ML13098A256
| ML13098A256 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 04/01/2013 |
| From: | Todd Fish Operations Branch I |
| To: | Exelon Generation Co |
| Jackson D | |
| Shared Package | |
| ML12328A009 | List: |
| References | |
| TAC U01867 | |
| Download: ML13098A256 (214) | |
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 SEPARATELY, what is the effect, if any, of this condition on the RPS low RPV level scram function. A. lE 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 KE~y Answer........_
-'¥-~--~-- ---"--
Choice Basis or Justification Correct: A CORRECT - LT -101 C & D will sense a HIGH level. LT -101 A & B are still available to detect and initiate scram on actual LOW level condition. Distractors: B INCORRECT - While LT-101C & D will sense a HIGH level, LT-101A & B are still available to detect and initiate scram on actual LOW level condition. Plausible if candid~te _misunQI?T!?tands R1:§I()~gl~~rrarlgel1'1l?l1t. C INCORRECT - Instruments would sense HIGH level- Plausible if candidate misunderstands level detector theory of operation. D ~NCORRECT - 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 ___ ~ _____ arrangement of instrument lir1~~s and condel1sing (;~amber lines. Psychometrics Level of Knowl~~cl9~_L __ " " " " " " " Difflcul!y___ _" ~ Time Allowance (minutes)
" - ~""""""""""""" "" ---~ c-RO HIGH I 10CFR55.41 (bll?l Source Documentation Source: D New Exam Item Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0
_01"--T~~am Bank (4397) "" --""""""""""""----"""" " ------" Referen~I?{!:))~ _____ t-_Mc:~-3=~5_=2_S==~h=-::2=--_"""""""""""""""""__~_____ Learning PLOT-5002B-3d Objective: KIA System: 212000 RPS Importance: RO 1 SRO 3.7/3.9 KIA Statement: K1.02 - Knowledge of the physical connections andlor cause- effect relationships between REACTOR PROTECTION SYSTEM and the following: Nuclear boiler instrumentation ~~;;;~~~:;::~:L~:-~~~~~ Sh 2Requlred~- ..
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 (1) 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 Answer Question # 2 RO Choice Basis or Justification
-----------------~~- -
Correct: D CORRECT - If a PCIS Group" isolation signal is received while any TIP detectors are outside of their shield, the detector(s) will withdraw to the "in 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, 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. C INCORRECT - Shear valve does NOT automatically actuate. Psychometrics LeveLoLl5r'loV\iledge __ ...... ____j)lfficultY.-__ ~___bJ.t:neJ\uo.wa!1ce (minl!tes) _ ---~--- .. RO- - - - - - ----- MEMORY 10CFR55.41 (b)(9) Source Documentation Source: New Exam Item ~ Previous NRC Exam: (PB 2009)
~ Modified Bank Item 0 Other Exam Bank: 0 IL T Exam
........ --~-
Bank Reference s : GP-B.B COL, SOlEX~ Learning PLOT-5007F-3a, PLOT-5007G-41
- Objective:
K/A System: 223002 PCIS/Nuclear Steam Supply Importance: RO / 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
~~~~~~:~:-:-:-:_~~t_~_A_:_L_S___:-_-_-_-~rl~N~O~N_*~****_E~-_-_**_-_______
-~~_---~~-_--~-*~~-------**-.-_______________.
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 20021 (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 Key Question # 3 RO Choice Basis or Justification Correct: D - Each or Not) has both normal 20D21 (Div I) and alternate 20D24 (Div II) 125V DC power to solenoids. A loss of one supply will not prevent SRV (ADS or Not) operation manually. -. - - -------~-------- ....... Distractors: A INCORRECT - Each SRV (ADS or Not) has both normal20D21 (Div I) and alternate 20D24 (Div II) 125V DC power to solenoids. Plausible because of common misconception that only ADS SRVs have alternate power. B - Each or Not) has both normal20D21 (Div I) and alternate 20D24 (Div II) 125V DC power to solenoids. Plausible because candidate may confuse the effects of the ADS logic power loss with valve nn\,",,,,,r loss. C - Each SRV (ADS or Not) has both normal 20D21 (Div I) and alternate 20D24 (Div II) 125V DC power to solenoids. Plausible if candidate makes conceptual error for ADS versus non-ADS valve solenoid power. _ --______1 Note: App~'B'Jir~2 3 SRVs are prot~~t~(:LQuEU()_c~~l:>lEU_ul1s. Psychometrics Level ot~n()\I\IIJ(jgJ Riffi~~I!Y_~~_ -----"Dme ~1I~~~n~e_(rl1i~nu=te:~s=--,--)-l---~~~~~~-- RO - HIGH 3.5 3 10CFR55.41(b)(7) Source Documentation Source: D New Exam Item cgJ Previous NRC Exam: (PB 2002) Modified Bank Item D Other Exam Bank: 0 ___~,---===--~_IL~_T Exam Ba_I1K~O__~_~ __ ~__ _~ ____ ~________ C-5 Blowdown Valves Power Monitor G - 2b KIA System: Relief/Safety Valves I Importance: RO / SRO
~~-~-----
2.8/3.2 KIA Statement: K2.01 Knowledge of electrical power supplies to the following: SRV solenoids REQUIRED MATERIALS: 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 Basis or Justification ~~~~~-~---- Correct: o SBLC Power Supplies are: 2A: E124-R-C, 2B: E224-R-B, 3A: E134-W-A, 3B: E234-R-B. Alarm DOS B4 has no relevance because no SBLC pumps are powered off the E-43 bus, Alarm 001 C 1 indicates that the E 12 bus is locked out and unavailable, so 2A SBLC pump does NOT have power to the motor. Distractors: A 2A has no power, 2B, 3A, 3B have power. Plausible if candidate does not know SBLC Pump power supplies.
~~ - ~-~~-- ~~~~~ ~~--~--~---+--~~~~~~~~~~~~~~--+-----~ ~~------ ~~~~~~~~~~---
B 2A has no power, 2B, 3A, 38 have power. Plausible if candidate does not know SBLC Pump power supplies. C 2A has no power, 2B, 3A, 38 have power. Plausible if candidate does not know SBLC Pump power supplies. Psychometrics ~~ Level of K~()Vv'le_dgf:) _--'2if!icult y . . . . . _ - - Time AliowancelminuteS)!~---: RO HIGH 1
- 10CFRSS.41(b)(7)
Source Documentation Source: [2J New Exam Item 0 Previous NRC Exam: 0 Modified Bank Item 0 Other Exam Bank: 0
~~---~~- ..- .........--~~ ..~~-'=='~.
I LT. Exam Bank
~--~-...... . ..... ~--.-
B_efete!1c;e(s t_ i PLOT S011 M-1-S-46
~b~2~__ -~.--~'
Learning Objective: I PLOT 5011 _ ....-.. - - - r KIA System: I 211000 - Standby Liquid Control Importance: RO / SRO 2.9/3.1 KIA Statement: K2.01 - Kn~vvl~Qge of tDe~!ectfriC91 RQw~~upl'lie~ 10 thel.()lIowll}~L SBLC PU!Dl's. ~ __ ~ REQUIRED
----- ~~- -
MATERIALS: NONE 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 "An 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 Question # 5 RO Choice 8asis or Justification Correct: A CORRECT - On a loss of RPS "8", the "8" S8GT train and fan will start. Drywell pressure will start to rise due to a PCIS GRP 3 isolation signal (AO ___....;.....2......5.1Ql?V\J V~r1t outboard 2"~~nt valve will clq~~t . _______ Distractors: B INCORRECT - 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 DW VentQLJt~()ard 2" \J~n! valv~'II\Iill clos~L __________ C INCORRECT - The "A" SBGT fan does not receive a trip signal on a loss of RPS"B". D INCORRECT - Drywell pressure will start to rise due to a PCIS GRP 3 isolation signal (AO-251 0 DW Vent outboard 2" vent valve will close). Psychometrics _Level of K.nowledg~_._ ... HIGH Source Documentation Source: k8J New Exam Item D Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0 ILT Exam Bank Re~r~'lc~(~L ____ J'LOI~§_Q9~A and PLOT-50_0c .7 .._G.=---____.. . . .___.__.__.._ Learning PLOT-5009A-6c Objective: KIA System: 261000 Standby Gas Treatment System RO/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 Oft-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 BeaJing Cooler.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key ~-- Question # 6 RO
-Correct:
-- Choice
-----T------ A CORRECT -.During the LOOP, TBCCW pumps trip, forty seconds later Basis or Justification RBCCW will provide cooling to essential loads which include station air compressors al1dCRD Pumps. __________
Distractors: B INCORRECT - This is a non-essential load. Plausible if candidate believes I 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 s~ndid~t~J~elieves the coolers will still have cooli!l9 wa1~!-"--- _ _ Psychometrics L~"-~L()LlSrl()"",-IE3<:!~___ - - - - - - -- _QiffJc::ulty _______ Time Allowance (minutes) -- --------- r - RO MEMORY 10CFR55.41 (b)(4) Source Documentation Source: 0 New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 [gIILT Exam--- Bank (2971) Refe~!l~e(s):___ ____ ON-118 Loss of TBCCW System:l)ro_c~sture m _ _ _ _ _ _ _ _ __ Learning PLOT-5035-3b Objective: ~A system__ J400000 com~onent c:n9 wate,__ KIA Statement: K3.01 - Knowledge of the effect that a loss or malfunction of the CCWS will have on 1:::: ---- -
~~:;;~
the following: Loads cooled by CCWS
- - - ---------------------1 REQUIRED MATERIALS:
NONE Not~sC!I"lc:LComments: ____________ ______________ _
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 APRM system 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 Key
....... --~-~-~~~ .......-
-~-~~--~ ..... -~.~
Basis or Justification Correct: D I CORRECT - APRM downscale (~3.2 %) in MODE 1 will generate a control
.* rod withdraw block and downscale alarm 211 C-2 only.
i Distra~;;s~~-IA~ I NCORRECT= APRM downscale (~3.2 %) in MO-OE1wlllgenerate 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 _ . .-tt__
~--
C
~.~~_t _ _ _....._ _ SiiTlulated ThE?rmal Power Ijig~_~~~_
. INCORRECT - A scram vote signal is only generated for:
APRM Inop Trip
. High Neutron Flux Simulated Thermal Power High Psychometrics Level of Knowl~c:l~_~_~___~ Difficul!y Time Aliowan~~-.imil}!:lteslJ~ RO MEMORY I 10CFR55.41 (b)(7)
Source Documentation Source: D New Exam Item ~ Previous NRC Exam: (2007ILT) D Modified Bank Item D Other Exam Bank: 0
. _______...._ _ ~_._~. ~~ __ __ l __ ~O I LT Examl3!r*~~__ . . . _~_~__ ~
Reference{ s): PLOT 5060, ARC 211 C-2 Learning PLOT 5060 - 3a Objective:
-~~.--~
KIA System: 215005 APRM/LPRM - TI~mportanc~:~~RO 1 SRO
....._._ _ _ _ _ _ _ _ .~~ ____..i_ _ _ _ ~ __ ._~ __ ~ __ ~_~ __ ._ _ _ ~. ___ . _._ _ _ _ _ ._.~~ .. _ _ _ _ _ _ _ __......
~ ~ __ 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
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 _-->-{1:...z..)_ _ 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 Key Question # 8 RO Choice Basis or Justification Correct: The Static Inverter is current limited. If a fault develops it will automatically 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 The Static Switch will transfer to alternate source in order to maintain 20Y050 panel energized. B 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 ~f~n~~I~_d.9.t? __ t_. ____ DifficuJW rTII'11~ Aliowanc~Jm inlJ1~sl - r----~--- RO - HIGH 10CFR55.41 (b )(7) Source Documentation Source: o New Exam Item [gJ Previous NRC Exam: (PB 2007) Modified Bank Item [gJ Other Exam Bank: (PB 2010 Cert)
... -----.~----- ------'----""=".
I LT Exam Bank R~f~~~I'1.f~t~l __ _ ARC-220 F-5 Learning PLOT-5058-5c Objective: K/A syste~:~~---GI262002 - Uninterruptible po\Al'erS~I~~ly-li.I~port.a. nce: RO/SRO (A.C.lD.C.)
- 3.1 /3.4
******* _~~ .. ____ ~_~~,,_. _______******** _~_~_.J ___ ~ _________
K/A Statement: K4.01 - Knowledge of Uninterruptable Power Supply (A.C. I D.C.) design feature(s) and/or interlocks whi~h .p.rovk!e f()IJhe following:... TrC!.n~fer fr()!!!J:>referre.<!..p.owe,r to _~I!ernate~power supplLe,s. REQUIRED MATERIALS: fNONE ___ ~ _ __ Notes and Comments: - ~~~---
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 Question # 9 RO Choice Basis or Justification Correct: B 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 __ -l---,w,,,,i,*t*h.out seal condensatep~_. 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 ___ +.. ____ !hr().l:Ish se§lls:_~ _ _ .
--1 C I 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
..-.~~-. ---t .. -----..-.-..... . on airborne contamination ..-...:...:'------...::-, .......-. ----~ .. .
D 1 Incorrect - HPCI operation is permitted to continue without Gland Seal U condenser system, as identified in SE-lO, Att 9. Plausible because the
. . cancll<:tC!te mClYJ'!Q! kno-,!, this fact. ..__ ____ ___ .._
Psychometrics Lev~l.C>i!<ilowlec!ge_ -- ---" DiffiCLJI~)I___ ~___lI'me ~11()\,/IJan~~{minuteS}j_____130 HIGH .
- 10CFR55.41(b)(7)
Source Documentation Source: r2J New Exam Item D Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0 ILT Exam Bank Reference(!;L SE-1 O,Alt~,.J..If_S.A.FiPar~ ~~4J__ _ Learning PLOT 5023 4b Objective: KIA System: 1206000 HPCI RO/SRO 2.8/2.9 K/A Statement: K5.02 - Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT IN~IECTION SYSTEM: Turbine shaft sealing REQUIRED MATERIALS: 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 vEmted?
(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
....- -.....- . . .-~ ..... ~~ .... - ....... ~.~
Choice Basis or Justification Correct: 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 _~~~.: __~..t-.system is filled c:1.!!<!,,-ente(t.~ .____.. .__... ...~......~~_ Distracters: B INCORRECT - SO 14.1.A-2, Core Spray System Alignment for Automatic or Manual Operation, verifying CIS discharge pressure in conjunction with
+._.~.....L.....J.L...~ _L.i.n
.... . e Vent Accumulator Low Level alarm is clear is r~uJred'.
INCORRECT - SO 14.1.A-2, Core Spray System Alignment for Automatic or Manual Operation, verifying CIS discharge pressure in conjunction with
....._~_.+_.._._.~.. +.v~..~~ifyi~ th~ Line .YE:lJtf\ccur'!1l.Jla.t9..rJ.-0~_L~"~~c:1rmif)c:learis~~~quired.
INCORRECT - SO 14.1.A-2, Core Spray System Alignment for Automatic or Manual Operation, verifying CIS discharge pressure in conjunction with verifyir'1..9..!hE; Lin~Vent Accumulator Low Level alarm is clear is requjred .
-.~
Psychometrics Leve.1 ()tK~no\,4Vl~c:jge_~~~_ .... Difficlj1ty ...- . . . ~ Time Allowance (minutes).... -
~ ---.~~ .--
r~-~* RO MEMORY I 10CFRSS.41 (b)(14) Source Documentation Source: !2J New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank Reference(§i)_:_.+_S~0.1~t.:1.A, ARC 224 A-4 Learning PLOTS014 - 4e Objective: KIA System: 209001 LPCS ROISRO 2.S/2.S KIA Statement: KS.OS - Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SP~A)'. SYSTEM_:§tsterrt.Venti.Qg REQ~J~~l). MA.rERI~L§:. .[NONE_._ Notes and Comments: --- - ----~
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 11. Per SO 32.1.S-2(3) HPSW System Startup and Normal Operations, under normal conditions, (eg: torus cooling for HPCI testing, Shutdown Cooling operations), a trip of the in*-service HPSW pump(s) supplying cooling water to RHR heat exchanger impacts the ability to (1)
As described in T-101 BASES, during transient operations (eg: LPCI mode), a trip of the in-service HPSW pump(s) supplying cooling water to RHR heat exchanger impacts the ability to (2) A. (1) minimize radioactive leakage (2) rapidly remove decay hea1t 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 Question # 11 RO Choice Basis or Justification Correct: A CORRECT - Per SO 32.1.A-2(3), under normal conditions, HPSW is applied to RHR Hx to minimize radioactive leakage to the environment. Per procedure T-101 "RPV Control", HPSW needs to be applied to the in service RHR heat exchanger as soon as possible to promote rapid removal oJ de~~l'J:1~CJt Distractors: B INCORRECT - See discussion above. Plausible if candidate does not know T-101 BASES regarding HPSW operations. C INCORRECT - See discussion above. Plausible if candidate does not know T-101 BASES regarding HPSW operations. Vibration concern is _associated with HX Outlet valve. D INCORRECT - See discussion above. Plausible if candidate does not know T-101 BASES regarding HPSWoperations. Vibration concern is
---~----
associated with HX Outlet valve. -~------.---~---- Psychometrics Level of Knowled~
- ----------_ . . . .- - - r--~----.-
Difficulty Time Allowance (minutes)
- RO -----
MEMORY 10CFR55.41 (b)(8) Source Documentation Source: C8J New Exam Item Previous NRC Exam: 0 _,,:1 D Modified Bank Item Other Exam Bank: 0 _ ____. _OILT_Exam ..§~IJKD ____ _ Refe~l"!ce(s): ___ (T-101J?aSeS, Design Basis Document P-S-09.__ . Learning ! PLOT 1560 Obj 9 O~j~~~i~~: _____ -L. KIA System : 203000 RHR/LPCI: Injection Mode Importance: RO / 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 RHRlLPCI: INJECTION MODE (PLANT SPECIFIC) : Component cooling water systems ~:~~~~~::::~L&
I::!Ch i~ Bottom, HPSW Ihe:omponent cooling water system for the RHR S}'!~~.!!1!~ the LPCI M o d e , . _______ _
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 high exhaust pressure should elevated containment
-+--'------------------
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.
------t--~<---<--- ---------------~
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 because of direct pressurization of the torus. Psychometrics ... Level of KnOWIJ\I!I~e~~_LI_ MEMORY I Source Documentation Source: rgJ New Exam Item D Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank ---~~-~----~~- -
~~J~renc~(~t T-102 Sh 2 and -~--~=~~~.~~----~-
T-102 Bases - Learning PLOT 5013 EIO 10q Objective: KIA System: 217000 - RCIC Importance: RO I SRO
<--".-,-"----~----------<---<~" ---- <." ----------<
3.5 I 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 -----"~"--~-------.-- 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 Answer 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 1" line and are designed for removal of non-condensibles at power or air removal for refueling or hydro test conditions. There is industry OE that confirms that bulk boiling of coolant due to lack of shutdown cooling will result in going
--'.. .~_~~1-'goL.r--'---ea-'---t:...::..e'---'rthan 212 F(i!!d_pressurizil]gjl'l§lBEY withJh!? vel"lts()penl~ ~ i 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 i
... ~ _ _ . -j--- .~.~_.+.~~:~::~i~~sS~~~:nrepti0~ rega~ingnde~Yheat-,,"ing absorbed by I B 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 I'
, candidate has misconception regarding natural circulation flow via RHR r-__________~_~--_+'~i~i~n~.------- .
