ML13154A319
| ML13154A319 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/25/2013 |
| From: | Todd Fish Operations Branch I |
| To: | Exelon Generation Co |
| Jackson D | |
| Shared Package | |
| ML12328A009 | List: |
| References | |
| TAC U01867 | |
| Download: ML13154A319 (212) | |
Text
{{#Wiki_filter:Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 1. Unit 2 is at 100% power.
A leak develops on the instrument line connected to the 3B Condensing Chamber that depressurizes the reference leg and seats the Excess Flow Check Valve (EFCV) in that line. Actual RPV level remains steady at 23 inches and drywell pressure remains below 1 .0 psig. Using P&ID M-352, Sheet 2, PROVIDED SEPERATELY, what is the effect, if any, of this condition on the RPS low RPV level scram function. A. IF actual RPV level lowered to 1 inch, a full scram would be initiated. B. IF actual RPV level lowered to 1 inch, a half scram ONLY would be initiated. C. As a result of depressurizing this instrument line, a full scram would be initiated. D. The associated level transmitters would NOT be affected due to the operation of the excess flow check valve in the instrument line.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
. . ____.__._...___..__..___...____.._._ . _._~ ..._A__n_sw I...__ __e.r.__~::.~_ ..._ . _._._ ..._ ..._.
Question # 1 RO Choice
-.-.----.~~--~----~-.----~--~---
Basis or Justification Correct: CORRECT - LT-101C & D will sense a HIGH level. LT-101A & B are still
.
- available to detect and initiate scram on actual LOW level condition .
. Di~;racl-;;~~-r- ~--CO-R-RE--CT- Wh-il-e-L-T~1*01C-&Dw-i-lI-s;~~~-; HIGH--le-v~*I,-LT=101A& B are still available to detect and initiate scram on actual LOW level condition .
._._~_. __.. . _______..+___ I PI~ysibl~ if ~andLd_atemisynd~ista~ds BPSJggi2..Cl.,!~~g~menL.
i C I INCORRECT - Instruments would sense HIGH level - Plausible if
.
- candidate misunderstands level detector theory of operation.
D INCORRECT - Leak is DOWNSTREAM of Instrument Line EFCV. Ball check valve closes, isolating the instrument line, ambient losses will depressurize the reference leg side of the instrument, LT101 C, D will sense a high level. Plausible if candidate does not understand physical ___.. __ ~. ___~ __*. i!f!angem~!lLof in~trurr1ent Jines and cond~J1_~Jr:l9..~hamber linf?s. Psychometrics L~y.~Lgf K!lOWI~c!9..e__ +___ ... __ ~fficu~L____..~.Time_AII()wance (f!Ilnute~l . RO HIGH 10CFR55.41 (b )(7) Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0
. __~[8J-!~I..r::~a'!lJ3an~.l4}97)-.--- ....
_~:;~~::e(s):_ ---I ~~~;-:0%~B~3d---"--"--'----""'----'--' Objective: I KIA syste~~"- j2120()O-RPS -- .. --.>" . - ...- - - - .....--. -*---T,mportance: RO I SRO I 3.7/3.9
._ ... _ _. _.. _.... - - _ . _ - . _ - _.....----.--.--.....- - . - - . - -....- ... ~-'----.-- ...
KIA Statement: K1.02 - Knowledge of the physical connections and/or cause- effect relationships between REACTOR PROTECTION SYSTEM and the following: Nuclear boiler instrumentation
-~-!~~I~!-*-*~---~--:--~-*:---~..:t~---~L--S:.*.~-.-.:. -=1-::: Sh 2~~~ir~~~=-=-
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 2. If a Group II isolation is actuated with a Traversing In-Core Probe detector in the core, the inserted detector withdraws to the "in-shield" position and the associated will close. In the event the detector fails to withdraw, the TIP Shear Valve (2) actuated.
A. (1) TIP Ball Valve ONLY (2) will be automatically B. (1) TIP Ball Valve ONLY (2) can be manually C. (1) TIP Ball Valve AND TIP Purge Valve (2) will be automatically D. (1) TIP Ball Valve AND TIP Purge Valve (2) can be manually
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 2 ------------------
-----~----
RO Choice----r---~-___+~------------------ Basis or Justification Correct: CORRECT -If a PCIS Group II isolation signal is received while any TIP detectors are outside of their shield, the detector(s) will withdraw to the "in- I shield" position and the associated ball valve will close. The isolation signal [ also closes the TIP purge valve. In the event the detector fails to withdraw, I the TIP Shear Valve (XV-2-07-102) will be manually actuated lAW SO 7F.7.A, 'Traversing In Core Probe System Isolation in the Event of Containment Isolation"
-.,,~.~-----------.. '-",--~----- .- -----------
Distractors: A INCORRECT - The detector withdraws to the "in-shield" position; SV-109 also closes. Shear valve does NOT automatically actuate. B INCORRECT - SV-109 also closes. NOT automatically actuate. Psychometrics Source Documentation Source: D New Exam Item [g] Previous NRC Exam: (PB 2009) I6l Modified Bank Item Other Exam Bank: 0 ILT Exam Bank----------~-- Reference(sL_ GP..8.ELG9L, S97f,-Z:A_______ Learning PLOT-5007F-3a, PLOT-5007G-41 Objective: KIA System: 223002 PCIS/Nuclear Steam Supply Importance: RO I SRO Shutoff 2.7/2.9 KIA Statement: K1.13 - Knowledge of the physical connections and/or cause-effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the: Traversing in-core probe system REQUIRED MATERIALS: NONE f-------------------I------ ------------------------ Notes and Comments: ---'------------------ -- - -----~-
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 3. Unit 2 is at 100% power.
A loss of the Division I 125V DC power supply 20D21 (2PPA) has occurred to the Safety Relief Valve (SRV) Solenoids. The following conditions exist:
- Annunciator 227 C-5 BLOWDOWN VALVES POWER MONITOR alarm is received.
- Division II 125V DC solenoid power is energized and available.
- Both Divisions 125V DC ADS Logic power are energized and available.
Based on the above conditions, which one of the following is correct regarding the capability to manually open ADS and non-ADS SRVs from the Control Room? A. Only ADS SRVs can be manually opened. All Non-ADS SRVs are without power. B. Only the Division II ADS and non-ADS SRVs can be manually opened. C. All ADS SRVs can be manually opened, but only three (3) of the Non ADS SRVs can be manually opened. D. All ADS and Non-ADS SRVs can be manually opened.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 3 RO
, ~.. .. '=ri-"~'~~ ---- ~--~--.- ---,-~---------,-
Choice Basis or Justification Correct: I D CORRECT - Each SRV (ADS or Not) has'bothnorm~aI20D21(Div I) and i i alternate 20D24 (Div II) 125V DC power to solenoids. A loss of one supply _______ _ I I will not prevent SRV (ADS or Not) operation _m-"ntlan~c ____ _ Distractors: A . INCORRECT - Each SR.V (ADS or Not) has, both normal 20D,2,1 (DiV, I) and
- i alternate 20024 (Div II) 125V DC power to solenoids. Plausible because of
- I common misconception that only ADS SRVs have alternate power.
.---~---- ." B
- INCORRECT - Ea"ch-SRV (ADS or Not) has-both nOITnal20D21 (Divl)and
- i alternate 20D24 (Div II) 125V DC power to solenoids. Plausible because i ! candidate may confuse the effects of the ADS logic power loss with valve I
[---~-------._--I~~Ji~E}r I C INCORRECT - loss. """"""'- ------ - - - - - ,----~-, Each SRV (ADS or Not) has both normal 20D21 (Div I) and
- . , i ! """a"""I.te.rna.t.e makes conceptual . . 2"OD.2.4,("D""".i,V II) 1.
error . .2. 5"",V," for ADS . DCversus power non-ADS to sOlenoids.' valvePlausible solenoidifpower. candidate __, ~ ___ ~ __i_ _ _ , NQte..:.puri_r:!9j\J2p'R' fire, ~ SHVs .§I~E}protected due to cable runs. ., Psychometrics Source Documentation Source: o New Exam Item [g] Previous NRC Exam: (PB 2002) Modified Bank Item 0 Other Exam Bank: () i 0 ILT Exam Bank 0
~~~!~!~~~ ~~~~~~~5_~~wdownv.;;;~~we--;Mo~~---=-=-__~_ _
I KIA System: '~---1239bo2-Ref~f/safetyvalves Importance: RO 1 SRO
. . . ~,___________~ _______.,_ . . L ____,____ 2.8/32 K/ A Statement: K2.01 Knowledge of electrical power supplies to the following:
SRV solenoids
~-'--**********-~--*~'**'*--------***3******-
REQUIRED
~~-~~---
MATERIALS:.......
........- - - - .. -~------ ----~
NONE
.. ----------~--- ....... ----~~-~----- ....-
Notes and Comments:
.---~--"------------ ....... --~--~--- _.---- -- ........ ---~ .. ~~----...... ~---- --" - -.. --~"- *... - - - - - --- ~~~"-~-
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 4. Given the following:
- Units 2 and 3 were initially operating at 100% power
- Both units scram when a loss of offsite power (LOOP) occurs
- All 4 EDGs start and re-energize their associated busses
- Two minutes later, the following alarms are received:
- 001 C-1 "E12 BUS DIFFERENTIAL OR OVERCURRENT RELAYS"
- 005 B-4 "E43 BUS DIFFERENTIAL OR OVERCURRENT RELAYS" With no operator actions, which Standby Liquid Control Pumps have power available to their motors?
A. 2A, 3A ONLY B. 2B, 3B ONLY C. 2A, 2B, 3A ONLY D. 2B, 3A, 3B ONLY
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
~- .... --~
Answer Key Question # 4 RO Choice ~,---~-t-- Basis or Justification Correct: Power Supplies are: 2A: E124-R-C, 2B: E224-R-B, 3A: E134-W-A, 3B: E234-R-B. Alarm 005 B4 has no relevance because no SBlC pumps are powered off the E-43 bus, Alarm 001 C1 indicates that the E12 bus is locked out and unavailable, so 2A SBlC pump does NOT have power to motor.
---~~~---"
Distractors: 2A has no power, 2B, 3A, 3B have power. Plausible if candidate does not I* know SBlC Pump power supplies.
-~~~~~--~~-.~--~-- "..- - - - i - - - - - "
has no power, 2B, 3A, 3B have power. Plausible if candidate does not SBlC Pump power supplies. has no power, 2B, 3A, 3B have power. Plausible if candidate does not SBlC Pump power supplies.
-_.... _--" - -..... --~~
Psychometrics le"'E!!~tKnowle~g~~ ______ J:?ifficu!t~______ _ Tirn_E!_AlIgwange (l11in~tE!~lL - - _ .__ RO -~ ..._ - - - - - - - - - . -- HIGH I I 10CFR55.41 (b)(7) Source Documentation Source: [gJ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0
-= IlT-Exam
...- Bank
__ ~_.J. _ _+ ______ PLOT 5011 _ M-1-S-46 learning PLOT 5011 Obj 2a Objective: KIA System: 211000 - Standby Liquid Control Importance: RO I SRO 2.9/3.1 KIA Statement: K2.01 -: ~nowleQ~e of the._E!!ectricalJ>.ower s.!!QP-'!E!sJo th~Jo[Q.wil'lg~ SBj..C_ f>~lT!ps:. REQUIRED MATERIALS: LNONE _ __ Notes and Comments: - - - - - - - - - - ..... -.--~-~- ..... - - . - - - ... - - - . - .
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 5. Unit 2 is operating at rated power:
- Drywell pressure is .6 psig and slowly rising.
- Drywell venting is in progress and the "A" SBGT train and fan are placed in service in accordance with SO 7B.3.A-2 "Containment Atmosphere Pressure Control and Nitrogen Makeup" to vent the Drywell.
- Subsequently drywell pressure is .5 psig and slowly lowering.
- Ten minutes later a loss of 'B' RPS occurs.
For the above conditions, which one of the following describes the status of the SBGT system and drywell pressure? A. Both the "A" and "B" SBGT trains and fans will be running, drywell pressure will start to rise. B. The "A" SBGT train and fan will continue running, "B" SBGT train and fan will remain shutdown. Drywell pressure will continue to lower. C. The "A" SBGT fan will trip; the "B" SBGT train and fan will start. Drywell pressure will start to rise. D. Both the "A" and "B" SBGT trains and fans will be running. Drywell pressure will continue to lower.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer
":,c:,.,,..." # 5 RO
~-----~"--~-------~~~~~~~---~-~--~"--~~~~~~------~~~~~~ ~----~~~~~--~~~----~~~~~-"~~ ~ ----~----- ~ ~ -- -~~------
Choice Basis or Justification Correct: CORRECT - On a loss of RPS "B", the "B" SBGT train and fan will start. Drywell pressure will start to rise due to a PCIS GRP 3 isolation signal (AO 2510 OW Vent outboard 2" vent valve will Distractors: B i INCORRECT - On a loss of RPS "B", the "B" SBGT train and fan will start. I Drywell pressure will start to rise due to a PCIS GRP 3 isolation signal (AO I 2510 DW Vent outboard 2" vent valve will
~~~~--~---.I-C- ~IINCORRECT - The "A" SBGT fan does~:~~n:~:~o-t=rC::e~~~""c:e=Li'--v-e a trip signal on a loss of RPS"B".
,--, ~~~ ~~ _........_ - _.. - ...... ~.~ .. - _... ., - '-
- 0 INCORRECT - Drywell pressure will start to rise due to a PCIS GRP 3
___L li~olati~:_~~~al (AO-2510 OW Vent outboard 2" vent valve will close). Psychometrics Level of Knowledge HIGH
~ ~ ~ ~ __ ~___gifficl:J~ __ "___~
Time Allowance (minutesL !1 OCFR:504; (b)(7)~~ Source Documentation Source: lSI New Exam Item D Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0 ILT Exam Bank B~f~r~nc~(sJ: _ PLOT-5009A and PLOT -5007G Learning PLOT-5009A-6c Objective: KIA System: 261000 Standby Gas Treatment System Importance: RO 1 SRO 3.2/3.4 KIA Statement: K3.03 - Knowledge of the effect that a loss or malfunction of the Standby Gas Treatment System will have on the following: Primary Containment Pressure REQUIRED MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 6. Unit 2 is operating at 100% power when a complete loss of Off-Site power occurs.
All EDGs start and power their respective 4KV busses. One minute later, which of the following components will have cooling water flow available? A. Station Air Compressors. B. Instrument Nitrogen Compressors. C. RWCU Non-Regenerative Heat Exchangers. D. Condensate Pump Motor Lower Bearing Cooler.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 6 RO Choice
- - - - - - - - , _........... _ - + - - - - - - - - - - ............ - _ . _ - - -
Basis or Justification Correct: A CORRECT -.During the LOOP, TBCCW pumps trip, forty seconds later RBCCW will provide cooling to essential loads which include station air
. . _.___________--.:-_ _-+-.::...co.-:..:..I!lpressors and CRQJ)..lJl!!p~: .. _________.. ______ . .
Distractors: B INCORRECT - This is a non-essential load. Plausible if candidate believes otherwise. C INCORRECT - This is a non-essential load. Plausible if candidate believes otherwise. D INCORRECT - Condensate pump coolers are cooled by TBCCW, which loses power on a LOOP, and is NOT backed up by RBCCW. Plausible if
..__.__... _ _ _-,---.__~.--=::.:...:..=.:.=.:..=...:b=e::..:cli=ev.::..ce=s::...:t:.:....:h=e . .c.=.o=-o.::.:I::=e.:r..s=. will still have cOQling'v'Yater.______ __
Psychometrics f-.~el of Knowledge Time Allowance. (r1}inut~ltl-r1*0CFR:5~41
. .___ Diffic.l:'!!Y_ .... ... ...... 1 (b)(4)
MEMORY I Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 D Modified Bank Item 0 Other Exam Bank: () ILT Exam Bank Referenc~( s): ON-118 Loss of TBCCW §ystem - Procedure Learning PLOT-5035-3b Objective:
----=--=-=---.. . . . . . --..
KIA System: 400000 Component Cooling Water Importance: RO I SRO 2.9/3.3 KJ A Statement: K3.01 - Knowledge of the effect that a loss or malfunction of the CCWS will have on the following: Loads cooled by CCWS
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 7. Unit 2 is operating at 25% power when the following occurs.
* #2 APRM fails downscale (not INOP).
Which of the following describes the expected plant response? A. Alarm ONLY. B. Alarm, Rod Block, AND Half scram. C. Alarm, Rod Block, AND Full scram. D. Alarm AND Rod Block; NO scram signals
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 7 RO Choice Basis or Justification Correct: CORRECT - APRM downscale (::: 3.2 %) in MODE 1 will generate a control rod withdraw block and downscale alarm 211 C-2 only. Distracters: - APRM downscale (::: 3.2 %) in MODE 1 will generate a control rod withdraw block and downscale alarm 211 C-2 only. B INCORRECT - A scram vote signal is only generated for: APRM Inop Trip High Neutron Flux
- _ _ ~ ~m Simulated Thermal Power High _____. m C INCORRECT - A scram vote signal is only generated for:
APRM Inop Trip High Neutron Flux Simulated Thermal Power High Psychometrics Level of Knowledge Qifficul~_ . . . Time Allowance (minutes) i RO I MEMORY 10CFR55.41 (b)(7) Source Documentation Source: D New Exam Item ~ Previous NRC Exam: (2007 ILT) D Modified Bank Item D Other Exam Bank: 0 Reference(s): Learning PLOT 5060 - 3a Objective: KIA System: 215005 APRM/LPRM RO/SRO 3.7/3.7 KIA Statement: K4.01 - Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following: Rod Withdrawal Blocks MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 8. Given the following:
- The 20Y050 supply from the Static Inverter is in a normal lineup
- A fault occurs on the 20Y050 Panel that results in an excessive current condition (>300 amp setpoint)
The Static Inverter _---10....(1::..L.)__ and the 20Y050 Panel_->.(2=)<----" A. (1) deenergizes when the input breaker (CB1) trips on overcurrent (2) deenergizes B. (1) receives a shutdown signal that opens both breakers (CB1 and CB2) (2) deenergizes C. (1) Static Switch swaps to the Alternate Source (2) remains energized while the fault clears D. (1) Static Switch is prevented from transferring to the Alternate Source (2) remains energized while the fault clears
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 8 RO - - - - - - _ . _ - -_._-----_. Choice Basis or Justification Correct: C The Static Inverter is current limited. If a fault develops it will automatically i transfer to the Alternate Source which can supply the larger current
! necessary to clear the fault and then transfer back to normal DC supply
+when fault clears.
Distracters: A i The Static Switch will transfer to alternate source in order to maintain
. 20Y050 panel energized.
The Static Switch will transfer to alternate source in order to maintain 20Y050 panel energized. D The Static Switch will transfer to the alternate source in order to maintain 20Y050 panel energized. Psychometrics Level of HIGH Knowledg~ . I______ QiffiCLJltL~____ __ TTime_All()~~c~tminLJt~~) .. RO 10CFR55.41 (b)(7) Source Documentation Source: New Exam Item ~ Previous NRC Exam: (PB 2007) D Modified Bank Item ~ Other Exam Bank: (PB 2010 Cert) ILT Exam Bank
-'--~""'--~-'---'-~ - - _ . . . . _ - - _ ......._---.._--- ---- .
__ C-'-----.L..._ ..... ____+-ARC-220 F-5 Learning PLOT-5058-5c Objective: KIA System: 262002 - Uninterruptible Power Supply Importance: RO / SRO (A.C.lD.C.) 3.1 /3.4 KIA Statement: K4.01 - Knowledge of Uninterruptable Power Supply (A.C. / D.C.) design feature(s) and/or interlocks ~~;~!~::~~~g l~~~~fro=~~~erre~~~~fi~~o:rSU~p!ies.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 9. A fire occurred in the Main Control Room requiring evacuation. The following conditions exist on Unit 2:
- The crew is executing SE-10 "Plant Shutdown from the Alternative Shutdown Panels"
- The URO is performing SE-10 Sheet 2 and SE-10, Attachment 9 to establish control at the HPCI Alternative Shutdown Panel
- The HPCI Gland Seal Condenser Vac Pump (20K002) will NOT start
- The HPCI Gland Seal Condenser Cond Pump (20P028) will NOT start For the above conditions what is the operational implication on continued operation of HPCI?
HPCI operation: A. may continue. The HPCI room airborne contamination levels will not rise. B. may continue. The HPCI room airborne contamination levels will rise. C. must NOT continue due to the excessive HPCI room airborne contamination levels. D. must NOT continue due to the potential to damage the HPCI turbine shaft seals.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key I I
---~~-~~.
Question # 9 RO Choice Basis or Justification Correct - SE-lO, Att 9 and FSAR sec 6.4.1 identify that HPCI operation may continue without gland seal condenser condensate and vacuum pump(s). Knowledge of steam turbines and steam supply to HPCI turbine is required for candidate to determine the impact of continued operation
+~ ___+...:..with9l:Jtgl~1'l9 seal condensate pump.
Distracters: A Incorrect - Shaft sealing will NOT function normally as the steam will leak into the surrounding room vice being condensed by the gland seal steam condenser sub-system. Plausible because the candidate may not understand how turbine seal steam functions via design steam leakage out C Incorrect - HPCI operation is permitted to continue without Gland Seal condenser system, as identified in SE-lO, Att 9. Plausible because the candidate may not know this fact but may deduce the impact of the failure on airborne contamination levels. ---------- --..... , o Incorrect - HPCI operation is permitted to continue without Gland Seal condenser system, as identified in SE-10, Att 9. Plausible because the
.~ ...~.. ~~~_...L.....~ _ _.L...::..C~1'l9i9(3tE:l_may not know this fact.
Psychometries Level of Knowledge Diffic~ ____ c--Time Allow~_nge {1TI!I'l.lJte~ RO HIGH 10CFR55.41 (b)(7) Source Documentation Source: [gJ New Exam Item D Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0 ~----- ILT Exam Bank RE:lfE?rE?I'l~~L ___ ~E:-10, AtL~,JlFS~13 para 6.4. L___ ~_.~_ Learning PLOT 5023 4b Objective: KIA System: 206000 HPCI RO/SRO 2.8/2.9 KIA Statement: K5.02 - Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM: Turbine shaft sealing REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 10. Per SO 14.1.A-2(3) "Core Spray System Alignment for Automatic or Manual Operation", which of the following methods must be used to verify the Core Spray System is adequately filled and vented?
(1) Verifying Core Spray Discharge Pressure is::: 50 psig (2) Verifying "A(B) CORE SPRAY LINE VENT ACCUMULATOR LOW LEVEL" alarm is clear A. 1 ONLY B. 2 ONLY C. 1 OR 2 D. 1 AND 2
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
~--- ........ _--------_._---_....... - - --------- ......... ~~-------~--~ ........... ---.-..... -----~-
Answer Key Question # 10 RO Choice Basis or Justification - ....... _ - - ----- - Correct: - As per SO 14.1.A-2, Core Spray System Alignment for Automatic or Manual Operation, verifying CIS discharge pressure (both locally and in the MCR in conjunction with verifying the Line Vent Accumulator Low Level alarm is clear is required verification that the ___ ~ __ ~ ___ ~, ...... ____ ~'__--=-"--':.::.--=c=----:..=---=:: is filled and__________
...=..==__==_ vented. .........._______ ~____ . ___________ _
Distracters: INCORRECT - SO 14.1.A-2, Core Spray System Alignment for Automatic or Manual Operation, verifying CIS discharge pressure in conjunction with
"~~~1'i!!9.the Line Vent~2~.lJ..!!lulator Lo\y~evel alarm is clear is required.
C INCORRECT - SO 14.1.A-2, Core Spray System Alignment for Automatic or Manual Operation, verifying CIS discharge pressure in conjunction with
,,~r:lfying the Lll"l~_Vent A~~lJ_r:nulat9r Lo~L~vel alarm is clear is reqlJir~d_. __ _
A INCORRECT - SO 14.1.A-2, Core Spray System Alignment for Automatic or Manual Operation, verifying CIS discharge pressure in conjunction with Line Vent Accumulator Low Level aICi!,!!is clear is required. Psychometrics _L~,,~_I oflSl"lowledge I-~-~_~-Pj.tfi~~_--i--Ilme A~owal"lc:;e (minut~S)LJ ___~ RO . . __ MEMORY
- I I 10CFR55.41(b)(14)
Source Documentation Source: [g] New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 _.. . . _____~~1--D-ILT Exam Bank ___ ~~_ Referenc~(sL _____ SO 14A.1 ARC 224 A-4 Learning PLOT5014 - 4e Objective: KIA System: 209001 LPCS Importance: RO I SRO
..........._ _~_. _ _ _ _ _ .. _. ______ ~ _ _ _ .........L_______ 2.5/2.5 KIA Statement:
K5.05 - Knowledge of the operational implications of the following concepts as they apply to LOW PRESS'-!BE CORE SPRt-.Y j3YSTEM :_§.y~t~m VenJi.r!9....~_ _____ __ _ ___ ________.. BEQIjII!I:D MATERIAL~ fNgl,jE- __________ ~_ ~_ . Notes and Comments: - -
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 11. Under NORMAL conditions, (eg: torus cooling for HPCI testing, Shutdown Cooling operations), HPSW is applied to RHR heat exchanger in order to (1) ; during transient operations (LPCI mode), per T-101 BASES, HPSW is applied to RHR heat exchanger in order to (2)
A. (1) minimize radioactive leakage (2) rapidly remove decay heat B. (1) minimize radioactive leakage (2) minimize radioactive leakage C. (1) minimize vibration of RHR heat exchanger tubes (2) rapidly remove decay heat D. (1) minimize vibration of RHR heat exchanger tubes (2) minimize radioactive leakage
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer I Question # 11 RO Choice Correct: 1 A i Basis or Justification CORRECT - Under normal conditions, HPSW is applied to RHR Hx to minimize radioactive leakage to the environment. Per procedure T-101 ____.. . ____ ___ ~_~~~~a~~~:~!'~~:nS: ~~~i~~t~ep;~~!~ ~~~~d~~;~~~c; ~~:~~:_~t Distractors: 1 B *. IN.. CORRECT - See discussion above. Plausible if candidate. does not _~~~: T-101BA~ES regarding HPSW operations. ._ C i INCORRECT - See discussion above. Plausible if candidate does not
- know T-101 BASES regarding HPSW operations.
1--*--------*------***-***** .. -L-~--- . . ----.. . - . -------.. . -- . . --.. . - . . . ----.. ---... D . INCORRECT - See discussion above. Plausible if candidate does not
... __________ .1 know T-101 BASES regarding HPSW operations.
Psychometrics Source Documentation Source: IZI New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank J3.elerence(~): T-J()L~C)~~§.,[)esig~ B()~is .P9_g.l_I11~rlt P-s-q.§l __ Learning PLOT 1560 Obj 9 Objective: KIA System 203000 RHR/LPCI: Injection Mode Importance: RO I SRO
--_._- 3.0/3.1 .... _ - - _ . .
KIA Statement K6.10 - Knowledge of the effect that a loss or malfunction of the following will have on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC): Component cooling water systems
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 12. Unit 2 was manually scrammed due to a leak in the torus.
Torus level is 10 feet and lowering. Unless otherwise directed by TRIP procedures, RCIC must be secured at: A. 9.5 feet, to prevent direct pressurization of the torus. B. 9.5 feet, to prevent exceeding the pump vortex limit. C. 6 feet, to prevent direct pressurization of the torus. D. 6 feet, to prevent exceeding the pump vortex limit.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 12 RO Choice Basis or Justification - --------- Correct: D CORRECT - T-102, Step T/L-16, RCIC is secured if it is aligned to the Torus in order to prevent vortexing. This limit is to be adhered to unless TRIP procedures direct the use of RCIC regardless of the limit Distracters: A INCORRECT - This is the level and reason for securing HPCI under these conditions. RCIC is not secured at this torus level because the energy the RCIC turbine exhaust can add to the containment is small and the turbine would likely trip on hjgh exhaust pressure should elevated containment
~ ~ __ ~~~L'~=~~
occur. Plausible if candidate confuses HPCI and RCIC limitations. B INCORRECT - This is the level for securing HPCI to prevent direct pressurization of the torus. RCIC does not get secured until 6 feet torus level. Plausible if candidate confuses HPCI and RCIC limitations. C INCORRECT - This is the correct level for securing RCIC, but incorrect reason. RCIC is secured at 6 feet torus level due to vortexing and not
..... ~ ____,_____~ __ ~~b_~e_c_a_u__ .....f.c:jire_ct.p!~~sLJriz_C!tio}!<?f the tOf.!J~:
se_.o . . ____~_____ __ Psychometrics
~evel of~no~_ed9.~__ L_ __ ___ ):>~ffic;u tty ________~ Time Allow...§l!1~~~ (rl'lj!1LJ!~~J_ ~----- -
RO MEMORY I 10CFR55.41 (b )(7) Source Documentation Source: ~ New Exam Item Previous NRC Exam: 0 o Modified Bank Item Other Exam Bank: 0
--------r - - ' = -I'L T Exam Bank R~ference{~):_ T-102 Sh 2 and T-102 Bases Learning PLOT 5013 E/O 10q Objective:
KIA System: 217000 - RCIC Importance: RO I SRO 3.5/3.5
~----.---.
KIA Statement: K6.03 - Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): Suppression pool water supply
---~---- . _ - - - . - . - - _....
REQUIRED MATERIALS:
------~~-~-------- ---------_._----
Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 13. Unit 2 is in MODE 4, twenty-four hours after shutdown, following extended full power operation.
- 2B RHR pump is operating in the Shutdown Cooling Mode.
- Reactor Coolant temperature is 135 degrees F on a very slow downward trend.
- No Reactor Recirculation pumps are in service.
- Reactor water level is being maintained at +30 inches.
- MSIVs are shut.
Which one of the following describes the Reactor Coolant temperature response if the operator secures the 2B RHR pump? (Assume no additional operator action is taken.) Reactor Coolant temperature will: A. Lower until equilibrium is reached with ambient drywell temperature. B. Lower until equal to HPSW temperature in the RHR heat exchanger. C. Rise until bulk boiling occurs, and reactor pressure rises above atmospheric pressure. D. Rise until bulk boiling occurs, with reactor pressure steady at atmospheric pressure.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 13 RO Choice Basis or Justification ~.--.----~---~ Correct: C CORRECT - Decay heat will cause RPV coolant temperature to rise and eventually reach boiling. Reactor pressure will increase above atmospheric pressure (NOTE: Even if examinee assumes RPV head vents are open
- pressure will still increase since the head vents are on a 1n line and are I designed for removal of non-condensibles at power or air removal for i h~~~Yd ditions.~~
~
ref.ue.
