AA-1007 G6-8 ~ - __. - - ____~ NIA Importance:
SRO 4.3 I ~~~ ~_._ - KIA Statement the significance
~____ of each annunciator or ______ alarm. Table PBAPS 3-1 (EAL MATRIX) . ~- ___ ~__- - __ _______