D INCORRECT - R~actor ~ressure will increase above atmospheri~-.--~l Plausible if candidate believes head vent will relieve sufficient energy. to. L._ _ _ _ _ _ _ ~~.~_ _._____"_ _ _ _ ---1.i-C:..re=_v~e:..:..n::..:.t..c..:...:re'__=_s~,ure/ten1perature rise. . __ Psychometrics I Level of t5l"1S!..VIIledge I ~~ __~~iffiClJltY~~__~JJmeAllolIV~~l"Ice (I'!lirlutest . ~__~ RO .. ~~. HIGH I I 10CFR55.41 (b)(14) Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 J8tlLT ;~~'!I~(1nkJ4266)
.13~fe.Ce.1"I9§l(S): _~~~+_O._N.:-1g§L GP:-t2 Learning PLOT5010 - 9.k.6 Objective: ~~.,~~------- - - - - -....
KIA System: 205000 Shutdown Cooling Importance: RO I SRO 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
;~~~!~~~sf=ato~se'~9~L .. - _n __ ~ __ . _ . _ __ ~.~
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 14. Given the following conditions on Unit 2:
- Battery Charger 2AD003 is placed in the Equalize Mode in accordance with SO 57B.1-2 "125/250 Volt Station Battery Charger Operations"
- During the charge, AC power to the charger is lost due to a momentary loss of power to the E-12 bus
- Power is subsequently restored to the E-12 bus Which one of the following describes the status of the 2A Battery Charger one minute after the E-12 bus is reenergized?
2A Battery Charger __________ A. is charging at the "float" charging rate B. is charging at the "equalize" charging rate C. is deenergized and must be manually returned to service D. is energized but the DC output switch must be manually closed
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 ~ Question # 14 RO
~~ ~~-----~~~~~~~~---
cor~;=t~i~T-B*:m N:2~: SO 578. ~_~~i~uO:o~u:t::::t~nAC input power, the i I ~attery charger will return to the same mode it was in once power is restored. IF the battery charger was in the Equalize mode, THEN
~~ _____ ~~~~~~ ___ \_~_~ __ +-c::t:h~~:~e~~::---,t=im~e~r,---w=-=-I=-==--'IIpick up wh~E~it was intert'::lpJed ANQ~!11~_()ut: u Distractors: A The charger will return to the equalize charge mode. Plausible if the applicant remembers the charger will automatically restart but does
! not remember it will return to the same mode it was in prior to the uiu~()~~~l()~s .
. The battery charger will automatically restart 15 seconds after the I E12 bus is restored. Plausible if the applicant does not remember
___ tb§l1!~_~~~a rger""iII a ut0f!1C!!i~§lII~Ies!§lrt. D I Plausible since procedure precaution requires waiting 15-20 seconds i after closing AC input switch before closing DC input switch when placing charger in service to prevent blowing fuses in battery charger. Applicant may believe charger design would prevent
, automatic restoration (DC switch closure) following a loss of AC i power to the char er for the same reason.
Ps chometrics Level of Knowledge Difficulty Time Allowance RO _~_JEnin~l!~~L~_ MEMORY 10CFR55.41 (b)(7
)
Source Documentation Source: 0 New Exam Item C8J Previous NRC Exam: (PB 2011) C8J Modified Bank Item 0 Other Exam Bank: 0
~::~~~;:~~~f,~}t~i~a~ --~~~-=~ -'~- -~-
Objective: KIA System: 263000 - D.C. Electrical Distribution Importance: RO / SRO 2.5/2.8 KIA Statement: A1.01 - Ability to predict and/or monitor changes in parameters associated with operajillil t~e D.C. Electri~C!1 J)jstripution cont~<?ls includi!!g: Battery chargingLdischargiIJ!a rate. REQUIRED MATE~L~!-S~_1'NONE; _____ __ _ 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
-~-- .. - *.. -.. ~--- ... ~----- ... -~~---~--
Answer I Question # 15 RO
~ ~ ~ ~~ _ ~ Choice _---=--==r _______ ~~asis ()! Justificati~n__ _~_~~
Correct: A I CORRECT- SE-16 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
~~_ __
j t~ErTl.erg.erl~ Diesel GI:merators dueJQ.gricLund~r'{ol!<!ge." Distracters: .. 1
. ~~-~ *-~C-~
B
*t
. INCORREC.T- Part (1) is...c.orrect. but part (2) is incorrect - plausible s.in
. AO 50.1 relates to grid disturbances, but is associated with Main Generator perturbations .
INCORRECr=-part (1)lsincorrect, but part (2) is correct - plausible --- -~- . I because candidate needs to recognize that 220kV is LOW for the supply to
..ce.
- 2SU.
D iINCORRECT- 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 _~~Y_~LQI-'Sngwl~gflE3 . J- ~ __Diff!~l:J~_ _ ~i-~Ijm~~!I()wan~~(rTlinute~Lf-___ RO MEMORY
- I 10CFR55.41(b)(10)
Source Documentation Source: C8J New Exam Item D Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0 .13~f~!~:~~;~):-~_~ lS~1~1~~;m~~:~Y~~~~~~E-16 Base___s.. ______ Learning i PLOT 5053 EIO 10i Objective: ... _ --L -I-~***-~-- -. KIA System~__1~6~~001 - AC Electrical Distribution Importance: RO/SRO 3.1/3.4 KIA Statement: A2.09 - Ability to (a) predict the impacts of the following on the AC. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ~_~_xc:~.eclingu"'oltag~ Iimi!Ci!i0.Ds ~~ ....- - . - --_....._ - - - - - - REQUIRED MATERIALS: None Notes and Comments: _ _ _....... _~ _ _ _ _ _ _ L
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 _~(..:.J.1):-, and the crew should respond with (2)
A. (1) alarm ONLY (2) applicable Alarm Response Cards B. (1) alarm and rod block ONLY (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 I I Question # 16 RO Choice Basis or Justification I Correct: D CORRECT - 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 are all satisfied . _ _ _ _ ... _
.... "'-.:..::..._c---=-
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; understand trip functions and/or logic scheme. I C INCORRECT -Full Scram is expected - Plausible if candidate does not fully! _._.. ...L~u_..n. . _derstand trip ~~nctions and/or logic scheme. I Psychometrics _~~yel Of~r1Q\AlIt?.9~lm____.J)ifflCUlty.. I Time A"owance If!1inu!es1.J. RO HIGH' I I 10CFR55.41(b)(6) Source Documentation Source: i D New Exam Item D Previous NRC Exam I ~ Modified Bank Item D Other Exam Bank __ ~JhT Exam Bank Reference( s) :__ ..J.6~_C-£11--'?=-1.§l~nd G:1._____.__~____.__. Learning PLOT-5060C-3b, -5a, -6a Objective: _.1.---_.. _--- . KIA System: 215003 - Intermediate Range Monitor I Importance: RO/SRO System 3.7/3.8 KIA Statement: A2.04 - Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (lRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the
£~I}!i)~9ue.nc~s of_!~()~e abnQrm~al conditip~~()!_()perati~m~ _ m_ J.IP.s"c::§ile or down sc~lt?trie_s; REQUIRED --- MATERIALS: --
NONE ~---- Notes and Comments:
~--- -- ~-
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 17. The following conditions exist on Unit 2 following a LOCA:
Parameter Time =0 sec Time =60 sec RPV Level: -60 inches, lowering -160 inches, lowering i RPV Pressure: 850 psig, lowering slowly 800 psig, 10weringsloV\llyJ Drywell Pressure: 1.4 psig, rising slowly 2.0 psig,~ising slowly
- Other than scram actions, NO operator actions have been taken
- The plant has responded as designed Beginning at time = 60 sec, which statement below describes the subsequent RPV pressure response?
RPV pressure will ... A. immediately begin to lower rapidly due to ADS valve actuation. B. lower slowly over the next 10 minutes due to the LOCA condition ONLY. C. lower slowly until 105 seconds have elapsed, then lower rapidly due to ADS valve actuation. D. lower slowly until 9 minutes have elapsed, then lower rapidly due to ADS valve actuation.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 17 RO ~" ~,,-,~ ---~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~r*****-~-- Choice
~----~-~.~ ~~. -,..,.. ~.-~~~--..... +.. . . . . ._ -- - .
~~-~~---~~, --,~--~,
Basis or Justification Correct: C CORRECT - ADS initiation conditions will be 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 _ll1anipulated. _ _.~ ____ ~_~ Distracters: A INCORRECT - See above discussion of actuation setpoints. Plausible if candidate does not know~p~cificl:\[)S ini!iatiol'l_<::-,"itE~r§._ B INCORRECT - See above discussion of actuation setpoints. Plausible if candidate does not know specific ADS initiation criteria. D INCORRECT - See above discussion of actuation setpoints. Plausible if candidate does not know specific ADS initiation criteria. Psychometrics Level of Knowledge Difficulty Time Allowance
- RO I*****
(minute~L_ --~-- ~~--r HIGH 10CFR55.41 {b}{7} Source Documentation Source: [8J New Exam Item D Previous NRC Exam Other Exam Bank ______ .......... L ..L -_ _ _ + ___ M-1 __ ~ Sheets 2 and 3 Learning PLOT-5001G-9.k.4 Objective: KIA System 218000 - Automatic Depressurization Importance: RO I SRO System KIA Statement A3.08 - Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including: Reactor Pressure
- !t~~!~~::~:~~ALS: tONE :---= ~----===----.-. .- ~ _~~-~---- ~
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 18. Due to a loss of off-site power, the E-2 Emergency Diesel Generator started automatically and loaded the E-22 4kV bus 5 minutes ago.
Which one of the following describes (1) the current Mode of Operation of the E-2 Diesel Generator and (2) the ability to control Diesel Generator speed and output voltage from the Main Control Room? The E-2 Diesel Generator is operating in _(1)_ mode; speed and voltage controls are (2) A. (1) droop (parallel) (2) disabled B. (1) droop (parallel) (2) enabled C. (1) isochronous(unit) (2) disabled D. (1) isochronous(unit) (2) enabled
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 18 RO Choice Basis or Justification Correct: C CORRECT - Automatic start on bus under voltage picks up the DB1IDB2 relays, which will place the governor in UNIT mode (TSR relay drops out) and disables the control room controls for the governor and the voltage regulator (LSA ri3l<:iypl<?k~Lmmm~ _ _mm_m Distractors: A INCORRECT - (1) Incorrect - see above discussion. (2) Correct-Plausible because candidate may confuse Droop and Isochronous modes of opera!i0n an<:t!()rmay not understand function of DB1IDB2 relaycircuitrL B INCORRECT - (1) Incorrect - see above discussion. (2) Incorrect Plausible because candidate may confuse Droop and Isochronous modes of operation and/or may not understand function of DB1/DB2 relay c i r c u i t r y *
- m m ____________ mm D INCORRECT - (1) Correct _. see above discussion. (2) Incorrect Plausible because candidate may confuse Droop and Isochronous modes of operation and/or may not understand function of DBlIDB2 relay
--~~-~
circu~ __ _ Psychometrics Level of KnOWIi3c19El~~___ ~[)ifficu Itym Time AI'()IJ'.I<:iflce (minuti3!il RO
- - - - - - - - - - - - - - - - - - - - - --- mm MEMORY ! 10CFR55.41 (b)(8)
Source Documentation Source: New Exam Item D Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0 rg] IhI Exa!11J3ank (~72043)~ __m__ _ Referen9i3(S):___m_ SQ§2A.1.B EDG QQElratiQfl~(F'EEl(;§ll!~Orlm~}) mm m Learning PLOT5052-3c, f, g, h Objective: KIA System: 264000 Emergency Diesel Generators Importance: RO / SRO 3.0 I 3.1 KIA Statement: A3.01 - Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESEL/JET) including: Automatic staning of compressor and emergency generator ~~;;-::~~~::~S~~JNONE_m _ _ m__ m m m m m m m m m m -__m_- m_~~mm-mmmmm-----
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 19. A transient at Unit 2 resulted in the following:
- The HPCI System was manually initiated for level control by arming and depressing the HPCI Manual Initiation push-button
- The arming collar (23A-S1 05) was left in the "ARMED" position.
- The HPCI turbine subsequently tripped due to the reactor water level exceeding +46 inches.
- RPV level is +52 inches, slowly lowering
- DW pressure is 1.75 psig, slowly rising How will the HPCI system respond to the following conditions?
HPCI will ... A. re-inject when Drywell Pressure reaches 2.0 psig (assume RPV level is then +50 inches). B. re-inject if the PRO depresses the Manual Initiation pushbutton again (assume RPV level is then +50 inches; DW pressure is 1.80 psjg) C. remain tripped until RPV level reaches +29 inches (assume DW pressure is then 1.85 psig). D. remain tripped until RPV level reachj3s -48 inches
Peach Bottom Initial Reactor Operator LicensE~ NRC Examination April 2013 ________________________~_____ ___ Answer Key Question # 19 RO Choice Basis or Justification C~rr;ct~-----IC The HPCI turbine trips auto reset and DO NOT se;~in-.lith;trip condition clears AND an initiation signal is present, THEN the HPCI turbine (pump) will cycle on and off as the high RPV level auto resets itself. With the arming collar left in the "ARMED" position the system _____ __________ jrIiJi9tig_~ __sJgnal remains pnasent.__________ ___ _ Distracters: A With the arming collar left in the "ARMED" position the system initiation signal remains pnasent. If drywell pressure reaches 2 psig it will have no additional effect on HPCI. B With the arming collar left iin the "ARMED" position the system initiation signal remains present. Pushing the Manual Initiation p!-!§hbutton again will have no additional~!f~~L~~J""tPc;L___________ o With the arming collar left iin the "ARMED" position the system initiation signal remains present. If RPV level were to lower to the Lo-Lo setpoint (-48 inches) it would have no additional effect on HPCI, as re m _______ LniectiollJ'tarted when the high level trip auto-reset._ Psychometrics Level of Knowledge Difficulty Time Allowance RO (minutes) ------ ... - ---------------- ---- HIGH 10CFR55.41 (b)(7) Source Documentation Source: D New Exam Item [gJ Previous NRC Exam (PB 2008) D Modified Bank Item D Other Exam Bank [gJ ILT Exam Bank ---------------- -'3eft?!_ence_(~): SO 23.7.C-2 Learning PLOT-5023-4c Objective: KIA System ---- 206000 - ~Ci-~~----~~ - l~:orta~~_;~::~o K/A Statement ~1J_!L____ Al:>iIi!Y to 111~_"'~911y operate and/or monitor systeIllJ?umpsj!!thec;QntrolJ"Qom. _______ _ REQUIRED MATERIALS: NONE Notes and Comments:
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.
~~~~~-~--+ ..... ~~~........ ~ .. ~------- ~
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" pushbuttons. The auto start can be aborted by pushing any speed LOWER or RAISE pushbutton and the turbine speed will be controlled at the the turbine is at when the~u!tpJ'1waspushed 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. o INCORRECT - 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. Psychometrics Level.2LI<~I}2\J\1I~r;j~_L .. ~~ __~iffi<:;lJlty __~_ I Tirt'le Allo\J\l~l1c~J.rt'lil:lutesJ . RO HIGH 10CFR55.41 (b )(7) Source Documentation Source: D New Exam Item l8J Previous NRC Exam (2008 NRC) D Modified Bank Item D Other Exam Bank
~~~ _ D ll:.J~x~m B~I}I< ~
1 Referencetst___~O 6C,J .A-?:... ~_~_~~ .~-~~~ .... -.~-- ........ ~ Learning i PLOT -5006-4c Objective: KIA System 259002 - Reactor Water Level Control Importance: RO / SRO 3.7/3.6 KIA Statement A4.02 - Ability to manually operate and/or monitor in the control room: All individual component controllers in the automatic mode. ~::~~~~~::~s:= INON~_~ =
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 21. Per the UFSAR, which of the following are the Safety Design Basis of the Emergency Diesel Generator (EDG) and Auxiliary system(s)?
- 1. Support safe shutdown load requirements for BOTH units assuming simultaneous DBA accidents.
- 2. Support safe shutdown load requirements for BOTH units assuming a LOCA on one unit and a 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 Key Question # 21 RO~~~~~~--~~~-~ ~~-r--- Choice Basis or Justification Correct: C CORRECT - As specified in the UFSAR, Para 8.5.2, the Safety 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 PQlo\Ier (~90P)L§1~d~aJluf"~igne ~t~nd~y~cjies~L~n~r~!()~. Distractors: A INCORRECT - Design Basis for EDG and Auxiliary does NOT allow for simultaneous LOCA accidents. One EDG failure is allowed. Plausible if candidate does n()! recc:iJlthe~~ sR~'?ificJ!E:!~ign_~§1tt.lr~~. 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.
~~~~--r----
D INCORRECT - Basis and plausibility previously explained. Psychometrics Level_~U<nowl~Q~_+ ~ .. ~ ~_ _ J::>ifficul~_~__ 1 Time AIIO~§1~c~Jr'T'linutestJ~B9_ Memory I L 10CFR55.41(b)(8) Source Documentation Source: IZI New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank ReferencE:!(sL_ UFSAR Ch Para 8.f5.2; DBD P-S-07 ~ Learning I PLOT- 5052 Obj 1 Objective: KIA System: RO/SRO
~.~ .--~-.~ .. -~- -~.--------.~~
3.9 14.0 KIA Statement: 2.1.27 - Conduct of Operations: Knowledge of system purpose and I or function.
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 Question # 22 RO Choice Basis or Justification Correct: C CORRECT - An electric ATVVS 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 fuses to insert
~
control rods.
...._._ ..- - - - - - - .. ~~~-- .. ---
Distractors: A INCORRECT - T-212, "RWCU System SBlC Injection" is NOT required due to SBlC already injecting into the RPV.
~~~~~~~~ -~-~~-~--+----~~ f --- ~~~~~~ - ~~--~-~-~~~~~ ~~~.~- ~.-~- ..
B INCORRECT - T-216, "Control Rod Insertion by Manual Scram or Individual Scram Test Switches" is required to be performed for a hydraulic A TWS. This is an electric ATWS. l-D-~h-N-c6RRECT - C-losfng-HV--2:~~3--56"Char-gingWaterHeader-l3lock-VaTve to I
- HCUs" per T-246, "Maximizing CRD Flow to the Reactor Vessel" at 1000 I I psig RPV pressure will have no impact on driving control rods.