. .I.ing or . ro test c. . on. . T.. here is industry OE that confirms that bulk boiling of coolant due to lack of shutdown cooling will result in going
_______.. . . . . ____-;-- re~ter thall21 ?£?n(jpressuri~i!!9Jl1e REV with the vents Distractors: A INCORRECT - With the RHR pump tripped there is no longer shutdown I cooling flow from the reactor vessel to the RHR heat exchanger. Plausible if candidate has misconception regarding decay heat being absorbed by
~~~(jt:Y~ell cooli!!a§j'~tt!rn_*__ ~~~~~~~~~___~~_~___~~~__ _ __ ~_~~~_ ~~
B INCORRECT - With the RHR pump tripped there is no longer shutdown cooling flow from the reactor vessel to the RHR heat exchanger. Plausible if candidate has misconception regarding natural circulation flow via RHR INCORRECT - Reactor pressure will increase above atmospheric. Plausible if candidate believes head vent will relieve sufficient Psychometrics __Le\l~1 of Knowl~cj9t!___ HIGH
- ~~~~~ __~DJffic~ ~-
~ ~ ~ ~ ~ --+Iirn~ J\l!owal"lc;~_irnil1_l!t~~L __ " '0-RO . ~ .~
10CFR55.41 (b)(14) Source Documentation Source: New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0
.-~-~---t~~~~~ rgJJJ-J_~)(?m Bal"l~ (4266L____ ~ _____~~~_~
Rt!!~t!!1C;~ts): ~ ON-1 GP-12 Learning I PLOT5010 - 9.k.6 Objective:
---r~portance: ROI SRO
~~- -~-- -~ --~-~~--
KIA System: 205000 Shutdown Cooling -
~ __ _ _
~ ~~~~~~~~~~~ _________ ~______ __ _ _ .
~L __ 3.7/3.7 KIA Statement: A 1.06 - Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls incluc:lil"lg:ReactorJ~~ratlJr~! Cll!oderato!,_y~sse~!I~~_
REQUIRED MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 14. Given the following:
- Unit 2 is operating at 85% power when a design basis LOCA occurs.
- The output breaker on 125VOC battery charger 2B0003-1 trips open.
- Prior to tripping, charger 2B0003-1 was supplying the Division II 250 VOC bus.
Assuming no operator action, how will the plant respond to this event? The Division II 250 VOC bus will: A. remain powered at rated voltage supplied by battery charger 2B0003-2. B. remain powered at rated voltage supplied by the 2B station battery ONLY C. immediately de-energize until battery charger 2B0003-2 is placed in service. O. remain powered at rated voltage supplied by the 2B station battery AND the in-service 20 charger
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
.~-----~- .....
Answer
--------~-
Key Question # 14 RO Choice Basis or Justification
~------ ~----"""~--------i-------------~------- .. I, Correct: 0 CORRECT - 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 2B and 20 batteries. The batteries are designed to supply loads during a DBA for 2 hours .
Distractors: . --TTNCORFfECr-.:charger-280003-=2 must be manually placed in service...only one charger can be in service at a time, The question stem states "assuming no operator actions. B i INCORRECT - when the output breaker for charger 280003-1 trips, the I charger no longer supplies power to the Division II 250 VDC bus. The bus r ...... _ _ _ _ _ _ _ _ _ _ .~i~___ i loads would thenl:>~_~~~Lb.Y. BOTH th~ 2E!.f:l!l~~Qj;l?tter~§____ .. _ i
'1',. C ~NCORRECT - the battery will fully support all loads for approximately 2 hours with no battery charger; the bus will remain energized.
I,
"----.-~----- .... --~~
Psychometrics Level of Knowledg~___ Difficultv i Time Allowance (minutes\~_ RO
! - __ ~. . L i---~--------~-~LI 10CFR55.41(b)(7)
'i ...._ .__
HIGH Source Documentation Source: o New Exam Item [g] Previous NRC Exam (2007) o Modified Bank Item 0 Other Exam Bank ILT Exam Bank
... -~---+-----"=-----------.---.---.-----.~ ..-------- _ . - . - ~~---- .... _ - -
Reference( s): PLOT 5057, print E-26
...._ - - _...-_...._ - _..-...+----........ -.-.~--- .... --.--~ ............------........- - . - - -...... -.~----.- ...... --.
Learning PLOT-5057-7a Objective: KIA System 263000 DC Electrical Distribution Importance: RO 1 SRO 2.5/2.8 KIA Statement A 1.01 - Ability to predict andlor monitor changes in parameters associated with operating the D.C.
~~~~:;ij;~~~~~~~:L~~'O~ cF~:~n~Uding Ba=e~ch~~ngldisc~ar9ing. :te Notes and Comments:
-- ~ - -~ -- -- -
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 15. The plant electrical system is in a normal configuration.
The Transmission System Operator reports that supply grid voltage to the 2SU Offsite Source is 220 kV. This will result in (1) voltage on the 4kV buses, and requires use of (2) to mitigate the consequences of this condition. A. (1) under (2) SE-16 "Grid Emergency" B. (1) under (2) AO 50.1 "Response to Main Generator Perturbation caused by Grid Disturbance" C. (1) over (2) SE-16 "Grid Emergency" D. (1) over {2} AO 50.1 "Response to Main Generator Perturbation caused by Grid Disturbance"
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 I ~~"~""~~--~~-~-""""""""""""--"--"~- -----~~----""- --~-~-~-"~- ------~~-~- 19~estionfhf~(J~~-. , --""""""""""""--~-T--------'---- Answ._r_B_a~_"s~~i_SO-~-r-J~~tifi~~tion
--c-*-~-=-=-:-------c=----~------------~------------------------~---------~~
Correct: A 6 Entry Condition 1.2 identifies <225kV on 2SU as requiring entry into the procedure. Per SE-16 Bases, "If voltages are below the values listed above and a LOCA occurs, the 4 kV buses may transfer to ___ --+-t~h"~~[I~rgency Diesel Gener:.ator~due to grid undervol~~" __ u Distracters: B INCORRECT- Part (1) is correct, but part (2) is incorrect - plausible since AO 50.1 relates to grid disturbances, but is associated with Main Generator
, 2e!1lJrt:>a..tic~ns..__ .__~__ ~"~_ ~ ""~
C i INCORRECT- Part (1) is incorrect, but part (2) is correct - plausible because candidate needs to recognize that 220kV is LOW for the supply to 2SU. INCORRECT- parts (1), (2) are incorrect - plausible since AO 50.1 relates to grid disturbances, but is associated with Main Generator perturbations AND candidate needs to recognize that 220kV is LOW for the supply to 2SU. Psychometrics Source Documentation Source: [2J New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 __ DJLT Exam Bank
~Refereilce{~_~~ SE-16,__(3rid EmergencY,§lnsL?E-16 Bases~ _______ _
Learning
- PLOT 5053 E/O 10i Objective:
KIA System: 262001 - AC Electrical Distribution Importance: RO / SRO _~~ __~"_~____________________ ~ __~ __________________ ~.~_. _______~ ____ ~". _____________ 3.1 I 3.4 KIA Statement: A2.09 - Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Excee(jing'yolta~lil11itatl(>n_s ____.---___~__ ~ _______________._._____________.____ ~~__~ _________________.____~ __. REQUIRED MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 16. The following conditions exist on Unit 2:
- A reactor startup is in progress
- Critical data has just been completed
- The "B" and "E" WRNM channels simultaneously fail downscale and are displaying a Critical Self Test Failure The plant will respond with an _--'-'-'--_ and the crew should respond with (2)
A. (1) alarm ONLY; no rod blocks or scram signals (2) applicable Alarm Response Cards B. (1) alarm and rod block ONLY; no scram signals (2) applicable Alarm Response Cards and SO 62.7.A-2 "Receipt of Rod Blocks" C. (1) alarm, rod block, AND half scram ONLY (2) applicable Alarm Response Cards, SO 62.7 .A-2 "Receipt of Rod Blocks", and GP-11 E, "Reactor Protection System - Scram and ARI Reset" D. (1) alarm, rod block, AND full scram (2) T-100 "Scram"
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
~.---~~~~~~~-----~~~-----~ .......... - - - - - - - - - -.. --.__ .........._ - - - - - _ ........ _-_.. -~----------- ........ ---~-- - --~ -----~-- ......_---- ._-,
Answer Question # 16 RO
--~-.---.-. *------------T----~**********---------- .. ---~**********-------~-~
Choice
- - - ..- ------,--.. . . . .---+----------.. -.-------~-----Basis or Justification
. ~-------.----------.- . -~--~-
- Critical self test failure is a "trip" signal. One in each trip system will generate a full scram, annunciators for WRNM and RPS, and a control rod block. WRNM "B" is RPS B. WRNM "E" is RPS A. RPS Logic for WRNM is 1 out of 4 taken twice. Rod Block, RPS Scram, and Alarm all satisfied.
Distracters: A INCORRECT - Full Scram is expected - Plausible if candidate does not fully understand trip functions and/or logic scheme. B INCORRECT -Full Scram is expected - Plausible if candidate does not fully I understang trip functions and/or logic scheme. l~__ . . ___ _~;~=~;~i~~~;:~£~~~~o~~:;;~~:i~~ii~~~did;e-d~~s notf: Y Psychometrics __ Lev~LQfKnowledg~_~____ . ___ Pi1fi~LJJ!L _____LTil11e AIIQ~an~~Jl11jI'lLJtesL f--.. RO HIGH
- 10CFR55.41(b)(6)
Source Documentation Source: D New Exam Item D Previous NRC Exam [g] Modified Bank Item D Other Exam Bank ILT Exam Bank ARC-211 B-1 and C-1
-,----'----_......_ + - - - - -
PLOT-5060C-3b, -Sa, -6a
.--~------------ ..~------ ........ --.... ~-~-- -------- _.
KIA System: 215003 -Intermediate Range Monitor Importance: RO I SRO System 3.7/3.8
- - - - - - - - _........ __ ...... ---~--~- ..- . - - -
KIA Statement: A2.04 - Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the
~;ci~~~~~A~d~~=~tQonni~c~~~tion~ or o~eration~ _ scaleor d,,-w,,-"cal~trips Notes and Comments: - - - ---
~--- ~ - --- . -- -~-- . __._-_ ..
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 17. The following conditions exist on Unit 2 following a LOCA:
i Parameter: = Time 0 sec = Time 60 sec = Time 200 sec I RPV Level -60 inches -165 inches -195 inches I nt::'1'\ I RPV Pressure 850 psig ig 8§O psig I Drywell Pressure 1.4 psig 2.9 psig 4.5 psig I ! 2A RHR Pump OFF RUNNING RUNNING I I 2A, 2B CS Pumps OFF RUNNING RUNNING
- Other than scram actions NO other operator actions have been taken
- NO other ECCS pumps are operating Based on the latest conditions, which statement below describes the expected RPV pressure response? RPV pressure:
A WILL remain steady. B. SHOULD be lowering. C. WILL lower 105 seconds after RPV level reaches -160 inches D. WILL lower 9 minutes after RPV level reaches -160 inches
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 r~~~---- I~~~tlon Answer Key
-.-..... ~-~ ...
# 17 RO Choice Basis or Justification Correct: B CORRECT ADS initiation conditions are met with 2 psig drywell pressure and 160 inches RPV level with ECCS injection available - this is followed by a 105 second timer to allow for level recovery or operator intervention. Since NO procedure actions have been taken other than scram actions, the ADS inhibit switches have NOT been manipulated. ADS should have initiated but has failed to c.~.-.- . . . -..-.-- .... - .....- ... _-_.1-99 so asindic?tedJ)Y..B.Py..prE:!§§ur~l)1E:!?9y atl3?O_psi9:.. ..
Distractors: A ! INCORRECT - Plausible if candidate does not know specific ADS initiation criteria.
-... --~- ----r--......;- .--.. . .- -.. . ...- ....-....- -.. -.
~.-
C . INCORRECT - ADS initiation conditions are met with 2 psig drywell pressure and
. 160 inches RPV level with ECCS injection available - this is followed by a 105 second timer to allow for level recovery or operator intervention. By the given timeline, more than 105 seconds have already elapsed. Plausible if candidate
.. ___.._ _ .. 9ges notknqll'v'_~E:!c;ific A_QS irli!i.ati<J..n_~d.tE:!ria ...._._..._....__.._
D 'INCORRECT -ADS initiation conditions are met with 2 psig drywell pressure and 160 inches RPV level with ECCS injection available - this is followed by a 105 second timer to allow for level recovery or operator intervention. Nine minute timer is associated with LOW RPV level and nominal Drywell pressure, this indicates possible steam line break outside the drywell. Plausible if candidate does not know
. s.p~.clfic AQSi£litj§tionc!.i!eria. . .._ ........._.__._
Psychometrics
.b.~\lel of Kn()~I~~~l HIGH
~""- - . _.Qiffi~_l!I!Y_ _ --r.- Time i\it()"!.?nce(m inutes1_ I** ..*
RO 10CFR55A1 (b)(7) Source Documentation Source: [8J New Exam Item o Previous NRC Exam D Modified Bank Item o Other Exam Bank
....... --'--.~.
ILT Exam Bank
.Refer~nce(l)J.. _.+.cM.-'--._1-§i....51.,§_h~(3!§.~~ncL3 __._. ____.. ~...__ .....~~ ...._
Learning Objective: PLOT-5001G-9.kA
..... ...._ - _......- ... -~ ....... - .....
KIA System 218000 -~_uto~atic..~~press~r~z~tion_~yste: ... I_:Portance: RO/SRO 4.2/4.3 KIA Statement A3.08 - Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including: Reactor
- _ . . .Pressure
. . .-_._._ . .- - -....- ..... ~-,--.- ....- - - . -.... ~- .. --.. ... ...
~ - ~~.- ._.............
REQl!IR_ED~M~A_TE_R_IA_L..~: ....._ ..__.-+_:.:...~.::.:=-.. Notes and Comments: .-- ........ --~- ....
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 18. The E-1 Emergency Diesel Generator is started in accordance with SO 52A.1.B, "Diesel Generator Operations".
The Operator can check for normal field flash by observing E-1 EDG (1) and that the E-1 EDG is at rated speed by observing (2) A. (1) voltage between 4.16 and 4.40 kV (2) alarm 001 "G-4 E1 DIESEL RUNNING" AND frequency between 58.8 and 61.2 Hz B. (1) frequency between 58.8 and 61.2 Hz (2) alarm 001 "G-4 E1 DIESEL RUNNING" ONLY C. (1) alarm 001 "G-4 E1 DIESEL RUNNING" ONLY (2) frequency between 58.8 and 61.2 Hz D. (1) alarm 001 "G-4 E1 DIESEL RUNNING" AND frequency between 58.8 and 61.2 Hz (2) alarm 001 "E-1 DIESEL GEN NOT RESET'
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer
# 18 RO Choice .------~-.-----.-.-
Basis or Justification Correct: A CORRECT - Field flashing will result in EDG terminal voltage of approx. 4kV. If the field does not flash there is no terminal voltage generated. EDG rated speed can be verified by receipt of MCR alarm 001 "G-4 E1 DIESEL
......._. _ _ _...... _ _--+ _ _ _-t--R_Uf\lNING",ll,ND qbservin~QQ. fr~queng~et\l\leen§~JtCl_l1d 61.?l-iz.
Distractors: INCORRECT - (1) Field flash is NOT related to the speed (frequency) of the EDG .. (2) Plausible because receipt of MCR alarm 001 "G-4 E1 DIESEL RUNNING" occurs at 855 but is NOT the rated EDG "'...."""'..... INCORRECT - (1) Field 'flash starts at the LSS (250 rpm). Plausible because receipt of MCR alarm 001 "G-4 E1 DIESEL RUNNING" occurs at t - - - - - - . - . - - + - - -.. 5 I3Et\/1. (2) Cqrrect
----+._8_5_ ______.. . __... ______ ....._... _. __
INCORRECT - Plausible if candidate confuses alarm 001 "E-1 DIESEL GEN NOT RESET" as being related to normal EDG start sequence. Alarm is actually received if EDG trips. EDG rated speed can be verified by receipt of MCR alarm 001 "G-4 E1 DIESEL RUNNING" AND observing _ _ _ _ _ ...1._ _ _ _ _ _ .1 _~PGJr'~quency_!>~.1ween 5~~~..Clnd 61.?!j~. ___ .._.. __ ._. ____ Psychometrics _b~Y~J of KnowledfLe___ ~Lfficulty ___._ Time AllovyCln~~_(rninu~~~2 I
~~~~-.
RO - MEMORY I 10CFR55.41 (b)(8) Source Documentation Source: [8J New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0
......_ _ _ _ _ _....... _ _-+ _D ILT E)(.Cl'!l Bank .________.. . . __ ._._... _ ..__ ._____ _ ~::;~~n~~2-:--+~CO~;;~3~DG Q~Cl!ions _ _ . .
Objective: II KIA System: LU"+UU'U EmergencY,Diesel Generators Importance: RO / SRO 3.0/3.1 KIA Statement: A3.01 - Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESEL/JET) including: Automatic starting of compressor and emergency generator
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 19. Unit 2 is operating at 100% power with the following conditions present:
- Instrument Air Header pressure as read in the Main Control Room on PI-2425A and B located on panel20C12 is as follows:
Based on the above, Instrument Air Header Pressure is (1) and procedure (2) will provide direction for correcting the condition. A. (1) Low (2) SO 36B.B.A-2 "Instrument Air System Routine Inspection" B. (1) Low (2) ON-119 "Loss of Instrument Air" C. (1) High (2) SO 36B.B.A-2 "Instrument Air System Routine Inspection" D. (1) High (2) SO 36B.1.A-2 "Unit 2A and B Instrument Air System Startup"
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Choice. Basis or Justification cor;;~t~-l~B r-C~ORRECT~i;:;;t;um;~t~AirH;~d~r pressure is low. *ON-119 will provide
, ! direction to address the IA header pressure condition. ON-119 entry Distract;;~-. ",-- i-;~~~~~~:~:;~~;;;;:s:i;r~:"~e~~!~s-u-re-i~I~~-,h-o~;~~rON-119 will i
- provide direction to address the IA header pressure condition, not 36B.8.A
~~~--+-c ~~b~~~~~:~:;:~;n~~i;!~!~~:~~~~~;~iS i "Instrument Air System Routine Inspection"._
low. ON-119 will provide
, direction to address the IA header pressure condition, not SO 36B.8.A-2 L------'-., D I INCORRECT - Instrument Air Header pressure is low, however ON-119 will I
. provide direction to address the IA header pressure condition, not SO 36B.1.A~2 "Unit 2A and B Instrument Air System Startup" _J Psychometrics RO 10CFR55.41 (b)(10)
Source Documentation Source: D New Exam Item D Previous NRC Exam IZJ Modified Bank Item D Other Exam Bank: 0
--~-i~~~
ILT Exam Bank ... ~~ ..... ~-~~-~~~-~------~~~~-~--- Reference(sl~~_~~~ ON:'1J~! ARC 216~~~D_-~3_~__~. . ~__~~. _~~ . . . _~._._~ . . __ .. _~~_~ ..~_~ Learning PLOT-5036-9j Objective: KIA System: 300000 Instrument Air Importance: RO / SRO 2.6/2.7 KIA Statement: A4.01 - Ability to manually operate and/or monitor in the control room: Pressure gauges MATERIALS:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 20. Unit 2 reactor startup is in progress.
- 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.
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 License NRC Examination April 2013 Answer Question ......# 20 RO
- - - - - -.. ~---.---,
Choice Basis or Justification Correct: B CORRECT - When the "AUTO START' pushbutton is depressed then the RFPT will 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 speed will be controlled at the speed the turbine is at when the button was pushed. t-- **- 1i-------******** Distractors: A INCORRECT - Plausible due to speed range of 400-600 rpm is the minimum governor control speed achieved by depressing the "SLOW" or "FAST RAISE" push buttons. The auto start can be aborted by pushing any
, speed LOWER or RAISE pushbutton and the turbine speed will be
_____ S;gntrolled at the speed the tutbine is at V'{b~1l th~I.Jt1on w~~J'l!shed _ C \ INCORRECT - Plausible due to speed range of 400-600 rpm is the
. minimum governor control speed achieved by depressing the "SLOW" or "FAST RAISE" pushbuttons.
D INCORRECT - The auto start can be aborted by pushing any speed LOWER or RAISE pushbutton and the turbine speed will be controlled at the _ _ _ _ _ ---'-_ _ _ ......_ ...1___ speed the turbine is at when the button was pushed. __ _ _ ..... __ _ _ _ _ .. . Psychometrics ._- L.~'.t_ElI..2it5nowl~s!g\! -[---- _._J?ifficul~. . _~L:ril11e AIIQ.vycmc~Jmi_nut~LI.... RO HIGH I . 10CFR55.41(b)(7) Source Documentation Source: D New Exam Item L2J Previous NRC Exam (2008 NRC) D Modified Bank Item D Other Exam Bank ILT Exam Bank --_....._--- Referen~~:..___ SO 6C.1 . .A___ . -2____.. . ____..___...... ___._.___.. _ ...___._ Learning PLOT-5006-4c Objective: KIA System 259002 - Reactor Water Level Control Importance: RO 1 SRO 3.7/3.6 KIA Statement A4.02 - Ability to manually operate andlor monitor in the control room: All individual component controllers in the automatic mode. t~~;~~~~:!~::~AJ,S'=~tON~-=~ . - ----..
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 21. Which of the following are design functions or features of the Emergency Diesel Generator (EDG) and Auxiliary system(s)?
- 1. Support safe shutdown load requirements for BOTH units during simultaneous DBA accidents.
- 2. Support safe shutdown load requirements for a SINGLE unit during LOCA with a simultaneous Loss of Off-Site Power.
- 3. Allow for failure of ONE EDG.
- 4. Provide sufficient fuel for 14 days of continuous EDG operation at Design Basis Event conditions.
A. 1 and 3 B. 2 and 4 C. 2 and 3 D. 1 and 4
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 21 RO Choice Basis
- - - ....... _-_ or.. _
Justification Correct: C CORRECT - Design Basis for EDG (and Auxiliary) System is to ensure that sufficient power is available to provide for the functioning of required emergency safeguard and selected non-safeguard systems for safe shutdown of both reactor units assuming a Loss of Coolant Accident (LOCA) in one unit, a Loss of Offsite Power (LOOP), and failure of one Distractors: INCORRECT - Design Basis for EDG and Auxiliary does NOT allow for simultaneous LOCA accidents. One EDG failure is allowed. Plausible if candidate does not recall the~e specific de.sigl'![eatlj.r:.es. B INCORRECT - Sufficient fuel is provided for 7 days operation. 14 days is plausible because this is TS LCO allowed action time for having one EDG out of service . ..... _ - - - - - - - INCORRECT - Basis and plausibility previously explained. Psychometries L.evelof I<QQwledg~ I-j- .._ _.... __. RO Memory 10CFR55.41 (b)(8) Source Documentation Source: [gJ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank BefE:..r:.ence(~t . DBD P-S-07._______ Learning PLOT- 5052 Obj 1 Objective: KIA System: 264000 EDGs Importance: RO / SRO
......- - - - _3.9 14.0 KIA Statement: 2.1.27 - Conduct of Operations: Knowledge of system purpose and 1 or function.
REQUIRED MATERIALS: Notes and Comments: - - - .... _-_ .... - - - ........ _..__ .. _.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 22. Unit 2 was at 100% power. A low RPV level transient occurred. The following conditions exist:
- Reactor power: 55%
- RPV pressure is 1000 psig
- RPV level is -75 inches and steady
* 'A'SBLC Pump is injecting into the RPV
- HPCI and RCIC are injecting into the RPV
- ALL Individual Scram Test Switches at Panel20C016 were momentarily placed in the SCRAM position and returned to the UP position
- T-220 ,"Driving Control Rods During Failure to Scram" has been directed
- All Full Core Display Blue Scram Lights are NOT lit For the above conditions which one of the following Equipment Operator actions must be performed to cause a reactor shutdown?
A. Lineup SBLC Tank to the RWCU Precoat Tank per T-212, "RWCU System SBLC Injection" B. Lift leads to defeat ARI initiation logic and install jumpers to defeat RPS scram signals per T-216, "Control Rod Insertion by Manual Scram or Individual Scram Test Switches" C. Remove group scram solenoid fuses per T-213, "Scram Solenoid Deenergization" D. Close HV-2-3-56 "Charging Water Header Block Valve to HCUs" per T 246, "Maximizing CRD Flow to the Reactor Vessel"
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 22 RO Choice Basis or Justification
~~T------'-~-----********~~*-----~*****~---------------*----
Correct: C CORRECT - An electric ATWS exists. T-213, "Scram Solenoid Deenergization" is required to be performed. The first part of T-213 is to place all Individual Scram Test Switches in the SCRAM position. Since this was already completed the next step is to remove group scram solenoid to insert control rods. Distractors: - T-212, "RWCU System SBLC Injection" is required due to SBLC already injecting into the RPV. ~~- . . . --.~--~ B 1-fNCORRECT - T-21S, *Control Rod Insertion bY-Manual Scra," or i Individual Scram Test Switches" is required to be performed for a hydraulic A TWS. This is an electric A TWS. o INCORRECT - Closing HV-2**3-56 "Charging Water Header Block Valve to HCUs" per T-246, "Maximizing CRD Flow to the Reactor Vessel" at 1000 i psig RPV pressure will have no impact on driving control rods. _L_ . . ____ ._ . ___.~~ . ~_ ...___ ~.. _~ . __~__~_. Psychometrics Level9f Knowledg~mL ____ RiffiCUI!Y__ --1ILme Allowance (mir1lJl~sL ~-~--~ .*. RO
-~-~---~-~ ... ~.
HIGH i 10CFR55.41 (b)(12) Source Documentation Source: [gI New Exam Item 0 Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: () ILT Exam Bank ~F3~fer~r1_~~~1_~ __I-101 T-213 Learning PLOT-1560 Obj 13 Objective: KIA System: 2000 RPS Importance: RO / SRO 3.8/4.0 KIA Statement: 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 23. Unit 2 scrammed due to low RPV level. The following conditions exist:
- RCIC auto started to restore level, which reached a maximum at
+35 inches
- RCIC is now in manual control with the flow controller dialed low (0 gpm)
- RPV level is -10 inches and lowering slowly
- RPV pressure is 940 psig, controlled by EHC
- RCIC discharge pressure is 660 psig
- RCIC turbine speed is 2800 rpm
- RCIC indicated flow is 0 gpm
- Torus and CST levels are normal With no further operator action, what is the result of leaving RCIC in its current configuration?
RCIC will _ _ _ _ _ _ __ A. trip on turbine overspeed B. pump CST water to the Torus C. suffer exhaust check valve damage D. trip on high turbine exhaust pressure
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Choice Basis or Justification
.~-~-----~--.~~
Correct: B CORRECT - Based on the given conditions, RCIC is running with the minimum flow valve open. Since RCIC suction is lined up to the CST and the minimum flow discharge is to the torus, CST water will be pumped to the torus. Distractors: A 1 INCORRECT - RCIC will trip on overspeed under certain conditions if the I controller is in AUTO, I.e. in CST-to-CST mode and MO-23-24 (common I return to the CST) closed due to high Drywell pressure or HPCI suction swap from the CST to the Torus. With the controller in MANUAL none of Joe congJtigns thJ:i.1I~~Cid to.Jl_I'l()v~~t?~9 event ~~~p.!esent. C INCORRECT - Exhaust check valve damage is not a concern above 2200 rpm. D INCORRECT - RCIC will not trip on high turbine exhaust pressure under the given conditions. RCIC is designed to run on min flow for extended periods. Psychometrics Level of Kno'y!I~~gg~ RO HIGH 10CFR55.41 (b)(7) I Source Documentation Source: D New Exam Item C8J Previous NRC Exam: (2009 NRC) D Modified Bank Item D Other Exam Bank: 0 ILT Exam BankQ___ ~ ReferenQ~is ): M-359 Sheet 1 SO 13.1.C Learning PLOT-5013-9.k.7 Objective: KIA System: 217000 - Reactor Core Isolation Cooling Importance: ROISRO System 3.5/3.5 KIA Statement: A4.11 - Ability to manually operate and/or monitor in the control room: Condensate storage tank level REQUIRED M~"tE;RIALS--= ~ - ~11f\lON~_ Notes and Comments:
-- - --~ ---- - ---
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 24. A reactor startup from cold conditions is in progress.
RPV pressure is being raised to 150 psig per GP-2, "Normal Plant Start-up." While monitoring nuclear instrumentation it is expected that withdrawal of control rods from position 36 to 48 will result in: A. little or no indicated power change due to low control rod worth B. little or no indicated power change due to high control rod worth C. a substantial rise in indicated power due to relative proximity to neutron detectors D. a substantial rise in indicated power due to not yet establishing positive pressure control with bypass valves.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key
------------.. -----~-- .. - - - - - -
Question # 24 RO Choice-----,-- .. _-_._- ... - . _ - _ .. Basis or Justification Correct: A CORRECT - Control rod worth is low from core positions 36 to 48 as described in GP-2 "Plant Startup". Distracters: INCORRECT - Plausible because control rod worth is low, not high, from core positions 36 to 48 as described in GP-2 "Plant Startup". INCORRECT - Plausible because neutron detector proximity does have an effect on count rate but only at control rod notch locations 16 through 22. o I INCORRECT - Plausible beca~~~ the ca~dicJ~te could determine that pressure changes will cause power to rise, however, there is no significant effect at this low RPV pressure with control rods being moved from position 36 to 48. Ps chometrics Level of Knowlecjge Difficulty, Time Allowan"" {minutesL 11 OCFR:S041 b) 1) MEMORY Source Documentation Source: rgJ New Exam Item D Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0
. O_I_b..T.I~xarn Bank Reterence_(~L__ _ . GP.:_£_._ .. __. __.. . .__ _
Learning PLOT-5060C Obj 11 Objective: KIA System: 215003 -Intermediate Range Monitor Importance: RO I SRO System 3.7/3.7 KIA Statement: A 1.02 - Ability to predict and/or monitor changes in parameters associated with operating the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM controls including: Reactor P2V'!E:;r indica.ti9n res"p0rls~ to !O.c:l position cbange~ _.. . . REQUIRED PJiATERIALS:_ 1f-!9NE - -. Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 25. 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" has been implemented as directed by T-101 "RPV Control" 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 License NRC Examination April 2013
~~~- .......
Answer
~~-~~~---~ ..... -~~-.