I_._._____ L __ Psychometrics Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: 0 ~;~~!;:(S):-i;~~9~~~~1i
- 0 IlT Exam Bank
-=--.. . . -~ ~-~--- -.~-----.
KiA System: 1212000 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.
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 O. trip on high turbine exhaust pressure
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key I Question # 23 RO 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: INCORRECT - RCIC will trip on overs peed under certain conditions if the controller is in AUTO, i.e. in CST-to-CST mode and MO-23-24 (common 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 t~e conditions that lead to an ove~~p~ed event are present. 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 t.h.e given conditions. RCIC is designed to run on min flow for extended I periods. Psychometrics Level of Knowled~e Diffi~lJlty_ITime Allowancetminutesll RO HIGH ! 10CFR55.41 (b)(7) Source Documentation Source: D New Exam Item [gJ Previous NRC Exam: (2009 NRC) D Modified Bank Item D Other Exam Bank: 0 IL"fE:xamJ:?C!I1I< 0 Reference(s): M-359 Sheet 1 SO 13.1.C Learning PLOT -5013-9.k. 7 Objective: KIA System: 217000 - Reactor Core Isolation Cooling Importance: RO/SRO 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 MATERIALS: NONE 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. 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: B INCORRECT - Plausible because control rod worth is low, not high, from core positions 36 to 48 as described in GP-2 "Plant Startup". C INCORRECT - Plausible because neutron detector proximity does have an effect on count rate but only at control rod notch locations 16 through 22. o INCORRECT - Plausible because the candidate 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. i Psychometrics I Level of 1Sn<:>'v\fI~cjg~__ Diffi<;lJltL.____ . . . Till1~AllowC3n~~jr'Tlinut~~... _ RO I MEMORY 10CFR55.41 (b)(1) Source Documentation Source: k8:J New Exam Item Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0
.OILT Exam Bank Ref~r~nc~J s)~. ___ GP-2 Learning PLOT-5060C Obj 11 Objective:
KIA System: 215003 - Intermediate Range Monitor Importance: RO I SRO i System I .__ I 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 (lRM) SYSTEM controls including: Reactor power indication respons~Jo rod position changes REQUIRED MATERJALS: I NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 25. Given the following:
- Unit 2 was initially operating at 100DA) 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 I
i 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. r- ------ --- Distractors: i B The outboard MSIVs are supplied by Instrument Air. C The outboard MSIVs are supplied by Instrument Air. D The inboard MSIVs are supplied by Instrument N2. Psychometrics Source Documentation Source: D New Exam Item ~ Previous NRC Exam (2008I\1RC) D Modified Bank Item D Other Exam Bank ILT Exam Bank Reference(s): ON-119;M-333;M-3S1 ;M-372 sht1' T-261 Learning PLOT-S001A-7b Objective: KIA System 300000 Instrument Air . Importance: RO I SRO 3.1/3.2 KIA Statement K1.0S - Knowledge of the connections and I or cause effect relationships between INSTRUMENT AIR SYSTEM ~_rlQ1h_e folloYv'i!l~___Mai~~Heam Isolation Valve air REQUIRED MATERIALS: NONE 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 "A" 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 Licens~~ NRC Examination April 2013 Answer Key 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 directs the oQ~.r..C!tQIt()_'I'~rify water level is 1:>~irJ.~controlledI:>Y DFCS.
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 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 inp~t ...... ___ . _.. ___ D I NCORRECT -Plausible if candidate does not know that the DFCS will automatically shift to single element control on loss of a steam flow input. Psychometrics DlfficuHy ! Time Allow_.nce (minutes) I RO 10CFR55.41 (b )(7) Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank Reference{s): $~Q6C.1.D-2, ARC201 H-1 Learning PLOT 5006-7i Objective: KIA System: 259002 Reactor Water Level Control Importance: RO / SRO 3.3/3.4 KJA 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 Lnputs 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.
- A relay failure causes a Power-to-Load Unbalance trip.
- The POWER LOAD UNBALANCE TRIP (206 B-1) annunciator goes into alarm.
Based on the above conditions, which one of the following describes the plant response? A. Control valves partially close to compensate for the unbalance. B. Generator lockout ONLY. C. Generator lockout and turbine trip ONLY. D. Generator lockout, turbine trip and reactor 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 TSVrrCV 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. Distractors: A INCORRECT - EHC does not automatically compensate for P/L imbalance. B INCORRECT - a generator lockout will result in a turbine trip. D INCORRECT - The PLU circuit will produce a generator lockout/turbine trip, but the reactor does not automatically scram. Psychometrics
.... Level of Kno",leclge . Diffic;ulty I Time Allowance ('!lin utes) I RO HIGH i I 10CFR55.41 (b)(7)
Source Documentation Source: o New Exam Item 0 Previous NRC Exam (2007) o Modified Bank Item o Other Exam Bank Reference{s):
... ~~ ..- ILT Exam Bank GP-2, ARC 206_B-1 ... I? Bases 3.3.1.1 Learning PLOT5001 B - 5f Objective:
K/A System 245000 Main Turbine Gen. / Aux. jlmportance: RO / SRO 3.4/3.5 K/A Statement K1.08 - Knowledge of the physical connections and/or cause- effect relationships between MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and the following: Reactor/turbine2t~ssure contl"()l~ygel1J_ REQUIRED MATERIALS: '.~.~1. 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 __~'"'"---__
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 Answer , 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 E311~rgiz~~ and ener91zeto function. Distracters: INCORRECT - Power supply is correct; the Backup Scram Valves are normally de-energized. INCORRECT - Power supply is incorrect. D INCORRECT - Power supply is incorrect; the Backup Scram Valves are normally de-energized. I Psychometrics Level of Kn()~I~diJE3 Diffi~UltYI Time AliowanceJ'!1jl1uJE3§)i RO MEMORY i I 10CFR55.41 (b)(7) Source Documentation Source: o New Exam Item k8J Previous NRC Exam: (PB 2009) o Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank Referen~E3(s): E-26 ShE3~L1, M-1**§-:-54 Sb~E3L8 Learning PLOT -5003A-2c Objective: KIA System: 201001 - CRD Hydraulic I Importance: ROISRO 3.5 I 3.6 KIA Statement: K2.03 - Knowled~e of electricC)lp()w~r~sl!pplies to the foliowing:E3§1ckupl:)CRAM valve solenoids. MATERIALS: 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 I 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
. i SU}:tJP'!Q()Il1- ...~~ . --..~.
Distractors: A I Incorrect, Plausible ... however the Reactor Building (RB) floor drain sump will overflow to the equipment sump before allowing contaminated water to overflow intcUh~§l.JmpI()QXll..~_ .. B Incorrect, Plausible ... however the sumps are not sealed and would overflow into the room if the overflow to the Equipment Sump did not exist. C Incorrect, Plausible ... however the overflow will not go directly to the Waste Collector Tank, will overflow to the equipment sump. Equipment Sump Purnps~v.tili tben p~mp it Q"'E!! to tb~\lYas~J~ollect()! Tank. i Psychometrics Levelgf KnoVVIE!clge j I DifficLJlty. : Time Allowance (minute~) 1 .. RO MEMORY I i 10CFR55.41(b)(13) Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 Modified Bank Item 0 Other Exam Bank: () J8LILT Exam Bank Referenci3(s): M-369 sheet 1 Learning PLOT5020 - 6d Objective: KIA System: RO/SRO 2.7/2.8 KIA Statement: K3.04 - Knowledge of the effect that a loss or malfunction of the RADWASTE will have on following: Drain sumps REQUIRED MATERIALS: NONE Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 30. In the event of a demand signal failure, Hecirculation 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 ANQ 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 Key Question # 30 RO Choice Basis or Justification Correct: o CORRECT - Recirc Pump speed is limited by scoop tube lockup circuit and mechanical 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 l~c!<~()_~ir<::lJi!~L~o~~Jf~s;tlv~ly limits sp~~c!~)(c::ursions. C INCORRECT - Partially correct answer - No mechanical stops on the scoop tube - plausible because scoop tube lockup circuit is part of the li,!!itirl9_~glJip!l1~J'1!' along-"Ylth the Jorg~l1_ Po~_i!loE!er mechaJ'1ic::~L~t2PS. Psychometrics Level of Knowlegge J [)iffic::l.Il!Y_____ Time Allo""~_nc~(minutes) 1-10CFR:5~41 (b){6) MEMORY Source Documentation Source: [8J New Exam Item Previous NRC Exam: 0 Modified Bank Item o Other Exam Bank: 0 OILT Exam Bank Reference{s): PLOT§QQ2_,_-':a 4~_~j::~00292 (Jordan Positioner_§p~cificati()n) Learning PLOT 5002-3t Objective: KIA System: 202002 Recirculation Flow Control Importance: RO I SRO 2.9/2.9 KIA Statement: K4.07 - Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature(s) and/or interlocks whlctl provid_e for the followi!"l~___1'.11nimum and maxi,!!!!,!! pump !:)peed setpoints REQUIRED MATERIALS:_ -lNONE _~ _ 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 FROM location BBB-5 TO location U-52, and is currently over an area of empty fuel storage racks (HH-26 and approx 50 surrounding cells are all empty)
- 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 Fuel Pool map (from M-1324) is shown on next page.
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 (U-52) B. be placed back in its original storage location (BBB-5) C. be placed in the nearest storage location (HH-26 or surrounding cell) D. remain suspended in its present location
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Peach Bottom Initial Reactor Operator License NRC Examination April 2013 I Answer Key I
-- I Question # 31 RO Choice Basis or Justification Correct: C CORRECT - 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 be placed in the nearest available underwater storage location if time is vital. The ARM alarming and GP-15 evacuation make time vital . ....... _----------- . - - --------- - -- --------- Distracters: A 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 case sincetbej\HM is alar!l11n~.__ . 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 case sil'1_~eJb.EU~.BM is alC!I'!1infL . . . ___ . o 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 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
......U)..§J9I~_leaving theRE;fueJlngBridge:
Psychometrics Level of Kno\'VJ~9ae __ ~____ DifB211I!y'JJme AlloV'.'~rlc~ (IT1i1'l~t~~)1 RO HIGH . ! 10CFR55,41(b}(12) Source Documentation Source: IS] New Exam Item D Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0 ILT Exam Bank ..... ------ ------- Referenc:~(s): FH-74 Learning PLOT 5019-9.k.2 Objective: KIA System: 234000 Fuel Handling Equipment Importance: RO I SRO 2.9/3.4 KIA Statement: K5.03 - Knowledge of the operational implications of the following concepts as they apply to FUEL
- ~~~~~~~ ~~~I:~!~~/ W~1te~~~~ shlelci against Iadiati~~
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 Reslrictors ANQ 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 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
...... +~~...~.~.... ~.~_j-_-'--rec~_s::...:t.*r.ictors ar~ C3r1E:r1girl~~r~_d SC3fE:JyF~~a..!lJ re .
Distractors: A Incorrect - see above. Plausible as the flow restrictors are part of the protection scheme. B Incorrect - see above. Plausible as the MSIVs are part of the protection scheme. Incorrect - see above. Plausible as the flow restrictors and the MSIVs are each part of the protection scheme, but neither is sufficient alone. Psychometrics Level of Knowledge -f ~_~_.~_~ .._.D ...... ~iff.~.i..c_.. ulty Time Allowanc~Jt"rIif1ute~L~ MEMORY Source Documentation Source: [gJ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 Dl'=TE:~aml:!a.D~~ Refer~f1~E:(~): . UFSAR Ch Sec 4.1 and 4.5 al'lgCh 11-,-~ec 14.6.5 Learning 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 BA.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 i Question ID# 33 RO Choice Basis or Justification Correct: B CORRECT - Based on the Routine Inspection, the expected range for Off Gas System Flow is 20-45 scfm. 120 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 -120 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 I INCORRECT -The normal range of off-gas flow is 20-45 scfm. Psychometrics
~evel of Knowle(j~ __ l SRO HIGH N/A Source Documentation Source: o New Exam Item D Previous NRC Exam Modified Bank Item 0 Other Exam Bank (200B PB Cert)
ILT Exam Bank Reference(s): SO BA.6.A Placing Standby SJAE in Service SO B.B.A-2 Off-Ga~§y~tem Routine Insp~~JioD_ Learning PLOT5008-9.k.7 Objective: Knowledge/Ability I 271000 Off-gas RO I SRO i 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". B. run both Recirculation Pumps manually back to 45% per ARC 214 B-3 (G-3) "A(B) 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 Answer Key Question # 34 RO Choice Basis or Justification Correct: B 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 steady at 98%. Operator needs to manually initiate runback of Recirc PUfTl2S to_~t5°(~~~ed~__ __~ __ Distractors: A INCORRECT - Runback is required, but to 45% not 30%. Plausible if candidate confuses run back 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.
D INCORRECT - Runback has failed as discussed above. Plausible because i failing to perform the run back may result in a scram on low level, but the runbac~is d~sjg[ledtQJ?recILJdt?this scral'l1!:)o sh_()l.lld be~initiated FIR§T. Psychometrics Level oLKflo\l\lIt?_ct.RE! DifficLJlty~_ .. ~~ __ ~I Time AliowCl[lcejminute~L RO HIGH i 10CFR55.41 (b}(7) Source Documentation Source: o New Exam Item Previous NRC Exam: 0 o Modified Bank Item o Other Exam Bank: 0 J8ULT Exam Bank Referel1_cE!(s): OT-1 00 Reactor Low Level,~f\RC 214 B-3 Learning PLOT-5006 Obj 3i Objective: KIA System: 259001 Reactor Feedwater Importance: RO I SRO 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: 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 Choice Basis or Justification Correct: C CORRECT - Both Reactor Building and Refuel Floor Ventilation isolate (Group 3 signal) when either Reactor Building ventilation exhaust RIS-3 : 452A or C AND B or 0 are reading.::: 16 mr/hr I OR Refueling Floor ventilation exhaust RIS-3-17-452A or C AND B or 0 are reCicJIng ;:1iLrl"lr/hr. Oistractors: A INCORRECT - Both Reactor Building and Refuel Floor Vent isolates. Plausible if candidate believes ONLY RB will isolate due to higher Reactor BuildiQ9JiiRad COJ1J:fttion... __ . B INCORRECT - Plausible if candidate believes that Reactor Building and Refueling Floor ventilation hi radiation setpoints are different. o 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 0 are reading.::: 16 mr/hr OR Refueling Floor ventilation exhaust RIS-3-1 T*452A or C AND B or 0 are r~~qing =:J~~r/br: Psychometrics Level of Knowledge ,
--------- -- . -- --_._-+ --
HIGH I _____.__.. .O=. iffi cult Y __ 1Time Allowaflee (mil1ute,,) RO 10CFR55.41 (b )(9) Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 o Modified Bank Item D Other Exam Bank: 0 ___ 0 ILT Exam Bank Reference(sl___ ARC 3180-4, TiLlable 3.3.6.1-1 Learning PLOT5040B-3a Objective: KIA System: 288000 Plant Ventilation Importance: RO I SRO i 3.8/3.8 KIA Statement: A3.01 - Ability to monitor automatic operations of the PLANT VENTILATION SYSTEMS including:
- -*-* Isolation/initiation signals REQUIRED MATERIALS:
Notes and Comments: 1 NONE
-~..*.. - ...
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 OVERRIDE (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 Page 6 of 6 FIGURE 1 DRYI"lELL CHILLED WATER (DWCt-l) SATURATION CURVE
.- 350 325
- r.,
~
~
-:: 300 ../
L V
-.- 275 /
~
/
/
;... 250
~
- ..t.
/
;;::
225 /
':,;,"
V 200 o 10 20 30 40 50 60 70 80 D\VC\V RETURK HEADER PRESSCRE 01'\ PI-20262 (PSIG)
- IF TI-80146 is out of service, THEN use RT-O-40C-530-2 to determine DW BuD: Average Temperature.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 36 RO Choice ~~------. Basis or Justification Correct: A CORRECT - Per T-102 "Primary Containment Control" with torus level at 19 feet, drywell spray is not permitted due to covering the vacuum breakers S_t()russpr~Ylsr1Ot all~YJ~d iftorus level 2! 21 f~e!L Distractors: B INCORRECT - Torus (but not Drywell) 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 sRr~YJ.!s_l1ot allowed if t()Fl.J.~_lev~15J1 f~~t),---_ C INCORRECT - logic to spray is satisfied. Plausible if candidate does not I understand logic inputs. D INCORRECT -logic to spray is satisfied. Plausible if candidate does understand logic inputs. Psychometries Level of Knowledg..e I
. Diffi9UI~. __ I TirneAliowance .{minutes] .: RO HIGH 'I 10CFR55.41 (b)(7)
Source Documentation Source: o New Exam Item 0 Previous NRC Exam: IZJ Modified Bank Item 0 Other Exam Bank: 0 D_ILT Exam Bank Refer~nce(s): . . ARC-224 D-=--2~-.I-J 02 Learning PLOT-5010-4s Objective: KJA System: I 230000 - RHR/LPCI: Torus/Suppression Importance: RO / SRO Pool Spray Mode 3.6/3.3 KJA Statement: A4.09- Abili!Y!<:)_l11cmu~!Iy'~perate and/or monitor in the cOil!rolrQ9.m: Indicating lights and alarms. REQUIRED MATERIALS: Notes and Comments:
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 37. Unit 2 is operating at 100% power
- Drywell Pressure unexpectedly rises to 1.3 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 enter T-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 Question # 37 RO Choice Basis or Justification Correct: CORRECT - A GP-4 Manual Scram is required at 1.2 psig in Drywell, 1.3 psig is above 1.2 psig Distractors: INCORRECT - Not required unless both seals on a Recirc Pump fail. However, requirement to scram at 1.2 psig still applies. 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 Level O~:~Viled~ + I:)iffi(;ulty_~ I Time Aliowance{rTlil'1!ltes) T~- I I RO 10CFR55.41(b)(10) Source Documentation Source: New Exam Item Previous NRC Exam: D Modified Bank Item D Other Exam Bank: 0
.18U~r~~~m Bank ReferenceJs): OT-101 Learning PLOT1540-6 Objective:
KIA System: i 223001 Primary CTMT and Aux. Importance: RO I SRO I 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: NONE 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: 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 supply. _____ ~___ ~_ 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. D INCORRECT - Plausible if candidate does not understand above described power supply arrangement. Psychometrics ,--L_e_v_e~___o___~___~--=~_o;_~_ed__g_e____ I! _ _ _ DiffICLJIty_ TTimeAliowance (minutes) i-- 10CFRi5041 (b)(6) Source Documentation Source: D New Exam Item Previous NRC Exam: 0 IZl Modified Bank Item Other Exam Bank: 0 ILT Exam Bank --------- IELOT-5060, ~ M-1-S-34
~
Refere!lc:e(s): 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 --,----- 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 current. M\lVs n?_f!lJ!!I!_l!r1~hang~Q. __ Distractors: INCORRECT - see explanation above - MWs will not change, MVARs will rise. INCORRECT - see explanation above - MWs will not change, stator current will rise. INCORRECT - see explanation above - MVARs will rise. Psychometrics Level of KI'l()""JE!.~9.~_~.___ ..... _...J;>iffic;ulty I Time Allowance (minutes) RO HIGH I 10CFR55.41 (b)( 4) Source Documentation Source: lSI New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0
. D ILT Exam Bank R~ferenc. ____e(s_)_
__ '_. ____..-_- - [PLORT-12-04D Learning - -r PLOT-50500bJ9k Objective: KIA system:j7ooo00 Gener~t;rV~ltage and-E-I;ctric Grid Importance: RO I SRO I 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. PerT-102 Bases, with torus level in its normal range, which one of the following describes (1) the Primary Containment Pressure Limit - A (PCPl-A) and (2) the consequence of exceeding this limit?
A. (1) 49.1 psig (2) inability to operate SRVs B. (1) 49.1 psig (2) loss of containment integrity C. (1) 60 psig (2) inability to operate SRVs D. (1) 60 psig (2) loss of containment integrity
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 40 RO Choice Basis or Justification Correct: C 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. Distractors: A INCORRECT - Plausible as 49.1 is the LOCA Peak Pressure (Pa). B INCORRECT - Plausible as 49.1 is the LOCA Peak Pressure (Pa). D INCORRECT - Plausible as high containment pressure could ultimately I result in a loss of containment integrity, but the containment design J pre~sureis 62 psig, which is above 6.Qp~Jg:. Psychometrics Level of Kno\l\ll~<:l~~ .... f.. Pifficul!Y..~.~~_~ r* Time .Allowance
. ....(minutes)
.... ,! RO I
MEMORY I 10CFR55,41 (b)(9) I Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank Referel'1~~(sL T-102 Bases, PCPL-A Bases in T-BAS ............... Learning PLOT 5007-5a Objective: KIA System: 295024 - High Drywell Pressure Importance: RO I SRO 4.1/4.2 KIA Statement: EK1.01 - Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: Drywell integrity: Plant-Specific
Peach Bottom Initial Reactor Operator liCenSE! NRG 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 EGGS 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: D CORRECT - per FF-01 "Fire Brigade" Distractors: A INCORRECT - The size of the fire is not defined by FF-01. B INCORRECT - This is a requirement from ON-114 to scram the reactor. C INCORRECT - This is associated with the time limit for performing EAL classifications. Psychometrics Level ot ~nQyvl~dg~ J _j)lffi~~ ...... _ - - --- Time Allow~I'l~~ (minute~~l I RO MEMORY I 10CFR55.41 (b)(1 0) Source Documentation Source: D New Exam Item [gJ Previous NRC Exam (PB 2008) Modified Bank Item Other Exam Bank ILT Exam Bank Referer1c~(~): __ FF-01 Notes Learning PSEG-0214L-03 Objective: KIA System 600000 - Plant Fire On Site Importance: RO I SRO 2.9/3.1 KIA Statement AK1.02 - Knowledge of the operational implications of the following concepts as they apply to Plant ~ Fire On-Site: Fi~~fighting. REQUIRED MATERIALS: -******f----~----~-~ NONE Notes and Comments: . .*. *.-* -=:-=~..~-----
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 DWff-2 is PROVIDED ON THE NEXT PAGE. Initiation of drywell sprays in order to control drywell temperature will initially result in a 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 OW SPRAY INITIATION LIMIT 800 -- _550 ... L
'( ~SOO J
- Q..
2450 / w t w 4DO I c
~360 I
w
~300 --
J
,/ ~250 = =200 / !!It °150 I
100
/
0 2
- 8 & 1 o 12 16 DRYWELL PRESSURE CPSIO)
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 42 RO Choice
~~-~~~-~
Basis or Justification Correct: D - Based on the given conditions, drywell atmosphere is super heated. Per T-102 Bases for step DWIT-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 I curve limitation is to ensure that even if an evaporative cooling condition exists, the capacity of the Torus to Drywell vacuum breakers is not exceeded. Distracters: A INCORRECT - see above. Plausible if candidate believes that the DWSIL curve is based on preventing conditions that would result in evaporative (;()Qli..,g anc:Lr~igpre~stJr~ ~re~duc;ti()..,. ~ ~~ ~ ~ _~~__ ~ ~~~ 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 deficiencyl'litJndamentals of fluid thermodynamics. 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 _~L~g~JJcJenc;yin_ f!!n_g~_I'!1~..,~I§_()UI!!i(tth~rmodynami(;~. ~~~ Psychometrics Level of Knowledge I Difficulty ~~~~~~~~~~Tirne AliowangE!lIT!inutes)
- RO HIGH
- 10CFR55.41(b)(14) I Source Documentation Source: rgj New Exam Item D Previous NRC Exam: 0 D Other Exam Bank: 0 ILT Exam ..............
Bank Reference(s) : T-1 02 StE?P DWIT-_~1~~3~~~~_~_~~ Learning PLOT2102 DBIG - Obj 6 Objective: 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: ~~~~~~~~~:~::~;:$=r3l~:=~&~ables .=-=__~ _~_
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 1 - ------- -1 I Question ~_~~_~s::> i Choice Basis or Justification Correct: D CORRECT - Per step 2.8 of ON-113, pumps can remain in service if below I 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, __ Ilec~§~lta!illg tri2pil!9~fJhe pu~____________ ______ B INCORRECT - Not limited by time in ON-113. Plausible because loss of RBCCW could lead to overheated bearings on the Recirc Pumps, _l1ecessitatill9Jrippir1fL()fth~J~L1I"Tlj>_s. ___ _ C INCORRECT - Not directed by ON-113. Plausible if candidate incorrectly believes that seal purge flow provides adequate cooling to recirc pump seals. Psychometrics L~v~L()f KnoV\lI~dg_~_ -- ------ _ __Diffl~~lty Time AlloVlfCl_rlC;~ (rl1inut~s11-10CFR5~~ 1(b)(1 0) MEMORY Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 [gJ Modified Bank Item D Other Exam Bank: 0 _ D-'~T §.x.9rnJ:3Cl_rl~_ _ ___________________ Reference(s): ON-113 Learning PLOT-DBIG-1550 Obj 2 Objective: KIA System:-- 295018-=--Part~l-o-rC~~pl;t~ Loss of - - -l--Import~~~~~---R-O I SRO-Component Cooling Water 3.4 I 3.6 KIA Statement: AK2.02 - Knowledge of the interrelations between Partial or Complete Loss of Component Cooling Water .Cl_nciJb_e_f()U2\1\{il"!fr£lCl_n!9f~rC!tioI"lS. _ _____ __ _ ______ _
-~~t~~~~~~:::-~~LS~-.-.-__-t~9~ ~--~--____ E ---------_______
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 temperature lowers D. Drywell ternperature rises
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key i Question # 44 RO i Choice Basis or Justification Correct: A CORRECT - Lowering Torus level will cause the HCTL to be more lr:~~~~tiv__e__.._._._ .._.__.. 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. D INCORRECT - Drywell temperature has no effect on HCTL. Psychometrics Level of KnowledgE! f DifficLJlty_ I Time Aliowarl2E!.Jminutes) j RO HIGH I 10CFR55.41 (b )(9) Source Documentation Source: o New Exam Item rgJ Previous NRC Exam (2008 NRC) o Modified Bank Item 0 Other Exam Bank RefereDc:;e(s}:_ TRIP/SAMP CLJr\.IE!s,Table and L..il"Y1its_-:.J~ases Learning PLOT-21 02DBIG-6 Objective: KIA System 295026 - Suppression Pool High Water RO/SRO Temperature 3.5/3.7 KIA Statement EK2.06 Knowledge of the interrelations between Suppression Pool High Water Temperature and the followir'lg: ~LJr>P.!E!~~i()rl_P()gLhevel. 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 Answer Key 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 first s~~C3l"Tlmi'lgJh~ reactor. Distractors: A INCORRECT - Not a reason for scramming the reactor during performance I of SE-10 "Alternative Shutdown". It is anticipated that drywell pressure will I I rise grater than 2 psig, especially if there is a loss of off-site power, due to 1C1~k()Ldry'JVell cooli'l~ . . C INCORRECT - Condensate pumps are secured prior to scramming the I reactor during performance of SE-10 "Alternative Shutdown". RPV level I control will be with HPCL D 'I* INCORRECT - SE-1 0 strategy includes closing the MSIVs and allowing the i SRVs to open on setpoint until the crew can take positive control of SRV I __~j)~~~atLonfrom the Alternative Shutdown Panel. _. I Psychometrics Level of KnoV'Jl~cige~ _J2ifficulty~ __~ [Time AIIQ~_C!I1ce <f!111!l.!tes)_L RO MEMORY I 10CFR55.41 (b)(9) Source Documentation Source: cgj New Exam Item 0 Previous NRC Exam o Modified Bank Item 0 Other Exam Bank
-+~~--=...,..
ILT Exam Bank Ref.erence(s):.
~~~ .... ~------- . UFSAR Chapt~~~L .
Learning PLOT 1555 DBIG Objective:
-~T KIA System 295016 - Control Room Abandonment
- Importance: RO 1 SRO I 4.1 14.2 KIA Statement AK3.01 - Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT: Reactor SCRAM REQUIRED MATERIALS:
~-****f~**~----*****-~-*-~-~-~~-
NONE N()tes and Comment~~_ ~~~ ___~ __~
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 - PerT-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 pr()I"I'I()J~l"I'Ii>5LQgJl~~L(:!L~ri~uJiQ!1.ofJt1.~_bQl"()-"~Un!lt~RPY*.~~~ Distractors: INCORRECT - Per T-101, Step RC/P-14, depressurization during an ATWS is not performed until the ENTIRE SBLC tank has been injected-I Plausible if candid.~e confll~~e"~Qd2~~~su~~_g!JidCll'!ce. ~~ I INCORRECT - Plausible if candidate confuses Hot Shutdown Boron Weight I I (21 % of SBLC Tank) with Cold Shutdown Boron Weight (Entire SBLC Tank i iV()LlIl1l~l . ~_ D I INCORRECT - Plausible if candidate confuses Hot Shutdown Boron Weight
* (21 % of SBLC Tank) with Cold Shutdown Boron Weight (Entire SBLC Tank I yolu.ITLeL !
Psychometrics _Level ()f Know!~gg~ I Diffi~ulty__ I Time Allowanc~ll"1'1inutesl J RO I I HIGH I i . 10CFR55,41 (b)(9) Source Documentation Source: I:8J New Exam Item D Previous NRC Exam D Modified Bank Item D Other Exam Bank I IL- T Exam . Bank Reference(s): I TRIP/SAMP Curves, Table and Limits - Bases, T-117 and Bases, T-101 and
- Base!S.-:::- Definitioll()f "t;_()! Sh~tgown Boron Weigbt" ~_ _ _
Learning PLOT-DBIG-2117-6, Objective: KIA System I 295037 SCRAM Conditions Present and Importance: RO I SRO I 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: Notes and Comments:
~~
I NONE
-~
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 Answer Question # 47 RO Choice Basis or Justification Correct: B 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 radiation dose to station personnel is not changed by the isolation, however, the severity of the release to the pu~lic is minimized:~ i o INCORRECT - This distractor is based on a procedural caution about how I SBGT is aligned, but is not the basis for the isolation. Psychometrics
~RJfficulty Time Allowctnce (rr1irllJJ~~) .. RO 10CFR55.41 (b)(9)
Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 D Modified Bank Item ~ Other Exam Bank: (LORT)
~ .... -~~~ .. ILT Exam Bank
+~.~.=".-~~
ReferE3rlc;~~L UF"~!-_f3_ Chapter 5___~ Learning PLOT-5007G-12 Objective: KIA System: 295023 Refueling Accident Importance: RO I SRO 3.3/3.6 KIA Statement: AK3.03 Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS: Ventilation isolation
- ~:~~~~::::~AL~S~:_~_~_-~J~NI~O~N~E=*******_~___~.~_~_. ~_-_~~___________~___________
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: D CORRECT - Load set signal enters a low value gate, causing control valves to close. With the reactor at rated power, pressure will initially rise and byp(lss 'JClJ"-~~LlJ\fillo~I1~. 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 Level of Knowledge I -- - ...... DifflcuJty_ I Time Allowanc:~tll'lj!1~t~s) RO HIGH
- 10CFR55.41 (b)(4)
Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 cg] Modified Bank Item D Other Exam Bank: t.- ___=-"'". ILT Exam Bank Refer~nC:le(s): PLOT-500 1DL;_§QJJ~.1.A Learning PLOT-5001 DL Obj 6a Objective: KIA System: 295025 High Reactor Pressure Importance: RO I SRO 3.8/3.8 KIA Statement: EA 1.02 - Ability to operate andlor monitor the following as they apply to HIGH REACTOR PRESSURE: Reactor/turbine pressure regulating system REQUIRED MATERIALS: Notes and Comments:
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 StU 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 response of the following Main Generator components?
- 1. Generator output breakers will _ _
- 2. Exciter field breaker will _ _ __
A. remain closed open B. open remain closed C. open open D. remain closed remain closed
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 49 RO Choice Basis or Justification Correct: C CORRECT - on a main generator lockout trip, both output breakers will open and the exciter field breaker will open. In the stem of this question, having the unit initially at 25% power with the #1 Aux Bus transferred to a Startup Feed, is there to sway the candidate to thinking that this condition affects mail} g~n~ra!C?~tr1J:>_ rl3spc?I}~~-"-JLdoE?s not _ Distractors: INCORRECT - see C above INCORRECT - see C above D INCORRECT - see C above Psychometrics Level ()fK.'1owl~QR~_ ). . . _J:>Lffj~lJ.l!Yl_Til11eAliowance trni.'1lJtes ) RO HIGH I 10CFR55.41 (b)(7) Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item Other Exam Bank: 0 OJLI_E::xam ~~.'1k _ ____ _ Refert3nC;t3(S): ARC 220 B-1 2 GenR~~~ Learning PLOT-5050 Obj 3 Objective: KIA System: 295005 Main Turbine Generator Trip Importance: RO 1 SRO 2.7/2.8 KIA Statement: AA1 .04 - Ability to operate and/or monitor the following as they apply to MAl N TURBINE GENERATOR TRIP: Main generator controls REQUIRED MATERIALS: NONE 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 Question # SO RO --~--~~ Choice -~ .. -~~--~ Basis or Justification Correct: 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 I _._i_~E.?9!r1JQ ri~eJ~Jl:!~ighj~I"E!~sure scram se!Qoint. ~ ~___ _._ . Distractors: A 1INCORRECT - BPVs accumulators will eventually run out due to design i valve actuator leakage. C INCORRECT - With bypass valve operation, an immediate reactor scram is not expected at this power level. o INCORRECT - Scram on turbine trip is bypassed at this power level. Psychometrics
. Level of Knowledge~~~~_. ~~ DiffiCUII)'____1_Tim_~lIowance tminutes) I RO HIGH i 10CFRSS.41 (b )(7)
Source Documentation Source: o New Exam Item ~ Previous NRC Exam (PB 2002) o Modified Bank Item Other Exam Bank ILT Exam Bank ReferencE.?!?t ARC 206 C-S Learning PLOTS001 B - Sf Objective: KIA System 29S006 SCRAM Importance: RO I SRO I _. 3.7/3.7 KIA Statement AA 1.03 - Ability to operate andlor monitor the following as they apply to ~~t~;~~~~~:::~~:t;=R+~~~~rol~re~sure_regul~=9~stem
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 TORUS PRESSURE 320~~--~.--~.--~.--~.--~.--~.--~--.
~ ..... -!. ~~.- w ..... ":~"-,.,,. ~ ~.r" ~ ...... ., ~~:~ .......... ~ .1"'' ~.~.; ~""'*~"N ~:~ **. ~ .* ::"~.-
300 :::::;:::::::i:::::::t:::::::i::::::::t::::::::::::::::::::::::::::::::
. . . . . , , . 60 PSIO AND ABovE
,. .,. . 280 .-_. - ~ ~~ ~- .. ....
~-:~~ * ... "': .. ,. - .. ~ .....: ...... " ..... : . " ***** ~ ........ ,,-~ - .:~.,v .... ~~ r~. ~*' ~ *
* "'~ .. ~ ~ _, ~ ~ ~. ~ ~T'. ~ .. _ ~ .. ~ ..... ~ .. "~~ .... ,, ...... ~ .~ ....... ~. ,,~. '0"' ....... ~;...'" ~ ....... ;. ~., ~ ~.
u.. l60 ... " ........ ""................ . *.*. ~ ....... ;....... 30 10 SQ 9" P"T G
* \ ..... ~ .. ;~"'~.+ ....:~ ........ ~" .. ~ .. ~~~ ... ; ........ ,.~.~ .. ".... ~.~~~ .. ~-~ ........... -... -.... ~. .... ~ .'> ....
0...
- 2 240 nD
- f
- ::::::i::::::::~:
- 2 240 nD
~ ...
¥.,,~. ~~~
** -,,, ... ,, ~"..,
. /
............ " ."'................. ~
....... L.......~.::~::::OJ
~ ~
~
.... ~ ........ ~.
. 20
"':- ~ ... ~~ ** : ....... ".
TO 29.99 PSI G 10.5 FEET TO 12.29 FEET ~ zoo F-"~':~:;':":':-: : .:~::~:~:':":'::~:::"'"'::.r.;"...........*..... i. -._
- . . _. . *4*1_*'-"-'....4w...:.;,,;,;..L 10 TO 19.9S PSIG (I) I eo S TO !l.SS P:;1 G
~ IS0~~~~~~~--~~~
3 TO 5.9S PSI G CJ t- 140 12D o TO 2.99 PSIG 7.sue S,OOO 10,000 11.000 12.000 1 PUMP 15.00G 6.000 1B.COO 20.000 22,000 24,000 :2 PUMPS no r--. . . . ' . ., TORUS PRESSURE I 300 I**..*:..***..*:* ..***..:*.. *m.:.. m... :::::;::~;:::::::~:::::::1::::::::1 so PSIG AND ABOVE u: "0 E.**C::I*C.*.:r .::.+.:**;::**. :;:::. ::c**** 1
- 260 I ..... :" ...... ~ ........ : ....... ~ ........:. * ........ : ....... 30 TO 59.99 PSI G
......, ~._~~.: .. _ ........ "..._~ ....... t. .............. ~~". ... ~~ .. "~~.,...-.. ..
240 .... ;....... :......................... ...... . . 0... , . , .. ,:. ....... ~ ........ ; ....... ~ ........ ~ ....... ~ ........ ~ ....... ; ....... 20 TO 29.99 PSIG
~ 220 ..... ;.........;........ t.. ::::::::::::::~:::::::~::::::::~:::* .. *::::::::. Q ('
t 2.3 FEE T TO 14.49 FEE T t- zoo ::::::::::~::~::::::::::::::::L:;;:*~*"':*:L. .... ;~:::::::: ...::::, 10 TO 15 ** 9 PSI,", (I)
~
180 , : ; ,.. P ... : ..... ) ' ....