Key -.---~- ...... ~---~ Question # 25 RO ~------ ...... ~ -----~-- Choice ~~--~ .... -- Basis or Justification Correct: A The inboard MSIVs are supplied with Instrument N2 from both the 'A' and
* 'B' 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 MSIVs but not the outboard MSIVs. Distractors: B The outboard MSIVs are supplied by Instrument Air. C The outboard MSIVs are supplied by Instrument Air. The inboard MSIVs are supplied by Instrument N2. Psychometrics Levelc>LKf!QwlE3c:l.9E3 ~. Di1fic:~I!y__~_~ I Time Aliowance.lrTlinut~~) RO .. Memory 1 OCFR55.41 (b)(7) Source Documentation Source: New Exam Item [gJ Previous NRC Exam (2008 NRC) D Modified Bank Item D Other Exam Bank flef~nce(sL --lO~11~_;;:;~3B_;~~~~J; M~3_7_2~ht1' T-261 Learning PLOT-5001A-7b Objective: KIA System 300000 Instrument Air Importance: RO I SRO 3.1 /3.2 KIA Statement K1.05 - Knowledge of the connections and / or cause effect relationships between INSTRUMENT AIR SYSTEM and the followin.9: Main Steam Isolation Valve air -REQUIRED MATERIAis:~ ~*-JNIO~E _ ~---~ _ Notes and Comments: ~
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 26. Unit 2 is operating at 100% power when the Digital Feedwater Control System (DFCS) experiences a loss of one main steam flow input due to failure of the "An Main Steam Line Flow Transmitter (DPT-2-6-51A).
Which one of the following identifies (1) the expected DFCS response, if any, and (2) what action the operator should take? A. (1) Lowers feedwater flow to match the lower steam flow signal. (2) Enter OT-100 "Reactor Low level" B. (1) Automatically transfers to single element control. (2) Verify reactor level is being maintained by DFCS per ARC 201 H-1 FEEDWATER FIELD INSTRUMENT TROUBLE. C. (1) Automatically transfers to single element control. (2) Manually transfer DFCS back to three-element control lAW SO 6C.1.D-2, "Reactor Feedwater Automatic Level Control". D. (1) Remains in three element control. (2) Manually transfer DFCS to single-element control lAW SO 6C.1.D 2, "Reactor Feedwater Automatic Level Control".
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 26 RO Choice Basis or Justification Correct: B CORRECT - as described in SO 6C.1.D-2, the DFCS will automatically default to single element control upon loss of a steam flow signal. ARC 201 __ . . . --,....::.H..:...-1 dire~tsJb~_9perator to verlfy~at~r:.level is beinRcontrolled byJ?PCS. Distracters: A INCORRECT - Plausible if candidate does not know that the DFCS will automatically shift to single element control on loss of a steam flow input I and RPV level will remain ".~---- ... ~-. the same. C INCORRECT - as described in SO 6C.1.D-2, the DFCS will automatically default to single element control upon loss of a steam flow signal, however, the system will NOT allow transfer back to 3-element with a failed steam
- flow input.
D rINCORRECT~-Plau~ible if candidate does notk~ow that the DFCS will automatically shift to single element control on loss of a steam flow input. Psychometrics Level of KnoV'll~.99.e I Diffi~_ulty. -J. Time Allowance (minutes) * . RO HIGH ! T 10CFR55.41 (b )(7) Source Documentation Source: C8J New Exam Item D Previous NRC Exam: () D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank .. _.- ..... _ .Refec~n~~ts): SO 6C_11.Q:~, ARC 201 H-1 Learning PLOT 5006-7i Objective: KIA System: *-lI~~rt~nc;:- RO I SRO
- 1. ___ 3.3/3.4 KIA Statement:
A2.01 - Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of any number of main steam flow hpljts__ REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 27. A Unit 2 startup is in progress with the following plant conditions:
- Reactor power is 25%.
- Generator output is 200 MWe.
- Annunciator TURBINE STOP VLV. CLOSURE & CONTROL VLV FAST CLOSURE SCRAM BYPASS (210 A-2) is lit.
- A relay failure causes the Power-to-Load Unbalance lockout to actuate.
- The POWER LOAD UNBALANCE TRIP (206 B-1) annunciator goes into alarm.
Based on the above conditions, which one of the following describes the expected plant response? A. Reactor scram ONLY. B. Generator lockout and turbine trip ONLY. C. Generator lockout, turbine trip and reactor scram. D. The turbine remains online; the reactor does NOT scram.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- - - - - - - - -.. -.-~.. -
Answer Key Question # 27 RO Choice ~------ Basis or Justification Correct CORRECT - If the PlU circuit energizes, a generator lockout and turbine trip will occur. Since reactor power is < 29.5% RTP (turbine 1st stage pressure is < 138.4 psig, equiv to 28.9% RTP), a reactor scram will not occur as a result of the TSVITCV closure. The turbine bypass valves will rapidly open, preventing a scram from high reactor pressure/neutron flux. The end result will be the reactor at 25% power with the turbine-generator
----- ----..... ~
off-line.
-f- -------i--~:---
Distractors: INCORRECT - The reactor does not automatically scram. C INCORRECT - The reactor does not automatically scram. D INCORRECT - The PlU circuit will produce a generator lockout/turbine trip. Psychometrics level of Knowleqg~ .___ HIGH DifficlJlty - Time Allow~ce lminutes) 1* 10CFR;S041 (b)(7) Source Documentation Source: New Exam Item [g] Previous NRC Exam (2007) Modified Bank Item D Other Exam Bank _DllT EX<lI'l1_Bank ____~ __ _ ReferenceJ§): GP-2, ARC 206 B-1 TS Bases 3.3.1.1 ._. . . ________.______ _ learning ! PLOT5001B- 5f Objective: KIA System ! 245000 Main Turbine Gen. / Aux. Importance: RO / SRO 3.4/3.5 KIA Statement K1.08 - Knowledge of the physical connections and/or cause- effect relationships between MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and the following: Reactor/tufl?il1~_p_ressure contr.Q1 s.1~t~rl1___ REQUIRED MATERIALS: . -.-.--f. NONE Notes and Comments:
- -- ~~"---------- ---~
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 28. Unit 2 Backup Scram Valves (SV-2-3-140A and SV-2-3-140B) are powered from (1) and are normally (2)
A (1) Safety-Related DC (2) de-energized B. (1) Safety-Related DC (2) energized C. (1) 120 VAC RPS (2) de-energized D. (1) 120 VAC RPS (2) energized
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 28 RO Choice ......... Basis
~--~ ..
or Justification Correct: A CORRECT - The Backup Scram Valves are powered from 125 VDC panels 2PPA (Div. I) and 2PPB (Div. II), respectively. They are normally de
.........f~- ........ p~.nergj~~~_Gi.r1c:1er1~rgjz~Jo functi()n.
Distracters: B i INCORRECT - Power supply is correct; the Backup Scram Valves are
. normally de-energized.
C INCORRECT - Power supply is incorrect. D INCORRECT - Power supply is incorrect; the Backup Scram Valves are normally de-energized. Psychometrics Lev~,-oLlSno\IV~c:l9.~ Difficlj!ty____ __ f. Til1'1~~lIOW~rlC~(l1'1inut~sJi RO MEMORY 10CFR55.41 (b)(7) Source Documentation Source: o New Exam Item rgj Previous NRC Exam: (PB 2009) o Modified Bank Item 0 Other Exam Bank: 0
-+-~ ~.- . ILT Exam Bank ........ ~-~---~
Referenf~(sJ: E-26 sheet 1 M-1-S-54 sheet 8 Learning PLOT-5003A-2c Objective: KIA System: 201001 - CRD Hydraulic Importance: RO I SRO 3.5 I 3.6 KIA Statement: K2.03 - Knowledg~.ol~I~~trical p()\IV~r_supplie~.!()th~J()Jlo\Al'~~_E:3§1Ckup§9R.J\M_ valve solen()ids. REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 29. Given a constant input of contaminated water to the Unit 3 RB Floor Drain Sump AND a failure of both associated sump pumps, contamination levels in the room will:
A. rise because the sump will overflow to the room floor. B. not be affected because the sump is sealed and excess input will backup to the source. C. not be affected because the sump will overflow directly to the Waste Collector Tank before the top of the Floor Drain Sump is reached. D. not be affected because the sump will overflow directly to the RB Equipment Drain Sump before the top of the Floor Drain Sump is reached.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 29 RO Choice Basis or Justification Correct: D Correct - The Reactor Building (RB) floor drain sump will overflow to the equipment sump before allowing contaminated water to overflow into the SLlI!!PJOOI"ll.. _____ .....- _ ------- _-- .._. - - - _ Distractors: A Incorrect, Plausible ... however the Reactor Building (RB) floor drain sump will overflow to the equipment sump before allowing contaminated water to
-----i----- ()\ferfJ()w iQ1() !he_s~'!'Pfoom. ______..
Incorrect, Plausible ... however the sumps are not sealed and would I B overflow into the room if the overflow to the Equipment Sump did not exist.
-----T- I Incorrect, Plausible ... however the overflow will not go directly to the Waste I C Collector Tank, will overflow to the equipment sump. Equipment Sump
_______L____ _ f>umpswill th~_n_p~rl1pJLoy~rJ()JI1~ V\I~~1e_goll_~ctor Tank. Psychometrics L~,,_~Lof I5noVIII~clge_ . ___ j)iffiflJj~ ___ -------- - __ ILrlJ e ~QV\l.al"1<::~(mil1utestl RO MEMORY I 10CFR55.41 (b)(13) Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 _____ ~.l~I~~~_I!1_ E3~n_~ Referenc;e{s2: M-369 sheet 1 Learning PLOT5020 - 6d Objective: KIA System: 268000 R;d~aste -- ---- ----------TI~~~rtance: ._. RO/SRO _ ... __ . . . ____ . _ _ _ _ . L_ _._. . ._ _ _ _ _. . _ _ _. _____________J _____________ 2.7 I 2.8 KIA Statement: K3.04 - Knowledge of the effect that a loss or malfunction of the RADWASTE will have on following: Drain sumps
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 30. In the event of a demand signal failure, Recirculation Pump minimum and maximum speed is limited by:
A. Mechanical stops on the Scoop Tube ONLY B. Mechanical stops on the Jordan positioner ONLY C. Scoop Tube Lockup circuitry AND Mechanical stops on the Scoop Tube D. Scoop Tube Lockup circuitry AND Mechanical stops on the Jordan Positioner
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 30 RO -~~~~-~~ Choice ~~--~~-+~~-~-~----~~~~~~~~~~~~~~~~ Basis or Justification Correct: D - Recirc Pump speed is limited by scoop tube lockup circuit and stops on the Jordan positioner. Distracters: A . INCORRECT - No such mechanical limiter - plausible because there is a
, mechanical limit on the scoop tube positioner.
B INCORRECT -Incomplete answer - plausible because Jordan Positioner mechanical stops are part of the limiting equipment, but the scoop tube LI()c;~up ~gr~ujLalso eff~~tiyelylil'l1lts~Reed excursions. C INCORRECT - Partially correct answer - No mechanical stops on the scoop tube - plausible because scoop tube lockup circuit is part of the limiti~g ~qlJi~ITlt;3rl!, ~1()!1ll~it~th~~ordan Positioner mechanical stops. _ ~ Ps chometrics ~~ Level of Knowledgt;3 Difficulty~ __ ~ Time Allowance (l'Tlinutes} RO MEMORY Source Documentation Source: [2J New Exam Item D Previous NRC Exam: 0 I D Modified Bank Item D Other Exam Bank: 0 ~::~~~~~(S) *** I D ILT Exam Bank
*.1{{~f;:~~~~E~--:::-0-0-2--9-~2-*-(~J****()****rcl~~F>()§i1ioner Sp~~Jfication)
Objective:
~-.-.~-.-L.- - .---.-~~-
KIA System: ! 202002 Recirculation Flow Control Importance: RO/SRO 2.9/2.9 K/A Statement: K4.07 - Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature(s) and/or interlocks whi9h provid~ f.9r tht;3 follo\IVin9.:"_~ Minimum~nd m.§l~im~u~m~l!ITlP spee(!.setpoints REQUIRED MATER~ALS~~ __ IJ!9~E __ __ Notes-and Comments: - ---- - ----- - - -
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 31. Fuel handling activities are being conducted in the Unit 2 Spent Fuel Pool in preparation for the upcoming refuel outage. The following conditions occur:
- A spent fuel bundle is moving to a new pool location and is currently over an area of empty fuel storage racks
- Fuel pool level has lowered by 1.5 feet for no known reason
- The Fuel Floor Area Radiation Monitor (ARM) is alarming
- The Main Control Room has entered T-103 "Secondary Containment Control"
- A GP-15 Evacuation of the Refuel Floor has been directed Per FH-74 "Actions in Response to an Unexpected Loss of Fuel Pool, Reactor Cavity, or Equipment Storage Pool Water Inventory", prior to leaving the Refuel Floor and securing the Refueling Bridge the spent fuel bundle must:
A. be placed in its new designated storage location B. be placed in its original storage location C. be placed in the nearest storage location D. remain in its present suspended location
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key
-- -- -~-~~--~ ...
Question # 31 RO corr~c~~ice -',l-ct~~~~~:~;:;~~;~~;=~;~~~~~~~;::~~~e~~~~~f prior to leaving the Refuel Floor and securing the Refueling Bridge the
- spent fuel bundle must be placed in the nearest available underwater storage location if time is vital. The ARM alarming and GP-15 evacuation t--**-- --...-'--
make time vital. Distracters: A INCORRECT - Per FH-74 "Actions in Response to an Unexpected Loss of i Fuel Pool, Reactor Cavity, or Equipment Storage Pool Water Inventory",
. this action is only performed if radiological conditions permit, which is not
._____+_th__ e_ case since the ARMi~ala!.I"l1ing: _______ .__
B INCORRECT - Per FH-74 "Actions in Response to an Unexpected Loss of Fuel Pool, Reactor Cavity, or Equipment Storage Pool Water Inventory", this action is only performed if radiological conditions permit, which is not the_~as~~i'2c;e the ARM is aIClfll1il1.9.: D
- INCORRECT - Per FH-74 "Actions in Response to an Unexpected Loss of
- Fuel Pool, Reactor Cavity, or Equipment Storage Pool Water Inventory",
I this action is not an option. Even if time is vital and no storage location is available, at a minimum the bundle would be lowered as low as possible
. i bef()re_If:}~vingJb~B~J.ueling__ E!d~ge. ___ _
Psychometrics i
~ev~L()f!<:.I1_o~I~c!g~___ Difficulty _ IJime Allowance (rnin_lJtes) RO t
HIGH
- 10CFR55.41 (b)(12)
Source Documentation Source: ~ New Exam Item Previous NRC Exam: 0 Modified Bank Item o Other Exam Bank: 0 __ DJbT Ex~ml?;:Ink ________ . Referenc~{sJ:.._ FH-74 Learning PLOT 5019-9.k.2 Objective: KIA System: 234000 Fuel Handling Equipment ROISRO 2.9/3.4 KIA Statement: K5.03 - Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLlNQJ~:9UIP~ENT: Water Cl~Cl~_hi~lc!.C!.~Clinst radiation REQUIRED MATERIALS: - NONE- -- Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 32. Which of the following design features ensure reactor vessel water level remains above the top of the fuel during a DBA Main Steam Line (MSL) break accident outside Primary Containment?
A. MSL Flow Restrictors ONLY B. Main Steam Isolation Valves ONLY C. EITHER MSL Flow Restrictors OR Main Steam Isolation Valves D. BOTH MSL Flow Restrictors AND Main Steam Line Isolation Valves
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 32 RO Choice Basis or Justification Correct: I D Correct - per UFSAR Ch04, Sec 4.1, the MSL flow restrictors limit steam flow during the time required for MSIVs to close, ensuring RPV level remains above TAF, thus protecting the fuel barrier. The MSL flow restrictors are an Engineere~~C!fetYE~iltlJre. Distractors: A Incorrect - see above. Plausible as the flow restrictors are part of the protection scheme. Incorrect - see above. Plausible as the MSIVs are part of the protection scheme. C ! Incorrect - see above. Plausible as the flow restrictors and the MSIVs are I each part of the protection scheme, but neither is sufficient alone .
......... ~. . ..~~.........J_.~..~...... _ _ ~. ___ ~~ _ _ _ ~_
Psychometrics Level of Kno.!,!,leqge Difficulty ......_---- - ------ -- _ Time ,l\JlQ!'!'~!1_~~(ll1inutes) --~ .... RO MEMORY I 10CFR55.41 (b)(3) Source Documentation Source: [gl New Exam Item Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank Refererlce(s}: .... ___LLJES,l\-'3~~.94, Sec 4.1 and 4.5 ~11~ Ch 1 Sec 14.6.5 Learning I PLOT 5001A-3a, b Objective: KIA System: 290002 Reactor Vessel Internals Importance: RO I SRO 2.9/3.1 KIA Statement: K6.20 - Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR VESSEL INTERNALS: Main steam system REQUIRED MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 33. Given the following:
- Unit 2 is operating at full power.
- The 2B Steam Jet Air Ejector (SJAE) was placed in service using SO SA.6.A-2, "Placing the Standby SJAE in Service and Placing the In Service SJAE in Standby".
- Ten minutes later the URO notes that FI-4020 (lower indication) "Off-Gas System Flow" on Panel20C007A is reading 120 scfm and steady.
This value of Off-Gas flow is (1) than normal and will result in (2) Main Condenser vacuum. A. (1) higher (2) improving B. (1) higher (2) degrading C. (1) lower (2) improving D. (1) lower (2) degrading
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question I D# 33 RO Choice Basis or Justification
~~~-~~~~~~~~~~~~~~~~~
----~----t-~---------~---~------~~~~~~~~
Correct: B CORRECT - Based on the Routine Inspection, the expected range for Off Gas System Flow is 20-45 scfm. 104 scfm is well above the expected range and based on the ACMP for Unit 2 (Elevated Main Condenser Air In Leakage) condenser vacuum will be degrading. Distractors: A INCORRECT -104 scfm is above the expected range and will correspond to degrading (not improving) condenser vacuum. C INCORRECT -The normal range of off-gas flow is 20-45 scfm. D INCORRECT -The normal range of off-gas flow is 20-45 scfm. Psychometrics SRO N/A Source Documentation Source: D New Exam Item Previous NRC Exam D Modified Bank Item ~ Other Exam Bank (2008 PB Cert) D ILT Exam ~an~ _____ _ Reference(s):
- SO 8A.6.A Placing Standby SJAE in Service
~ ~_~9~.§--A=g9ff-Q_~~_§y~t~rn _Routine 11!~2.ectiQr-l __ _
Learning i PLOT5008-9.k.7 Objective: i Knowledge/Ability 1271000 Off-gas Importance: RO I SRO KIA ~~~~~~ ___________ ~ ~ ___ _ 3.1 /3.1 (Description of K&A, from catalog) A 1.08 - Ability to predict and/or monitor changes in parameters associated with operating the OFFGAS SYSTEM controls including: System flow REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 34. Unit 2 was operating at full power when the following transient occurred:
* 'C' Reactor Feedwater Pump tripped.
- Feedwater flow is 13.0 E6 Ibm/hr and rising slowly
- A and B RFP speed are rising slowly
- RPV level is +16 inches and lowering slowly.
- Reactor Power is 98% and steady.
Based on the above plant conditions, the Reactor Operator must immediately: A. run both Recirculation Pumps manually back to 30% per ARC 214 B-3 (G-3) "A(B) RECIRC FLOW LIMIT".
- 8. run both Recirculation Pumps manually back to 45% per ARC 214 8-3 (G-3) "A(8) RECIRC FLOW LIMIT".
C. lower power in accordance with GP-5, Power Operations, until water level is restored. D. perform a plant shutdown in accordance with GP-4, Manual Reactor Scram.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 I - i Answer Key Choice Basis or Justification
-r----~---------
Correct: CORRECT - RPV level 17" with a RFP less than 20% flow should result in Recirc runback to 45%, which has failed as evidenced by power being I steady at 98%. Operator needs to manually initiate run back of Recirc ______ JEul11p~!()5~!~ ~m~ed. Distractors: A . INCORRECT - Runback is required, but to 45% not 30%. Plausible if I candidate confuses runback setpoint necessary for condition. C INCORRECT - Runback has failed as discussed above. Plausible because a GP-9IS directed for an OT-100 (Low Reactor Level) condition. GP-5 power reduction cannot be performed fast enough to preclude scram on low level. o INCORRECT - Runback has failed as discussed above. Plausible because failing to perform the runback may result in a scram on low level, but the runb~c_Us~~~l9Ded t()J)~clude this s(;;ram so should be iniJiflted FIRST._ Psychometrics Source Documentation Source: D New Exam Item Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0 [gJ_~~IJ~}(§lm ~C!I1K___ . ___ _ Referen~~(§J:_ I. OT-100 Reactor -. Low ...--------. ..---- Level, ARC 214 B-3 Learning
- PLOT -5006 Obj 3i Objective:
KIA System: Importance: RO I SRO 1 __ 3.7/3.7 KIA Statement: A2.01 - Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump trip REQUIRED MATERIALS: ----r******.**---*** NONE--------
---~~~~~~-
Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 35. Unit 3 is operating at 70% power with the following conditions:
- Elevated Main Steam Line Radiation levels due to a suspected fuel clad leak.
- A steam leak develops in the HPCI Room that cannot be isolated.
- Reactor Building ventilation exhaust RIS-3-17-452A, B, C and Dare reading 20 mr/hr.
- Refueling Floor ventilation exhaust RIS-3-17 -458A, B, C and Dare reading 8 mr/hr.
Which one of the following describes the response, if any, of the Reactor Building Ventilation System to the above conditions? A. ONLY Reactor Building ventilation isolates. B. ONLY Refueling Floor ventilation isolates. C. Both Reactor Building and Refueling Floor ventilation isolates. D. Neither Reactor Building nor Refueling Floor ventilation isolates.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 35 RO Basis or Justification
------~----------------------------
C CORRECT - Both Reactor Building and Refuel Floor Ventilation isolate (Group 3 signal) when either Reactor Building ventilation exhaust RIS-3-17 452A or C AND B or D are reading.::: 16 mr/hr OR Refueling Floor ventilation exhaust RIS-3-17-452A or C AND B or Dare _________________ c________.-~~~dLn-R.::: 16 mr!bt: ___________ _ Distractors: A ! INCORRECT - Both Reactor Building and Refuel Floor Vent isolates. Plausible if candidate believes ONLY RB will isolate due to higher Reactor Buildin9J:iLRacL~Qndition . B INCORRECT - Plausible if candidate believes that Reactor Building and Refueling Floor ventilation hi radiation setpoints are different. D INCORRECT - Both Reactor Building and Refuel Floor Ventilation isolate (Group 3 signal) when either Reactor Building ventilation exhaust RIS-3-17 452A or C AND B or D are reading.::: 16 mr/hr OR Refueling Floor ventilation exhaust RIS-3-17-452A or C AND B or Dare r~adin9_~_16 mr/hr. _____~________ Psychometrics I Level ofl$r:lo_wlecj9E3 __ [)lfficulty _I Time Allowa~~{minutesl I RO I HIGH 10CFR55.41 (b)(9) Source Documentation Source: D New Exam Item Previous NRC Exam: 0 [Zj Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank Reference{s): _ ~RQ._318--':::>-=-4l..TS T~bl~}}-,'§J-= 1 Learning PLOT5040B-3a Objective: K/A System: 288000 Plant Ventilation Importance: RO / SRO 3.8/3.8 KIA Statement: A3.01 - Ability to monitor automatic operations of the PLANT VENTILATION SYSTEMS including: Isolation/initiation signals REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 36. The following conditions are present on Unit 2 following a LOCA:
- Recirc Pumps are tripped
- Reactor level is -25 inches and lowering
- Reactor pressure is 850 psig and lowering
- Drywell pressure is 8 psig and rising
- Drywell temperature is 250 degrees F and rising
- Torus level is 19 feet and rising
- DWCW return header pressure is 26 psig
- Drywell cooling fans are tripped
- The "B" Loop of RHR is NOT available
- SYSTEM I RHR CONTAINMENT SPRAY SELECT IN MANUAL OVERRI DE (224 D-2) is in alarm
- Performance of T-204 "Initiation of Containment Sprays Using RHR" has just been directed.
Based on the above conditions, containment Spray logic (1) spray initiation. Procedurally the above conditions allow (2) NOTE: T-223 Figure 1 "DWCW Saturation Curve" is PROVIDED ON THE NEXT PAGE. A. (1) permits (2) spraying the Torus ONLY per T-204 "Initiation of Containment Sprays Using RHR" B. (1) permits (2) spraying the Drywell and Torus per T-204 "Initiation of Containment Sprays Using RHR" C. (1) does NOT permit (2) restoring Drywell Cooling per T-223 "Drywell Cooler Fan Bypass" D. (1) does NOT permit (2) spraying the containment per T -205 "Initiation of Containment Sprays using HPSW"
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 T-223-2 Rev. 6 6 of 6 FIGURE 1 DRYWELL CHILLED WATER (DWCW) SATURATION CURVE 350 325
~
~ .."",
v 300 V
./
.- 275
.,.,....,c O!
L V 250
/
/
- .£:
- ZZ5
~,
,...., I 200 o 10 20 30 40 50 60 70 80 D\VC\V RET,(JRN HEADER PRESSlJRE ON PI-2016.:! (PSIG)
... TI 80146 is out of service, use RT-O-40C-530 2 t.o determine DW Bulk Temperature.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Choice Basis or Justification Correct: A CORRECT - Per T-1 02 "Primary Containment Control" with torus level at 19 feet, drywell spray is not permitted due to covering the vacuum breakers
~(t9rus sP@Y!~ not allowed if torus level ~ 21J~et).__ ~
Oistractors: B I INCORRECT - Torus (but not Orywell) sprays are permitted. Plausible
, because applicant may NOT recognize that with torus level at 19 feet, drywell spray is not permitted due to covering vacuum breakers (torus SPEC3Y isnoLCillo~ed if torus leveL~?1 fe~1t INCORRECT - logic to spray is satisfied. Plausible if candidate does not understand logic inputs.
-logic to spray is satisfied. Plausible if candidate does not logic inputs.
Psychometrics Level c>f15Qowleqge ..1 HIGH Source Documentation Source: o New Exam Item 0 Previous NRC Exam: [g] Modified Bank Item Other Exam Bank: 0 Reference(s): ARC-224 0 ...2; T -102 Learning PLOT-501 0-4s Objective: KIA System: 230000 - RHRlLPC~To~~s/Suppression -Tl~~ort. ance: ROISRO Pool Spray Mode I 3.6/3.3 K/A Statement: A4.09 - Ability_tc>'I}lCi~nLJ()IIy"'c>p~fate and/or Q1QDitorirlJhecontrol f9Q.m: Indicating lights andalarms. REQUIRED MATERIALS:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 37. Unit 2 is operating at 100% power
- Drywell Pressure unexpectedly rises to 1.2 psig and is trending up.
- OT-101, "High Drywell Pressure" has been entered.
The operating crew must IMMEDIATELY: A. perform GP-3, "Normal Plant Shutdown". B. perform GP-4, "Manual Reactor Scram". C. scram and enterT-101, "RPV Control" ONLY. D. scram and enterT-101 "RPV Control" AND T-102 "Primary Containment Control".
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 37 RO Choice Basis or Justification Correct: B CORRECT - A GP-4 Manual Scram is required at 1.2 psig in Drywell. Distractors: A INCORRECT - Not required unless both seals on a Recirc Pump fail. However, requirement to scram at 1.2 psig still applies. C INCORRECT - T-101 is not required to be entered until drywell pressure reaches 2.0 psig .
... ---- -----_.. --+-_._
D INCORRECT - T-101 and T-102 are not required to be entered until drywell pressure reaches 2.0 psig. Psychometrics RO 10CFR55.41 (b)(1 0) Source Documentation Source: D New Exam Item D Previous NRC Exam: D Modified Bank Item Other Exam Bank: 0 ILT Exam Bank
.-~----- -
Refe~e~e_{s 1_ OT-101 Learning PL0T1540-6 Objective: KIA System: 223001 Primary CTMT and Aux. Importance: RO I SRO 4.6/4.4 KIA Statement: 2.4.49 - Emergency Procedures I Plan: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. REQUIRED MATERIALS: -------- ._---_. Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 38. Unit 2 is at 100% power when the 2A RPS MIG Set output breakers spuriously trip.
What effect does this malfunction have on the '2A' Rod Block Monitor (RBM) AND associated Operator Display Assembly (ODA)? A. De-energize the '2A' RBM AND associated ODA B. De-energize the '2A' RBM ODA ONLY. C. De-energize '2A' RBM ONLY D. NEITHER '2A' RBM OR associated ODA will de-energize.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 38 RO Choice .... ~.~ ---- - Basis or Justification Correct: C CORRECT - Loss of 2A RPS bus de-energizes the 2A RBM, but the associated ODA remained powered because power is fed from 120 V uninterruptible power (20Y50 panel) which is powered from either emergency power or DC backup. Loss of the 2A RPS bus will not impact this Distractors: A INCORRECT - Plausible if candidate does not understand above described power supply arrangement. B . INCORRECT - Plausible if candidate does not understand above described power supply arrangement. o INCORRECT - Plausible if candidate does not understand above described power supply arrangement. Psychometrics Level of Knowhedge ! MEMORY i Source Documentation Source: New Exam Item Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: 0 ILT Exam Bank Referen~~(~:_ PLOT -506()~_M-1-S-34 Sht 38 Learning PLOT5060-7b Objective: KIA System: 215002 Rod Block Monitor Importance: RO I SRO
-~---~------ -~-----
3.0/3.2 KIA Statement: K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the ROD BLOCK MONITOR SYSTEM: RPS REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 39. Unit 3 is operating at 100% power.
- The Main Generator automatic voltage regulator is in control.
- A grid disturbance results in steadily lowering grid voltage.
How will the Main Generator initially respond to the above conditions? MWatts will (1) MVars will (2) Stator Winding Current will (3) A. (1) rise (2) lower (3) rise B. (1) remain steady (2) rise (3) rise C. (1) lower (2) rise (3) remain steady D. (1) remain steady (2) remain steady (3) rise
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 39 RO Choice .----- .---- ......-. -----r**-- ----+ ... Basis or Justification Correct: B CORRECT - Lowering Grid Voltage will cause the automatic voltage regulator to raise generator terminal voltage (overexcitation) in an attempt to maintain grid voltage steady. This will result in additional VARS, which raises statorc.LJ!rent. MWs remain un~h.a_r1R~cL Distractors: A INCORRECT - see explanation above - MWs will not change, MVARs will rise. C INCORRECT - see explanation above - MWs will not change, stator current will rise. D INCORRECT - see explanation above - MVARs will rise. Psychometrics Level of KnOwled9..E!_J ------ - _Difficulty ----. Ti_rn~_t.llowC!n~~_(minut~sl - - RO HIGH I 10CFR55.41 (b)(4) Source Documentation Source: [gJ New Exam Item D Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: () ILT Exam Bank Reference(st PLORT 12-04D = . . . ._._ Learning I PLOT-5050 Obj 9k Objective:
-.---. f ... ------
KiA System: 700000 Generator Voltage and Electric Grid i Importance: RO 1 SRO Disturbances
- 3.3/3.4 KiA Statement: AK1.02 - Knowledge of the operational implications of the following concepts as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following : Over-excitation.
REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 40. Per T -102 Bases, which one of the following describes (1) the drywell high pressure limit and (2) the consequence of exceeding this limit?
A. (1) 60 psig (2) inability to operate SRVs B. (1) 60 psig (2) loss of containment integrity C. (1) 62 psig (2) inability to operate SRVs D. (1) 62 psig (2) loss of containment integrity
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 40 RO Choice ---~--- Basis or Justification Correct: A CORRECT - 60 psig is the PCPL-A limit - as discussed in the Bases, this supports ability to operate SRVs, and is "utilized to ensure the pressure capability of the Primary Containment". (See PCPL-A Bases discussion). If containment pressure is >60 psig, and Instrument Air supplying containment pneumatics at the minimum pressure of 85 psig, then there may not be sufficient differential pressure across the SRV bellows to open the valve.
------I--~,~-=: -------~-- -~~~-------- ---------~--
Distractors: B INCORRECT - Plausible as 60 psig is close to the Containment Design pressure of 62 psig. C INCORRECT - Plausible as 60 psig is close to the Containment Design
. pressure of 62 psig.
I Psychometrics _~evel of Knowlecjg~ ___ __ Difficulty _ -t Time Allowance (minutes ) I i RO MEMORY 10CFR55.41 (b)(9) Source Documentation Source: ~ New Exam Item D Previous NRC Exam: 0 D Modified Bank /tem D Other Exam Bank: 0 ILT Exam Bank Reference(s): __ T-1021!~se~,PCPL-A Bases in T-BAS Learning PLOT 5007 -5a Objective: KIA System: 295024 - High Drywell Pressure Importance: RO I SRO 4.1 14.2 KIA Statement: EK1.01 - Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: Drywell integrity: Plant-Specific REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 41. 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 License NRC Examination April 2013 Answer Key Question # 41 RO Choice Basis or Justification Correct: CORRECT - per FF-01 "Fire Brigade". Distractors: A INCORRECT - The size of the fire is not defined by FF-01.
-------J B INCORRECT - This is a requirement from ON-114 to scram the reactor .
. INCORRECT - This is associated with the time limit for performing EAL
- classifications.
Psychometrics ___ Qifficu~ I Time AII2\IVanceJrnin_ut~:;)I_. . RO I I 10CFR55.41(b)(10) Source Documentation Source: New Exam Item [g] Previous NRC Exam (PB 2008) D Modified Bank Item D Other Exam Bank i D ILT Exam Bank .~::;~~;:eW~--f~~~~;~e:L_03 Objective: I KJA System 1600000 - Plant Fire On Site RO/SRO 2.9/3.1 KIA Statement AK1.02 - Knowledge of the operational implications of the following concepts as they apply to Plant ~~~E~~":~::~~~=~:tti()N~~--***
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 42. The following drywell conditions exist on Unit 2 following a small-break LOCA:
- Drywell temperature is 270 degrees F and rising slowly
- Drywell pressure is 8 psig and rising slowly Which one of the following is correct regarding initiation of Drywell sprays for these conditions?
Drywell Spray Initiation Limit (DWSIL) Curve DWfT-2 is PROVIDED ON THE NEXT PAGE. Initiation of drywell sprays in order to control drywell temperature will initially result in a (1) reduction in drywell pressure due to (2) cooling of the drywell atmosphere. A. (1) slow (2) convective B. (1) slow (2) evaporative C. (1) rapid (2) convective D. (1) rapid (2) evaporative
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 CURVE DW/T-2 DW SPRAY INITIATION LIMIT 800 -"........, _560 ...... ,L
~SOO J
- Q..
.2450 /
w w 4DO I a
- 360 J
w
~300
-- ,/
~260 J
=
CD 200 / Q
-= 150 I 100
/
o 2 4 6 8 10 12 14 16 DRYWELL PRESSURE (PlIO)
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 42 RO Choice Basis or Justification Correct: o CORRECT - Based on the given conditions, drywell atmosphere is super heated. Per T-102 Bases for step DW/T-13, initiation of sprays will initially result in evaporative cooling and rapid pressure reduction, followed by convective cooling and controllable pressure reduction. Purpose of DWSIL curve limitation is to ensure that even if an evaporative cooling condition exists, the capacity of the Torus to Drywell vacuum breakers is not
~~~--t--~--~--f-exceeded. ----~.-----.
Distracters: A INCORRECT - see above. Plausible if candidate believes that the DWSIL curve is based on preventing conditions that would result in evaporative coolinga~nd.-!~i(tpJ'"essur~J~ducti()n .~_._~ B INCORRECT - see above. Plausible if candidate believes that the conditions would result in evaporative cooling, but that the pressure reduction would be slow, indicating the candidate has a knowledge
~ deficienc.Y~Ln fl.lndam~l1~l§_olJluid tb~rm~clynamic:§.
C INCORRECT - see above. Plausible if candidate believes that the conditions would result in convective cooling, but that the pressure reduction would be rapid, indicating the candidate has a knowledge dE3fic:l~r]c:y.in fundamentals of fluidmermodyn~rnics .. ,--------------------------------------------------------------------- ..~ Psychometrics Source Documentation Source: [gI New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item Other Exam Bank: 0 _.01 ~T E.)(~I11_!?~nk Reference(s):_ T -1Q?~E:)C!§es1 ~§.t~"Y[T-1~ ..... Learning PLOT2102 DBIG - Obj 6 Objective: I KIA System:
- 295028 - High Drywell Temperature Importance: RO I SRO 3.7/4.1 KIA Statement:
EK2.01 - Knowledge of the interrelations between High Drywell Temperature and the following: DrY'v'.'eli Spr_~y:Jv'ark I&II~~_ REQUIREDMATERIAL.§:~~ 1~§.!~~am Tables Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 43. With both Recirculation Pumps running at a speed of approximately 1400 rpm, a sustained loss of Reactor Building Closed Cooling Water (RBCCW) occurs.
In accordance with ON-113 "Loss of RBCCW", the recirculation pumps A. must be tripped immediately B. must be tripped within 1 hour C. may remain running provided CRD seal purge flow is maintained D. may remain running provided pump temperatures remain within procedural limits
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 43 RO
-----[
Choice Basis or Justification Correct: CORRECT - Per step 2.8 of ON-113, pumps can remain in service if below temperature limits. Distracters: A INCORRECT - This action is not required by ON-113. Plausible because loss of RBCCW could lead to overheated bearings on the Recirc Pumps, 1-------- + tr necessitating ipQirl9 oftJJ_~[>LJmp§,-_ B INCORRECT - Not limited by time in ON-113. Plausible because loss of RBCCW could lead to overheated bearings on the Recirc Pumps, neces~~tatin9Jri[>piI]92fJ~e~lJ_rn2s. ____ INCORRECT - Not directed by ON-113. Plausible if candidate incorrectly believes that seal purge flow provides adequate cooling to recirc pump seals. Psychometrics Lev~1 of Knowled9E3 DifE<2ljl!y Time Aliowan<2E3Jminutes) RO MEMORY 10CFR55.41 (b)(10) Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 lSI Modified Bank Item D Other Exam Bank: () ILT Exam Bank RefereJ'1<2E3!~): ON-113 Learning PLOT-DBIG-1550 Obj 2 Objective: I KIA System: 1295018 - Partial or Complete Loss of RO/SRO Component Cooling Water 3.4/3.6 KIA Statement: AK2.02 - Knowledge of the interrelations between Partial or Complete Loss of Component Cooling Water_and the JQ.llowing:J'lant 0r~at~o.ns.- ____ REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 44. A transient occurred on Unit 3 that resulted in the following plant parameters:
- Reactor pressure: 900 psig
- Drywell pressure: 18 psig
- Drywell temperature: 235 degrees F
- Torus pressure: 16 psig
- Torus temperature: 145 degrees F
- Torus level: 15 feet Which one of the following conditions will cause the margin to the Heat Capacity Temperature Limit (HCTl) to be reduced?
A. Torus level lowers B. RPV pressure lowers C . Torus temperatu re lowers D. Drywell temperature rises
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 44 RO Choice Basis or Justification
~------------~~-~--------~---
Correct: A CORRECT - Lowering Torus level will cause the HCTL to be more restrictive. Distractors: B INCORRECT - Lowering RPV pressure will cause the HCTL to be less restrictive. C INCORRECT - A lowering Torus temperature will raise the margin to HCTL. INCORRECT - Drywell temperature has no effect on HCTL. Psychometrics Lev~Lpf !$!lQ\II{I~~g~ 1--- Difficllity TimeAIIQ\II{~flce ll'l1i!1ut~S)+_1 OCFR:5~41 (b)(9) HIGH I Source Documentation Source: New Exam Item ~ Previous NRC Exam (2008 NRC) Modified Bank Item D Other Exam Bank ILT Exam Bank ~ --~- ----- ---------- R~f~renc~_(Si): ___ TRIP/~A~P CuIY~s_'_I~_~Le_ and Limits - Bases Learning PLOT-21 020BIG-6 Objective: KIA System 295026 - Suppression Pool High Water . Importance: RO I SRO Temperature 1__ . 3.5/3.7 KIA Statement EK2.06 - Knowledge of the interrelations between Suppression Pool High Water Temperature and the following_:_§LJPpre.ssionpool Le"e1____ REQUIRED MATERIALS: NONE~~ --~- -------- Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 45. The Shift Supervisor has decided to abandon the Main Control Room and has entered procedure SE-10 "Alternative Shutdown".
Prior to leaving the Main Control Room, the reactor is scrammed to: A. ensure that Drywell pressure does not go above 2 psig B. facilitate depressurizing the RPV at less than 80°F/hr C. allow RPV level control with Condensate D. avoid lifting any SRVs until control is established at the Alternative Shutdown Panel
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 45 RO Choice Basis or Justification Correct: B CORRECT - a main goal of SE-1 0 "Alternative Shutdown" is to take the
- RPV to a cold shutdown condition. This cannot be accomplished without Distractors: A I~?~~-~~~~~~Il~~:tr:~:::~n for scramming th~ reactor durin~~ performan~~
of SE-10 "Alternative Shutdown". It is anticipated that drywell pressure will rise grater than 2 psig, especially if there is a loss of off-site power, due to lack of dl}'VVE3IL~QQlinR~_~ INCORRECT - Condensate pumps are secured prior to scramming the reactor during performance of SE-10 "Alternative Shutdown". RPV level control will be with HPCI. I INCORRECT - SE-10 strategy includes closing the MSIVs and allowing the I SRVs to open on setpoint until the crew can take positive control of SRV LQf>E3rat!<?Ilfr"om _t~e Alternative Sh~tdQ,,!ne~rlel._ 0 _ ...~
- Psychometrics Level of Knowledge_ ~ Diffic;ljlty I,' ~Iime Aliowance(l'l1in_utE3~)1 RO MEMORY I 10CFR55.41 (b)(9)
Source Documentation Source: 19] New Exam Item D Previous NRC Exam D Other Exam Bank ILT Exam Bank RefE3~ence(~) ~ UFSAR ChaptE3r_t1 0 Learning PLOT 1555 DBIG Objective: KIA System 295016 - Control Room Abandonment Importance: RO I SRO 4.1/4.2 KIA Statement AK3.01 - Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT: Reactor SCRAM REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 46. Unit 2 has an ATWS transient in progress.
- RPV level is being maintained in band of -195 inches to -172 inches using HPCI.
- The CRS has assigned injection of 21 % of Standby Liquid Control (SBLC)
Tank Level as a critical parameter milestone. When the SBLC milestone is reached, then: A. RPV level can be raised to +5 to +35 inches and the reactor will remain shutdown under hot standby conditions. B. RPV pressure can be reduced and the reactor will remain shutdown under hot standby conditions. C. RPV level can be raised to +5 to +35 inches and the reactor will remain shutdown under all conditions. D. RPV pressure can be reduced and the reactor will remain shutdown under all conditions.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
~.~ .....- . - - - -
Answer Key --_ .... _--- - - .... ----~-- Question # 46 RO Choice -~-- ..... Basis or
~~---~- ....
Justification Correct: A CORRECT - Per T-117, Step lQ-31 , lQ-33 and NOTE #33,21% of SBlC tank is Hot Shutdown Boron Weight - achieving this milestone permits raising RPV level to +5 to +35 inches, which is desired at this point to nrnrT1rt'TC '!1~xiD.~nc! _g!~~!ll?ut!p~."QfJb~J-?O~9!!.1flJ_~,~ ReV~_ .. Distractors: B INCORRECT - Per T-101, Step RC/P-14, depressurization during an ATWS is not performed until the ENTIRE SBlC tank has been injected Plausible if candidat~g<:>nfu~~s-'evel ClI*l<:l_P!~~_sl.J~eguidance. C INCORRECT - Plausible if candidate confuses Hot Shutdown Boron Weight (21% of SBlC Tank) with Cold Shutdown Boron Weight (Entire SBlC Tank L'-'-2!~rl'1e) D i INCORRECT - Plausible if candidate confuses Hot Shutdown Boron Weight
. (21% of SBlC Tank) with Cold Shutdown Boron Weight (Entire SBlC Tank VQ1LJmel_ ____ _
Psychometrics I level of Knowledg~ RO HIGH 10CFR55.41 (b)(9) Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam o Modified Bank Item 0 Other Exam Bank
. 0 IlT Exam Bank Reference(s): I TRIP/SAMP Cu~~~,T~-ble and Limits - Bases, -117 and Bases, T T-101 and Bases - Definition of "Hot Shutdown Boron VV~i9.~f' ___ _
learning PlOT-DBIG-2117-6, Objective: I KIA System 295037 SCRAM Conditions Present and ! Importance: ROISRO Reactor Power Above APRM Downscale or 3.2/3.7 Unknown KIA Statement: EK3.04 - Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Hot shutdown boron weight REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 47. Unit 3 plant conditions are as follows:
- Mode 5
- Core Shuffle 1 is in progress
- A leaking fuel bundle is dropped in the Spent Fuel Pool
- Refueling Floor ARMs are in alarm
- Reactor Building and Refueling Floor Ventilation Systems isolate
- Standby Gas Treatment System initiates For the above conditions, which one of the following describes the reason for the Reactor Building and Refueling Floor Ventilation Systems isolations?
A. Prevents an off-site radiation release B. Provides a filtered and elevated release C. Maintains radiation exposure to station personnel ALARA D. Prevents ductwork failure by routing the release through hardened ducts
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 47 RO Choice ------------ Basis or Justification Correct: CORRECT - Aligning the release through the SBGT System provides filtration and elevates the release through the main stack versus the vent stack Distractors: A INCORRECT - This does not prevent a release but lowers the radioactivity
. of the release and elevates it as described above.
C ,INCORRECT - The specified condition is related to radiation alarms. The i radiation dose to station personnel is not changed by the isolation,
, however, the severit of the release to the is minimized.
INCORRECT - This distractor is based on a procedural caution about SBGT is aligned, but is not the basis for the isolation. Psychometrics i Lev~19f~Q9wll3dJl~_ ___~ffi~~J!y ---- ,--- Time Allowance (minutes) RO I HIGH 10CFR55.41 (b )(9) Source Documentation Source: D New Exam Item Previous NRC Exam: 0 D Modified Bank Item IZJ Other Exam Bank: (LORT) Reference(s): Learning Objective: KIA System: 295023 Refueling Accident Importance: RO / SRO 3.3/3.6 KIA Statement: i AK3.03 - Knowledge of the reasons for the following responses as they apply to REFUELING . ACCIDENTS: Ventilation isolation fREQUIRED
- - - - ' - ' - MATERIALS:
------------- NONE
-+----------------------- - - - - - -
Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 48. Unit 2 is operating at 100% power.
EHC Logic fails such that Load Set fails towards 0%. Which one of the following describes the EHC System, turbine, and reactor pressure response? Control Valves will (1) , Reactor pressure will initially (2) , Bypass Valves will (3) A. (1) close (2) lower (3) open B. (1) open (2) lower (3) remain closed C. (1) open (2) rise (3) open D. (1) close (2) rise (3) open
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 48 RO .. ...
~ ~.~-~-,--~~- ...
Choice .--~.~~ ..~- Basis or Justification Correct: o CORRECT - Load set signal enters a low value gate, causing control valves to close. With the reactor at rated power, pressure will initially rise aD.eJi:>ypass 'It'~b{~~will opeD~_~~ Distractors: A INCORRECT - Reactor pressure will initially rise on Control Valve closure, not lower. B INCORRECT - All 3 parts of this distractor are opposite to the real system response. C INCORRECT - Control Valve closure will occur, not opening. Psychometrics
-U~v~l()f Knowleqge l~~ ~~--~- Diff'~9~J1L ....1 Time AII()Y!'an~~(mir1~!~_~LI RO . ...
HIGH I 10CFR55.41 (b)(4) Source Documentation Source: New Exam Item D Previous NRC Exam: ()
~ Modified Bank Item D Other Exam Bank:
ILT Exam Bank Referen~~~(§t~ PLOT-5001 SO 1B.1_._A~~~_~~ Learning . -~ 1 PLOT-5001 DL Obj 6a Objective: KIA System: 295025 High Reactor Pressure Importance: RO I SRO
~_ I 3.8/3.8 KIA Statement: EA 1.02 - Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: Reactor/turbine pressure regulating system REQUIRED MATERIALS:~INONE~-~- - -~- -~--
Notes and Comments: _~~~_~~_T~y.I:?ILcL9ill:TandLORT ~~!1kqLJ~§tions referenced above
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 49. Unit 2 is on line operating at 25% power with both unit auxiliary buses being supplied by the main generator.
- The Plant Reactor Operator transfers loads on the #1 Unit Aux bus to the S/U feed (#11 breaker shut) using procedure SO 53.A-2 "Transferring Unit 2 Aux loads from Unit Auxiliary Transformer to Startup Feed Buses".
- Prior to transferring the #2 Unit Aux bus a Main Generator lockout occurs on generator Differential Overcurrent.
Assuming no further operator action, based on the above conditions, what is the status of the following Main Generator components?
- 1. Generator output breakers will _ _,
- 2. Exciter field breaker will - - - -
- 2. Voltage regulator will _ _ _ __
A. remain closed, open, transfer from manual to auto B. open, remain closed, transfer from auto to manual C. open, open, transfer from auto to manual D. remain closed, remain closed, transfer from manual to auto
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 49 RO f- - ------- Choice Basis or Justification Correct: C CORRECT - on a main generator lockout trip, both output breakers will open, the exciter field breaker will open, and the voltage regulator will transfer from auto to manual. In the stem of this question, having the unit , initially at 25% power with the #1 Aux Bus transferred to a Startup Feed, is I there to sway the candidate to thinking that this condition affects main '
. g~r1~rator trituespJ)r1~~,-ltgoesIlO1.
Distractors: A INCORRECT - see C above B INCORRECT - see C above D INCORRECT - see C above Psychometrics Level of Knowledge I RO HIGH I 10CFR55.41 (b) (x) Source Documentation Source: l8J New Exam Item 0 Previous NRC Exam: 0
. 0 Modified Bank Item 0 Other Exam Bank: 0
.... ~
._.___._ .. jIOJ!.-Il'::.)(am Bank__ __.____ . . . .___
B~Le!~l}ce(st______ ARQ.J20_B-1 __2 G~11 R~I::i~ . . __ . . . . _.. . . . Learning PLOT-5050 Obj 3 Objective: KIA System: Importance: RO / SRO 2.7/2.8 KIA Statement: AA1.04 - Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: Main generator controls REQUIRED MATERIALS:~:__.*f * .N9NE ..... Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 50. A reactor startup and power ascension is in progress on Unit 3. Reactor power is 18%.
The Turbine Generator has been synchronized to the grid and loaded when the following transient occurs:
- The running EHC pump trips.
- The standby EHC pump fails to start.
- The Main Turbine trips on low EHC pressure.
Based on the above conditions, which one of the following is correct regarding pressure control and plant response? A. Six Turbine Bypass valves will open and remain open to control pressure and NO reactor scram is expected to occur. B. Six Turbine Bypass valves open for several minutes to control pressure and then the reactor will scram on high pressure. C. Reactor scrams IMMEDIATELY on high pressure/power and NO Turbine Bypass valves open due to the loss of EHC hydraulics. D. Reactor scrams IMMEDIATELY on the turbine trip and Turbine Bypass valves will open for a short time following the scram.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key
~-"-"- ~-----~~~----- --~ .....
Choice ~ .*.. . . .
~ -~-~---~ ...........-
Basis or Justification Correct: B CORRECT - Scram on turbine trip is bypassed at this power level. 6 BPVs are expected to open for 18% power. BPVs have separate accumulators to allow opening for some time without EHC fluid pressure. When accumulator pressure gets low the BPVs will close and reactor pressure will
~t>~.9ll"lt() rise tQJ~_~ highf>rE!~~~r~ElC:;~.rn setp9L'lL.
Distractors: A INCORRECT - BPVs accumulators will eventually run out due to design valve actuator leakage. C INCORRECT - With bypass valve operation, an immediate reactor scram is not expected at this power level. D INCORRECT - Scram on turbine trip is bypassed at this power level. Psychometrics Level of Knowlegge 1
!liffiPultY Time A""""",ce (minute~ll 10CFR~041 (b)(7)
HIGH I T Source Documentation Source: D New Exam Item ~ Previous NRC Exam (PB 2002) D Modified Bank Item D Other Exam Bank o ILI. Exa!!lJ3ank ReferenC:;E;(~ ARC 206 C-5 Learning PLOT5001 B - 5f Objective: KIA System 295006 SCRAM Importance: RO / SRO 3.7/3.7 KIA Statement AA 1.03 - Ability to operate and/or monitor the following as they apply to SCRAM: R.eactor/turbin~.2r~~su~!.~g.ul~~l'"lgsystem REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 51. Using the figures on the next page, which of the following sets of conditions allow safe operation of a loop of RHR in the LPCI mode at all flow rates?
Torus Level Torus Pressure Torus Temperature A. 11 feet 9 psig 200 deg F B. 12 feet 5 psig 195 deg F
- c. 13 feet 11 psig 220 deg F D. 14 feet 7 psig 180 deg F
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 TORUS LEVEL RHR 320 ~ .............:........ ~ *** ~ H ** ........ ~ . . . . . . . .: . . . . . . . .:. . . . . . . . : . . * * . 300 ...... **.. *.. *;* .... *.. i** .. ***;** ..... *; .... *.. ~*.**** ..~.** ...... ** .. ..
'1 TORUS PRESSURE
- m ........ +.......:....... ........ ....... ........ .......:......:: 60 PSI G AND
~ ~ ~ ~ ~BOVE u: 280 f-: :::: ::::: ::t::::::;:: :::::t::::::~:::::::r:::::::~::::::T: :::: 1
- 260 I= ... :.w ....:..... m;.... r..~, . . ..... *'1* .. *** .... * .. ::130 TO 59.59 PSIG
- w ..... .4 .. ~~,. ~. '" ....... ,. .... " ... {._~~~ * ** -'.,. ~ ..... ~ ......... ~"".I_ *
- _ ** _~ ... __ ~. ~ ~. I~ ~... *
..... 240 il.
- i::::::::i::::::::i::::::::~::::::::~::
..... L......~~:~4:;;;J 20 TO 29.99 PSI G
~ 220 ................................................ ~ ...... "~ ....... :....... .
W ..... ~~ .. ~ .... <-.~ ........ ~ .. ~~".~-~- .... ~~~; 10£ FEET TO 12J9 FEET . 10 TO 19.99 PSI G
"" 160 . G TO 9.89 PSI G
~ ISO ~-:-:-+~~~~-i--\.~~
C) 3 TO s.sa PSI G I- 140 120 100~~--*~--~--~--~--~--~~~-- 7.S0C a,coo 9,OnO 10,000 11,000 12,000 1 PUMP 15,OO,} 18,OOC 18,000 20,000 22,000 24,000 Z ?UMPS RHR LOOP FLOW (GPM) 320 r--. . . . , , . . . TORUS PRESSURE 300 I=m ..:..**....;....*..*: ........:.. ***.. -;-:::;;:i::::::::t:::::::l:::::::1 so PSIG AHD ABOVE r~' -. -; ........ --:.--~~ .... -~ ............":" . ".. ~- . :"...... --.. ~ . . -.. ~ -- ~:"-"~ ~ ~ . . _ .... ~ u: 280 :~: ~:1~~:::: :1:::::: ::~:: :::::t::: ~:::~:: :::~ :t::::~::~:: ::: ~:1 ~ ::~::: j
- 260 . :- ...... :........ 30 TO 59.99 PSrG 2~O Ll.
- E W
220 .,. .. ?.:." ...... _"' ....~.. __ ...... ,. .. t. .. " .. __ ... _,,._._ .. ,, __ t.. __
- f- 200 r:=::7:7=::-t:=+~~~~-.;--=..-.;.;.:,: ... 10 TO 19.98 f'S!G 12.3 FEET TO 14.49 FEET CIl 18()
=>
0::. ISO 0 I- 140
. , 0 TO 2.99 PSIS 120 100 7,500 8,001 9,000 10,000 11,000 , 2,000 1 PUMP 15.000 lG,CCO 18,000 20,000 22.000 24,000 Z PUMPS RHR LOOP FLOW (GPM)
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
----- ......... ~~~ .~ ..-~
Answer -Key
......--~-. .. -.---.~.--
Question # 51 RO Choice Basis or Justification Correct: o CORRECT - This meet the criteria of the RHR NPSH curves for two-pump operation at all flow rates, as shown on T-102, Sheet 3.
. i Distractors: A INCORRECT - This does not meet the criteria of the RHR NPSH curves for '
I two-pump operation at all flow rates, as shown on T-102, Sheet 3.
! Operation is in the unsafe region of the curve when flow is above -20,000
. i....~~.~~ ..... -tgPrl1' Plausible if can~Jg~te plo!si l1c;orrec!ly:
B INCORRECT - This does not meet the criteria of the RHR NPSH curves for two-pump operation at all flow rates, as shown on T-102, Sheet 3. Operation is in the unsafe region of the curve at all flow rates. Plausible if C *~*I' ~r:g~~~~~~t~_~h~~r~;:y~~t-~eet th~ criteria of th~~ RHR NPSH curves for two-pump operation at all flow rates, as shown on T-102, Sheet 3.
, Operation is in the unsafe region of the curve when flow is above -17,000
.J
~~ _~~l~_ ~ gp.f11-l'lausible. if candidate'pIQ~Jnc()~r~E!~~>'..~_~_
Psychometrics Level of Knowledg~~ Dif!iC;lJlty~ __ __Til11~ Allowanc~_ll!lif1ute~) ! RO HIGH
-1 10CFR55.41 (b)(5)
Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 rgJ Modified Bank Item Other Exam Bank: () ILT Exam Bank Reference_(~}: _ T-1 02 ShE3~!?LembeddeciLrlguesti,,-1}) Learning PLOT-2000 Obj 11 C Objective: KIA System: 295030 Low Suppression Pool Water Level Importance: RO I SRO 3.9 I 3.9 KIA Statement: EA2.02 - Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: SuppreSSion pool temperature REQUIRED MATERIALS: None Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 52. Unit 2 was operating at 100% power.
- The jet pump mixer for Jet Pump '11' becomes displaced.
This failure will cause a sudden RISE in: A. Core Plate Flow B. DP on Jet Pump '12' C. 'A' Recirc Loop Drive Flow D. 'B' Recirc Pump Speed
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Choice Basis or Justification Correct: C CORRECT - Jet Pump 11 is in the 'A' Recirc Loop A INCORRECT - Core Plate flow will LOWER due to drop in core flow B INCORRECT - Jet Pump 12 shares a riser with JP 11, so its DP will LOWER D INCORRECT - Failed Jet Pump is not in this loop. "B" Recirc Speed will be unaffected. Psychometrics I Difficulty_____ Time AliowanceJrl1inutesll RO I 10CFR55.41 (b)(2) Source Documentation Source: D New Exam Item Previous NRC Exam: 0 i D Modified Bank Item Other Exam Bank: 0
--- --t ILT Exam Bank Reference(~L__ .QN-100 Learning . PLOT-1550 DBIG Obj 4 Objective:
KIA System: 295001 Partial or Complete Loss of Forced Importance: RO I SRO Core Flow Circulation 3.0 I 3.1 KIA Statement: AA2.04 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Individual jet pump flows REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 53. Unit 2 is experiencing a loss of Instrument Air transient.
Per procedure ON-119 "Loss of Instrument Air", the Backup Instrument Air Compressor 2DK001 will automatically start AND the Backup Air Control Valve (AO-80250D) will automatically open at to supply the
----'-="--
Instrument Air header. A. (1) 80 psig (2) A ONLY B. (1) 90 psig (2) B ONLY C. (1) 80 psig (2) A and B D. (1) 90 psig (2) A and B
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
----~ .... ~-~ .. -~-~-~--- ......._--- ~ ---~~~- ... -
Answer Key Question # 53 RO...._------, ...- -_. Choice Basis or Justification Correct: B CORRECT - Per ON-119 Note on page 3, the Backup Instrument Air Compressor 2DK001 and AO-80250D will open when both the 'A' and 'B' Instrument Air receiver pressures drop to 90 psig, and will supply the "B" header ONLY.
---------' ......~-.----
Distractors: A . INCORRECT - 80 psig is a plausible distractor based on ON-119 i referencing 80 psig as pressure to check instrument air dryer operation. C INCORRECT - 80 psig is a plausible distractor based on ON-119 referencing 80 psig as pressure to check instrument air dryer operation and a block valvl? is_norl11iiilly~lQ~edvv.hj~llPl"l?\lentsfE3~cjing the_'!>.' header. D INCORRECT - 80 psig is a plausible distractor based the fact that a block valve is normally closed which prevents feeding the 'A' header. Psychometrics Level of l<!u:>~edgl? Difficyltr____ . ~- ..... ...Iime AIiOwanCI?(I11Jl1utes) . . . ! RO MEMORY 10CFR55.41 (b)(4) Source Documentation Source: o New Exam Item 0 Previous NRC Exam
~ Modified Bank Item 0 Other Exam Bank ILT Exam Bank
~--""=~------
Referl?l1c;l?(s): ON-119--- --. Learning PLOT-5036-4a Objective: KiA System 295019 Partial or Total Loss of Inst. Air Importance: RO 1 SRO 3.5/3.6 KiA Statement AA2.01 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSSgF INSTRUMENT AI R: .lns!r~m~n!air~y~!~I11..Pressure ~~;;~~:~~~!-::~~L~:J=~~~-~-~. .- _
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 54. Per T-104 "Radioactivity Release Control" Bases, the (1) ventilation system will NOT be restarted if T -200 "Primary Containment Venting" is in progress due to (2) ?