* ~ **. -'~ ... w~""".--:~ __ . *.:~ ... ~~*._ ._~.! *. ______ ;~__
m~' ..... ;, .......
~~
... ~'~_~.w 'v~1' ... ..
G TO 9.99 PSIS 0::: 1SO t--* .... ;........:... * ....; .. * .... ~** ...... ~** ..... ~ ........r... ..; ........ 3 TO 5.39 PSI G
~ 140 t=:::::~:: :::: :t:::::::;:::::::~::::::):::::::~::::::) :::::::'.::':'"
~ ...............:........;.......{........t.......' ....... +...... ...... 0 TO 299 PSIG
- ~ t:::::: ~:: :::::t:::::::j:::::::t::::::i:::: :::j::::::1=:........ '1
- 7,500 a,coo 9,000 10,000 11,000 12,llOO 1 PUMP 15,DOC IS,eDn 18,000 2Q,OOO 22,000 24.000 2 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. Oistractors: A 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 when flow is above -20,000 gpl11.. plausible if candidate incor~t:;~tly:. 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
. candidate pl()ts il1c:()rrectly:... ...
C 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 when flow is above -17,000 gpm. Plausible if c(3nciid.(3~J)l<:>ts incorr_e~!IY~_ Ps chometrics
.Lev~l_of Kn0V>.1lt:;cige Oiffi~l:D!Y. Time Allowanc~(rninutes) RO HIGH 10CFR55.41 (b Source Documentation Source: New Exam Item 0 Previous NRC Exam: 0
~ Modified Bank Item 0 Other Exam Bank: 0 OILT Exam Bank ReferenceJs): T-102 Sheet 3 (t:;I11J:>t:;.ci.CiElct inqLJt:;~1J()n)
Learning PLOT-2000 Obj 11 C Objective: KIA System: 295030 Low Suppression Pool Water Level Importance: RO/SRO i __ 3.9/3.9 KIA Statement: EA2.02 - Ability to determine andlor 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 Answer Question # 52 RO Choice Correct: C
~~~=- ~_~ ~ ~ ~~~()SiS or Justifi~~ti~~- -~_-~ -
I CORRECT - Jet Pump 11 is in the 'A' Recire Loop Distractors: A I INCORRECT - Core Plate flow will LOWER due to drop in core flow 1 B II NCORRECT - Jet Pump 12 shares a riser with JP 11, so its DP will ILOWER_ _____ _ D ilNCORRECT - Failed Jet Pump is not in this loop. "B" Recirc Speed will be I unaffected.
,i J _
Psychometrics Level ofKn9.wl~dgE:! __ t Diffi~LJI!y_ _Tim~,l\l!()JV...?nc~J_~inutes) i --- RO HIGH i 10CFR55.41 (b)(2) Source Documentation Source: D New Exam Item Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0 [81IL,.. .Exam Bank Referenc~(s): __ ON-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 (1) 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 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 valve is nOrrT'1§lJly_cJ()se~t\J\lI1J~h_preve_ntsfeedin_gJhe 'A' header. o 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 Kno\iVledge MEMORY
-- Diffi~LJ!!Y_~~
r----
! Time Allowance (minutes) i RO I 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
+~--'""="'---~
R~f(3rence(sL ~~~ .I ON-119
~--- ~~~~~ -- - .
Learning ! PLOT -5036-4a Objective: KIA System 295019 Partial or Total Loss of Inst. Air I Importance: RO/SRO 3.5/3.6 KIA Statement AA2.01 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Instrument§lir syst~l"1"lpres~_ure REQUIRED MATERIALS: NONE Notes and Comments:
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 Answer Key Question # 54 RO Choice Basis or Justification I Correct: B 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 tb~_RadJl.'Cl§)1e~u i l~iQg_"-~n!iICl~()Q i ntCl~~*. __ _ 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. D 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 Ra~waste Buil_dj_nJl_~(3ntilation intakE~.,NOT becal!se of sarTlpling. Psychometrics Level of Kn()""I(3Q.ge I Diffi~!-!Ity_ !-- Time Allo""aQce .(llJifl.lJ.tes1 RO _. ~-- ------- - --- --... -"._------ MEMORY i 10CFR55.41 (b)(10) Source Documentation Source: New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 J81J LT Exam Bank ---._.. __ ...... Reference( !;): T-104 Bases Learning PLOT-1560-9 Objective: KIA System: I 295038 High Off-site Release Rate Importance: RO I SRO 3.7/4.7 KIA Statement: 2.4.6 - Emergency Procedures/Plan: Knowledge of EOP mitigation strategies. REQUIRED MATERIALS: NONE Notes and Comments:
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-2-2-3-11 OB (Panel 20C003-02) + 10 inches
- Wide Range on LR-2-2-3-11 OA (Panel 20C004C) + 20 inches
- Narrow Range on LI-2-6-094B (Panel 20C005) 0 inches
- Wide Range on LI-2-2-3-85A (Panel 2:0C005) -15 inches Using the SE-11 Attachment C "Instrument List" ATTACHED SEPERATEL Y, determine which of the following statements is true.
Actual reactor water level: A. cannot be determined B. 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
*******ChOi~;------r****-----------*--*B~SiS or Justificat;;~--
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 Distractors: I I A f _ -1___ offsite_power and will continue to accu ...~~lYll'ldica!~J~~ctor water le",eL I INCORRECT - RPV level CAN be determined. U-2-2-3-85A is a post-accident monitoring instrument and per SE-11 Attachment C "Instrument I List", is powered by DC following a loss of offsite power and will continue to ____~______ _ ~ accl!rat~Jyjl'l.di<;~t~ rea_cto...~ater level. __ . I B ~NCORRECT - LR-2-2-3-110A is NOT a post-acciden~ mo~itoring
. Instrument and per SE-11 Attachment C "Instrument LISt", IS not powered b [)C foliowing_<UQss _atoffsiteup_olllf~r'____ u ____ m C INCORRECT - While Narrow Range U-2-6-094B is powered by DC following a loss of offsite power, its range is () inches to +60 inches and is indicating at minimum value due to actual reactor water level being < 0 inches.
Psychometrics Level of Knolllfl~cjg_~ [)ifficulty ______ Time Allowance (minutest RO HIGH I 10CFR55.41 (b)(7) Source Documentation Source: L8J New Exam Item 0 Previous NRC Exam: 0 Modified Bank Item 0 Other Exam Bank: 0 OILTEx~I'l"IBank il Reference(s):_ SE-11
Attachment:
.......C .. :- - Learning PLOT 1556 Obj 2.b Objective: KIA System: 295021 Loss of Shutdown Cooling Importance: RO I SRO 3.7/3.9 KIA Statement: 2.4.3 - Emergency Procedures I 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, "RCtC 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 T-----~- --+---- -............--------.------ Basis or Justification Correct: C CORRECT - as described in first NOTE of SE 13.1. Distractors: A INCORRECT - Plausible because this component is in the RCIC turbine steam flow path. B INCORRECT - Plausible because this component is in the RCIC turbine steam flow path. D INCORRECT - Plausible because the potentiometer exists and is used for routine overspeed testing, but is not used by SE 13. Psychometrics L~\I~L of_KnQwleci.9.~__ J _______Qiffic.l.JI.ty Time Allowance {Ill in...ute.§>> _. - --- RO MEMORY 10CFR55.41 (b)(7) Source Documentation Source: k8J New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0
. _OI1-T Exam Bank ______ ~ __ _
Refer~I1<:;~(~): SE-13.1-2 ----_._----_........ Learning PLOT5013-9h Objective: KIA System: 2... 95004 Partial or Co;';plete Loss of DC rw"r J;;;;~~rtance: RO/SRO 4.4/4.0 KIA Statement: 2.1.30 - Conduct of Operations: Ability to locate and operate components, including local controls. REQUIRED MATERIALS: NONE 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 - - - ' - ' - ' - - and a Group II (2) 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
-----_ .. - - - ............... __.. _-_Answer...... .
Key Choice - ..... --.-~-. Basis 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
----.---..----.-+----.-...---t altern~te off-~i~ower is.9"'a_i!<!!:>I~,~QG.~.Q_(\JQT ~tart. __
Distractors: B INCORRECT - Plausible because the candidate could believe an outboard isolation occurs. An Inboard half-isolation occurs due to loss of 20Y033,
\I\Illi.c:;hl()~E:!~weLwhertJhE:!_~R!:>us de~_~D~mi~t?~_r110mentarily ..
C 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, Plausible because the candidate could believe that E22 and E42 could be
...........*___ . ____ . L _ _ _ _ _._. feed from #2 SU bus. They arE3.fE3~Qrl().rr11ally from either 343 SU or #3SU .
Psychometrics Level of Knowledge . J?lfflcul!y._ . __ Time Allowance (r11iDlJtE;~J I RO MEMORY i I 10CFR55.41 (b)(7) Source Documentation Source: D New Exam Item Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0
.. ~ ILT Exar11..l3~n~ __ .
Reference(s): SO 53.].9, GP-?<:; _____ . Learning PLOT5054-7b Objective: KIA System: 295003 Partial or Complete Loss of AC Importance: RO I SRO 3.5/3.7 KIA Statement: AA2.04 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF AC. 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 I
--------------------- --~-------- I
- Question # 58 RO Choice ---------,------- -- ---------------------
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
---------------r------------- _rl0t to~x~_~c:jt~oo degre~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 other conditions, and 1500 degrees F is the Steam Cooling clad temp maintained in T -117 . C . INCORRECT - 1800 degrees F is the Zero Injection Clad Temp for Steam Cooling used in T -111. Plausible because this method can used to establish but NOT dLJrirl9~M~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 lather conditions. Plausible because Spray Cooling can be used to establish
- ACC under other conditions, and as previously described 1800 degrees F can be used to e~tablish ACC duringI:J11 conditions.
Psychometrics ___ Level of Knovvlec:jg~__ ~. ___ _ Dif!i~LJI~____ I Time Allowa_nce l'!linute~2J RO MEMORY I i i10CFR55.41(b)(10) Source Documentation Source: New Exam Item 0 Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: () _ _ 12~U'=-IJ::)(~m EJ§lDK__ _ Reference(s): T-117.B<:isi~~I:EJf\§ Learning PLOT 2117DBIG - 6 Objective: KIA System: 295031 Reactor Low Water Level Importance: RO 1 SRO 4.0/4.3 KIA Statement: EK3.04 - Knowledge of the reasons for the following responses as they apply to RE:f\QIOR LOW WATER LE:YE!-:Steam coo!i_rl9_ REQUIRED_. MATERIALS: NONE_. 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-101 "RPV Control"? A. Manual operation of SRVs ONLY B. Manual operation of Bypass Valves ONLY C. Both SRVs and Bypass Valves D. Neither SRVs or Bypass Valves
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 i Answer Key I ~-~~~~"-- ~~"-" Question # 59 RO Choice Basis or Justification Correct: A CORRECT - On a Group 3 is.olation 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 loss of Instrument 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 los~ of Instrument Nitrogen." 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 c,m be operated manually due to the pneumatic supply accumulators. Plausible if candidate does not know ADS SRV accumu'ator~LlPp'y~will support ADS SRVgperation. Psychometrics Level of Kn~)V\lI~c!g~ Difficulty Time AII()VJ~nceJl11inLJ~e~L~ "~R_O HIGH I 10CFR55.41 (b)(x) Source Documentation Source: cgJ New Exam Item D Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0
~~OILT Exam Bank Reference;~l __ ,..-101 Bas~~,- St~(! RCiP-4 Learning PLOT-5001A-5w Objective:
KIA System: 295020 Inadvertent Cont. Isolation RO/SRO 3.7/3.9 KIA Statement: AK1.01 - Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION: Loss of nort11C!1 heat sink REQUIRED MAT_E_R~A!-~:_~ __ -~rNONE __ ~~~~~ -=- ~- 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?
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key 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 0 F for Unit 2 and 200 0 F for Unit 3. Distractors: A INCORRECT - Unit 2 setpoint temperature is too low. Should be 230°F. B INCORRECT - The answers are reversed for both units. D INCORRECT - The Unit 3 setpoint temperature is too high. Psychometrics L~"E?L()fJS!lowled~~_ Difficulty _ J TiITI~~I()~c:Jnce (rl1if1lJtes) l RO MEMORY I I 10CFR55.41 (b)(7) Source Documentation Source: rsJ New Exam Item Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0
.... --------------j _DJl..IJ:!Cim Bank. -. . - . . -.-.~~~--
Reference(sl ______T~~I1§P~C Tables 3.3.6.1-1 Learning I PLOT-5007G obj 11 Objective: I __ L KIA System: 295032 High Secondary Containment Area Importance: RO I 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 REQUIRED MATERIALS: NONE Notes and Comments: This Question verifies candidate knowledge of a difference in s_e...!Qoints betwe~n.tJlJit~ 2 and 3 for 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. B. 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 I Question # 61 RO Choice---------r---------- - --- ---- - Basis or--------------------
------------------------ Justification Correct: I C 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. o 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 __ L§igJ1jfLcc~ntly!11itigaletbis~ff~ct.__ __ _ Psychometrics Level oJ_Knowledge - __-'~iffi~l.llty _ _ ---- Time Allowance (minutes) RO MEMORY 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 [gJ ILT Exam Bank R~ference(s): _Qt-.J-,1_0Z __________ _ Learning PLOT1550 - 4 Objective: KIA System: 295022 Loss of CRD Pumps Importance: RO I SRO 3.7 I 3.9 KIA Statement: AK3.01 - Knowledge of the reasons for the following responses as they apply to LOSS OF CRD PUMPS: Reactor SCRAM
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_-->=-,--_ 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 B CORRECT - OT-106 directs a manual scram at 24.0 inches hg. The RPS scram setpoint is 23.0 inches hg. Distracters: A INCORRECT - Plausible because 25.4 inches hg is the "COND LO VAC" alarm (206 0-2) setpoint. The Main Turbine and RFP Turbine Trip setpoint c--t:~~~~~:~.(~~usible because the RPS Scram Setpoint is 23 inches hg. I The Main Turbine and RFP Turbine Trip setpoint is 20 inches hg.
- O*--rINCORRECT-- Plausible because OT-106 directs a manual s~ram at 25.4
- inches hg if power is above BPV capability and unable to maintain load
. ___ "gfeat~!lb~_n I 300 MWe. The RPS ScrarTl Setpoint is 23 inchestlR Psychometrics
.~evel of Knowledge __ l_ DifficLJlty Time Allowance(minutes)
-1 RO MEMORY I 10CFR55.41(b)(10)
Source Documentation Source: !ZI New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: ()
.. ~ ........= ...I-.-.
LT Exam Bank ._ ..... _---- ... Reference(s): OT-106 Low Condenser Vacuum Learning PLOT 1540 DBIG Obj 2 Objective: K/A System: 295002 Loss of Main Condenser Vac Importance: RO 1 SRO 3.4/3.5 KIA Statement: AA 1.03 - Ability to operate and/or monitor the following as they apply to LOSS OF MAIN CONDENSER VACUUM: RPS REQUIRED MATERIALS: Notes and Comments:
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 "RWCU Outlet Valve"
Peach Bottom Initial Reactor Operator liCenSE! NRC Examination April 2013 Answer Key Question # 63 RO Choice -----r--~---~-- 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 10weri!l.9~V level. D INCORRECT - Plausible since the candidate has to realize that the MO-2 i 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 stopl~eJQVlferil'l~J~pV le\{ej.
Psychometrics Level of Kno'vV!~qg~ Difficulty Time Allowance (Illinut§) RO I 10CFR55.41 Source Documentation Source: [gJ New Exam Item 0 Previous NRC Exam: 0 i I o Modified Bank Item D Other Exam Bank: 0 ___ L ILT Exam Bank Reference(!)L ,OT-100, SO 12.1.A-2 Learning [-PLOT 5012 Obj ge Objective: KIA System: f295009 Low Reactor Water Level Importance: RO/SRO __ J 2.9 I 2.9 KIA Statement: AA2.03 - Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL: Reactor water cleanup blowdown rate
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
- Torus 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 PRI-SSilRE SUP tSS ON PtH::SSUH:- liM I
- .1 L -,--~----:------------~---,
28 2 A******* 20 16 1 7 -t******l--t---+--+---t---t-----jt---H 8 1 C 11 12 L3 14 ~S It=; 1/
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key Question # 64 RO -""""""-"-" Choice Basis or Justification 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 ABOVE the curve is UNSAFE and r~ClI:'i~~s a_blC?wdown. 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, _"~-'\tQI!~lJiril1g~ blo",,99wn __ __ _ " " _ _ " o INCORRECT - plausible if the Candidate does not recognize that the 2 inch drywell vent path will be isolated on a Group 3 signal of drywell Pf~SSur(3::2 p~ig. Psychometrics I L~vel_i>f Kno""I~c:tg(3 __ ~_ Difficul!y_ - -- ....
~~.~
Time Allow~~c~(minu!(3~) RO HIGH I 10CFR55.41 (b)(9) Source Documentation Source: [SI New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0
+ "
ILT Exam Bank
"""""""'=='-" - " "
Referen~eJs ): T-1 02 Ba§~§, CLI!"ePS/P- t{(3mbedc:le<ii!1JbisqlJ~f)tion} __ 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 1 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 _( 1)__, CRD ACCUMULATOR LO PRESS/HI LEVEL alarm is due to _(2)_. A. (1) insert (2) low gas pressure on accumulators B. (1) insert (2) high water level on accumulators C. (1) NOT insert (2) low gas pressure on accumulators D. (1) NOT insert (2) high water level on accumulators
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Question # 65 RO Choice - - -- -r Basis or Justification Correct: 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
......._.. +~_.. ~._ ...... rRrQ<;are reqLJ.ir..egJQ_ ~~.i!!t~grated In order to an§..werttl~_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 i alarm pushbutton indication works (depress pushbutton, if light goes out,
~LV\I'.al~UssueJfJ.igtlt r~mains illu!Dirlatec!,~§JlS.~LJE:!J . ..........~ . . .
i C t*INCORRECT - Rods WILL insert - see discussion above. Plausible if
. . . . ~__.~I~_mm_ ~~~~~1;~~~~~;~;~~~~~:i!~~:;;~~k~~~~~~r~~~~~ui~S:~~~~r~_RD D INCORRECT - Rods WILL insert - see discussion above. Plausible if
- candidate does not know how local accumulator alarm pushbutton
._.Urlgi~C3ti()n.~orks.
Psychometrics LevE:!I_9Lt5.nowl~g.ge. ____.. Diffi~lty ~----------
.Iime.AlloV\fanc~.(rll~n.utesL~_ . . . RO HIGH I 10CFR55.41 (b)(6)
Source Documentation Source: [gJ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 ILT Exam E3caD~Jl._. Reference{!5L T-216-2,j\RC 211 E-2 Learning PLOT5003A-8c Objective: KIA Syste....m ~: ~
... ..... ...... f9-5015I nco~~I~ete~SCRA-M-~--*
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 Equipment Operator non-Tech Spec rounds each shift.
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 Answer Key
~-.~-----~------- - ._. ................ ---~~-~~~-~-
Question # 66 RO Choice Basis or Justification ~~~-~~~~~-~~~~~ ~~~~~ ~~~~~ ~~~~~~ ~ ~~~~ ~~ - - ~ ~ - ~ - - Correct: 8 CORRECT - Coordinate plant activities with the Load Dispatcher is an RO activity per OP-AA-101-111, "Roles and Responsibilities of On-Shift Personnel.
--~--~~--~~-
Distractors: A INCORRECT - This is an SRO duty, specifically a duty for the Field Supervisor per OP-AA-1 01-111, "Roles and Responsibilities of On-Shift Personnel. C INCORRECT - This is a primary duty of an STNSRO per OP-AA-101-111, "Roles and Responsibilities of On-Shift Personnel. . D INCORRECT - This is Field Supervisor duty perOP-AA-101-111, "Roles and Responsibilities of On-Shift Personnel. Psychometrics Source Documentation
. . - . --t ~~d:~:~~~:~~t~m Source:
Learning II
~ New Exam Item
§
~RlflLence(s):_____ OP-P.t\:JQJ":Jl1 PLOT1529 Obj 1d
~~~_~~ ___ ~_~ __
D Previous NRC Exam: 0 D Other Exam Bank: 0 Objective: KIA system:-rG2.1~ C~nd~ct of Operati~~;-- Importance: RO I SRO 3.4/4.1 KIA Statement: 2.1.8- Ability to coordinate personnel activities outside the control room. REQUIRED MATERIALS: NONE Notes and Comments:
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 in 8LL situations. B. The site-specific procedure prevails in ALL situations. 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--~~_~=~t~-=~=~=~=~~~_ _~aSiS ~~~stif~c~ion .... ~~..-~ - Correct: C CORRECT - as stated in HU**AA-104-101, "Whenever a conflict arises between a standard procedure and a site-specific procedure, then the standard procedure shall prevail except when the site-specific procedure
............... ~.~~~ -+c.1ir"~~t~_'!~ti()_l1sJh~tel1sur~c()r11J?liallce \.I\ILtl1..i~LJ@l()_ryr~quirements".
Distracters: A i 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-specific procedure will typically overrule. _.L ~ Psychometrics Level of KnowlecJg~ i
~~~"t" ~-
MEMORY Source Documentation Source: D New Exam Item [8J Previous NRC Exam: (PB 2007) I J D ~~d:~~mB=:~~~~__ D Other Exam Bank 0 Refer~r1ce{lSl______ LHLI::-M-194-1 01 Learning PLOT-1570-8 Objective: KIA System: I G 2.1 Conduct of Operations Importance: RO I SRO I 4.6/4.6 KJA 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 psig
- 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 enterT-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 Basis or Justification Correct: 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-101: "scram condition with power above 4% or unknown". ~ ~~--------- -- Distracters: A INCORRECT - The prerequisite for GP-4 states "plant conditions require a 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
+..._~_.....+ ..m
.....a,.. ,n.:.uCJI1y'~cramJb~r~a9toLfplC!c~the mode switch in shutdown): .
B INCORRECT - This would rely on the Rector Protectiom 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 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-110direction.
Ps chometrics Levelofl5n_owledge_ Difficult)'_ Tim<>l\ll()wance (minuteS)_r . _. RO _ _ I HIGH 3.0 3 j 10CFR55.41(b)(10) Source Documentation Source: New Exam Item [g] Previous NRC Exam: (PB 2011) o Modified Bank Item Other Exam Bank: 0 ______ ..... 011TE_xa~ank RefeI~1'!2~(S): ARQ-_20§Q:-1 ;_OJ- 119; 9P-.~ T*JQL Learning PLOT-1529-2 Objective: KIA System: G 2.2 Equipment Control Importance: 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_§ystem,_and~.!ll:!derstand how operator actions_C!!!d dire~tives affec!.plant and system conditions. REQUIRED MATERIALS: . f NONE - -- ~ Notes and Comments:
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 i 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 partially correct. o INCORRECT - plausible if the candidate does not know that OPEX has demonstrated this phenomenon. Psychometrics Level of Knowle_dg~_. RO .. I MEMORY 10CFR55.41(b)(10) . Source Documentation Source: r81 New Exam Item 0 Previous NRC Exam 0 o Modified Bank Item 0 Other Exam Bank _,"=,,_,I_L
- c. T Exam Bank Referenc~{E)L GP-2(3!~~N:g§_4, Urt2l~nned Control Rod Withdrawals While Shutdown __
Learning PLOT 5003A Obj 9a Objective: KIA System G 2.2 Equipment Control Importance: RO I SRO 2.6/3.9 KIA Statement 2.2.18 Knowledge of the process for managing maintenance activities during shutdown o£~rC3.tioJ1.~--,~uch as risk assessment~,wo[kprioritization,etc. ~!~~~~:~~:~~;:LS: l-~:!,at~;io Ind-;;stry OPEJ(' i~cluding 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 Key Question # 70 RO Choice -~"---""""~""-"- Basis or Justification Correct: B CORRECT - The Group III PCIS isolation will trip and isolate Reactor Building ventilation. The failed filter inlet dampers will prevent SBGT from maintaining Reactor Building negative pressure. This will result in an unmonitored and unfiltered ground-level release. Distractors: INCORRECT -SBGT would not be exhausting Reactor Building air to the Main Stack. INCORRECT -Reactor Building ventilation dampers close on a PCIS Group i III isolation and isolate the Reactor Building from the Vent Stack. D INCORRECT -The release would not be through a monitored path. Psychometrics Lev:el of Knowle(jgE}~~ HIGH Source Documentation Source: New Exam Item C8J Previous NRC Exam (PB 2008) D Modified Bank Item Other Exam Bank
-~~""~+-",.,.......
ILT Exam Bank Refer~nge( s): T-103 E3~~E}~J~te~$~G-:2L __ " Learning PLOT-5009A-6b Objective: i KIA System Radiation Control Importance: RO 1 SRO 3.4/3.8 KIA Statement G2.3.14 - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emerg~ncY_C::()I"'I(jitions or activities. 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 (5th ) fuel assembly is loaded near a WRNM.
Peach Bottom Initial Reactor Operator LicensH NRC Examination April 2013 Answer Key Question # 71 RO I 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 INCORRECT -ON-124 entry would only be required if the count rate had I doubled two times between CCTAS steps.
~_J_~ _______ ~l~~ _~_ __~ __________________________________________________
Psychometrics Leyel <2Ll<llowledg_e___ Difficulty I 'TilrT1~_~II()VlfC!nc~(rT1_il1ut~s ) RO t 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:
~ __ ______L8JJ~X_~X~1'l1 Bank
_HE!fe~E!n~E!i?): ON-124_ ____ ___ ________ ___ __ _ Learning PLOT-PBIG-1550-2 Objective:
-- - -- -- ----~~- --
KIA System: G 2.3 Radiation Control Importance: RO I 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, access toJ~kecj ~igh radiation a,=-e~s! ~Jigning filters,_ ~tc,-, ___ REQUIRED IV!ATE_RI~L.~~ [t-.lO~E ___ _ ___ _ Notes and Comments: -- --- - --- -- -----
Peach Bottom Initial Reactor Operator License NRC Examination April 2013
- 72. According to T-BAS (INTRO) "Introduction to TRIPs and SAMPs - Bases",
TRIP and SAMP procedures provide operators with (1} 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 _.. ____ Choic~ ----~l~~-- *--~-*-~*-----~-B~~~~~ Ju~tifi~~tion .. Correct: B . CORRECT - as discussed in T-BAS (INTRO) documentation, TRIPS are symptom-based, and no risk or probability threshold is assigned. Every _ effort has been made to§l<l.Q!ess an'y mechanisti~::I.lIyp()ssible condition. Distracters: A INCORRECT - TRIPS are not bounded by Design Basis Analysis Events plausible since the candidate may incorrectly believe the TRIPS protect
...__+-_.",-a_i_ns_t._DBA accidents and SAMPS are forJ~~y()nd-DBA 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 scenarios. D INCORRECT -TRIPS are NOT rule-based. Plausible since "rule-based" is a common term associated with Human Performance activities, and sounds similar to "Event Based or Symptom Based". Psychometrics ... Level()Lf5Q.()wl~<l9.~_ .- .... ____Plfficul:tY Time Allowance (minutes) I RO . MEMORY I 10CFR55.41(b)(10) Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 I 0 Modified Bank Item 0 Other Exam Bank: 0 _______ ~.--DJ'=T Exam Bank .. .________ . ______ ... _. Reference(sl:___ I T-BA~(LN_TROL ......_ Learning PLOT 2000 DBIG Obj 3 Objective: KIA System: G 2.4 Emergency Procedures / Plan Importance: RO I SRO 3.3/4.0 KIA Statement: G2.4.18 - Knowledge of the specific bases for EOPs. REQUIRED MATERIALS: NONE Notes and Comments: . . . ._ .. __._._..___ .... L.. __ .. _~_._._
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 (i.e., provide additional time for SLC, RPS, etc. to perform their safety functions). From T-117 Bases: ADS initiation would complicate efforts to maintain RPV level within required level ranges. FURTHER, rapid
- and uncontrolled injectio~ ?f I~rge volumes of relatively c?ld, un-borated I water from low pressure Injection systems may occur. With the reactor I either critical or shutdown on boron alone, the positive reactivity addition .
due to boron dilution and ternperature reduction may result in a reactor power excursion large enough to cause substantial fuel damage. ADS is il1hiQi!~d tOQl'"~E!I"I!Jhis from happening. Distractors: A INCORRECT -ADS initiation would not cause large irregular neutron flux oscillations... it would cause a rapid reduction in reactor power due to voids. B INCORRECT -During an AnNS Torus temperature may exceed 110 degrees F before boron injection anyway due to SRV operation ...this is not the reason forJl1blbitil1i} 6[)§~___ ___ ___ _ C INCORRECT -Depressurization due to ADS initiation must also be accompanied by elevated Drywell temperature for this to occur. ..this is not tllE?_rE!c!~()D for inhibitl!!9..f.D§._____ _ Psychometrics I Level of _KI1_~\IVle~~E! __ ----- __Diffic;LJlty Ti m~AllglJI{a!1~E! _(Ill i~Lltes ) RO MEMORY I 10CFR55.41 (b)(5) Source Documentation Source: D New Exam Item [gI Previous NRC Exam: (2008 RO) D Modified Bank Item D Other Exam Bank: 0 01 LT Exam _13_,mk ______________________ Ref~rE!n_c:;e{s): ______ I-!- ~17_I3_a_~~§_ _ __ _____ _ Learning PLOT-2117-6 Objective: ___ _ KIA System G 2.4 Emergency Procedures / Plan Importance: RO / SRO 3.6/4.4 KIA Statement G2.4.22 - Knowledge of the bases for prioritizing safety functions during abnormal/emergency ~~~!::~~~~§~S:- -J~~~~~--_ . _._.
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 equal. Distractors: B INCORRECT - a failed second stage seal would cause a seal stage 2 low flow alarm since flow would bypass the bleed off line (and flow sensing switch) and be diverted out the failed seal. Plausible because candidate lJllaY-'!1i~undE~r~!?llg Recirc Seal construction. C 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. D 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 Recirc Seal construction. Psychometrics
~~ ~~_ Diffic;~I1Y ~~~_~_~~Iil1'l~Allowance Cl11inu~~l RO 10CFR55.41 (b )(7)
Source Documentation Source: New Exam Item 0 Previous NRC Exam: o Modified Bank Item ~ Other Exam Bank: (PB LORT) ILT Exam BankJL_ Reference(s): ARC 214 A-1 Learning PLOT 5002 Obj 9.k.5 Objective: KIA System G2.4 Emergency Procedures I Plan _I~p:rtan"": :~ :::0 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 _ ~~ s..Y..§tel11 int~g~it~,~c;~f'ltcijnl11en!conditiof'ls~radioact!vity release c~ontrol, etc ..
~:t~~~~~~~~::~~LS:.~ ..
.~~ ..
~ .~:~ . :~;;ti~~-matche the
. s. . th.e.
~_J~~~een Reci~~£,ul!!p mech~rttca!~eals r~C!ctor
- b. .
KIA ;~~ use.. of the~;-Ia. and tionship .. coolant
Peach Bottom Initial Reactor Operator License NRC Examination April 2013 system integrity, containment conditions, and potential radioactivity release. The seals are designed to prevent reactor coolant system leakage in the event one of two seals fail and also limit reactor coolant leakage in the event both seals fail. Reactor Recirculation Pump seal integrity is maintained by monitoring seal cavity pressure, seal cavity temperature and staging flow. The candidate must have knowledge of the status of Recirc Pump Seals and related parameters (First and Second Stage Seal Pressure) in order to correctly assess the status of the seal's safety functions of reactor coolant system integrity,
~ontai!1l1!ent _~_<?_I'1_(jl!t0I'1!!,and ra_cIJ~activity release control.
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: CORRECT - Per FF-01, Fire Brigade, "The Control Room should request I D other work 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. Oistractors~1 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 . ...
. ~.-.~ ..
Il B . INCORRECT - any individual with Vital Area access permission can
-- .......... - .. ~-~-- ~.---------------
i provide access escort. Plausible because Security and Operations and the I I work groups most closely associated with fire brigade and security
-- .jC-+;~~~~~E~~~:::~!;v~:~~~;~h~:!~~;!e:::e:-~Y:!~~~i:~s~~~m"nts i provide access escort. Plausible because Security is most closely linked with Vital Area access control.
Psychometrics __ Level ofJ'Snq~ledg~__ . ~___ ~_ Diffi~lIl!y . Time Allowc:iDceUl!liDLJJes) --------~ ..- RO ------ --- ~ MEMORY I 10CFR55.41 (b)(1 0) Source Documentation Source: r8J New Exam Item Previous NRC Exam: 0 I 0 Modified Bank Item Other Exam Bank: 0 Referencei~-tPo~LT ExaITlBan~_.__.__ . .. Learning ! PLOT 1565 Obj 2 Objective: KIA System Conduct of Operations RO/SRO 2.5/3.2 KIA Statement G2.1.13 - Kn2V\f1e._d.9.~()ffaclli!y re..gLJlr_e.ment~_ for~()_ntrolILDgyital /(;~mt!olled access. REQUIRED MATERIALS: NONE Notes and Comments: Applicable to Licensed Operators and NOT General Employee Knowledge because Licensed Operators have SOLE responsibility to direct actions out of procedure FF-01, Fire Brigf:!~!._ .. __ ~ __ ~
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 PoolIRPV 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 "Refuel Operations" 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. B. (1) +467" (2) add water with Condensate Transfer per GP-6. C. (1) +481" (2) maintain RPV level between +480 to +488 inches per GP-6. D. (1) +481" (2) lower RPV level and maintain RPV level between +470 and +477 inches per GP-6.
ATTACHMENT 3 LT-70 CORRECTION TABLE FOR +474" purpose of this Attachment is to the wi th ~.he INDICATED - - ".... _ - level value in order to maintain ACTUAL level a necessary compensate for effects on indicated Reactor 1 when Reactor water is above or below calibrated co~di (80°F) . TEMP of :ND. LEVEL TEMP of IND. LEVEL 70 474.6 II6 4 .4 72 474.5 118 4 0, 4 04.4 o 470.0 76 474. 469. 78 474.1 124 469.4 no 474.0 126 4 2 473.8 1 8 468. 84 473.7 o 468. 86 73.5 13 68.4 88 473.3 134 468. ] 90 4 1.36 46 . 92 473.0 1 8 467. 94 472.8 140 467.3 96 472.6 142 67.0 98 472.4 :44 4 100 47 . 146 466. 10 472.0 148 466. ]
}04 471. 8 o 465.8 106 7 .6 108 471. 4 o 471. 1 6 t; 6 .9 II 470.9 158 464.
II4 4 0.7 160 164.2 IF Reactor water' is 120°F, TH§l0 CATED should be maintained at +470.0" order to inta.ln Z\CTUl\L RPIl leve.! at +474".
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 I f - -- - - ----------- --- ----- -- Answer Key I Question # 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 Incorrect - RPV level per Attachment 3 is high, the distractor is plausible because if RPV level were +467" water would need to be added to restore the level to band. C Incorrect - RPV level is high, but the RPV level band per GP-6 is +470 for
+477 inches.
Psychometrics
~E!Y~LoJJSnowledge _ _________ J:~if!~ulty________ _lim e_6I1<:>wC!r1_~~{l'r1 in LJtes) SRO HIGH I 10CFR55.43(b)(7)
Source Documentation Source: ~ New Exam Item D Previous NRC Exam D Modified Bank Item D Other Exam Bank OILT ~xaIlU:3C!rl~_ __ R~fer~I'1~~(s):_ GP-E3,"~fuE!LQpE!!§l1io_ns". __ _ 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:
~~~~:~!~!~~LS:_~J~~=.*.*- ..-.. . . ..
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-I02 "Primary Containment Control" "SRV Tail Pipe Limit" Curve
*** CURVE T/L-I SRY TAIL PIPE LINtT
** 17.8
- 17.2 I
~--- ...--~..------~**
- 18.8
;:: lS.4
* ** ~
16.0 CONTI NUE EFFORTS TO * .... RESTORE TORUS LEVEL * ,.. w w lS.8 I.... TlL-ZI -'
*I.. IS.2 L (YES)
... ....a 14.8 14.4 0 700 400 800 800 1000 1200 RPY PRESSURE (P310) t (Nil) L If T-l0l HAS NOT ALREADY BEEN ENTERED.
THEN 1. MAIIUALLY SCRAM THE REACTOR USINO OP-4
- 2. ENTER T-l01 ANO EXECUTE IT CONCURRENTLY WITH THIS PROCEDURE 1....-_ _ a. _PERFORM
____ RPV_DEPRESSURIZATION _--._ _ _ _ _PER __ _--- **** -.-iJ.2-101 T-lm IIC-l L L T/L-t2 T-I02 "Primary Containment Control" "Heat Capacity Temperature Limit" Curve
- CURVE 1/1*1 I
IlEAT CAPACITY TENP LINIT SAFE REOION IS eELOW THE CURve
" . . . . . . . . . . . - . . . . - - * * * -_ I 2BO I ~,
t I REGARDLESS OF RX POWER, 260 240 -- r-- ~ IIPV PRESSURE I ENTRY INTD T-118 OR STEAN COOLING 290 I seCTI ON OF T-111 IS .!!2.! ReQUI RED. o HI 14.9 PSI G
*** 1-- 220 LOWER RPV PRESS TO MAINTAIN AN ~
OPERATING POINT BelOW CURVE T/1-1 719 I
** EXCEEDI NO COOLDOWN RATE LI MITS IF NECESSARY w
...=> 200
~ 75 TO 799.S PSI G
..... aoo
~ TO .99.9 P8IO II< ISO SOD TO 889.9 PSI 0
, L TlT-8 W
~ 100 TO 899.9 PSIIl
!i
** w lBII V 900 PSIO OR ABOVE 0-
** .=>,
170
./
** 0 lBO
- 160
~................-.