A. (1) PEARL Building (2) Radioactive Samples being handled here B. (1) Radwaste Building (2) Ventilation intake location C. (1) PEARL Building (2) Ventilation intake location D. (1) Radwaste Building (2) Radioactive Samples being handled here
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 54 RO..
--.-.-- -~-~~-
Choice-. ~,,-- .... - --~--.- ..... -. Basis or Justification Correct: CORRECT - T-104 Bases, Step RR-6 directs NOT restarting Radwaste Building ventilation if T-200 is in progress due to several of the vent paths having the potential of discharging radioactive materials into the vicinity of the Radvvast~J~~ilding ventilation intC!~e--"~_ Distracters: A INCORRECT - PEARL Building ventilation is restarted, specifically because of radioactive sample handling. C INCORRECT - PEARL Building ventilation is restarted, because of radioactive sample handling.
--.--~-- ..... -- ---~--~ ..... - ----~~- ~
o INCORRECT - T-104 Bases, Step RR-6 directs NOT restarting Radwaste Building ventilation if T -200 is in progress due to several of the vent paths having the potential of discharging radioactive materials into the vicinity of the Radwas1~~LJiLdl'lg ventilation ir:'tak~~gT because of s~l11pl~r:'g: Psychometrics Level of Knowled.ge __ . Diff~.ultL __ ._____ .__ _Tlrn~ AIIO\tVi3rl<:;~j!!1inutesJ~_t~__ ___ RO _ ___ MEMORY , 10CFR55A1(b)(10) Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 J8J ILT Exam Bank ------ - _g~f~!~n<:;e(lS)~___ _T -1Qi~~lS~S_ Learning PLOT-1560-9 Objective: KIA System: 295038 High Off-site Release Rate Importance: RO 1 SRO 3.7/4.7 KIA Statement: 2.4.6 - Emergency Procedures/Plan: Knowledge of EOP mitigation strategies.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 55. Unit 2 was in Mode 3, with Shutdown Cooling in service, preparing for a refueling outage when the station experienced a loss of all AC power (Station Blackout).
The following conditions exist on Unit 2: Instrument Reading
- Fuel Zone Range on LR-2-2-3-110B (PaneI20C003-02) + 10 inches
- Wide Range on LR-2-2-3-110A (Panel 20C004C) + 20 inches
- Narrow Range on U-2-6-094B (Panel 20C005) 0 inches
- Wide Range on U-2-2-3-85A (Panel 20C005) -15 inches Using the SE-11 Attachment C "Instrument List" ATTACHED SEPERATELY, determine which of the following statements is true.
Actual reactor water level: A. cannot be determined
- 8. is +20 inches C. is 0 inches D. is -15 inches
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 55 RO ------"- Choice Basis or Justification Correct: D CORRECT - U-2-2-3-85A is a post-accident monitoring instrument and per SE-11 Attachment C "Instrument List", is powered by DC following a loss of _()ffsit~()vy~r al1<tV'!!ILf()l1tinue to accur(31f3JYlD<:Jl~(3tEueactorwater level. Distractors: INCORRECT - RPV level CAN be determined. U-2-2-3-85A is a post-I accident monitoring instrument and per SE-11 Attachment C "Instrument ______ " " ' __ ~;;~~r~~~;d7c'!:r~c:;~I~:~:~:~~~~ of offsite power and w"1 continue to t-8
+- _
INCORRECT - LR-2-2-3-110A is NOT a post-accident monitoring _~:1;"C<~~~~:~:;:-~/O~:i~~~~;~ "Instrument list", is not powered
. C . INCORRECT - While Narrow Range U-2-6-094B is powered by DC I following a loss of offsite power, its range is 0 inches to +60 inches and is I indicating at minimum value due to actual reactor water level being < 0 inches.
Psychometrics I Level of Knowledg_~ .. _ - - - - - _"PJfficll~Y__"_ ""~ _Tll11e AltQ.~anceJl11inll~sLL RO HIGH I 10CFR55.41 (b)(7) Source Documentation Source: [gJ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank
"-~"'=-'--
Referer1~f3C~ t " " SE-11 Attachment C ---"""-" Learning PLOT 1556 Obj 2.b Objective: KIA System: 295021 Loss of Shutdown Cooling Importance: RO / SRO 3.7/3.9 KIA Statement: 2.4.3 - Emergency Procedures / Plan: Ability to identify post-accident instrumentation. REQUIRED MATERIALS: SE-11 Attachment C Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 56. Due to a loss of DC power, the Unit 2 RCIC system is being operated locally, using SE-13.1-2, "RCIC Manual Operations on Loss of 125/250VDC Bus 2DA-W-A ".
For these conditions, speed control of the RCIC turbine is obtained by local operation of: A. MO-2-13-16, Steam Isol Valve B. HO-2-13-4495, Inlet Control Valve C. MO-2-13C-4487, RCIC Turbine Trip Throttle Valve D. the EGM Control Box Bias Speed Setting Potentiometer
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 56 RO
~'--'~-'~"',--'"
l-:- Choice ._- ... _.__Basis.. or Justification CORRECT - as described in first NOTE of SE13.1. correct __ Distractors: A INCORRECT - Plausible because this component is in the RCIC turbine steam flow path. INCORRECT - Plausible because this component is in the RCIC turbine steam flow path. INCORRECT - Plausible because the potentiometer exists and is used for routine overspeed testing, but is not used by SE 13.
.- - - - - . - . . - ... --~---~ ..........- . ... ~---~.~
Psychometries =L~e__Y. . _e_~Q_~_,,~_. _ ~_QR_~_I~_-Qg_,,~_-_l_... __-_..._.~~__D_._iff_iC_.U_lty_,-_.._-.=~_"._~_T_lm___e_A_JI_OW_,.-_~-n_c_e_(_m_in_u_t_es_)----,-1_1_0_C_F_R_:_50_.4_1->.(......b),-,-(7-<)-..JI Ir Source Documentation Source: ~ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank . ReferenceJsJ SE-13.1-2 Learning PLOT5013-9h Objective: KIA System: 295004 Partial or Complete Loss o f DC Pwr Importance: RO 1SRO
. . . . . . ._-_. . . . - .-.,.--~-- . .
4.4/4.0 KIA Statement: 2.1.30 - Conduct of Operations: Ability to locate and operate components, including local controls. REQUIRED MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 57. Unit 2 is operating at 100% power with the electric plant in a normal lineup when the SU-25 breaker trips.
Which one of the following describes the effect of the SU-25 breaker trip on Unit 2? A fast transfer to their alternate sources will occur for 4 kV busses (1) and a Group II half isolation will occur. A. (1 ) E12 and E32 (2) Inboard B. (1 ) E12 and E32. (2) Outboard A. (1 ) E22 and E42 (2) Inboard B. (1 ) E22 and E42. (2) Outboard
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 57 RO Choice Sasis or Justification Correct: A CORRECT -In a 'normal' lineup, E12 and E32 are powered #2SU bus, which is energized via the SU-25 breaker. On trip of SU-25, these buses will fast-transfer. An Inboard half-isolation occurs due to loss of 20Y033, which loses power when the E12 bus de-energizes momentarily. Since
.~~!!!aJ_e of!.-sit~po~~lJ~_Ci'.t:~ilat:>le,J::PGs do NOT start.
Distractors: S INCORRECT - Plausible because the candidate could believe an outboard isolation occurs. An Inboard half-isolation occurs due to loss of 20Y033, wbigbJ.Q§espow..er _!Vhen the E 12 bus _g~.:e-"1.ergiZ:Els momentarily INCORRECT - Plausible because the candidate could believe that E22 and E42 could be feed from #2 SU bus. They are feed normally from either 343 SU or#3SU.. _--- ... _ .._ o INCORRECT - Plausible because the candidate could believe an outboard isolation occurs. An Inboard half-isolation occurs due to loss of 20Y033,
. which loses power when the E12 bus de-energizes momentarily. Also, I Plausible because the candidate could believe that E22 and E42 could be feed from #2 SU bus. Th_ElY_are feed nOr'!1CiIJyfrom either 343 SU or #3SU.
Psychometrics '. . Level of Knowledg~ +- RO I HIGH 10CFR55.41 (b)(7) Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 D Modified Sank Item D Other Exam Bank: 0 ILT Exam Bank Reference(s): SO §~-,I:~.__<:3E-~g_ Learning PLOT5054-7b Objective: I KIA System: 295003 Partial or Complete Loss of AC Importance: RO I SRO 3.5/3.7 KIA Statement: AA2.04 - Ability to determine andlor interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: System lineups REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 58. T-117, "Level/Power Control" requires level to be restored and maintained above -195 inches indicated.
Which one of the following identifies the basis for this value of indicated level? Maintaining level above this value maintains peak clad temperatures below (1) degrees F by the (2) method of ACC. A. (1) 1500 (2) Steam Cooling B. (1) 1500 (2) Spray Cooling C. (1) 1800 (2) Steam Cooling D. (1) 1800 (2) Spray Cooling
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 58 RO Choice ~~r- ~ .....--------'-~-~ -- ~~- Basis or Justification Correct: A CORRECT - as described in T -117 basis, ACC is assured by steam cooling down to -195 inches RPV level WITH INJECTION with clad temp not to exceed 1990 dE:!g.r~~s. F. Distractors: B INCORRECT - Conditions for Spray Cooling can NOT be relied upon in T
- 117 because Core Spray design bases assumes the reactor is shutdown.
Plausible because Spray Cooling can be used to establish ACC under I other conditions, and 1500 degrees F is the Steam Cooling clad temp i
--.-t--~-
maintained
.... ---~-
in. -.-----. T-117. .. --~-- . . . --~--- -- - ...
.~.--- -~----
C I* INCORRECT -1800 degrees F is the Zero Injection Clad Temp for Steam Cooling used in T -111. Plausible because this method can used to t3sJC3blisb.~CC-,.bLJ!f\JQ*LcJlj!irlg~JWS conditions and T -117: D INCORRECT - Conditions for Spray Cooling can NOT be relied upon in T 117 because Core Spray design bases assumes the reactor is shutdown. Plausible because Spray Cooling can be used to establish ACC under
- other conditions. Plausible because Spray Cooling can be used to establish II. ACC under other conditions, an . . d as..previousl Ydescribed 1800 degrees F can be used to establish ACe durirlg T-111 conditions. _
Psychometrics -- L~\,fel Qf Kn.QV!I~<!g~. L Diffi<2yl!y__ ~ _~_.ILlTle Allowance (rrrinljtes)_ RO MEMORY 10CFR55.41 (b)(1 0) Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0
~---~t~ -."'-.-~
ILT Exam Bank Reference(s): T-117 T-BAS -~---- .. ~--- .....------ -~--.~-- ...... - Learning PLOT 2117DBIG - 6 Objective: KIA System: 295031 Reactor Low Water Level . -tmportance RO/SRO 4.0/4.3 KIA Statement: EK3.04 - Knowledge of the reasons for the following responses as they apply to _____ ~~___R.. EAgIQ~ LQVIJ W~TER LEyEL~ S~arrr cooling___ REQUI~EI? MATERI~I..S: ____ rNONE ___ _ _~_ _~ Notes and Comments: - --
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 59. Unit 2 was operating at 100% power when an inadvertent Grp III Grp III PCIS Isolation occurred due to multiple instrument failures.
Assuming NO SCRAM ACTIONS and NO OTHER OPERATOR ACTIONS for 30 minutes, which of the following, if any, is (are) available to initiate a controlled RPV depressurization per T-1 01 "RPV Control"? A. Manual operation of SRVs B. Manual operation of Bypass Valves C. Both SRVs and Bypass Valves D. Neither SRVs or Bypass Valves
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
~--- ........- - - - - - - - .----.- ..... -----~-.... -- - - - - - - - - - _ .
Answer Key Question # 59 RO Choice Basis or Justification Correct: A CORRECT - On a Group 3 isolation eventually (approx 20 minutes) the Inboard MSIVs will close due to their accumulators bleeding down, rendering the Bypass Valves unable to support pressure control. However, the (ADS) SRVs can be operated manually due to the pneumatic supply
-------+_:.:
accumulators. Distractors: B INCORRECT - On a Group 3 isolation eventually (approx 20 minutes) the Inboard MSIVs will close due to their accumulators bleeding down. Plausible if candidate does not know that Inboard MSIVs will close due to _ lo~~ ofJn~!r.':lmerlLNitr<~g~n .______~.______ C INCORRECT - On a Group 3 isolation eventually (approx 20 minutes) the Inboard MSIVs will close due to their accumulators bleeding down. However, the (ADS) SRVs can be operated manually due to the pneumatic supply accumulators. Plausible if candidate does not know that Inboard MSIVs will close due to_!oss of Instrul1}~111Nitr:.()'@J1. D INCORRECT - On a Group 3 isolation eventually (approx 20 minutes) the Inboard MSIVs will close due to their accumulators bleeding down. However, the (ADS) SRVs can be operated manually due to the pneumatic supply accumulators. Plausible if candidate does not know ADS SRV ac(;umul~t()r_s_l!2PlY'vVm~llpp_o_rt_~DS.SRV_()p~~~tion. Psychometrics Levelg! ~I}owlf?c!gi=l QifficJ.llty___ Time Allowance (minutes) RO
--I HIGH
-ti ------ - -
***- - . - . . 0 - ' 10CFR55.41(b)(x)
Source Documentation Source: C8J New Exam Item Previous NRC Exam: () o Modified Bank Item Other Exam Bank: 0 __ _ ______ I___D ILTE_~m Bank _________ .R~ferenCe(!S):.... . . . . __+I--1-01. Bc:l!S~!S__'_§Jep_RCLE_::4_ Learning i PLOT-5001A-5w Objective: KIA System: 1295020 Inadvertent Cont. Isolation Importance: RO I SRO 3.7/3.9 KIA Statement: AK1.01 - Knowledge of the operational implications of the following concepts as they ___?J>PlyJoINt\QYi::RTENT CONTAINMEf',IIJ§OLATION : Loss of normal heat sink REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 60. Which of the following are the correct Unit 2 and Unit 3 High Reactor Building Main Steam Tunnel Temperature setpoints for the Group 1 Isolation?
Unit 2 A. 200°F
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Question # 60 RO Choice .. Basis or Justification
~-~ -~~~~ ,.""'-~~~-~~--. -~-- -------~---~-- ..
Correct: CORRECT - Per Tech Spec U2(3) Table 3.3.6.1-1 the High Reactor Building Main Steam Tunnel Temperature setpoints for the Group 1 Isolation are 230°F for Unit 2 and 200°F for Unit 3. Distractors: INCORRECT - Unit 2 setpoint temperature is too low. Should be 230°F.
- INCORRECT - The answers are reversed for both units.
D INCORRECT - The Unit 3 setpoint temperature is too high. Psychometrics ... L~\leLof Kn()'!"Vl~cl.ge I MEMORY Source Documentation Source: kSI New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank --.-.~ ~~-.--.. -- --_._. Reference(s): Te_c.b§3'p~c LJgpIJables 3.3.6.1-1 .. Learning PLOT-5007G obj 11 Objective: KIA System: I 295032 High Secondary Containment Area I Importance: RO/SRO Temperature 3.6/3.8 KIA Statement: EK2.04 - Knowledge of the interrelations between HIGH SECONDARY CONTAINMENT AREA TEMPERATURE and the following: PCIS/NSSSS
~l-REQUIRED MATERIALS: NONE Notes and Comments: This Question verifies candidate knowledge of a difference In
_____ se!.Q.()ints between Units l.~ncl ~ Jor same function.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 61. Unit 2 is at 8% power with reactor startup in progress.
- RPV pressure is at 165 psig for HPCI/RCIC testing.
- 2B CRD pump is out of service (unavailable)
- The 2A CRD pump trips an electrical fault.
- Charging header pressure is 100 psig and dropping.
Why does ON-107, "Loss of CRD Regulating Function", direct a Reactor scram if accumulator trouble alarms occur on withdrawn rods in this condition? A. To ensure all control rods are fully inserted before overheating can affect the mechanism seals and impact scram times.
- 8. To ensure rods can be inserted since normal rod insert and withdrawal functions are inoperable.
C. To ensure control rods can be inserted before the HCU accumulators depressurize and cannot complete the scram. D. In anticipation of tripping both Reactor Recirculation Pumps due to loss of seal cooling.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 61 RO Choice Basis or Justification Correct: CORRECT - per ON-107 Bases Distractors: A INCORRECT - This is NOT the bases for scram at this point. Plausible because cooling is lost and high CRD Temps can have impact on scram time. B INCORRECT - This is NOT the bases for scram at this point. Plausible because this is a concern for CR functionality, but NOT scram basis. D i INCORRECT - This is NOT the bases for scram at this point. Plausible because seal cooling is lost and seals will heat up, but scram will not
~igQifi~~!1t~rl1i!igat~Jhis_~ffegL_
Psychometrics Level of Kn2wle(jge _~ mm_m_n __ RLffic::ultY_____m_[ Tirn~j\"owance (minutest im RO MEMORY I : 10CFR55.41 (b)(6) Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank Referenc~(s): ON-107 Learning PLOT1550 - 4 Objective: KIA System: 295022 Loss of CRD Pumps Importance: RO I SRO 3.7/3.9 KIA Statement: AK3.01 - Knowledge of the reasons for the following responses as they apply to LOSS OF CRD PUMPS: Reactor SCRAM REQUIRED MATERIALS:------_._. NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 62. Unit 2 is at 40% power during a normal Plant Startup.
OT-106, "Condenser Low Vacuum" procedure is in progress due to a vacuum leak. IF condenser vacuum reaches (1) , THEN GP-4, "Manual Reactor Scram" will be performed in anticipation of the RPS automatic Reactor Scram at (2) A. (1) 25.4 inches hg (2) 20.0 inches hg B. (1) 24.0 inches hg (2) 23.0 inches hg C. (1) 23.0 inches hg (2) 20.0 inches hg D. (1) 25.4 inches hg (2) 23.0 inches hg
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 62 RO Choice .......... ---~~---~-~- Basis or Justification Correct: B CORRECT - OT-106 directs a manual scram at 24.0 inches hg. The RPS scram setpoint is 23.0 inches hg. Oistracters: A INCORRECT - Plausible because 25.4 inches hg is the "CONO LO VAC" alarm (206 0-2) setpoint. The Main Turbine and RFP Turbine Trip setpoint is 20 inches C INCORRECT -Plausible because the RPS Scram Setpoint is 23 inches hg.
- The Main Turbine and RFP Turbine Trip set point is 20 inches hg.
~~-.~~ .. ,. - -~~~ . . ~-~~~ ......
I 0 ; INCORRECT - Plausible because OT-106 directs a manual scram at 25.4 inches hg if power is above BPV capability and unable to maintain load ____~~_~ __ greater than 300 MWe. The RP!?§cr§lI11'§~1p2JQtis 23 inches hg. Psychometrics Le\lel~2f Kn0'llfl~<!g~ ~-f-~'
...9iffi(:;lJl!y_ _ _.._ Time Allowance (minutes)
- ~-~. ~. .
RO MEMORY 10CFR55.4 1(b)(1 0) Source Documentation Source: I [3J New Exam Item D Previous NRC Exam: 0 I D Modified Bank Item D Other Exam Bank: 0 Refecel1~~----=~~;~Ho;6:!"e~-;-v;;;;,-;;;-* Learning I PLOT 1540 OBIG Obj 2 Objective: I j - KIA System:
- 295002 Loss of Main Condenser Vac Importance: RO / SRO 3.4 13.5 KIA Statement: M 1.03 - Ability to operate and/or monitor the following as they apply to LOSS OF MAIN CONOENSERVACUUM: RPS
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 63. Unit 2 is in MODE 2 with a heat up in progress.
- Vessel level is being maintained by the Control Rod Drive Hydraulic System and the Reactor Water Cleanup System.
- The URO observes RPV level slowly lowering.
- The crew enters OT-100, "Reactor Low Level".
For the above conditions, which of the following actions is required? Throttle: A. open CV-2-12-55 "RWCU Dump Flow" B. closed CV-2-12-55 "RWCU Dump Flow" C. open MO-2-12-68 "RWCU Outlet Valve" D. closed MO-2-12-68 'HWCU Outlet Valve"
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 63 RO Choice Basis or Justification Correct: B CORRECT - Since RPV level is lowering, the crew has 2 choices for recovering level, either raising CRD system flow (not a listed option), or reduce RWCU dump flow (blowdown flow). The way to reduce dump flow is to throttle closed CV-2-2-55. Distracters: A INCORRECT - The way to reduce dump flow is to throttle closed CV-2-2 55, NOT open the valve. This would make the RPV level change worse. C INCORRECT - Throttling open the MO-2-12-68, RWCU return to the RPV, while the system is in the dump (blowdown) mode of operation will not change the dump flow rate which must be reduced in order to stop the lowering RPV level. . __________ . . INCORRECT - Plausible since the candidate has to realize that the MO-2 12-68, RWCU return to the RPV, is normally fully closed while in the dump (blowdown) mode of operation. Dump flow rate must be reduced in order to sJQPtb~IQ~eringJ~PV-.lev§!l____.. . Psychometrics Level of Knowledge HIGH I Difficulty --i 2LmeA~lwanceJminutes) i OCFR5~~ 1 1(b)(1 0) Source Documentation Source: ['8J New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 o ILLExam Bank Referen~i§)t _______LQ'-'-1QQ,_§9 12.1.A-2_ Learning
- PLOT 5012 Obj ge Objective:
KIA System: 295009 Low Reactor Water Level Importance: RO / SRO 2.9/2.9 KIA Statement: AA2.03 - Ability to determine andlor interpret the following as they apply to LOW REACTOR WATER LEVEL: Reactor water cleanup blowdown rate REQUIRED MATERIALS: None
---~---------.---
Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 64. The following drywell conditions exist on Unit 2 following a small-break LOCA:
- A Torus-to-Drywell Vacuum Breaker is stuck open
- Drywell and Torus Sprays cannot be placed in service
- Drywell pressure is 10 psig and rising 1 psig per minute
- Torus pressure is 10 psig and rising 1 psig per minute
- T crus Level is 15 feet and steady Based on the above conditions, and using the curve provided below, an Emergency Blowdown is ___.
A. required IMMEDIATELY B. required in 14 minutes C. required in 10 minutes D. NOT required if the 2 inch Drywell Vent can be established. CURVE PC/P-2
;JRI::SSUHE UPPF-?f:S$ ON PRFS U I ~.tl I 28 UJ 20
=
12~~~**~-r---+---+-~~--~--~r---H~ 1C
~I 1L 1 <1 1B
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 64 RO Choice Basis or Justification T******~*~+----~*-~-~*~ Correct: B CORRECT - Torus pressure will reach 24 psig in 14 minutes - this will plot on the UNSAFE side of PC/P-2 curve, requiring a blowdown. Candidate must calculate and apply the trend to the curve and know that the area __~_... +~___FA----"B=-O_=__V:..E the curve is UNSAF~ and rE!9!:l!r:.~...§IQI9wdown. Distracters: A INCORRECT -It will take Torus pressure 14 minutes to reach 24 psig which will plot on the UNSAFE side of PC/P-2 curve. Presently, Torus is on SAFE side of the curve. C INCORRECT -10 minutes is too soon. Torus pressure will reach 24 psig I in 14 minutes -10 minutes will still plot on the SAFE side of PC/P-2 curve, iN9TJ:.equiril"lg~<!..QIQ'.II.Icj.9wn* I __ n n n . _ u u _ o I INCORRECT - plausible if the Candidate does not recognize that the 2 i inch drywell vent path will be isolated on a Group 3 signal of drywell l_PiE!_~l:iure ~ ~sig:.._ _ n ___ ~______ n n ___ ** Psychometrics ... Level of Kn()""I~_c!9..E!.~+- __ ___ Qifficulty___~_LTime Allowance(l1"1tnlJte~) .1 RO HIGH I 10CFR55.41 (b)(9) Source Documentation Source: k8J New Exam Item D Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: 0 OJ.LI Exam Bank~_~ _.___ ~~._ ____._._ Re~rE!!!~~{§): . I-*1J)~ BCI~!?! Curye PS/P-JJE!rl1 be c:ic:iE!c:i.i!lJ hi s..9ljesti()nJ Learning PLOT2102 DBIG - Obj 6 Objective:
---.~----+----
KIA System: 295010- High Drywell Pressure Importance: RO I SRO 4.2/4.2 KIA Statement: 2.4.47 .. Emergency Procedures I Plan: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. REQUIRED MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 65. A scram occurred on Unit 2.
The following conditions currently exist:
- RPV pressure is 940 psig.
- The 2A CRD Pump is in service.
- The scram discharge volume is drained.
- 211 E-2, "CRD ACCUMULATOR LO PRESS/HI LEVEL" alarm is annunciating for 3 control rods that are at position 06.
- An EO reports that depressing the affected accumulator local Accumulator Trouble pushbutton on Panel 2AC078 results in the light remaining energized.
Which one of the following is correct regarding the ability to insert the 3 control rods using Individual Scram Test switches? Control rods will - - - - - - , CRD ACCUMULATOR LO PRESS/HI LEVEL alarm is due to A. insert, low gas pressure on accumulators. B. insert, high water level on accumulators. C. NOT insert, low gas pressure on accumulators. D. NOT insert, high water level on accumulators.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Choice Basis or Justification A CORRECT - Control rods will insert due to both reactor pressure.::: 940 psig and the CRD Charging Water available. Alarm response for 211 E-2 provides direction for local operator checks. Locally depressing the affected accumulator local Accumulator Trouble pushbutton and the light
. remaining energized is indication of low HCU accumulator pressure.
; System knowledge plus knowledge of the NOTES in the ARC and T-216
~proc(lre_r~qlJir.~d toJ)~ir'ltC3.9!"atedJ!1 ()i(j~Lto.cms'N~r Jh§l question.
Distractors: B INCORRECT - Rods will insert, but alarm is not caused by water in accumulator - Plausible if candidate does not know how local accumulator alarm pushbutton indication works (depress pushbutton, if light goes out, ____;-VI.Iater i~~!1_~!_~J!g!ljremains iliumir'lCl1edL9(ls iS~lJ~)_ C 'INCORRECT - Rods WILL insert - see discussion above. Plausible if candidate does not know that 940 psig reactor pressure and lor CRD ____ L_~hargil1..9water ()r~~_sure i~Cl.f:~eR.t~pJ~lqr: ~Ql1trQlrod il1~ertion. o INCORRECT - Rods WILL insert - see discussion above. Plausible if candidate does not know how local accumulator alarm pushbutton indication works. Psychometrics . _L-evel of KnoVl.'ledge_ Difficulty_ . - Time Allowance (mLnl:l~e_s) RO HIGH 10CFR55.41 (b)(6) Source Documentation Source: IZI New Exam Item 0 Previous NRC Exam: () o Modified Bank Item 0 Other Exam Bank: 0
.__ . __._. ._ ...____ ~--DJI..Tl:~~rn. Bal1k_O_______ ____ _ ~:::7;:-"-(sL --I !~b~~2~O::~~~
1 E-2 Objective: KIA System: Importance: RO I SRO 4.2/4.0 KIA Statement: 2.4.50 - Emergency Procedures I Plan: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 66. According to OP-AA-101-111, "Roles and Responsibilities of On-Shift Personnel," which one of the following is an activity that a Licensed Reactor Operator will perform?
A. Authorize fire protection impairment permits. B. Coordinate plant activities with the Load Dispatcher. C. Maintain oversight during transient conditions. D. Review non-licensed operator rounds each shift.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key
~.~------- ... ~-----~---------
Question # 66 RO Choice Basis or Justification Correct: B CORRECT - Coordinate plant activities with the Load Dispatcher is an RO activity per OP-AA-101-111, "Roles and Responsibilities of On-Shift A~tf:~~~~~:CT~~-Thi;i;an SRO duty, specifically~' duty f~;the ~Field ..
. I Supervisor per OP-AA-1 01-111, "Roles and Responsibilities of On-Shift
. !.E.~r§()nnel. ...
C I INCORRECT - This is a primary duty of an STAISRO per OP-AA-101-111,
* "Roles and Responsibilities of On-Shift Personnel. .
NCORRECT - This is Field Supervisor duty per OP-AA-1 01-111, "Roles Responsibilities of On-Shift Personnel. Psychometrics Source Documentation Source: [g! New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0
. _._._.+---",=,~~.I~.L~T._~.Exam Bank 0. __~_~ . . ..._~
Referen9~(s):~~_~~ OP-AA-101-111 Learning PLOT1529 Obj 1d Objective: KIA System: G 2.1 - Conduct of Operations Importance: RO I SRO 3.4/4.1 KIA Statement: 2.1.8- Ability to coordinate personnel activities outside the control room.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 67. According to HU-AA-104-101, "Procedure Use and Adherence", when a conflict arises between a standard procedure and a site-specific procedure, which procedure prevails?