L (YES) 10.6 11 12 13 14 16 lB TORIIS LEVEL (FT> 17.1 _ *
- _~ L PERFORM AN EMERGENCV SLOWDOWN USING T-112 L -_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _....J V E8-1 L... TIT-l0
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 T -117 "LevellPower Control" (YES) L L I B >-----.. ,",: L- (NO)
- i TERMINATE AND PREVENT RPV INJECTION USl NG T-240 (ATT. 1. FI G. 3)
L- LLl-21 I PERFORM AN EMERGENCY BlOWDOWN USING T-112 -.F-llD L L- LQ-22 V EB 1
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 77 SRO 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. While RPV Level is below -19S inches, it is only S 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 -19S inches, which it can. Therefore, T-112 is not warranted under these conditions. D Operation is on the SAFE side of the HCTL curve. Psychometrics Diffic:l:Jlty ___tIiDJe A!L()\I\I'<3_I1~~ (minutes) I SRO I 10CFRSS.43{b)(S) Source Documentation Source: D New Exam Item ~ Previous NRC Exam (PB 2008)
~ Modified Bank Item D Other Exam Bank ILT Exam Bank Ref~r~l1ce(s): T-1 02 and Bases ~~~~~~~~~~~~~~~~~~-~ -- --~- ~- --
Learning PLOT-PBIG-21 02-Sa Objective: KIA System 29S030 - Low Suppression Pool Water Importance: SRO Level/S 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"
'"0 CD Q)
(") AO 60A.1-2 :::r Rev. 0 IJj 8 of 11 o
~
ATTACHMENT 1 o PBAPS BACKUP STABILITY SOLUTION 3 3d. POWER FLOW OPERATION MAP
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'0 N
o 4,] Core Flow (% of Rated) w
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 78 SRO Choice **~***~**~l*---=~~===-===-==-:--~~~§~~~~i~~~~~~£~!i~~ __~- --- .... -_. - - .... . .. Correct: B 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___~'!Qcj§_ must b~_J!!~~r1ed sincel~~_?J3J~.e~irc;pl.Hl"1ptripped. Distractors: A 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. Clpl~tin-~-i~RegiOn 2 is correct, however, raising*recirc pump speed would I ~~t~e_~corr~c~ac~~~_if~operatin~:~_~~=i~_~ ~_~it~ i~di~ations of TH I. Psychometrics I Level ofK rl 21,t\iLedge 1 Dlffic;!Jlty' ___ .. _... . _lirt"l~~~()""anc~ (minutes) SRO HIGH i I 10CFR55.43(b)(5) Source Documentation Source: D New Exam Item [8J Previous NRC Exam: (PB 2011) [8J Modified Bank Item D Other Exam Bank: 0
._.._. DJ~Tl::xam_B_an~
_g~f~r_e_nge(s): lOT-112; fl.9 60A.1-2 Learning PLOT-PBIG-1540-3, -4 Objective: KIA System: 295001 - Partial or Complete Loss of Importance: SRO Forced Core Flow Circulation 3.2 _. . . . . _._ .. - ....... _ . . _. .. _ L . _. KIA Statement: AA2.02 - Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation: Neutron Monitoring _ __ __ _ __ REQUIRED MATERIALS: 1NONE Notes and Comments: It is the SRO's job function to determine the operating point on the Power-to-Flow map (or Backup Stability Solution Power Flow __ __ _ __ __ 9p~ration Map1.'-'1/l1lc~ is. an "iml11edi~e operator action" 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-2-1 0-17 is the "Shutdown Cooling Outboard Valve" MO-2-10-18 is the "Shutdown Cooling Inboard Valve" MO-2-10-2SA 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-2-10-2SA 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 _r~=_==-_~=_~=~-:-:---~~:~a~i~~?~}~stifi~~ti~~---- - -- Correct: C Correct - (1) Correct valves. MO-2-1 0-17 AND MO-2-1 0-18 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 J?~~~ Heat Removal sy§!emJ~L ____________________________________ Distractors: A I Incorrect - (1) Correct valves. (2) Per ON-125 an Alt Decay HT removal system is required to be placed in i service. You can only defeat high pressure signal per ON-125 if RPV _______ pressLlE~_J§_~L°...£sig. ________________________ _ B Incorrect - (1) MO-25 A does not close (isolate) on hi press (>70 psig). (2) Raising reactor level to > 50" is required if in MODE 3 and <70 psig. t- Not the case here. D 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 psig. _______________________________________________________ Psychometrics I ; _Level of Knowl~dge -- ~- -- --- Difficulty Time Allowance tl11il1_u!~_~ t, - - - - - - - -SRO -- i HIGH I I I 10CFR55.43(b)(5) Source Documentation Source: IZ] New Exam Item D Previous NRC Exam: 0 o Modified Bank Item o Other Exam Bank: 0 ________ l_D_.LLl.~xamJ~an~_______________ ___ Referellc~{~L ______ 9N..l?§~_C>.~s_2rUI"1_av~il~qi!ity_()t§..h..uJ9'O_V\fn_Cooling Learning PLOT-5010 Obj 3n Objective: KIA System: 295021 Loss of Shutdown Cooling Importance: SRO 4.1 KJA Statement: 2.4.31 - Emergency Procedures I Plan: Knowledge of annunciator alarms, indications, or response procedures.
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 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 hf:l§1charpressure is IOVII~!iD_9: Distractors: B This is the correct action per ON-121 for a drifting control rod only (I.e., NOT due to a loss of instrument air). Entry into ON-119 (and direction to scramjoverrides ON-121 actions for a drifting control rod. C This 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 any control rod begins to drift. _______..L _ ... ~_~_ Psychometrics Lev~I()USrl()wlE?dge Difficulty__ Time Aliowancf:lJrnLnutes)
- SRO HIGH I 3.0 3 I 10CFR55.43(b)(5)
Source Documentation Source: New Exam Item lSJ Previous NRC Exam: (PB 2009) Modified Bank Item D Other Exam Bank: 0 ___________~I~ TE:><§1m B§1r1k ________ _ Referel1ce{~L .__ ON-11 ON-121 Learning PLOT-DBIG-1550-2 Objective: KIA System: 295019 - Partial or Complete Loss of Importance: SRO Instrument Air 4.4 KIA Statement: 2.4.49 - Emergency Procedures I Plan: Ability to perform without reference to procedures those actions_ th~t!~qLlir~jmJllediate__or~~atio~~ sJ'~tem c0rllpg"-ents 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 Key
~-~-- ~ ~ ~~
Question
-- ~~~~~~
# 81 SRO
----~------ ---- ~~ ~.~ ~~,~~- ----~-~
Choice Basis or Justification 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 EAL FS1. Distractors: A Incorrect - The candidate may select EAL MU5 since it is for reactor coolant system leakage while operating. It is related to the >10 gpm unidentified leakage and the >25 gpm identified leakage. Not applicable in this case .
..... ~---------~- --------
B 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). Reactor cannot be as hig~~~.1% with onJy 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 of containment barrier. Psychometrics Level ()fJ<no_V'.'I~~.ge ~ __ . DiffiC:LJlty _f Time AllolIV_cance (minutes) ~.~ SRO HIGH 10CFR55.43(b)(5) Source Documentation Source: I2SI New Exam Item D Previous NRC Exam Modified Bank Item D Other Exam Bank D ILT E.x.-?rnY~Il~___~_____ Reference( s}: EP-AA-1007 EAL Matrix Learning PLOT Objective: KIA System 295024 High Drywell Pressure Importance: SRO 4.2 KIA Statement 2.4.40 - Emergency Procedures I Plan: Knowledge of SRO responsibilities in emergency plan in1.pJ~lll_ef1tation. REQUIRED MATERIALS: EP-AA-1007 EAL Matrix Notes and Comments:
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 IS 3.7.2. Previous 24 hours of hourly-recorded temperatures are: 90.2 90.3 I 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 89.9 90.2
- 89.8 89.9
.. 89.9 89.9 .
89.9 90.2 I What action is required, if any, to comply with IS 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. APPLICABILITY: 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 > 90°F and ~ 92°F. normal heat sink ;s
~ 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 Condi tion 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 condition B]. PBAPS UNIT 2 3.7-3 Amendment No. 244
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
... ~ ..... --.. ~~ ... -----.-~.----~- .................- . - - - . - - - - -
Answer -Key
------ ---- -------~--- ...
Question # 82 SRO Choice --,~- ............. --+----~------ Basis or Justification Correct: C Correct - TS LCO Cond B applies since stated current reading is > gooF. Averaging previous 24 hourly readings yields 90.10°F, which is in excess of 90°F. TS 3.7.2 bases states "If the water temperature of the normal heat sink exceeds gO°F when averaged over the previous 24 hour period,
! Condition C must be entered immediately. Required action is to shutdown
! lAW condition
-~T-~*"-~~~~---"""~~~--
C. Distractors: A i Incorrect - see above for TS LCO discussion - Plausible because candidate could confuse 1 hour in Required Action with once per hour in
~~:!~~~~nst~~:b:::;:~e~~~~~i~~snC~;:~~!_~:I:-~;i~~:~:::::teIY, B--
r candidate could incorrectly calculate conditions and incorrectly determine II that Condition
. - - -B -- applies.
o *Incorrect - see above for TS LCO discussion - Plausible because candidate could incorrectly determine average temperature as being below 90 F and degp~_n_qfl.Jr!I1E3rC3ction is required,_ Psychometrics Level_()f Knowled~_ ------
~gifflgl:llty ___ Time Allow..c!llc~Jminutes) SRO HIGH 10CFR55.43(b)(2)
Source Documentation Source: ~ New Exam Item Previous NRC Exam D Modified Bank Item Other Exam Bank Refere~~~-(~2:~--~l~i~~.I~:~~:~::~:----=_~=- Learning PLOT 1858 Obj 1 Objective: I ____ --I KIA System 295018 Partial or Total Loss of CCW Importance: SRO 4.7 KIA Statement 2.2.40 - Equipment Control: Ability to apply technical specifications for a system. REQUIRED MATERIALS: NONE Notes and Comments: At Peach Bottom, ESW provides CCW for a most of the ECCS compon~l}ts!b~t _~~quire CCW.
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.
- PRfTR-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, if any. Technical Specification section 3.6 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. 6 Page 5 of 11 6.2 Drywell Temperature Calculations NOTE IF all temperat.ure points in a given zone on TI-2::.01 are out ot service, THEN Dry,,-'ell Bulk Average Temperat.ure calculation on Data Sheet 1 will be INVALID AND EITHER TI -2 501 Point. 136 OR PR/TR-4805 must be used to calculate APPROXIMATE Drywell Bulk Average Temperature for entry int.o ON-120 and T-I02. Refer to Precaution Step 4.2.2 for limitations on T-223, T-I02 or SAMP-2 actions ,,*dt.h Invalid DrY'~lell Bulk Average Temperature. 6.2.1 IF TI-2501 or corresponding PMS computer points have at least. 1 val iel temperat.ure point. in each of Zones 1 throu9h 5, THEN CALCULATE DrY'/lell Bulk Average Temperature using calculatic,n on Data Sheet 1. OTHERWISE, MIA this
- 3tep.
6.2.2 IF TI-2501 or corresponding PMS computer point.s does NOT have at least 1 valid temperature point in each of Zones 1 t.lu:ough 5, THEN CALCULATE APPROXIMATE Drywell Bulk Average Temperature as follows. OTHERWISE, N/A this st.ep:
- 1. RECORD temperature reading from EITHER TI-2501 Point 116 OR PR/TR-4805 "Containment Temp" at Panel 20C003-02.
'F Instrument. Used
- 2. ADD 10:F to t,emperature recorded in substep 1 ahove to determine APPROXIMATE Drywell Bulk Average Temperature.
- F + 10'F =
6.2.3 VERIFY Drywell Bulk Average Temperature is less than 140=F. R _ _ _ .'_ _.'
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Choice Basis or Justification Correct: B 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 and OW ChilJ~rs to ~e~~()r~ OWJ(3rTl~eJQ~ 14Q~E* Oistractors: A
- INCORRECT - Calculated average OW temp is 141°F. Entry into ON-120 I. 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. INCORRECT - Calculated average DW temp is 141°F. No current entry condition into T-102 exists. ~ .. --------------------------------'-----------------------------------, Psychometrics Lev~1 of Kn_()wledg~. ~ __ Diffic;lo!!ty. r Time Allowance (minutes) Ii SRO HIGH -. . - _...- 10CFR55.43(b)(5) I Source Documentation Source: rgJ New Exam Item 0 Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: 0 ILT Exam Bank R~ference(~L .. RT 40C-530-2 ~Dryvyell T~'!1ReratlJ~e M0I'1~t()riI19" Learning PLOT-5040C Obj 10 Objective: KIA System: 295012 High Drywell Temperature Importance: SRO 3.9 KIA Statement: AA2.01 - Ability to determine and/or interpret the following as they apply to HIGH ORYWELL TEMPERATURE: Orywell temperature REQUIRED MATERIALS: Tectt~ec Section 3.6 CQr'I!ainlllent Systems 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 Blowdown". TABLE SC/L-2 WATER LEVEL-ALARM AND ACTION LEVELS ALARM ACT! ON LEVEL AREA INDICATION ST ATUS LEVEL UNIT 2 UNI T a: TORUS ROOM 6 IN. 100 IN. 100 IN. LI -2( 3)919 SUMP ROOM NONE 1 FT 7 IN. 1 FT 4 IN. LOCAL SIGN OR RCI CROON SIN. 2 FT 5 IN. 2 FT 5 IN. LOCAL 81 GN OR HPCI ROON 6 IN. Z FT Z IN. 2 FT 2 IN. LOCAL 81 GN A RHR ROON 6 IN. Z FT 11 IN. 3 FT 5 IN. LOCAL SI GN OR C RHR ROOM SIN. 1 FT 3 IN. 3 FT 5 IN. LOCAL SI GN B RHR ROOM 8 IN. 1 FT 5 IN. 3 FT 5 IN. LOCAL 81 GN OR o RHR ROOM SIN. 3 FT 4 IN. 3 FT 5 IN. LOCAL SIGN A CS ROOM 6 IN. 1 FT 10 IN. 3 FT 3 IN. LOCAL SIGN OR C CS ROON 6 IN. 3 FT 8 IN. 3 FT 1 IN. LOCAL SIGN B CS ROOM 8 IN. 2 FT 5 IN. 2 FT 4 IN. LOCAL SI GN OR o CS ROOM 8 IN. 2 FT 3 IN. 2 FT 10 IN. LOCAL SIGN
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer UUEitstJ()n # 84 SRO r--~--~~~~~---.-~-~~~~-----~--- ~- ---~~~~~~ ----~~~~~~-~. --- ~~~~~~--- --- ~~ --~ .. Choice f---**------*----****--,-----****----j------****-- Basis or
.....-.~-- .....
Justification
~-- ... - - - - - - . -.--- - - - - - .... - - .
Correct: B 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 I Incorrect -.'---P-r-oc-e-d~;e SE-9"Radioa~ti~; Liquid-Spill" provides direction to i i isolate the source of the spill. Direction to control room level is from T-103.
~c flnc~;rect -.GP4 w~uldbe performed if the condition w~s caused by a primary system leak. It is not a primary system leak.
o . 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 ______ ._ _____ __-.---ll~~~_~ ___..___ . . .______.. . _________.. . ______.___.. . .___ __ . . ____. ...~_______._ Psychometrics b~'II~l of KI'"10V'!l~dge_ ________Qiffic.!-JI!L____ _Tirll~ AI!()wal1~~J'1"1.irllJ!~s L. SRO HIGH 10CFR55.43(b)(5) Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 o ILT Exam Bank
!-T-103----- --Co;rtainment Control" and associated bases Learning PLOT 2103 DBIG Obj 5 Objective:
KIA System: 295036 Secondary Containment High Importance: SRO Sump/Area Water Level 4.6
.. _ _ _ .~ _ _ _ ...._ _ _ __ ..i~_ ....._ _ _ _ . __ ._~_ ....._ _ _.*._~ _ _ _.... _ _ ~_. _ _ _ _ ... _***. _ _ _ ._
Peach Bottqm 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.
r t'(.cAJ..","elj 'lt~ ~ rR\e~~ lY Given the above conditions, which of the following is tA8-~St8.t m8thQd to insect tbe-mB~Ufflber of eontroll pds?
+05' UJ...... ~~r 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 Question # 85 SRO Choice Basis or Justification Correct: - The conditions indicate that there is a hydraulic ATWS. 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: Incorrect - with a hydraulic ATWS T-213 will not be effective. Incorrect - with a hydraulic ATWS T-214 will not be effective. C Incorrect - T -215 will only allow insertion of one control rod at a time. Level of Knowled e _____ Diff!CultYmm___ _Time Allo"Yc:lnce (minut~~} SRO HIGH 10CFR55.43 b Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam o Modified Bank Item 0 Other Exam Bank _........._.~D_LL.I.Exam BalJ~~. _____m~ ____.. . . . _________~
~ef~.enCe(~~. T-2'1~, "Maximizing ~RQ.flow to the Reactor Ve.§_s..el" __
Learning PLOT 2101 Obj 7q Objective:
- -- ~-~ ~ ~~-~~ -
KIA System 295015 Incomplete SCRAM Importance: SRO ______~_. ___L_____. 4.1 KIA Statement 2.4.34 - Emergency Procedures I Plan: Knowledge of RO tasks performed outside the main control ro.om during an ~I"l"l~rgency andJ~~es-,=,I~~nt operational effects.mm ______ ~ ___ _ REQUIRED MATERIALS: NONE Notes and Comments: . m.*********** _ _ _ _ _
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 86. The following conditions exist immediately 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: D Correct - ADS will not initiate. Drywell pressure must be above 2 psig or RPV level must be below -160 inches for 9.5 minutes. Neither of these conditions are present. T-101 "RPV Control", step RC/P-4, directs a nitro en su I ber~~1Qred for long t~r~qp~ration of the APS 'VE'-~~~~ ... _ u Distractors: A Incorrect - ADS will not initiate without Drywell pressure above 2 psig or RPV level must be below -160 inches for 9.5 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 9.5 minutes. The ADS SRV
............. u._.._.. ~_~___~!accumulators do not provide a long term supply of nitrogen.