A. The standard procedure prevails. B. The site-specific procedure prevails. C. The standard procedure prevails except when the site-specific procedure directs actions that ensure compliance with regulatory requirements. D. The site-specific procedure prevails except when the standard procedure directs actions that ensure compliance with regulatory requirements.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key ----- ---~-------- Question # 67 RO Choice Basis or Justification Correct: CORRECT - as stated in HU-AA-104-101, "Whenever a conflict arises between a standard procedure and a site-specific procedure, then the I standard procedure shall prevail except when the site-specific procedure _~ _________________~_______+ __ __ __ actions that ensure_(;.Q_I!!~Q<:~ Vlt'it.r!!~.9_~latorY_reqLJifements". d'l_r_ects Distracters: A INCORRECT - plausible since the exception makes the rule. B INCORRECT - plausible if the candidate believes site-specific procedure will always overrule. D INCORRECT - plausible if the candidate believes site-speci'fic procedure typically overrule. Psychometrics Level.9f KnOWI_E!Qg~ __ ~_____ DifficLJI~y_ Time Allowance (l!!i.!.1lJt~s) j RO
~~.--
MEMORY
- 3.0 3 10CFR55.41 (b)(1 0)
Source Documentation Source: D New Exam Item ~ Previous NRC Exam: (PB 2007) D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank
~!=""--
~;;;~;~JS)i~'r;~:~OL- . ------- Objective: I
-----~ ---------
KIA System: G 2.1 Conduct of Operations Importance: RO I SRO 4.6/4.6 KIA Statement: G2.1.20 - Ability to interpret and execute procedure steps.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 68. Unit 2 is operating at 100% power when the following events occur (all times are in seconds):
- T=O - REACTOR HI-LO WATER LEVEL (210 H-2) alarms
- T=5 - URO attempts manual control of reactor water level
- T=15 - REACTOR WATER HI LEVEL TRIP (206 C-1) alarms
- A RFPT TRIP (201 G-4) alarms
- B RFPT TRIP (201 H-4) alarms
- C RFPT TRIP (201 J-4) alarms
- Reactor level indicates +48 inches
- Reactor pressure is 1028 psjg
- Reactor power is 100%
- T=20 - Reactor level indicates -5 inches
- Reactor pressure is 1028 psig
- Reactor power is 100%
What actions are required for these conditions? A. Perform GP-4 "Manual Reactor Scram". B. Trip the Main Turbine and enter T-100 "Scram". C. Scram the Reactor and enter T-100 "Scram". D. Scram the Reactor and enter T-101 "RPV Control",
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 68 RO Choice Basis or Justification Correct: D CORRECT - The given conditions indicate the main turbine did not trip on high reactor level as expected (which would have caused a reactor scram). Since the feedwater pumps tripped and RPV level has lowered below the scram setpoint of +1 inch, an ATWS condition has occurred. This is an entry condition for T-1 01: "scram condition with power above 4% or unknown".. . .
-------.-~-- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --- ----
Distracters: A INCORRECT - The prerequisite for GP-4 states "plant conditions require a I manual scram and sufficient time is available to perform pre-scram actions." There is insufficient time to perform GP-4 under these conditions. In addition, since a scram should have occurred, the operator is required to _______t--___ +f!1~J1ually scram the reactor (pl~c;~Jhe m~c:j~_syvjtc;I1J~Sl1u!clown}. B ! INCORRECT - This would rely on the Rector Protection System to scram the reactor, which violates the "Reactivity Management" Operations Fundamental (do not rely on the reactor protection system to protect the reactor during reactivity events). Since a scram should have occurred, the operator must manually scram the reactor (place the mode switch in shutdown). Plausible since the main turbine should have tripped on a high reactor water level. C I INCORRECT - Plausible since OT-110 "RPV High Level" directs entering T -100 if a scram condition occurs. However, a T -101 entry condition exists since the reactor did not automatically scram as expected. This overrides OT-110 direction. Ps~chometrics _Level of Knowledge __ ,--- HIGH ___ ~i~~~I!Y-----+TirneAIIO\N"~~{rl1ir1ute,,l 10CFR5:~ 1(b)(10) t Source Documentation Source: D New Exam Item r8J Previous NRC Exam: (PB 2011) D Modified Bank Item D Other Exam Bank: 0 ___________Dl~l"_Exam Bank_______________________________ Reference(s):
- --- -- - --------1--
ARC-206 C-1; OT-11 0; GP-4; T-101 Learning PLOT-1529-2 Objective: j Importance:
- --------------------- ----~-------------------- ----------
KIA System: G 2.2 Equipment Control RO 1 SRO 4.2/4.4 KIA Statement: G2.2.44 - Ability to interpret control room indications to verify the status and operation of a s'ystem, and understand how operator actiol"1~_and <:lir~.c;!L,,-es a...ff~.c;tp~~nt anci system conditions. ~~~;~~:~:~~~s: =--} NON~~_____~--- __ _ _____ _ _ _
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 69. GP-26, "Coordination of HCU, CRB, CRD, DBG, and PIP Work During a Refueling Outage" limits the number of HCUs that can be blocked.
This is because high Cooling Water pressure may result in Control Rods - - - - - A. Drifting IN ONLY B. Drifting OUT ONLY C. Drifting IN OR OUT D. Scramming.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
----~-- ..........
Answer Key
------- ~---------- .... __._._._-_ ..... _
Question # 69 RO Choice _ . - - _ ............ Basis or Justification Correct: C CORRECT Per guidance in GP-26 based on SEN 264, this limit is based on Industry Experience which has shown that excessive cooling water pressure can cause rods to drift either IN or OUT. Distractors: A INCORRECT - plausible as it is partially correct. B ' INCORRECT - plausible as it is partianycor-reCt~~-*- D ~ INCORRECT - plausible if the candidate does not know that OPEX has demonstrated this phenomenon. Psychometrics Level of Knowledge MEMORY Source Documentation Source: IZI New Exam Item D Previous NRC Exam 0 D Modified Bank Item D Other Exam Bank
.0 ILT Exam E3~nk Reference(~): GP-2§.,~Et'!:_.2_f)1_'_~npl~!1~~ 9_o_l'ltroLRQgWithdrawals While Shutdown Learning PLOT 5003A Obj 9a Objective:
KIA System G 2.2 Equipment Control RO/SRO 2.6/3.9 KIA Statement 2.2.18 Knowledge of the process for managing maintenance activities during _.. . shutdoVlln()perations,_§u~_b__C!~risk asse§~'!Iellt~,_l,'\/..9J:k prioritization~~tc~ ~~t~~::~~::::~LS: T~~:!lates to Industry oPg, influding Exelon Fleet OPEX.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 70. Unit 2 is operating at 100% power when a steam leak occurs in the Reactor Building.
- The Reactor Building exhaust duct radiation monitors reach the PCIS Group III setpoint.
- All systems operate as designed EXCEPT that both SBGT filter inlet dampers fail to open.
Which one of the following would result from this event? (ASSUME NO 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 Unit 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 License NRC Examination April 2013 Answer Question # 70 RO ~~.~--~--~~-,~ .. ~~-~ Choice ~~~~--~~. Basis or Justification Correct: - The Group III PCIS isolation will trip and isolate Reactor ventilation. The failed filter inlet dampers will prevent SBGT from Reactor Building negative pressure. This will result in an nlTl'I,r"",'" and unfiltered ground-level release. L .. Distractors: A INCORRECT -SBGT would not be exhausting Reactor Building air to the Main Stack. INCORRECT -Reactor Building ventilation dampers close on a PCIS Group
, III isolation and isolate the Reactor Building from the Vent Stack.
D INCORRECT -The release would not be through a monitored path. Psychometrics
.. Level of Knowledge. j _~_DifficuIJY.. ~~~ __ __~~ Time Allowance (minutes)~ I~ --- - ----
RO HIGH i Source Documentation Source: o New Exam Item ~ Previous NRC Exam (PB 2008) o Modified Bank Item 0 Other Exam Bank ILT Exam Bank .... ,..... _ - - - - - - - - - - - - - - - - - Reference~(sL ___J-103 Ba§)es (st§lP§QG-2)_ Learning PLOT-5009A-6b Objective: KIA System 2.3 Radiation Control .. ~~rl~portance: ROfSRO ____ L ____________ _ 3.4 f 3.8 KIA Statement G2.3.14 - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or
§l'!l~rg~l1~y conditions or activjties. . ~~~____ ~ ~ ..
REQUIRED- MATERIALS:
- - - ---- - ----- - - - _., . _. NONE
.. ...... ...................... ~------- - - ---- -- - ------ - - - - - - .
Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 71. Which one of the following requires entry into ON-124 "Fuel Floor and Fuel Handling Problems"?
A. New Fuel Assembly is dropped with no apparent damage. B. Refueling Floor Vent Exhaust Hi Radiation (218 A-1) alarms. C. An irradiated LPRM Detector is dropped in the ISFSI Cask Handling Area. D. Count Rate doubles as the fifth (5 th ) fuel assembly is loaded near a WRNM.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 71 RO Choice Basis or Justification Correct: A CORRECT - ON-124 requires entry and action for any dropped new fuel assembly. Distracters: B INCORRECT -Although this condition obviously requires action, it is not an entry condition into ON-124. C INCORRECT -ON-124 entry is required for a fuel assembly or single fuel
. rod dropped or damaged, but not for an LPRM detector.
D i INCORRECT -ON-124 entry would only be required if the count rate had
! doubled two times between CCTAS steps.
__________ L _______ _ Psychometrics Le\.lE3IQLKn(:~_VII~~ft~ __ ________ Difficulty_______ r- Time Allowance (minute§lL___R9 _ _ MEMORY I 10CFR55.41 (b)(1 0) Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: __ J8ULJ"J=1'~m~~Q~ ____________ _ Referen_c::e{s): ON-124 Learning PLOT-PBIG-1550-2 Objective: KIA System: G 2.3 Radiation Control Importance: RO / SRO 3.4 / 3.8 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, fNONE . ----_~=- -------_-~~~_ _ a_cc~§s t() 19cked high radiation areas, aligning_ filt~r~~tc~"- _____ _ __ n_ ~~t;~~~~~~:~::~ALS:_
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 72. According to T-BAS (lNTRO) "Introduction to TRIPs and SAMPs - Bases",
TRIP and SAMP procedures provide operators with instructions to manage critical plant parameters (2) the design basis of the plant. A. (1) symptom-based (2) within B. (1) symptom-based (2) beyond C. (1) rule-based (2) within D. (1) rule-based (2) beyond
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key
.-~~-",~.~- ..
~~
Question # 72 RO Choice -~~---r ~-"~~-"-+"""""""""~"~""""""""""""" Basis or Justification
--~--" "~~"--"--"~~----~~~"--"""""""""" "" """ "--~--~~-~~~--
Correct: B CORRECT - as discussed in T-BAS (INTRO) documentation, TRIPS are symptom-based, and no risk or probability threshold is assigned. Every effort h~sJ)~~.1} made to address C!r:!Y_mechanistically p(J~sible condition. Distracters: A INCORRECT - TRIPS are not bounded by Design Basis Analysis Events plausible since the candidate may incorrectly believe the TRIPS protect DBA accidents and SAMPS are for scenarios. C INCORRECT - TRIPS are NOT rule-based, and are NOT bounded by DBA analysis. Plausible since "rule-based" is a common term associated with Human Performance activities, and sounds similar to "Event Based or Symptom Based" and since the candidate may incorrectly believe the TRIPS protect against DBA accidents and SAMPS are for beyond-DBA i scenarios. i INCORRECT -TRIPS are NOT rule-based. Plausible since "rule-based" is
! a common term associated with Human Performance activities, and sounds
__ ~~_"_""""""""""""""""""~_~~_--'_~_~_.~ , similar to "Eve!)t Based or SympJQI11_E:t~~~'~~_~_~~~ Ps chometrics Level ofl<.l1.9~ll?qge _Pifficl.Jlty~_~ Time Allowance RO MEMORY 10CFR55.41 (b)(10) Source Documentation Source: I ~ New Exam Item Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: 0 1 1 _ ~ D ILT Exam Bank _~~[~ff?1'l9~(s--,--):__ T:!3ASJJt-JI~QJ __ ~ Learning PLOT 2000 DBIG Obj 3 Objective: KIA System: G 2.4 Emergency Procedures I Plan Importance: RO I SRO 3.3/4.0 KIA Statement: G2.4.18 - Knowledge of the specific bases for EOPs. REQUIRED MATERIALS: --" ~-----'~J--""""""""""""""""-~
" " " " " NONE
- -~-
Notes and Comments: ---------------- - --- - --
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 73. An ATWS is in progress on Unit 2. Per T-117 "Level/Power Control", a priority action is to inhibit ADS.
This is done to prevent __________ A. core damage due to large irregular neutron flux oscillations B. exceeding 110 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 License NRC Examination April 2013
~~-~ .... ~ ..... --
Answer Key Question # 73 RO Choice --~------ Basis or Justification Correct: D CORRECT - The ADS safety function is inhibited to give priority to other
, systems (Le.. provide additional time for SLC, RPS, etc. to perform their I safety functions). From T-117 Bases: ADS initiation would complicate efforts to maintain RPV level within required level ranges. FURTHER, rapid and uncontrolled injection of large volumes of relatively cold, un-borated water from low pressure injection systems may occur. With the reactor either critical or 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 ...~__ ._~ ~ __~_--rif1bibited to prevent!b~fror11_b~J~p~_ning"- __ ~~~ __
- trn:offi_ I ~~~::;:~t::~~:~:::~~~::~~~i:::~::~:::~~;:t:~~~dS.
I t::S I degrees F before boron injection anyway due to SRV operation ...this is not _~ ___ fJbEu~aso'l1or inhibitir1.g~6DS. __ ~ __~_~~_ C I INCORRECT -Depressurization due to ADS initiation must also be I accompanied by elevated Drywell temperature for this to occur ...this is not reason for inhi!>.itingbCl.§'____ . Psychometrics Level of Kn().....,I~gg~ ~
.._J?lfficul~L f-Iime Allo"!~'l~~(minutes) - ---- ------
RO MEMORY -I 10CFR55.41 (b)(5) Source Documentation Source: New Exam Item r8J Previous NRC Exam: (2008 RO) D Modified Bank Item D Other Exam Bank: 0 _j ILT Exam Bank Reference(?):
- T-117 Bases Learning PLOT-2117-6 Objective:
KIA System G 2.4 Emergency Procedures I Plan RO/SRO
........ -~--------. ---~
3.6/4.4 KIA Statement G2.4.22 - Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.__ ~:t~~~~~i~~~::~*~**--L-S--:.~.-.--- [[6~~
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 74. Unit 2 is operating at 100%.
"A RECIRC PUMP SEAL STAGE 2 HI FLOW' (Panel 214, A-1) alarm is received.
The following Reactor Recirculation Pump A" Seal parameters are reported:
- First Stage Seal (Seal Cavity #1) pressure: 1000 psig
- Second Stage Seal (Seal Cavity #2) pressure: 700 psig Which one of the following is correct based on the above indications?
A. #1 Seal is degraded. B. #2 Seal is degraded. C. #1 Seal pressure breakdown device is clogged. D. #2 Seal pressure breakdown device is clogged.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 74 RO .-~- ~~~-~~ --~---~-- ...................~--~-- Choice
~~~~-, .......... ~~ .......
Basis or Justification Correct: A
- CORRECT - higher than normal second stage cavity pressure can be the
. result of either: (a) partial failure of the first stage seal, or (b) clogged
, second stage leak off line. Since the second stage seal hi flow annunciator
- is alarming, cause (b) can be eliminated; a partial failure of the first stage seal is the cause of elevated pressure in the second stage cavity. If the
, first stage seal were completely failed, then both cavity pressures would be
~_ ~_ _i~_q~C!L_~~~~ __ ~~ ~~_~__ ~ .__~~~ ___ ... _~ __~__.. ~~~_~.~__ ~
Distractors: B i INCORRECT - a failed second stage seal would cause a seal stage 2 low I flow alarm since flow would bypass the bleed off line (and flow sensing
- switch) and be diverted out the failed seal. Plausible because candidate misunderstand Recirc Seal construction.
INCORRECT - a clogged first stage pressure breakdown device (PBD) would cause a higher dP between the first and second stage cavities; second stage cavity pressure would drop, and the low flow annunciator could be received (but not high flow). Plausible because candidate may misunderstand Recirc Seal construction. --- .. -- ---- INCORRECT - a clogged second stage pressure breakdown device would
; reduce flow in the bleed off line (past the flow sensing switch) and possibly
- cause a low flow alarm. Plausible because candidate may misunderstand i Recirc Seal construction.
Psychometrics LeveIQfJ5no~~cj@_ I __ Diffi(;l:ll~y_____~ ~ :ri~~_~JI()~~_nce{!TIinutes L RO ~ - ~ HIGH ! 10CFR55.41 (b)(7) Source Documentation Source: D New Exam Item D Previous NRC Exam: D Modified Bank Item C8J Other Exam Bank: (PB LORT) _D_J~-LE"-~I!I~?lilK1L______________ _ Reference(s): ARC 214 A-1 ~ebj:~~Je=~ l PLOT 5002 Obj 9.k.5 ________ _ ROISRO KIA System ' G 2.4 Emergency Procedures I Plan I Importance:
~_~~L ~_~~. ____.. 4.0 I 4.6 KIA Statement: 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant sytstem integr~'t2-contain'!l~nt condition~-, radi~~~"ity release control, etc.
REQUIBE[)_IIJI~!E~_IALS: ~O~E _ _~ ___ __ __ Notes and Comments: - --- ------ NONE
~-
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 75. In accordance with procedure FF-01 "Fire Brigade", if off-site fire departments are called to support the Fire Brigade, which of the following work groups, besides Security, can be used to perform security escort duties?
- 1. Operations
- 2. Maintenance
- 3. Radiation Protection A. 1 ONLY B. 1 and 3 ONLY C. 2 ONLY D. 1,2,3
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 75 RO Choice Basis or Justification Correct: o FF-01, Fire Brigade, "The Control Room should request groups on site to provide escorts." This is based on Ops and Security not having staffing required. Ops and Security are NOT precluded from escort duties. Distractors: A INCORRECT - any individual with Vital Area access permission can provide access escort. Plausible because Security and Operations are the work groups most closely associated with fire brigade and security activities. B INCORRECT - any individual with Vital Area access permission can provide access escort. Plausible because Security and Operations and the work groups most closely associated with fire brigade and security _________ +_____.. _._-:--~'~"'_'I1'_'_=_.=_'___R_.__P is another~!g_'!Q_\'VithJ~quent esco!t-!YPecll:-lty_assignments. C INCORRECT - any individual with Vital Area access permission can provide access escort. Plausible because Security is most closely linked with Vital Area access control. Psychometrics Level()LKnowleQ~ f --- Difficulty - I Time Allo'll@nce Jnlinutes) 1- 1OC FR5~~ 1(b)(10) MEMORY Source Documentation Source: (g] New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0
-----------------i---."'""' ILT Exam Bank_O __........__ ._ . ____ _
Refer~I1~~fs1 ___ FF-01 Learning
- PLOT 1565 Obj 2 Objective:
KIA System G 2.1 Conduct of Operations Importance: RO I SRO 2.5/3.2 KIA Statement G2*1 J ~ Knowledll~ gf fa...9ili!Y reqlJlr~!!!e~ts for contrQiU.rl9Jd!all controlled acces?.. REQUIRED ~ATI;~I~L.~: Notes and Comments:
--B~ONE NONE
- - - -- ~-- - ~
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 76. Unit 2 is in a refueling outage. The following conditions exist:
- The reactor head is removed.
- The fuel pool gates are removed.
- Inspections are in progress using the Reactor Cavity Work Platform (RCWP).
A loss of Fuel Pool/RPV cooling occurs causing:
- RPV temperature to rise to 142°F.
- RPV level on LT-70 to indicate 474".
Which one of the following shows (1) actual RPV level and (2) the action required for these conditions? (Refer to Attachment 3 of GP-6 on the next page) (1) Actual RPV level is _ _ _ __ (2) Required action for this condition is _ _ _ __ A. (1) +467" (2) evacuate the RCWP per GP-6, "Refuel Operations". B. (1) +467" (2) add water with Condensate Transfer per GP-6, "Refueling Operations". C. (1) +481" (2) maintain RPV level between +480 to +488 inches per GP-6, Refueling Operations" . D. (1) +481" (2) lower RPV level and maintain RPV level between +470 and +477 inches per GP-6, "Refueling Operations".
G?-6 41 44 ATTACHMENT 3 LT-70 CORRECTION 'I~ABLE FOR +474" The purpose of thi Attachment is to the Operator with tte proper INDICATED level value in order to maint in ACTUAL RPV level t +47". T~is is necessary to compensate for effects on indica Reactor 1 when Reactor water is above below calibrated condit (BO°F) . o~ TEMP l' IND. LEVEL TEMP of . LEVEL 70 474.6 1 6 470.4 72 474. 118 470. 4 474.4 20 470.0 76 474.3 2 469.1 78 47 469.4 80 474.0 126 469.1 82 473.8 128 468.9 84 4 3.7 o 46fl.7 86 473.5 468.4 88 473.3 134 4 .1 90 473.2 136 467. 473.0 38 467. 94 472.8 140 67.3 96 472.6 142 4 .0 98 472.4 144 46 100 472.2 146 466.4 102 4 .0 148 466 . 104 471. 8 150 106 471. 6 152 4 108 471. 4 154 4 .2 110 471.1 156 Hi4.9 112 470.9 1 8 4 14 470.7 160 4 4.2 Example: IF Reactor water is 20 0 F t THEN IND:;:CATED React leve:. should be maintained at +470.0" order to maintain hCTUAL RFV level at +474".
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 t..lUE~Stl(m # 76 SRO Choice Basis or Justification Correct: D Correct - RPV level per Attachment 3 is high and must be lowered to the band of +470 to +477" per step 5.4.18 of GP-6. Distractors: A Incorrect - RPV level per Attachment 3 is high, the distractor is plausible because if RPV level were +467" the RCWP would need to be evacuated. B nrl"lrro:>rt - RPV level per Attachment 3 is high, the distractor is plausible if RPV level were +467" water would need to be added to restore level to band. C nrl"lrro:>rt - RPV level is high, but the RPV level band per GP-6 is +470 for inches. Psychometrics I Level of KnoVIIled~~~ Difficulty Time Allowance (minutes) l .~. . SRO HIGH i 10CFR55.43(b)(7) Source Documentation Source: [2J New Exam Item Previous NRC Exam o Modified Bank Item o Other Exam Bank _DJLT EX?f"l"lJ3a!1K~_~ ______ ~ __ . ~ R~~r~rlc::~Jst__~~_~_qP-()JJ3~LJ~1 Oper§l!i9!!~"* Learning PLOT 1530 Obj 4 Objective: KIA System 295023 Refueling Accidents Importance: SRO 3.7 KIA Statement AA2.02 - Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS* Fuel pOQUf3'!.I:ll_________~~~~_ REQUIRED MATERIALS: NONE ......... -.--~-.~ .. Notes and Comments: ~.~~._ . .L.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 77. An ATWS is in progress on Unit 2.
RPV water level was intentionally lowered per T-117 "Level/Power 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 at 950 psig
- Torus temperature is 175 degrees F and rising
- RHR loop 'A' is in Torus cooling; 'B' loop is unavailable
- Torus pressure is 6 psig and slowly 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 "Level/Power 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 TIT-1 "Heat Capacity Temperature Limit." D. Perform Emergency Blowdown per T-112 due to being on the unsafe side of TIT-1 "Heat Capacity Temperature Limit."
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 T-102 "Primary Containment Control" "SRV Tail Pipe Limit" Curve
*** CURVE TIL-l
**I SRY lAll PIPE LIMIT 17.6=1=1=+=+=++=+=1=+=+=++==1 I
I UNUFE I I I I
...... 1616 CONTINUE EFFORTS TO I L.. TIL-21 RESTORE TORUS LEVEL * $.... IS~t=~~~~:t:t:t=+=+=4~
I
~ 1I*.2:t=t::t::t:
L.. (YES) !! 14.8+-+-+-+ 14.4 t:t=t=l::t::t:::t:=~=l::t::t\:j o 200 400 800 1100 1000 1200 RPV PRESSURE (PSIOI 6(NO) L.. IE 1-101 HAS ~ ALREADY BEEN ENTERED. THEN 1. MANU ALL Y SCRAM THE REACTOR USI Nil GP-4
- 2. ENTER T-l0l AND EXECUTE IT CONCURRENTLY WITH THIS PRDCEOURE ...... ____ ~ L..
- 3. PERFORM RPV DEPRESSURIZATION PER T-101 L.. ~T/-L--2-2-----------------'*.----------------~ \ ) RC-t I
T-I02 "Primary Containment Control" "Heat Capacity Temperature Limit" Curve i CURII -I I
- ilEAl CAPACITY lEMP UMIT
.,-_ .. _- ... ------_ ..... -...** ~,
SAFE tlEOlOH IS BELOW THE CURVE ZID "!'"T--,r-r-,.-..,....-.,...---; ZSO ~"""I::II:-t-+- t* REGARDLESS OF RX POWER. ENTRY INTO 1-118 OR STEAM COOLING 240 290 RPV PRESSURE I seCT! ON OF 1-111 IS NOT REQUIRED. o 10 74.9 PSIO I I LOWER RPV PRESS TO MAINTAIN AN -- ~ 220
** OPERATING POINT BELOW CURVE Tll-1 EXCEEDING COOlOOWN RATE lIMITS ...w...=>
210 75 TO 289.S PUG t.... TlT-B IF NECESSARY .. w 200 100 300 TO 499.9 PSIO 600 TO 898.9 PSI G 700 TO 899.9 PSIO
~
180 900 PSI a OR ABOVE I I I C> 170 180 I lliO
'. __ ......t.... -----
* <YES) lo.s 11 12 13 14 Iii 18 17.1 TORUS lEVEL {FT>
PERFORM AN EMERGENCY BLOWDOWN USING 1-112 ...... ~ t.... L-______________________________...J \ ) EB-l I.... TlT-l 0
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 T -117 "Level/Power Control" (yes) L L I B >.. L- (NO)
----.~:
i t TERMINATE AND PREVENT RPV INJECTION USING T-240 (ATT. 1. FIG. S) L- LQ-21 PERFORM AN EMERGENCY BLOWDOWN USING T-112 -.F-m> L L- LD-2Z VEB - 1
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key _____ ~~ _ _ _ ~ ******* ~00_~ Question # 77 SRO-~--~~~-~~~~~~~~~c--~~~-~- -~~~~~ Choice Basis or Justification Correct: C Torus temperature is -3 degrees F from HCTL and rising. If Torus temperature cannot be maintained on the safe side of HCTL, T-102 TIT-8 directs maintaining RPV pressure on the safe side of HCTL. Distractors: A 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. B While RPV Level is below -195 inches, it is only 5 inches below band and is rising due to HPCI injection. The criterion for T-117 LQ-20 is whether or not level can be restored and maintained above -195 inches, which it can. Therefore, T -112 is not warranted under these conditions. Operation is on the SAFE side of the HCTL curve. b~~L()!JSI"1°wledge HIGH Diflicutt'Y ---.1 Psychometrics Time ~11cl1N..mCe(minlJte~1 SRO 10CFR55.43(b)(5) Source Documentation Source: D New Exam Item I:8J Previous NRC Exam (PB 2008) I:8J Modified Bank Item D Other Exam Bank ILT Exam Bank --~ -~-~ ~~~~~---~- ~~ T-102 and Bases --~-----~-" PLOT-PBIG-21 02-5a KIA System 295030 - Low Suppression Pool Water SRO Level/5 3.9 KIA Statement EA2.03 - Ability to determine and/or interpret the following as they apply to Low Suppression Pool Water Level: Reactor Pressure. REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 78. Unit 2 was operating at 90% power with the OPRM System inoperable when the
'2B' Recirc pump tripped. The following conditions currently exist:
- A loop flow (FI-2-2-3-092B) is 46 Mlbm/hr
- B loop flow (FI-2-2-3-092A) is 5 Mlbm/hr
- Indicated Core Flow (FR-2-2-3-095 black pen) is 51 Mlbm/hr
- APRMs are oscillating between 48 and 51 % in 4-5 second random intervals Which one of the following is correct for these conditions?
AO 60A.1-2 "PBAPS Backup Stability Solution Power Flow Operation Map" is PROVIDED ON THE NEXT PAGE. The plant is operating in (1) The required action is to (2) A. (1) Region 1 (2) scram the reactor and enter T -100 "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" C. (1) Region 2 (2) 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-112 "Unexpected/Unexplained Change in Core Flow"
'"U (1)
Q) (") AO 60A.1-2 ::r Rev. 0 IlJ Page 8 of 11 ATTACHMENT 1 8: o PBAPS BACKUP STABILITY SOLUTION 3 POWER FLOW OPERATION MAP :::J
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- l I
. -:- -~~ :~;: ~L=~~,,---;;.7~ . . --,J 1 I'" ,.
"'0 I _ fIIIIIiIIt' I "."
... - .or tv 7C o Core Flow (% of Rated) w
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Question # 78 SRO Choice Basis or Justification Correct: The calculation of core flow 51-2(5) =41 Mlbm/hr 1102.5 Mlbm/hr =40% (alternatively, 41 Mlbm/hr can be found on the upper 'x' axis). Plotting 41 Mlbm/hr vs. 48-51% power shows the reactor is operating in Region 2. All GP-9-2 rods must be inserted since the 2B Recirc_~~mptripped. __ Distractors: If a core flow calculation and/or plotting error is made, the applicant could believe the reactor is operating in Region 1.
- - _ . _ . _ - - - + - . _ - - - + - - - - - - - - - - - - -- - -- _._._--_._------ ----_._-------
D If a core flow calculation error is made, the applicant could believe the reactor is operating in the normal operating region. C Plotting in Region 2 is correct, however, raising recirc pump speed would not be a correct action if operating in Region 2 with indications of THI. Psychometries _ Level of KnowleQg~_ Diffic~lty ______ Time Allowance ................... (minutes) SRO HIGH
-'------ - --- ---.~------.~ ---~-~---~-- -~--*t-*~--
10CFR55.43(b)(5) Source Documentation Source: New Exam Item ~ Previous NRC Exam: (PB 2011)
~ Modified Bank Item o Other Exam Bank: ()
ILT Exam Bank 1 AO 60A.1-2 Learning Objective: K/A System: 295001 - Partial or Complete Loss of Importance: SRO Forced Core Flow Circulation 3.2 KIA Statement: AA2.02 - Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of Fo!cecL9ore FI9\J\fgir~ulati9n :t>Je~!r~n_ f\t1_C:>rl_it0rinJJ___ _ __ REQUIRED MATERIALS: ilNONE N.O.te..s.. and comment.s.. :. It .is.. the SRO's job function to determin. e. the oper.a.ting point on the Power-to-Flow map (or Backup Stability Solution Power Flow
.____ .__ .......... . __________ i>pe~~tio!L"'1<:ip),~t~I~IJJs "imrl1~dia!eQP~rator ~~tion" of OT-112.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 79. Unit 2 is in Mode 3.
* "A" loop of RHR is in Shutdown Cooling using the 2A RHR Pump
- RPV pressure unexpectedly rises to 90 psig and stabilizes.
- Alarm 224 C-1 "SYSTEM I HI REAC PRESS SHUTDOWN COOLING ISOLATED" is received.
For the above conditions, which one of the following describes the effect on Shutdown Cooling system valves (1) and what action(s) is/are required to be taken (2) ? MO-2-10-17 is the "Shutdown Cooling Outboard Valve" MO-2-10-18 is the "Shutdown Cooling Inboard Valve" MO-2-10-25A is the "A RHR Loop Inboard Discharge Valve" A. (1 ) MO-2-10-17 AND MO-2-10-18 close ONLY. (2) Direct high pressure signal defeated lAW ON-125, "Loss or Unavailability of Shutdown Cooling". B. (1 ) MO-2-10-17 AND MO-2-10-18 AND MO-2-10-25A close. (2) Direct Rx water level raised to > +50" lAW GP-12, "Core Cooling Procedure" . C. (1) MO-2-10-17 AND MO-2-10-18 close ONLY. (2) Direct that Alternate Decay Heat Removal Systems be placed in service lAW ON-125, "Loss or Unavailability of Shutdown Cooling". D. (1 ) MO-2-10-17 AND MO-2-10-18 AND MO-2-10-25A close. (2) Direct a Reactor Recirc Pump to be started lAW GP-12, "Core Cooling Procedure".
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 79 SRO ~-~-~~~~~~~~--~-~- Choice ---,---~~-~~~~~- -+-~- -~---~~~~ -~ -~~~-~-~~~ ~ Basis or Justification Correct: C Correct - (1) Correct valves. MO-2-10-17 AND MO-2-10-18 get a close signal (PCIS Group II) .MO-2-10-25A will not get a close signal if RPV press is> 70 psig. (2) ON-125 "Loss of SDC" directs if in MODE 3 > 70 psig to line up Alt [)e~ClY HeClLRel11()valJ)~~!~I1!{§>-- __~~ Distractors: A Incorrect - (1) Correct valves. (2) Per ON-125 an Alt Decay HT removal system is required to be placed in
. service. You can only defeat high pressure signal per ON-125 if RPV
**-~---~~--'-----~**Br~;:!:~j!(;~~~~ Ad~es not close (Isolate) on hi press (>70 pslg).
! (2) Raising reactor level to> 50" is required if in MODE 3 and <70 psig.
Not the case here.------- . . -- - -------_ ...... , - - - - - - - ---_... Incorrect - (1) MO-25A does NOT close on hi press (>70 psig). (2) Restarting a RX recirc pump is an action required if in MODE 3 and <70
- ___ ~ ___ ._~~~ ___ ~L p~.i.fL_____ _..
Psychometrics SRO 10CFR55.43(b)(5) Source Documentation Source: cg] New Exam Item D Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: ()
--1--==--
ILT Exam Bank Beferen~~( s t __ ~QN-125 Loss or UnavailabHi!r.QL§hutdgwn Cooling~ Learning i PLOT-5010 Obj 3n Objective: I KIA System: I 295021 Loss of Shutdown Cooling
.____ J__~ ____.. . .__..~ ____._
KIA Statement: 2.4.31 - Emergency Procedures I Plan: Knowleqge of annunciator alarms, indications, or response procedures.
- 80. Unit 2 was operating at 100% power when a Loss of Instrument Air occurred. The following conditions exist:
Peach Bottom Initial Senior Reactor Operator'License NRC Examination April 2013
- 80. Unit 2 was operating at 100% power when a Loss of Instrument Air occurred. The following conditions exist:
- SCRAM VALVE PILOT AIR HEADER PRESS HI-LOW (211 D-2) alarms
- A INSTRUMENT AIR HEADER LO PRESS (216 D-3) alarms
- B INSTRUMENT AIR HEADER LO PRESS (216 D-4) alarms
- Scram air header pressure is 50 psig and lowering
- ROD DRIFT (211 D-4) alarms
- The URO reports control rod 22-23 is drifting in Which one of the following actions is required for these conditions?
A. Scram and enter T-100 "Scram" per ON-119 "Loss of Instrument Air". B. Use the EMER IN control switch to insert rod 22-23 to Full-In per ON-121 "Drifting Control Rod". C. Scram and enter T-100 "Scram" IF a second control rod drifts per ON-121 "Drifting Control Rod". D. Begin a rapid plant shutdown using GP-9-2 "Fast Reactor Power Reduction" per ON-119 "Loss of Instrument Air".
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Question # 80 SRO Choice Basis or Justification Correct: A Applicant must recognize that ON-119 entry is required based on (interpret) IA System alarms. ON-119 directs a reactor scram if any control rod begins to drift in due to decreasing scram air header pressure. The given conditions indicate that scram air header F>r~~~_LJr~lsL()werlrlg--,--__ ___ _ Distractors: B This is the correct action per ON-121 for a drifting control rod only (i.e., due to a loss of instrument air). Entry into ON-119 (and direction to ___+--_____+-=c..:::. :..::.:~,_o=_v=_ec:...cr_r:...cid:::=es ON-12t§ftiQns tor a__Qljftir19 control rod. C is the correct action per ON-121 for a second drifting control rod, but is overridden by the direction in ON-119 to scram on the first drifting rod. D This is required by ON-119 when instrument air header pressure cannot be stabilized above 75 psig, but is overridden by the requirement to scram if ___ ..... ______ 1.____._ __,__ any_____ control _
~~
rod begins to drift. Psychometrics Source Documentation Source: New Exam Item [gJ Previous NRC Exam: (PB 2009) D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank
-=="-:-=--:-- ---_.--_._--- ---
Reference_(s t_ ON-11 ON-121 Learning PLOT-DBIG-1550-2 Objective: KIA System: 295019 - Partial or Complete Loss of Importance: SRO 1 Instrument Air 4.4 KIA Statement: 2.4.49 - Emergency Procedures I Plan: Ability to perform without reference to procedures those actions that re9uice immediate 0rratio Q.of system _compgnents and controls. REQUIRED MATERIALS:
- - ----- - - - ~ -
NONE
- -- - - - - ~
Notes and Comments:
-~---- ----~ -
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 81. Unit 3 was at 100% power when a steam leak occurred in the Drywell.
- Alarm 310 F-1 "DRYWELL HI PRESS TRIP" was received.
- A full automatic reactor scram occurred due to high Drywell pressure.
- ALL control rods fully inserted EXCEPT for 3 control rods presently at position "24".
- Drywell radiation is presently 90 R/hour.
- Drywell pressure rose to 18 psig and suddenly lowered to 0.4 psig.
- Alarm 310 F-1 "DRYWELL HI PRESS TRIP" was able to be reset.
- Alarm 310 J-3 "HIGH AREA TEMP" was received.
- TRS-3*-13-139 point 22 (RB 165' General Area) is reading 100°F and rising.
- Torus temperature is 92°F and slowly rising.
What is the HIGHEST classification for the given conditions? EP-AA-1007 "Radiological Emergency Plan for Peach Bottom Atomic Power Station - Table PBAPS 3-1 EAL Matrix" is PROVIDED SEPARATELY. A. MU5 B. MA2 C. FA1 D. FS1
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Choice Basis or Justification
~"" """-- ~- ~~~~~ -"--r~""" ""-"-"~~~" "i-"""-"-----""""----"--"~-----"-~-""-"""-"---"-""-"-"-----"~-"----"~ """ --"--~"" --""~-" "-- """ ~ "-" --"
Correct: D Correct - Greater than 2 psig in the Drywell (Loss of Reactor Coolant System) and the sudden drop in Drywell pressure with Alarm 310 F-1 "DRYWELL HI PRESS TRIP" reset (Loss of Containment) makes the condition a loss of 2 barriers which results in a Site Area emergency per __________+ _____ "_"_i=E~AL="~F_S~"""1"~~. ______"_" Distractors: A Incorrect - The candidate~ may select EAL MU5 since it is for reactor coolant I system leakage while operating. It is related to the >10 gpm unidentified leakage and the >25 gpm identified leakage. Not applicable in this case. i B i Incorrect - EAL MA2 is for an automatic scram condition failing to """" shutdown the reactor as indicated by reactor power> 4%. The candidate may select this EAL based on the ATWS (3 control rods not full in).
--"-+-------+"--~~~-~.~~.-"-~~-""~--,~-~-
as 4% with 3 control rods out. C Incorrect - The candidate may select EAL FA 1 if it determined that only a loss (or potential loss) of reactor coolant system exists. There is also a loss
' - _ _ _ _ _"""""_L of Ps chometrics
"_~~v:~J_9JKn_owleclge __ "_____._ Difficu!1)1__ ____ Time Allowance SRO HIGH 10CFR55.43(b)(5) Source Documentation Source: rgj New Exam Item 0 Previous NRC Exam D Modified Bank Item 0 Other Exam Bank ILTExam Bank EP-AA-1007 EAL Matrix
--t-""~~."--
Learning PLOT Objective: KIA System 295024 High Drywell Pressure SRO 4.2 KIA Statement 2.4.40 - Emergency Procedures I Plan: Knowledge of SRO responsibilities in emergency plan REQUIRED MATERIALS: EP-AA-1007 EAL Matrix
--"" -----+-----"-------------"".
Notes and Comments: """""_"_ _ _ _"_""""""""L _ _ _ _"""""_ _ _ _ _ _
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 82. Unit 2 is at 100% power.
Unit 2 Intake Canal temperature is being monitored per TS 3.7.2. Previous 24 hours of hourly-recorded temperatures are: 90.2 90.3 90.2 90.2 90.2 90.2 90.2 90.2 ! 90.2 90.3 i 90.3 90.2 90.2 89.9 89.9 90.2 i 89~ 90.2 89.8 89.9 E~ 89.9 89.9 89.9 90.2 What action is required, if any, to comply with TS 3.7.2 LCO (on next page)? A. Restore ESW to operable status within 1 hour, or be in MODE 3 in 12 hours and MODE 4 in 36 hours. B. Continue to monitor Normal Heat Sink temperature AND verify 24 hour average is ~ gO°F. C. Be in Mode 3 in 12 hours and Mode 4 in 36 hours. D. No additional actions are required.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 ESW System and Normal Heat Sink 3.7.2 3.7 PLANT SYSTEMS 3.7.2 Emergency Service Water (ESW) System and Normal Heat Sink LCO 3.7.2 Two ESW subsystems and normal heat sink shall be OPERABLE. APPLICABI LITY : MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ESW subsystem A.1 Restore ESW subsystem 7 days inoperable. to OPERABLE status. B. Water temperature of B.1 Verify water Once per the normal heat sink temperature of the hour is > gO°F and :<;; 92°F. normal heat sink is
- <;; 90°F averaged over the previous 24 hour period.
C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or AND B not met. C.2 Be in MODE 4. 36 hours Both ESW subsystems inoperable. Normal heat sink inoperable [for reasons other than PBAPS UNIT 2 3.7-3 Amendment No. 244
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- i. ()ljestion # 82 SRO
**** _ _ _ _ *.~**** _ _ _ ** _ m Choice Basis or Justification Correct: Correct - TS LC o Cond B applies since stated current reading is> 90°F.
Averaging previo us 24 hourly readings yields 90.1 O°F, which is in excess of 90°F. TS 3.7.2 bases states "If the water temperature of the normal heat sink exceeds 90 0 F when averaged over the previous 24 hour period, Condition C must be entered immediately. Required action is to shutdown lAW condition C. istractors: A Incorrect - see above for TS LCO discussion - Plausible because candidate could confuse 1 hour in Required Action with once per hour in
.G()mpleti~rl tirT1~p~r:Ba~~I)LG()l"IgjtioI1C
. must be entered immediately:
B Incorrect - see a bove for TS LCO discussion - Plausible because candidate could incorrectly calculate conditions and incorrectly determine that Condition
.... ~-~-----
B a'ppli~~:__ ~~ __ . . . _.~ __..___~_ . . ~ __ .
--~ ~---
0 Incorrect - see above for TS LCO discussion - Plausible because candidate could incorrectly determine average temperature as being below 90 F and decide no further action il)_regLJired. I Psychometrics
.___Difficulty -. __ Tim-" Allow,,-"ce (rrlillutest 1 SRO 10CFR55.43(b)(2)
Source Documentation Source: [8J New Exam Item Previous NRC Exam D Modified Bank Item D Other Exam Bank ILT Exam Bank
~~!~r:~nc~(s}: ____ +_T __..S. 3.7.2 and Bases Learning PLOT 1858 Obj 1 Objective:
KIA System 295018 Partial or Total Loss of CCW SRO 4.7
-------- -----n----
KIA Statement REQUIRED MATERIALS: Notes and Comments: 2.2.40 - Equipment Control: Ability to apply technical specifications for a system. NONE
- At P~~chBottom, ESW provides CCW for -a mo~t ~f the ECCS
_ com'p~nents that regl.llre y_CW. _
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 83. Unit 2 is at rated power.
- A small steam leak has occurred in the Drywell.
- Computer point for DRYWELL TEMP INDICATOR TI-2501 ZONE 4 is invalid.
- PRrrR-4805 "Containment Temp" is reading 131°F.
Referencing RT-O-40C-530-2 "Drywell Temperature Monitoring" PROVIDED ON NEXT PAGE., calculate approximate Drywell Bulk Average Temperature and determine required actions. Technical Specification 3.6.1.4 "Drywell Air Temperature" is PROVIDED SEPARATELY A. Continue to monitor Drywell temperature, no further actions required. B. Direct standby Drywell Coolers and Drywell Chillers in service lAW ON-120 High Drywell temperature. C. Restore Drywell Average Air temperature to within T.S. limits in 8 hours or be in MODE 3 in 12 hours. D. Direct maximizing Drywell Cooling, bypassing OW Fan Trips using T-223 if necessary, lAW T-102, "Primary Containment Control".
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 RT-O-40C-530-2 Rev. {; Page :, of 11 6.2 Dry\'lell Temperature Calculations NOTE points in a given TI-2S0 are out of service, THEN 1 Bulk Average Temperature ion on Data Sheet. 1 will be AND EITHER TI**2501 Point. 136 OR 4805 must be used to calculate Drywell Bulk Average for entry int.o ON-120 and Refer to Precaution 4.2.2 for 6.2.1 IF TI-2501 or corresponding PMS computer points have at least. 1 valid temperature point in each of Zones 1 through 5, THEN CALCULATE 1 Bulk Temperature using on Dat,a Sheet 1. OTHERWISE, s 6.2.2 IF TI-2501 or corresponding PMS comput.er does NOT have at least 1 val id t_emperature point in each of Zones 1 through 5, THEN CALCULATE APPROXIMATE 11 Bulk Average Temperature as llows. OTHERWISE, NIA this step:
- 1. RECORD temperature reading from EITHER TI 2501 Point 136 OR PR/TR- 4805 "Ccmtainment Temp" at.
Panel 20C003-02.
'F
- 2. ADD lO'F to temperature recorded in substep 1 above to determine APPROXIMATE Drywell Bulk Average Temperat_ure.
6.2.3 VERIFY Drywell Bulk Temperature is less than 140cF. 1__.1 R ______ . .
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 83 SRO -*~*-*-~~-~***r*---~~-***- Choice Basis or Justification Correct:
- CORRECT - Calculated OW temp is 141°F. ON~ 120 entered with drywell temp above 140°F. ON-"120 provides direction to start standby OW coolers
~~~_-+ _____~.. ~+~a__n~_d~~~O_W_.C . . . ~h_iII~e.~rs to restore 0vvJertl1:>J:>~lo\IVJjQoE.~
1 Oistractors: A INCORRECT - Calculated average OW temp is 141°F. Entry into ON-120 is required. C INCORRECT - Calculated average OW temp is 141°F. TS 3.6.1.4 Orywell Air Temp requires OW average air temp to be.::: 145°F. No TS entry .
. ~-~---.~.~-~ ~ -~~~-.-~---
o INCORRECT - Calculated average OW temp is 141°F. No current entry condition into T -102 exists. Psychometries I LeveL()LKno\IVl~dg~~ _~_~iffi~l.Ilty~~_~_~~ J'lme AIJOWCll1ce~(rni!1utes) I SRO I HIGH i 10CFR55.43(b)(5) Source Documentation Source: 1:8:1 New Exam Item 0 Previous NRC Exam: () o Modified Bank Item Other Exam Bank: 0 Refer~n~~(~:~ RT-O-40C-530~2 ..... ~~~_'
"Orywell_____
.... M~__________
Temperature
~~ ...... ~_. _________. Monitoring" Learning PLOT-5040C Obj 10 Objective:
KJA System: i 295012 High Orywell Temperature Importance: SRO ________________ ~~ _ _ _ _ _ _ _ _ _ I
; M_~ _ _ _ _
3.9 KJA Statement: AA2.01 - Ability to determine and/or interpret the following as they apply to HIGH ORYWELL TEMPERATURE: Orywell temperature REQUI~~I:)!!1_A~~RIA~~~_JTec-h S~c 3.6-"-1.4 nD~_e_1I Air T~l!IperatlJre Notes and Comments: ~ __~ ~ ~
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 84. Unit 2 is operating at 100% power when the following conditions occur:
- Torus level is 13.5 feet and steady
- Torus Room level is 120 inches and rising.
* "c" RHR room level is 2 ft. and rising.
What actions are required for these conditions? A. Use SE-9, "Radioactive Liquid Spill" to lower water level in the RHR room. B. Perform a GP-3, "Normal Plant Shutdown". C. Perform a GP-4, "Manual Reactor Scram". D. Perform a T-112 "Emergency Slowdown". TABLE SC/L-2 WATER LEVEL-ALARM ANO ACTION LEVELS AREA A.LARM ACT! ON LEVEL INOI CATI ON STATUS LEVEL UNIT 2 UNIT 3 TORUS ROOM 6 IN. tOO IN. 100 IN. LI -2( 3)919 SUMP ROOM NONE 1 FT 7 IN. 1 FT 4 IN. LOCAL 81 GN OR ReIe ROOM SIN. Z FT 5 IN. Z FT 5 IN. LOCAL 81 GN OR HPCI ROOM I) IN. Z FT :2 IN. :2 FT Z IN. LOCAL 81 GN A RHR ROOM SIN. Z FT 11 IN. 3 FT 5 IN. LOCAL SIGN OR C RHR ROOM SIN. 1 FT 3 IN. 3 FT 5 IN. LOCAL 81 GN B RHR ROOM S IN. 1 FT 5 IN. 3 FT 5 IN. LOCAL 81 GN OR o RHR ROOM 6 IN. 3 FT 4 IN. 3 FT 5 IN. LOCAL 81 GN A CS ROOM SIN. 1 FT 10 IN. 3 FT 3 IN. LOCAL SIGN OR C CS ROOM 6 IN. 3 FT 8 IN. 3 FT 1 IN. LOCAL SIGN B CS ROOM 6 IN. :2 FT Ii IN. :z FT 4 IN. LOCAL SIGN OR o CS ROOM 6 IN. 2 FT 3 IN. 2 FT 10 IN. LOCAl 81 GN
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Choice Basis or Justification
........~.~~........~ ....... ~~~~~-~~~~~-~~~~-~.~.--.-
Correct: Correct -.A leak form the Torus Room into the RHR Room is not:
- Isolable
- a primary system
- in the same area These condition per T-103 "Secondary Containment Control" require a GP 3 shutdown Distracters: A Incorrect -. Procedure SE-9 "Radioactive liquid Spill" provides direction to isolate the source of the spill. Direction to control room level is from T-103.
C Incorrect -.GP-4 would be performed if the condition was caused by a i. primary system leak. It is not a primary system leak. D Incorrect - T-112 Emergency Blowdown would be performed if the I condition was caused by a primary system leak. It is not a primary system ' - -_ _ _ ........ _~~ __ ____'__~~._~L leak.
~ ~~____ ~.~_._~~~ ~~. ~~~_~~
Psychometrics
~_J~y(31 of Knowledge _ _ _D_i_ffi~._c_uJ!y Llll'T!~AIIO~~~C~J!!1ifll:ltesL-I- SRO .. ~ .. ~.~.
HIGH 10CFR55.43(b){5) Source Documentation Source: ~ New Exam Item D Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0 D ILT Exam Bank t:~i;:;) -l;~~~~o~~~g ~:~~m~nt Cont;';-anda~~,;;~t;d ba~es Objective: ' KIA s~;~~: .j ~~~~;~t:7L~~t~inm;~tH;9~= l,:po:n:: _;:0 KIA Statement:
~2.1.2QAbilitYJ()~interpr(3tand ~~ecute pr()ce~!Jre steps._~ __ ~_.~ ___ ....... __ ._~ ~~~~~~~;~::~AL~~JNO~=_=-=-==_:~-:-
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 85. The Unit 3 Reactor fails to scram on high Reactor Pressure. The following conditions exist on Unit 3:
- ARI automatically initiated as designed.
- The URO has taken actions to stabilize RPV level at + 10 inches.
- Reactor pressure peaked at 1150 psig, lowered, and is now stable at 100 psig.
Given the above conditions, which of the following will be most effective (fastest) method of inserting the maximum number of control rods? A. T-213, "Scram Solenoid De-energization" B. T-214, "Isolating and Venting the Scram Air Header" C. T-215, "Control Rod Insertion by Withdraw Line Venting" D. T-246, "Maximizing CRD Flow to the Reactor Vessel".
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Choice Basis or Justification Correct: Correct - The conditions indicate that there is a hydraulic A TWS. T -246 is used for low reactor pressure conditions. With low reactor pressure and maximum CRD flow all control rods should insert into the core. Distractors: A Incorrect - with a hydraulic ATWS T-213 will not be effective. B Incorrect - with a hydraulic ATWS T-214 will not be effective. I
- Incorrect - T-215 will only allow insertion of one control rod at a time.
Ps chometrics LeY~L of Krl9_wlf:lgae~ SRO HIGH 10CFR55.43(b){5) Source Documentation Source: ~ New Exam Item D Previous NRC Exam D Modified Bank Item D Other Exam Bank
-~-~~
I LT Exam Bank .. ---~ .. - ... ~. J~f:lff:li~!!c;~(st__J -246, "Maxin11&rla~RJ2.FI()w tolll~R,f:lactor Vessel" Learning PLOT 2101 Obj 7q Objective:
--~-.- ... -~-~ ..- - --_._. - . ~.--~-.
KIA System 295015 Incomplete SCRAM Importance: SRO 4.1 KIA Statement 2.4.34 - Emergency Procedures I Plan: Knowledge of RO tasks performed outside the main control
- ~~it~~~::::~C~ a~T:~~~~~~=~rmi~nal effecffi
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 86. The following conditions exist immediat~ after a large-break LOCA:
- RPV pressure is 100 psig and dropping.
- All RHR pumps are injecting into the RPV.
- All Core Spray pumps are injecting into the RPV.
- RPV level is -170 inches and rising rapidly.
10 minutes have elapsed since the above (initial) conditions, and no operator actions have been taken. Which one of the following identifies (1) the ADS response and (2) the action required, if any, for these conditions? A. (1) ADS initiated (2) Instrument Nitrogen must be restored to maintain a long term nitrogen supply per T-101, "RPV Control". B. (1) ADS initiated (2) no actions are required. The ADS accumulators will keep the ADS valves open. C. (1) ADS did not initiate (2) open all ADS valves per T-112, "Emergency Blowdown". D. (1) ADS did not initiate (2) Instrument Nitrogen must be restored to maintain a long term nitrogen supply per T-101, "RPV Control".
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Question # 86 SRO Choice Basis or Justification Correct: Correct - ADS will not initiate. Drywell pressure must be above 2 psig or RPV level must be below -160 inches for 9.S minutes. Neither of these conditions are present. T-101 "RPV Control", step RC/P-4, directs a nitrogen supply be restored for long term operation of the ADS valves. , Oistractors: I ;:-:1**** Incorrect - ADS will not initiate without orywe II pres~~re ab~~~2~;igor-- RPV level must be below -160 inches for 9.S minutes. RPV level will recover with the current injection rate of the RHR and Core Spray systems . .. -----~--.- ...... ...... ~.--.-------- ...... --~-~- ... B Incorrect - ADS will not initiate without Drywell pressure above 2 psig or RPV level must be below -160 inches for 9.S minutes. The ADS SRV accumulators do not provide a long term supply of nitrogen. C Incorrect - PRY level is above a level requiring an Emergency Blowdown. The ADS valve position will be changed based on the crews long term ablljty to coqtrQLBPV levt?1 abo,{t? -172"._ .....___ . _____. Psychometrics Levt?1 of Krl9wledg~ __ . .._____I:>ifficul!y__.___ - _Time AIIQwalJg~ (minl.ltt?~J_ - - SRO .. ~----~------ HIGH 10CFRSS.43(b)(S) i Source Documentation
~ New Exam Item D .__ _ --L 6~~d:~:~B;:~~t~___
Source: I Previous NRC Exam __ D Ot:=-~~am Ba.~~ ____ ._..... ___ _ Reference(s): T-101 "RPV Control" Learning - ~Co~~~~!~:\~nd SV S)'Stem Alignment!"r Norrl1a!Q~a~,,"--_ _ __ Objective: I
. - j - - - - -....- .......... - - - - - -..... - - - . - ... ---T--------
KIA System 218000 ADS I Importance: SRO 3.6 KIA Statement A2.02 -Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION System; and (b) based on those predictions, use procedures to correct, control, or mitigate the
~~~~!~~i:~~norm1~~~~ons_:~~ations: La~~~e~_~~ _~~_--==-_-_-
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 87. Unit 2 is in Mode 1 at 100% power.
A DC ground has occurred causing alarm "2A DC POWER PANEL LO VOLTAGE" (209 C-3). The Equipment Operator reports that voltage on 20021 is 90 VDC. Based on the above, (1) what procedure will be used to address this condition, and (2) what action will be required? E-26, Sh 1 "125/250VDC System -Unit 2" is PROVIDED SEPARATELY. A. (1) SO 57B.1-2 '125250 Volt Station Battery Charger Operations' (2) verify HPCI is operable and restore RCIC to operable status within 14 days. B. (1) SE-13 'Loss of a 125 or 250 VDC Safety Related Bus' (2) a Tech Spec 3.0.3 shutdown is required. C. (1) SE-13 'Loss of a 125 or 250 VDC Safety Related Bus' (2) verify RCIC is operable and restore HPCI to operable status within 14 days. D. (1) SO 57B.1-2 '125 250 Volt Station Battery Charger Operations' (2) a Tech Spec 3.0.3 shutdown is required.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer---Key
"~-~~--------- ... ~. -,---~----------------- ....
Question # 87 SRO Choice Basis or Justification Correct: B Correct - 20021 is the Division I battery. RCIC is powered from the Division I battery. The 'A' Core Spray Loop and the 'A' RHR pump are INOP requiring a Tech Spec 3.0.3 shutdown. Distractors: A Incorrect - 20021 is the Division I battery. RCIC is powered from the Division I battery. Since 'A' Core Spray loop and 'A' RHR pump are INOP requires a 3.0.3 shutdown not just a 14 day action statement. SO 1-2 does not resolve the low ,,_ol!§lf]~Lground condition. C Incorrect - 20021 is the Division I battery. HPCI is powered from Division II. Incorrect - 200212 is the Division I battery. HPCI is powered from Division SO 578.1-2 does not resolve the low voltage/ground condition. Psychometrics Level ofl~no!Vled~_J SRO HIGH
- 10CFR55.43(b){2)
Source Documentation Source: cgj New Exam Item D Previous NRC Exam D Modified Bank Item D Other Exam Bank _Dl~I~)(~rn Bank... ............ ____ R_~ference(sL___ SE-13,.I~sh_S.£ec 3.5 Learning PLOT-5057 Obj. 10 Objective: KIA System 263000 DC Electrical Distribution . Importance: SRO 3.2 KIA Statement A2.01 - Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of tt1g~e abnormal cOI1~Jtions or ~Reratio"-s~ Grounds
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Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 88.
- A Reactor startup is in progress on Unit 2.
- The Reactor Mode Switch is in Startup.
- Control Rods are being withdrawn when Alarm 210 H-5 "24/48 VOLT BUS 2E-2G TROUBLE' is received.
- An Equipment Operator sent to investigate reports that there are 0 volts at pane12AD045.
Given the above conditions, which one of the following is the correct action with respect to the Reactor Startup? A section of print E-24, 'Single Line Diagram + 24VDC Power System' PROVIDED ON NEXT PAGE. Technical Specification 3.3.1.1 "RPS Instrumentation" is PROVIDED SEPARATELY. A. Bypass the affected WRNM and continue the startup. B. Continue rod withdrawal and exit MODE 2 within 12 hours. C. Place the associated trip system in trip within 12 hours. D. Begin a shutdown and be in MODE 3 within 12 hours.
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Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 r-------------- - - - ----------------- --------- I Answer Key Question # 88 SRO
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Correct: Choice C Basis or Justification Correct - This malfunction will make 4 WRNMs inoperable. Only one can be bypassed. The other 3 WRNMs would produce a control rod block and not allow a continued startup. Tech Spec 3.3.1.1.A.1 or A.2 requires that the trip system be _____ place~Jn_1rlf> within~ hOlJ~~: ____ _ Distracters: A Incorrect - Only one WRNM could be bypassed. The other 3 WRNMs would produce a control rod block and not allow a continued startup. f----------------- -- Incorrect - Control rod withdrawal cannot continue because of the control rod I block. 1----* D Incorrect - Mode 3 is not required until the completion times of Tech Spec 3.3.1.1.A.1 or A.2 cannot be met. I 1 __- ___________________ _ Psychometrics L~y_~QLKnowl~cla~_ qifficult}' - - - Time_AII2wance(minutes2 SRO HIGH 10CFR55.43(b)(2) Source Documentation Source: [g] New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 _OI!:~~~~Il1B~n~ ______________ _ Reference(s): Tech Spec 3.3.1.1 ARC 210 H-5, 24/48 volt bus 2E-2G Trouble E-24 'Single Line Diagram + 24VDC Power
- - - - ------ ~--------------- --------------
Learning PLOT 5060C Obj 10.a Objective: KIA System:
--- - -----r-- 2150031RM Importance: SRO 4.3 K/A Statement:
2.4.45 Emergency Procedures / Plan: Prioritize/Interpret annunciator/alarm.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 89. Unit 2 is operating at 100% power with both RPS M-G sets in service.
- Annunciator 208 E-1 RPS 'A' M-G SET TROUBLE OR IN TEST alarms.
- Subsequent investigation in the E-12 Room indicates that RPS 'A' M-G Set output circuit breakers AC757A and AC757C have a loss of DC control power.
For an actual undervoltage condition, the 'A' RPS M-G Set output breakers (1) automatically trip, and based on this condition (2) Technical Specification 3.3.8.2 "RPS Electric Power Monitoring" is PROVIDED SEPARATELY. A. (1) will (2) no action is required B. (1) will (2) DC control power must be restored within 72 hours C. (1) will NOT (2) 'A' RPS must be transferred to the alternate source within 1 hour per SO 60F.6.A-2 "Transferring RPS Power Supplies" D. (1) will NOT (2) 'A' RPS must be transferred to the alternate source within 72 hours per SO 60F.6.A-2 "Transferring RPS Power Supplies"
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Question # 89
.................. ~-- .. -.------~-----
Basis or Justification .................. - ... ~-~.- .. _. Correct: Correct - The loss of DC control power prevents the RPS M-G Set output circuit C breakers from functioning. The RPS M-G Set output circuit breakers are in series. With both RPS output circuit breakers unavailable, auto trip capability is lost Tech Spec 3.3.8.2. Condition B requires 1 hour to remove the RPS M-G Set from service which will first require the A' RPS to be transferred to the alternate source per SO
. ..~f§!rring RPS Power Supplie~'_'_ ................ ~.
c_ _ _ _ _........................ _+--___+-"'S.:::.O:..-:F.:.:-:S.:.:...A.:.-:=..2_"T:....:r=aQ ..< . . . _ .. Distractors: A Incorrect - The loss of DC control power does prevent the RPS M-G Set output circuit breakers from functioning. Action is required for this condition. With both RPS output circuit breakers unavailable, auto trip capability is lost. Tech Spec 3.3.8.2. Condition B requires 1 hour to remove the RPS M-G Set from service which will first require the A' RPS to be transferred to the alternate source per SO SOF.S.A-2 "TrClI1~f§!rrinJL8PS Power§l!PpliE!~~____. B Incorrect - The loss of DC control power does prevent the RPS M-G Set output circuit breakers from functioning. The RPS M-G Set output circuit breakers are in series. With both RPS output circuit breakers unavailable, auto trip capability is lost. Tech Spec 3.3.8.2. Condition B requires 1 hour to remove the RPS M-G Set from service. D Incorrect - With both RPS output circuit breakers unavailable, auto trip capability is lost. Tech Spec 3.3.8.2. Condition B requires 1 hour to remove the RPS M-G Set from service which will first require the A' RPS to be transferred to the alternate ~~. . ._ _ _ _ _ __'__._ .._._... .1....-'.-'-..:......-'-c...L.=.:.§Q.£:>.Qf=.:~A-2 "Transferring RPS Power Sl.lJ.>p.li~s". Psychometrics Level ofJSl"lowledge ---c-_._ _ _D __iff_iC_I.l_lt..Lym ____ m.l, Time AliowancemLminu!t3~L Lm SRO HIGH I . 10CFR55.43(b)(5) Source Documentation D New Exam Item 0 Previous NRC Exam: 0 IZl Modified Bank Item 0 Other Exam Bank: () ~...................... _ _ _ _ _--L... _~~. ILT Exam Bank Reference(s): TS 3.3.8.2, E-2365
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Learning 5060F Obj 8 Objective: KIA System: 2000 RPS SRO 4.0 2.4.50 - Emergency Procedures I Plan: Ability to verify system alarm setpoints and operate controls
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Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 90. The following conditions are present on Unit 2 during an A TWS:
- Both CRD pumps are unavailable
- The CRS directs initiation of SBLC
- The URO performs RRC 11.1-2 "SBLC System Initiation During a Plant Event" and reports the following:
o SBLC pump discharge pressure is 1400 psig o SBLC tank level is 56 percent o RWCU is isolated Per T -101 "RPV Control" which one of the following is correct for these I conditions? A. SBLC is injecting; monitor SBLC tank level per T-101 step RC/Q-16. B. SBLC is NOT injecting; perform T-210 "CRD System SBLC Injection". C. SBLC is NOT injecting; perform T-211 "CRD System Non-enriched Boric Acid and Borax Injection". D. SBLC is NOT injecting; perform T-212 "RWCU System SBLC Injection". I L- ( YES) (NO) L
---------"\,
- RC/Q-1 S rr:------
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------~
SBle TANK lEVEL DROPS TO O~ I INJECT BORON INTO THE RPV USING:
- CRD VI A THE SBle TANK ( T-210)
- eRO VIA THE CONDENSATE PRECOAT U;_N_~:'T':~B~=:: _ _ _ _ ~ TANI( ( T-211)
- RWCU VI A SBlC TANK (T -212) t t,1-------------....
L RC/li1-16 L RC/Q-17 11 ....... - - - - - - - - - - - - - I Y
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 90 SRO
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Choice <<<--~-- Basis or Justification
----~
Correct: D Based on the given conditions, SBlC is not injecting into the RPV: 1400 psig pump discharge pressure indicates the SBlC pump discharge relief valve is lifting (due to a blocked flow path). T-210 and T-211 cannot be performed without at least one CRD system pump available. Therefore, T -212 is the only option available, which can be implemented even tho~gh RWCU is isolated. Distractors: A Execution ofT-101 step RC/Q-16 is based on SBlC injecting into the RPV. Based on the given conditions, SBlC is not injecting into the RPV. B The applicant must know that T-210 cannot be performed without at least one CRD system pump available. In other words, use of T-21 0 requires CRD piping and an available CRD pump. C The applicant 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 and an available CRD pump. Ps chometrics level of Knowledg~ Time Allowance (rTlinutes) SRO HIGH 10CFR55,43(b)(5) Source Documentation Source: New Exam Item ~ Previous NRC Exam: (PB 2009) o Modified Bank Item 0 Other Exam Bank: 0
----l-~D-l~T_Ex~r:!l~~I'l!<
B~fe!~n"~E~{~)~__ <<J:1Qland__E3~~e!S;~P&IQM:35!!,~§~eet 1 learning PlOT-5011-4h Objective:
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KIA System: I 211000 - Standby Liquid Control Importance: SRO ________ << _______ L __ 3.4 KIA Statement: A2.04 - Ability to (a) predict the impacts of the following on the Standby Liquid Control System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow. REQ!JIRJ;Q MATERIAL_S:-=_fNO~-~---- ~- -- ---~ Notes and Comments:
- ~- ~ - -
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 91. Unit 2 was operating at 100% power when a low RPV level transient occurred.
- HPCI initiated on low RPV level and immediately isolated on a steam supply line break.
- RCIC initiated on low RPV level and is injecting at 400 gpm.
For the above condition, determine the effect on Reactor Building Ventilation and the correct follow-up action, if any. Assume 5 minutes has elapsed since the low level transient occurred. Reactor Building Ventilation has (1 ) and _....>.{=..2},----, A. (1) isolated (2) direct use of GP 8.B, "PCIS Isolation - Groups II and III" to reset the isolation. B. (1) isolated (2) direct use of T-222 , "Secondary Containment Ventilation Bypass" to restore Reactor Building ventilation. C. (1) NOT isolated (2) no actions are required. D. (1) NOT isolated (2) direct alignment of SBGT per SO 9A.1.B, "Standby Gas Treatment System Manual Startup" to provide an elevated release path.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Question.# 91 SRO Choice Basis or Justification Correct: B Correct - The RPV low level condition will cause a GRP III isolation. The GRP III isolation will cause a loss of RB ventilation and a rise of the main steam line temperatures. These temperatures will rise above the alarm setpoint and require entry into T-103 "Secondary Containment Control". This will allow the crew to use T-222 "Secondary Containment Ventilation Bypass". On a low RPV level transient RCIC alone will not provide enough flow to recover RPV level to > +1 inch until 15 to 20 minutes after the f---_....._ _ _ _ _ +._ _ ~_ _+.....::s....:.c.....:ra:.c...m:...:c.-~....-...- - - - - - ..........._ .._ _.__. A I Incorrect - with the Group III isolation signal still in with RPV level < +1 inch, GP-8.B, "PCIS Isolation - Groups II and III" could not be used to reset
.- - f . _........+ the isolation.
C Incorrect - Reactor Building ventilation will isolate on the Group III isolation signal of RPV level < +1 inch. o - Reactor Building ventilation will isolate on the Group III isolation RPV level < +1 inch. Psychometrics
. Ley.el otKnowledge I j)ifficulty .. __~ ~Time ~!!2'y!~nce Jmifl.tJt~~L SRO HIGH 10CFR55.43(b)(5)
Source Documentation Source: L8J New Exam Item Previous NRC Exam D Modified Bank Item D Other Exam Bank _~ _ _ . __ DJLT Exam B~I1!._. Reference(s): I T-222 "Secondary Containment Ventilation Bypass" _____+.I-:.1Q~"Secondary COl}tainment Con~r~_.._'._._ Learning PLOT 2103 Obj 7 OJ:ljective: _ _ ~-------.- - -
-*--*--r---**
KIA System I 288000 Plant Ventilation I Importance: SRO L. ..______.-__._.-.. . ..1 ___._. 3.6 KIA Statement A2.02 - Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions Low water level REQUIRED MATERIALS: Notes and Comments: .:..::..=.:--. .
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 92. A refueling outage is in progress on Unit 2.
The following conditions exist:
- A core reload is in progress in accordance with FH-6C "Core Component Movement - Core Transfers".
- Core reload verification is in progress in accordance with NF-AA-330-1001 "Core Verification Guideline".
As the refuel Floor SRO you are asked to verify the orientation of a four cell section of the core that is displayed on the following page. Per NF-AA-330-1001 "Core Verification Guideline" and based on the given fuel cell diagram on the following page, which of the following (1) actions, if any, is required and (2) what is the core reactivity concern, if any? A. (1) no action required (2) no issues with the core reload B. (1) suspend core alterations ONLY (2) negative reactivity insertion C. (1) submit a Level 3 Reactivity Management Event ONLY (2) core thermal limit violation D. (1) suspend core alterations and submit a Level 3 Reactivity Management Event (2) overheating of the fuel clad
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
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Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer i Question # 92 SRO I Choice Basis or Justification
....- ... ~-.-- .. ~- ..- - ... ~~- .. ~.--- ... ~---- .. - _._----_.". -
Correct: D Correct - The fuel bundle in the lower right hand corner is not oriented correctly (bale handle indicator is pointing away from the control rod). Suspend core alterations lAW FH-6C and initiate a Level 3 Reactivity Management Event lAW NF-AA-330-1001. Positive reacti vity insertion, core thermal limit violation, and fuel clad overheating are al I concerns of a I mis-oriented fuel bundle lAW FASR 14.5 and Appendix J and NF-AA-330 I** i 1001. I Distractors: A Incorrect - Actions are required. See D above. I ... ---"~ ..- .. -.-...- ... - - _.... _.-_.._---- . _..__ .. __. _... .. ---- .... B Incorrect - Suspend core alterations lAW FH-6C AND initiate a Level 3 Reactivity Management Event lAW NF-AA-330-1001. Negative reactivity insertion is NOT a concern. A core thermal limit violation is a concern of a i mis-oriented fuel bundle lAW FASR 14.5 and Appendix J and NF-AA-330 1001. C Incorrect - Suspend core alterations lAW FH-6C AND initiate a Level 3 Reactivity Management Event lAW NF-AA-330-1001. Fuel clad I overheating is a concern of a mis-oriented fuel bundle lAW FASR 14.5 and i
._- Al2pendi~ ancJ~F-AA-330-1 001~ ____ ~_. . ...1
__.l::~yel ofjSnowledge HIGH _ ......__[)ifficulty
.-r Psychometrics i
Ti!"e Allowanc" (minll!es)j-1-()C~~~3(b)(6-I &7) Source Documentation Source: ~ New Exam Item D Previous NRC Exam 0 D Modified Bank Item D Other Exam Bank D ILT Exam Bank Reference(s): "Core - Core Transfers" NF-AA-330-1001 "Core Verification Guideline". Learning i PLOT-1535 Obj 2 l Opjectiv~~_~----L.______..__.....__. __ ..... ___._.. .__ --.. .- .. i~- KIA System...._ _ 234000 Fuel Handling Equipment Importance: SRO 3.7 KIA Statement K5.05 - Knowledge of the operational implications of the following concepts as they apply to FUEL 1::I.~NDL1!'JG EgUIPry1J=NT:£uel orientation . ._ _........___...._._ ..__......_.._....
.-E_-. _ _ ~~~~~~~::~::~Al.l!~_=-EN_ _~~:-_- =-~-_--_.==.-. -=-.._.-. .-_-..-_..--..
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 93. A loss of cooling to the Off-Gas Recombiner Condenser has occurred.
Using the chart determine the appropriate actions. Assume the loss of cooling began at T =O. Recombiner Condenser Pressure 15 10 0') 5 "Ci) Q. 0 12 16
-5 .
-10
-15
-20 A. The recycle valve failed to open; open the recycle valve per the ARC and return the Jet Compressors to service using AO 8.1-2, "Recovery from Off Gas System Isolation".
B. The recycle valve opened and is returning condenser pressure to normal; continue to monitor operations of the Off-Gas system per SO 8.8.A-2, "Off Gas System Routine Inspection". C. MO-2990A, "Steam Supply" has isolated; swap Off-Gas Jet Compressors using AO 8.1-2, "Recovery from Off-Gas System Isolation". D. MO-2990A, "Steam Supply" has isolated; reduce reactor power using GP-9 2, "Fast Reactor Power Reduction".
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key ~~~~-~~-~~--~-'~~'~------~-~~-~----~----~'~---~B~a~~o~r~~iC~~ ~.~---- --- Correct: o CORRECT - The recycle valve opened as indicated by the flat spot on the curve at approx. 7 psig. The rise in pressure indicates that the recycle valve was not enough to control Recombiner Condenser pressure. When Recombiner Pressure reaches 8 pSig, MO-2990 isolates. There are no alternate components in the Recombiner System that can be placed in service for this condition. This will cause main condenser vacuum to drop and require entry into OT-106 "Condenser Low Vacuum" and require a reduction. Distractors: :1 INCORRECT - The recycle valve did open to try and control pressure as evidenced by the flat spot on the graph. Returning a jet compressor to
- .. ~-- ....... -)~ B r~~~~~;~~~~~r;~ee~;c;c~:::~:::'not~-uc~essfully controlling
[ [ReCOmbiner pressure as indicated by the rise in system pressure to 8 psig
~____ andJ!1en the rapid droQ.?ls 1\,10-299Q iso!at~fL_ __ __
r------ I [" C [' INCORRECT - The MO-2990 is isolated but there are no alternate components in the Recombiner System that can be placed in service to
" restore the system.
Psychometrics _~~vel ()fJ5rlowledg~_L __._._~Difficul!L_~~_~ILf'!I~ AIIO_v.-'§1Il~I3_(minutes) . ~ SRO HIGH 10CFR55.43(b)(2) Source Documentation Source: [8J New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank Reference(s): SO 8.8.A-2 "Off-gas System Routine Inspection" AO 8.1-2 "Recovery from Off-gas System Isolation" OT-106 "Condenser Low Vacuum" GP-9-2 "Fast Reactor Power Reduction" Learning PLOT 5008 Obj 9d
-~~e~:~:m: "-'-""j"2710-0-0-0-ff--g-a~--" "l,~p~rta~ce:sRo-
__.. .~_ . _ . . ____.__.. .___. __ . . _______ . . ~___ . . __.__.__ . . ~._. . __~_._~ ____L________ . 4.2 KIA Statement: 2.1.25 - Conduct of Operations: Ability to interpret references materials, such as graphs, curves, tables, etc. REQUIRED MATERIALS: Notes and Comments: C~ _ _ _ _.-L~=::~ ... ~_~.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 94. FH-6C, "Core Component Movement - Core Transfers" is in progress on Unit 3.
Per FH-6C, which of the following activities can continue without SBGT being in compliance with the requirements of Tech Spec 3.6.4.3 "Standby Gas Treatment System"? A. OPDRVs B. Core alterations C. Handling of Fuel Casks in Secondary Containment D. Movement of recently irradiated fuel assemblies
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 94 SRO Choice Basis or Justification Correct: C Correct - Procedure FH-6C, "Core Component Movement - Core Transfers" lists the activities that require SBGT to be operable. The handling of fuel casks in Secondary Containment can continue without _~~BG_T ____.~_ .....____~ ___.. . _~_ Distractors: A Incorrect - OPDRVs are prohibited without SBGT per Tech Sp ec 3.6.4.3 "Standby Gas Treatment System" and FH-6C. B Incorrect - Core alterations are prohibited by FH-6C, "Core Component Movement - Core Transfers". o Incorrect - Movement of recently irradiated fuel assemblies are prohibited without SBGT per FH-6C, "Core Component Movement - Core Transfers". Psychometrics
._!-evel_()l!5!1s>w~dge~ __ SRO MEMORY 10CFR55.43(b )(3)
Source Documentation Source: ~ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank
---==----- ... --~.------ ~::;~~;gceisL_ I :~~i~~~~r~~;;cl'Q~ent Movement - Core Translers-"_
O~j~ctive: KIA System . f~neric--_conduct:1 o:ations SRO 3.9 KIA Statement: G 2.1.40 _ Knowl~gge...Qf r_efueJiI1~.Ci.c:lmini~trativei~quirements .REgUIRED M_ATI:~IAh.s-=-- _~ l.PtJONE __ Not~~ and CQD1m.ents:._____]. ~~ _ ____ ___
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 95. Which one of the following activities requires a Temporary Configuration Change (TCC) per CC-AA-112, "Temporary Configuration Changes"?
A. Installation and removal of a jumper in accordance with an approved surveillance test procedure. B. Changing a Control Room alarm setpoint that is NOT in direct support of a Maintenance Work Order. C. Installation and removal of Measurement and Test Equipment (M&TE) in accordance with an approved surveillance test procedure. D. A temporary configuration change included with an Operations Clearance that does NOT affect the system beyond the clearance boundary.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
... _ . _ ........ ~.~-.--
Answer Key
......- ... - -........ -~--.
Question # 95 SRO Choice Basis or Justification Correct: B Per CC-AA-112, this is NOT an excluded activity and therefore requires a TCC. Distractors: A Per CC-AA-112, this is an excluded activity and therefore does NOT require a TCC. C Per CC-AA-112, this is an excluded activity and therefore does NOT require a TCC. D Per CC-AA-112, this is an excluded activity and therefore does NOT require a TCC. Psychometrics r----------------,---------------~----c~------------------~-----------*--- Level ot I5I'lQ!VJegge........ _,__~._ J::>.Lffl(;ulty~__ .. i~TimeNl()wanc.~jminutE;~) I SRO Memory I I ! 10CFR55.43(b)(3) Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: 0
. --~I*
. ILT Exam Bank ~.- ...--...
RE;f.eren(;~s):__
- CC-AA-112 Learning i PLOT-1570-19 5?bje~~"'.~~ . -.----1 ..__ . .~. _.
KIA System: Generic - Equipment Control SRO 3.2 KIA Statement: G 2.2.5 Knowledge of the process for making design or operating changes to the facility. REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 96. An Equipment Operator needs to enter a locked high radiation area for 15 minutes to manually operate several Primary Containment Isolation Valves to satisfy a Tech Spec required action.
- The dose rate in the work area is 6000 mR/hr.
- The Equipment Operator's present annual TEDE is 800 mR.
What is the highest level of authorization required to complete this task? A. Unit Supervisor B. Shift Manager C. Radiation Protection Manager D. Plant Manager
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer
-~"-"~-~-- ........
Key
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Question # 96 SRO Choice Basis or Justification Correct: C Correct - The EO's TEDE will rise to 2300 mR after completing the task. Per RP-AA-203, the Admin Dose Control Level (ADCL) is 2000 mR. The procedure requires approval of the Work Group Supervisor and Rad Protection Manager to raise the worker's ADCL between 2000 mR and 3000 mR TEDE. Distractors: A Incorrect - While RP-AA-203 requires approval of the Work Group Supervisor, it also requires the Rad Protection Manager to raise the worker's ADCL between 2000 mR and 3000 mR TEDE. B i Incorrect - For purposes of dose extension, the Shift Manager would be the equivalent of the Work Group Supervisor. RP-AA-203 also requires the Rad Protection Manager to raise the worker's ADCL between 2000 mR and 3000 mR TEDE. D Incorrect - Plant Mgr approval is required for raising ADCL to between 3000 mR and 4000 mR. Not the case here. l_ Psychometrics Level ofK:n()'JVJ~~g~_ Difficytty ___~_~ _TllPe _,,!I()!ya_I'lc:~{minutes) ! SRO 1 HIGH i 10CFR55.43(b)( 12) Source Documentation Source: [2J New Exam Item 0 Previous NRC Exam: 0 Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank Reference(s): __________________..._________ ~~~_ I RP-AA-203
.._+M--- ._______________...________ Exposure
~_ ......___ Control and Authorization Learning PLOT-1730 Obj 4 Objective:
KIA System: None SRO 3.7 KIA Statement: 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 97. Refer to the photographs on the following page.
Based on the conditions shown, which one of the following describes (1) the status of the Stator Water Cooling (SWC) System and (2) the required action? A. (1) A loss of SWC exists (2) Reduce power per GP-5, "Power Operations" B. (1) A loss of SWC exists (2) Reduce power per GP-9, "Fast Reactor Power Reduction" C. (1) A loss of SWC exists (2) Perform GP-4, "Manual Reactor Scram" D. (1) A loss of SWC does NOT exist (2) Continue monitoring per OT-113 "Loss of Stator Cooling"
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 GE N AMPS RANGE SEL GENERATOR flROTEC [,ON CIRcurr ENERGIZED
Peach Bottom Initial Senior ~eactor Operator License NRC Examination April 2013 Answer Key t* B Incorrect - GP-4 "Scram" is required, not GP-9.
-~QlI* Inc~~~~ct=*the-~Iar~s receiv-~d-d~t~';'ake ~*v~iidloss of SWC and with main generator amps> 9,480 amps, a GP-4 scram is required.
.. ~ - .... ~------------------ .. -- - - - - - - - - - - - - -.. _------ ------- --- ---- -
_..level of~!1()\l\Ilec!gE: __~j---~ __~_ _ _ Time Aliowanc~lQ!i!ll.!te.s) SRO HIGH 10CFR55.43(b)(12) Source Documentation Source: [8J New Exam Item Previous NRC Exam: 0 o Modified Bank Item Other Exam Bank: 0 ILT Exam Bank -~-~-- ....-.. R~f~renc~{!>): OT-11 "LQ.s~ of ~tatorCo-,,--lil"1g:. Learning PLOT -5050A Obj 6 Objective: KIA System: None Importance: SRO 3.7 KIA Statement: 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. RE.QUIRED MATERIAL§-=-~l~ONE Notes and comments:~_ILmb~c:I...f>.tlQtos.__.. . ._____ __.. . _........
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 98. Unit 3 is operating at 100% power The following indications are observed:
- Main Steam Line radiation monitors (RR-3-17-252) indicate 1.3 E+3 mR/hr.
- Vent Stack Exhaust radiation monitors (RR-3979) indicates 3 E-7 !JCi/cc.
- Air Ejector Discharge radiation monitor (RR-3-17 -152) indicates 7.5 E+2 mR/hr.
- Main Stack Gas radiation monitor (RR-0-17-051A) indicates 3.7 E-6 !JCi/cc.
Which one of the following describes the reason for the above indications and what procedural guidance is required to be directed? A. A resin injection has occurred; lower power in accordance with GP-9-3, "Fast Reactor Power Reduction". B. A resin injection has occurred; lower power in accordance with GP-9-3, "Fast Reactor Power Reduction" rods ONLY. C. Fuel cladding damage has occurred; lower power in accordance with GP-9 3, "Fast Reactor Power Reduction". D. Fuel cladding damage has occurred; lower power in accordance with GP-9 3, "Fast Reactor Power Reduction" rods ONLY.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 r~~~------~-----********* ..
~--~---- ----~~----~-
Answer Key Question # 98 SRO Choice Basis or Justification Correct: C - All the radiation monitors are reading normal full power background except the Steam Jet Air Ejector (SJAE) Discharge radiation monitor (RR-3-17 152). It is reading higher than normal. Failed fuel causes the release of fission product gases (Xe, Kr, I) into the reactor coolant. Fuel leaks do not cause Main Steam Line radiation levels to rise. The % life of Xe and Kr are long enough to indicate on the SJAE discharge radiation monitors. Alarm 318 E-1 "AIR EJECTOR DISCHARGE RADIATION HIGH-HIGH" requires reducing reactor power using GP 9-3 as required to maintain off gas discharge radiation levels below 700 mr/hr as read on RR-3-17-152. The GP-9 power reduction would include use of inserting
-----------,- control
~.--~--------.
rods flow limit is reached. A Incorrect - The injection of a resin into the reactor will cause a rise in N-16 activity in the main steam lines. During operation, the dissolved 02 in the reactor reacts with the N-16 to form nitrates (N03). N03 is soluble in water and does not readily carry over with the steam. A change in pH causes the N-16 to combine with the free hydrogen to produce ammonia (NH3) and nitrous oxide (N20). Ammonia and nitrous oxide are more volatile; therefore more N-16 carries over with the steam. rise in N-16 only indicates on the main steam line radiation monitors because
~ ~~~.. __ ~
the _ _ _ _._ short ___ ~c ...... life of the N-1 D Incorrect - Alarm 318 E-1 "AIR EJECTOR DISCHARGE RADIATION HIGH-HIGH" requires reducing reactor power using GP-9-3 as required to maintain off gas discharge radiation levels below 700 mr/hr as read on RR-3-17-152. The GP-9 power reduction would include use of inserting control rods once the core flow limit is reached Psychometrics
.!:~,!~I()Lt<:l1owledge SRO HIGH 10CFR55A3(b)( 5)
Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank ....._--_. ReferenQ~(s) : ARC 318 E-.1.:._ ~__. Learning PLOT-5008 Obj 7i Objective: KIA System: None SRO 3.1 KIA Statement: 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring
~:t~~~=~:~ri~nt'*!N9NE~_.~~ __._.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 99.
- Unit 2 has experienced an ATWS condition.
- The HPCI System is injecting into the RPV at 2500 gpm.
- HPCI turbine speed is 2200 rpm.
- HPCI suction is lined up to the Torus
- RPV level is -195 inches and steady.
- Torus temp is 210°F and slowly rising.
- Torus pressure is 2.5 psig and slowly rising.
Using the HPCI NPSH Limit Curve below, based on the above conditions, the HPCI system pump NPSH is on the side of curve and the Control Room Supervisor should direct (2) A. (1) safe (2) maintaining HPCI flow at the current value B. (1) unsafe (2) maintaining HPCI flow at the current value C. (1) unsafe (2) lowering HPCI turbine speed to return operation to the safe side of the NPSH curve D. (1) unsafe (2) securing HPCI; any further turbine speed reduction will cause system damage LL 0 260 * . . . * . . , . , TORUS PRESSURE
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HPCI FLOW (GPM)
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 99 SRO Choice .. _-_
.... --.--.. -.---~---
Basis or Justification Correct: B - Torus temperature of 210°F and Torus pressure of 2.5 pSig places system operation above the lowest curve on the NPSH curve, which is on the unsafe side of the curve. Due to the ATWS condition, RPV level was lowered per T-117 "Level Power Control" to -195" which supports Adequate Core Cooling (ACC). If the HPCI system speed is lowered, RPV level will lower and ACC is no longer assured. T-117 step LQ-18 and/or LQ-19 directs exceeding NPSH curves in order to maintain RPV level no lower than -195". Distractors: A Incorrect - Torus temperature of 210°F and Torus pressure of 2.5 psig places system operation above the lowest curve on the NPSH curve, which is.... on the unsafe side of the curve.
~--\.~-~~.---\~.=
C Incorrect - Torus temperature of 210"F and Torus pressure of 2.5 psig places system operation above the lowest curve on the NPSH curve, which is on the unsafe side of the curve. However, lowering HPCI turbine speed, and thereby lowering system flow, will not bring system operation below the lowest NPSH curve. Incorrect - Torus temperature of 210°F and Torus pressure of 2.5 psig places system operation above the lowest curve on the NPSH curve, which is on the unsafe side of the curve. If the HPCI system is secured RPV level will lower and ACC is no longer assured. T-117 step LQ-18 and/or LQ-19 directs exceeding NPSH curves in order to maintain RPV level no lower than -195". Psychometrics Level of Kl1owled9.~J_ DiJficultY. _____ 1 Time AliowanceJl!lil'lutes) .~ ..._. SRO HIGH I 10CFR55.43(b)(5} Source Documentation Source: [gI New Exam Item Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0 ILT Exam --- Bank_.------------- Reference( s): T -117 "Level Power Control" step LQ-18 and LQ-19 T -102 Containment Control" Sheet 3 Learning PLOT-2117 Obj 5 Obj~~!i\l~: .. ~..... KIA System: None Importance: SRO
~~----~ ...*.. --.-.-~- .-~---.
4.0 KIA Statement: 2.1.32 Ability to explain and apply all system limits and precautions. REQUIRED MATERIALS-=-~~_~__ -[NONE_ -___ Notes and Comments: ~ _~ ~
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 100. During the performance of ST-0-032-301-3, "HPSW Pump, Valve, and Flow Functional and In-service Test" the 3C HPSW pump discharge differential pressure was in the ALERT Range. Per ST-0-032-301-3, the 3C HPSW pump _->-':"'--_ and status is tracked by _ . . . . l . : . L - - _ . A. (1) is immediately declared inoperable (2) a short duration time clock entry B. (1) is immediately declared inoperable (2) T.S. 3.7.1 Condition A - one HPSW subsystem inoperable C. (1) remains operable (2) initiating an issue to place the pump on increased test frequency. D. (1) remains operable (2) Potential Tech Spec Action (PTSA) entry
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 100 SRO Choice Basis or Justification Correct: C
- Correct - per limitations step 4.3.6 of the ST. If any pump has test results I in the ALERT range, the pump remains operable. Initiate an Issue to place
______+-tbt=~uIllP_~~ in9rt=ase~Jt=~t frt=qlJt=Jl~Y_ Distractors: A Incorrect - Pump is not declared inoperable until performance reaches the Action Range. A SDTC entry is not required per the ST. Only for SSCs ___~__+ . . . __._____+ t!l~!~e_l~~________.________ ___ ______________ _ B Incorrect - Pump is not declared inoperable until performance reaches the I Action Range. A SDTC entry is not required per the ST. Only for Systems, Structures, and Components that are inoperable. T.S. 3.7.1 only entered if
.. _ I Fumpi~lnoperC3b!t=. _ .......,________
D Incorrect - The degraded pump condition would not be tracked by a Potential Tech Spec Action (PTSA) entry. The PTSA would not need to be entered until the HPSW pump is declared inoperable. Psychometrics ......_J-.~v~IQfJ<119~!edg~ __ L_________Q~ffiC.lJI!Y_____ --1 ..Iir'Tle.6J!9'Na nC~(lllinutes L SRO ---------_ .. MEMORY I 10CFR55.43(b )(5) Source Documentation Source: [gI New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank Ref~rel1_C::~(~L __ ST-O-032-301 "HPSW PlJlll~_\ICiI\fe-'_andF=.loV\lEul1c::tional and In-!5ervice Test" Learning PLOT 5032 Obj 9 Objective: KIA System: Importance: SRO 4.3 KIA Statement: G 2.2.14 - Knowledge of the process for controlling equipment configuration or status. R_~g~ILREI?MATERIALS:=_ I N9~i---~~ __:: ----~-~~~~- Notes and__Q()mm~I1~: _______L. ______ _}}