C Incorrect - PRY level is above a level re~uiring an Emergency Blowdown. The ADS valve position will be changed based on the crews long term __ ~~ ability to control RPV lev~l abo'V.eu-172". Psychometrics
._Level ofJS!lQ.'I"J~Qg_~_ r- ................. _ Difficult.Y.__ u ___ [ Tilllet.II9'v'1!c:iI1Ce .(f1:linut~~l SRO HIGH , 10CFR55. 43(b)( 5)
Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam o Modified Bank Item 0 Other Exam Bank ILT Exam Bank Reference( s): T-101 "RPV Control" .. _. ___._ _.._..___~ SOJ§.1.A-?"§RV and SV§ystem Alignmentfqr Nor.!llC!L9.p~~EI~.ion"- __.___ Learning PLOT-5001G Obj. 10 Objective: KIA System 218000 ADS Importance: SRO 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 _~()_n~~qlJenc~§_of those abl}ormalconditions or gperations.: Large breakbQC.:..:.A ..c __....... __ .......
~~~l~~~~:!!~L~"_tONE- u~_~_ -~- .
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 20D21 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 '125 250 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
~~j~~I'l~~# 87 SRO Choice Basis or Justification
~--r"-~""""~~'~-~-~~--~"'~~---~--~----~-----~------~.-- ----~---.
Correct: B Correct - 20D21 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 - 20D21 is the Division I battery. RCIC is powered from the Division I battery. Since 'A' Core Spray loop and 'A' RHR pump are INOP this requires a 3.0.3 shutdown not just a 14 day action statement. SO _~ __'+ __~' _____1__ 5,TI3.:~? doe~!Lo.t resol"'~E!tt!e 10"""'QI!~~g.-:.o.lJrlcJ.Q.()"cJition. C Incorrect - 20D21 is the Division I battery. HPCI is powered from Division II. D Incorrect - 20D212 is the Division I battery. HPCI is powered from Division II. SO 57B.1-2 does not resolve the low voltage/ground condition. Psychometrics ~ LeveJof Kno'JIIledg~_ . . ____ Qlff~ulty___~~ c--Time ~lIo'Jllance(r!linuLesLj _~~___ SRQ~~_ . HIGH : 10CFR55.43(b)(2) Source Documentation Source: [8J New Exam Item D Previous NRC Exam D Modified Bank Item D Other Exam Bank ILT Exam Bank
--+------=,~~--
Refer~r1.C:~:"~ SE-1 .1ech_§eE9.c 3.5 Learning PLOT-5057 Obj. 10 Objective: KIA System . 263000 DC Electrical Distribution Importance: SRO L______ ~ _ .__ . . _~.~ 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
-J NON" ___ ~=~~.=_~_________..
t.bos~C1b.rlprm§!Lc:()ndition~o.cc>peratiQ!l_s: GrolJrl(:t~___.
~~~~::::~~~LS:
Peach Bottom Initial Senior Reactor Operator L.icense 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 panel 2AD045.
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 section 3.3 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.
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 Key !QUestion # 88 SRO I Choice Basis or Justification l--- -. Correct: C 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 ______ f?~~ced_ in tr~ithin 12 hOLlrs. ________________ _ 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. B Incorrect - Control rod withdrawal cannot continue because of the control rod block. 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. Psychometrics ___LE?\I~I of 1Sr1()wl~~_______ .g!ffic~Jty______ _.lim~~lIo""'~_r1~e.(l11i_nute~) - --- SRO HIGH 10CFR55.43(b )(2) Source Documentation Source: II ~ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 D ILT Exam Bank 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 PO.'ll'E?r . ____________ m __ Learning PLOT 5060C Obj 10.a Objective:
- - - - - - - - - - - - - - + - _ . _ - - - _ ._ _. _ - - - - _ . _ - - _ .
KIA System: 2150031RM -l'~-po-rt-a-n~;~---SR.O
... _... _......_. ____..__._____________ __ .. __ _______,1..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 Section 3.3 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 Key Question # 89 SRO i ...................~.-.~-.---.--..~--~.-~ .... ~*-T*~~*-~---- ~~~~..~. Choice Basis or Justification Correct: C Correct - The loss of DC control power prevents 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 which will first require the A' RPS to be transferred to the alternate source per SO ~._~_. ____~~~_~.. --+~ ____ t..§..QE*6:!.~:.~~~TJ:.a.DJil!~rring ~l'§ PO\Ner Supplies" ............ _. __ . Distractors: A I 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 i RPS output circuit breakers unavailable, auto trip capability is lost. Tech Spec i 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 60F.6.A:::?"Transferring RPS Power Suppli~~"-__ ~.
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 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 source ef SO ~QEJ3.A-2 -"TranstE?f!Lng.BP'§ £'..Q\NeL§.t!PPlit::~".__ ... Psychometrics LeveIQfJ<no)All(3qgE:1_ ~. DifficlJj!y Time Allowance IUt.9..S) SRO HIGH 10CFR55.43(b)(5) Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 [2J Modified Bank Item Other Exam Bank: 0 ILT Exam Bank Reference(s): TS 3.3.8.2, E-2365 Learning PLOT 5060F Obj 8 O~j~.ective: KIA System: 212000 RPS Importance: SRO 4.0 KIA Statement: 2.4.50 - Emergency Procedures I Plan: Ability to verify system alarm setpoints and operate controls identified ..... in the alarm res onse manual
'----- ----.-~-- ' - - - - - - - _. .
REQUIRED MATERIALS: Tech~~~~c s~C:!!()1"1_3.~ (Unit~l Notes and Comments:
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 90. The following conditions are present on Unit 2 during an A TWS:
- A Loss of Off-Site power has occurred
- ARC-001 C-1 "E12 Bus Differential or Overcurrent Relays" is in alarm
- All other 4KV buses are energized
* "B" CRD pump is blocked and NOT available
- 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 'B' SBLC pump has been started o SBLC pump discharge pressure is 1400 psig o SBLC tank level is 56 percent o RWCU is isolated Per T-1 01 "RPV Control", (1) which one of the following is correct for these conditions and (2) what direction is given to the URO? A. (1) SBLC is injecting {2} Monitor SBLC tank level per T-101 step RC/Q-16. B. (1) SBLC is NOT injecting (2) Start the "A" SBLC pump per RRC 11.1-2. C. (1) SBLC is NOT injecting (2) Perform T-211 "CRD System Non-enriched Boric Acid and Borax Injection". D. (1) SBLC is NOT injecting (2) Perform T-212 "RWCU System SBLC Injection". L. <YES) (NO) L.
- RC/Q-1S r=------ ------~ INJECT BORON INTO THE RPV USING:
II - N lEVEL DROPS TO 0..-11
- eRO VIA THE SBle TANK ( T-210)
I IF SBle TA K '" I
- CRO VIA THE CONDENSATE PRECOAT ILL-_HEN_T_RI_PT_HE_SB'::_PU_NP _ _ _ _ II TANK (1-211>
.... _ . . . . RWCU VIA SBle TANK (T-212)
\..... RC/Q-IS t \. . t
~-------------~---------------~
RC/Q-17
.... ,- - - - - - - - - - - - - . . .""......0- _. _. - - - - - _. _.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
---.. _.- _... - ---.. -~-~-~- -----~---~.--
Answer Key Question # 90 SRO Choice
-~--------- ---,---------i---.-.- --~--------~----------_. __ _
Basis or Justification Correct: D Based on the given conditions, SBLC is not injecting into the RPV: 1400 psig pump discharge pressure indicates the 'B' SBLC pump discharge relief valve is lifting (due to a blocked flow path). The 'A' SBLC pump is powered from E124-R-C and has not power. T-211 cannot be performed without at least one CRD system pump available. Therefore, T-212 is the only option availa~~'--'Nhi~h c~I1~~il11plemente.9_~":'~l'1t~0':lgh R.WC'yJ~_lsolated. 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 11 The applicant must know that the 'A' SBLC pump is powered from E124-R C and has not power, hence this option is unavailable. .
-.--~---~- -.~--~.. ,-- .... ---~-.- .... _.. . ..
C 1 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. E 12 is the power supply to the '2A' CRD pump and it is locked out hence '2A' CRD pump is
. -----~------~~-
unavailable . Psychometrics le,,-~I O~::wledge t.__.~__D_iffi_IC.Ult~ . J :nme A1iclwance (",;nut~ 110CFR~~~3(b)(5) Source Documentation Source: D New Exam Item D Previous NRC Exam
, [g] Modified Bank Item: (PB 2(09) D Other Exam Bank: 0 I ILT Exam Bank Ref~rence(s): ! T-101 and Bases; P&ID M-358, Sheet 1 T PLOT~501~~~--~-~-**-~--
Learning Objective: .- .. ~ ........ --~--~------ _. . . - - - - - - - -.. ----~.-- ........ -----~-. KIA System: 211000 - Standby Liquid Control Importance: SRO 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 abnormCiI conditions or op.~r§lJ:iQI1~: Inadequate system flow. _.. ___~ ___ ~ _~~~~~~:~::~::~ALs:_:INONE--- . ______ = .
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" 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 Distractors: A 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-=-'c=....
the isolation. C Incorrect - Reactor Building ventilation will isolate on the Group III isolation signal of RPV level < +1 inch. D Incorrect - Reactor Building ventilation will isolate on the Group III isolation signal of RPV level < +1 inch. ,.. __...__...................** _ .**.** ~_ *.................... L.._~. _ _ . __ .... _ _ _ _ _ _ **.* Psychometrics Level of KnOWled9.~_j HIGH ' Source Documentation Source: ~ New Exam Item Previous NRC Exam D Modified Bank Item Other Exam Bank ILT Exam Bank Reference(s): T -222 "Secondary Containment Ventilation Bypass" _.__.. . . . . . . . . . . . . . . __._~_+I-103 "Secondary Containment C9I'lt~91"_ _.____._.... _.._.. Learning PLOT 2103 Obj 7 9~lective:
.. ~.
KIA System 288000 Plant Ventilation Importance: SRO
............................... ~~-~- .. ...
~ ~.- --.-. -
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 or operations: Low reactor water l e v e l _ _ ~ __
~::~:~~;~:::~A~s;-_jNON~____~ --- ____ _
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".
A four cell section of the core is displayed on the following page. Per NF-AA-330-1 001 "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 Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Question # 92 SRO 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 reactivity insertion, core thermal limit violation, and fuel clad overheating are all concerns of a mis-oriented fuel bundle lAW FASR 14.5 and Appendix J and NF-AA-330 1001. Distractors: A Incorrect - Actions are required. See D above. 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 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 overheating is a concern of a mis-oriented fuel bundle lAW FASR 14.5 and
.~ ..~ ..........J ~~ndix J and NF-AA-~3~9-=j 0::.0::..::1..:....~_~....~.......................
Psychometrics Level of Knowl~cig~ ~ .... ._._ _ Diffi~LJlty ~~ Timei\IL0II!§lI'1Ce (lllif1LJt~t>> SRO HIGH 10CFR55.43(b)(6
&7)
Source Documentation Source: [g] New Exam Item 0 Previous NRC Exam () Modified Bank Item D Other Exam Bank ILT Exam Bank Reference( s): FH-6C "Core Component Movement - Core Transfers" NF-AA-330-1001 "Core Verification Guideline". Learning PLOT-1535 Obj 2
.9.!?i~~!t~~~
KIA System SRO 3.7 KIA Statement K5.05 - Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING Fuel orientation J~;g!JLRED M.A: ...:T.:.:. E 1A ......... ..j!.~:.::.:=..:..:=::.. ............._ _
. . ::cR..::.::.:.:L:S.:::..:
Notes and Comments:
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 5 0> rn 0 C-o 6Mil'1~tes 10 12 14 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 Question # 93 SRO Choice Basis or Justification --.-~~.--~ ..~......... ----------- Correct: - The recycle valve opened as indicated by the flat spot on the at approx. 7 psig. The rise in pressure indicates that the recycle 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
~()~~r_r~du.f.tioll"- _ . . . .___~~ __~_._ .. _
Distractors: A I 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
- service will not re'!l~91'Jhe .2r_o..~l~I'!!:~~ __ . . . . . . ____._ ..
B INCORRECT - The recycle valve is not successfully controlling Recombiner pressure as indicated by the rise in system pressure to 8 pSig then the as MO-2990 isolated. 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 s stem. Leve.l.<?LI5!:l.Q~ed_g~.__ . __ ~Qlffi~~_I~Y_____0ime Allow~!:I~_~_(m_i!:l.l:J~~~_)___§8.0~ __ HIGH I 10CFR55.43(b)(2) Source Documentation Source: [8J New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 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 Objective.: KIA System: 271000 Off-gas Importance: SRO 4.2 KIA Statement: 2.1.25 - Conduct of Operations: Ability to interpret references materials, such as graphs, curves, tables, etc.
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 with Standby Gas Treatment system INOPERABLE? 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 corr;ct:-rc~
~--------~--
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 SBGT. Distractors: A Incorrect - OPDRVs are prohibited without SBGT per Tech Spec 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".
----------J----------
D _____In_c_orrect - Movement of recently irradiated fuel assemblies are prohibited
~ ~~~ ~Without SBGT per FH-6C, "Core Component Movement - Core Transfers".
I Psychometrics Source Documentation Source: [gJ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 _ 0_1 LTExam BC!I1K _________________ _ J~~f~~enc~(~~___ ~ ~ _E!-I_-J3g'-~~Q()r~J:;~J!1l?()l!~!!Ltlllo\/~_~~I1! - Core_Tfar1?f~~s" _ Learning PLOT 5019 Obj 9c Objective: KIA System: Generic - Conduct of Operations Importance: SRO
--" - - -" - -- -, - ~---~
3.9 KIA Statement:
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 Choice *~Ih*~-~-~- Basis or Justification ~~-~ . Correct: B Per CC-AA-112, this is NOT an excluded activity and therefore requires a TCC. I Distractors: Per CC-AA-112, this is an excluded activity and therefore does NOT require a TCC. Per CC-AA-112, this is an excluded activity and therefore does NOT require a TCC. o Per CC-AA-112, this is c:m excluded activity and therefore does NOT require a TCC. Psychometrics Level ()LtSn()Y\'"ledg~ I QiffLclJJ!y_ ---- Time AI!()*v*!,Clng~tl11jl1lJtesJ. SRO Memory I 10CFR55.43(b)(3) Source Documentation Source: D New Exam Item Previous NRC Exam: 0 D Modified Bank Item Other Exam Bank: 0 ILT Exam Bank l
!**g-g-~-Jj?-
Learning PLOT-1570-19 ObJ.*ectiv.e..:.. .. ~ KIA System: .....~ 1- Gen~ric - Equipment Contr~1 Importance: SRO _...........________________J ____ 3.2 KIA Statement: G 2.2.5 Knowledge of the process for making design or operating changes to the facility. Rf:q~_1R_f: DPJ!~I.E Rc.c:I. c..:Ac=LS.=.c: ._____+. . .N. .:.O
. ::..:..N=E:...~...
Notes and Comments:
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013
- 96. While operating at power, OT-101 is entered due to rising drywell pressure with the following conditions:
- Drywell pressure is 0.7 psig and rising slowly
- Drywell temperature is 137 degrees F and rising slowly
- The drywell is being vented using SO 7B.3.A-2 "Containment Atmosphere Pressure Control and Nitrogen Makeup" Subsequently, drywell radiation suddenly spikes to 2.5 x 10-1 /-lCi/cc and continues to rise.
Per OT-101, what action must now be taken? A. Terminate venting. B. Re-align the vent path to the Torus. C. Perform a GP-15 "Local Evacuation" of the Radwaste Building. D. Direct Rad Protection to perform dose calculations from Main Stack data.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 96 SRO Choice Basis or Justification Correct: A Per OT-101 Bases, venting is required to be terminated to ensure the ODCM release limits are not exceeded. i Distracters: B This is not in accordance with the direction in OT-1 01, but may be confused !I with the guidance for venting in T -200. ' C Evacuation of the Radwaste Building is not required for this venting operation. This would only be true for T-200 venting. Offsite dose calculations are not required since terminating the venting operation ensures ODCM (and Tech Spec) limits are not exceeded. Psychometrics Level of Knowledge Difficulty Time Allowance
~ ~~~ ......................
(minutes)
~.~-.~ .. - .. --~ ... 1-*
SRO LOW 3.5 3 10CFR55.43(b)(4) Source Documentation Source: o New Exam Item ~ Previous NRC Exam: (2011 PB) o Modified Bank Item o Other Exam Bank: () ILT Exam Bank Referef"lc;~lsl_~ 0T::J9j;§Q}B.3*6-2 ___~~_~____ _ Learning PLOT-1540-04 Objective: KIA System: G2.3 - Radiation Control SRO 3.8 KIA Statement: 2.3.14 - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or ~~mE:rgE:f"lc;ygonditions or activities. ~ ~_~______ REQUIRED MATERIALS: NONE Notes and Comments:
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 Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Choice Basis or Justification Correct: C Correct - OT-113, "Loss of Stator Cooling" follow up actions explains that the alarms received do to make a valid loss of SWC and with main g~,=,erator am ps > 9,48Q§lIllP_s,.CiG P-!§crami~~~quir~d. Distractors: A
- Incorrect - GP-4 "Scram" is required, not GP-S.
-- ------------ ......... _j--_._----,-----_._---_............_-----
B - GP-4 "Scram" is required, not GP-9. Incorrect - the alarms received do to make a valid loss of SWC and with main generator amps> 9,480 amps, a GP-4 scram is required. Psychometrics _L~\leJ_()f . K'='()wl~.sig~ J_________Piff~lJliL_____ Tirr1e_~llow'!nc_~ (fTlinute~)J __ I SRO HIGH I 10CFRSS.43(b)(12) Source Documentation Source: rgJ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 Reference(s): "Loss of Stat()~_gQ()lil]9"_ Learning PLOT-5050A Obj 6 Objective:
- - - -t**-- -------------- .
KIA System: Importance: SRO 3.7 KIA Statement: 2.4.4 Ability to recognize abnormal indications for system operating parameters which are ent~-Ievel conditions for emergency and abnormal operating procedures. REQUIRED MATERIALS: NONE Notes and Comments: ------------- Imb~QR.~()!os
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 IJCi/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 IJCilcc.
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 ONL'v:.
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 Answer Key Question # 98 SRO Choice .. ~.---- Basis or Justification Correct: C Correct - 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 tllan 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-3-17-1S2. The GP-9 power reduction would include use of inserting control rods once core flow limit is reached. Distractors: 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 i carryover with the steam. A change in pH causes the N-16 to combine with the I free hydrogen to produce ammonia (NH3) and nitrous oxide (N20). Ammonia and nitrous oxide are more v....o.latile; therefore more N-16 carries over with the steam.
The rise in N-16 only indicates on the main steam line radiation monitors because
. L _* . *
- _ . _ _ ()fthe short half life ofJhe N-16.
I B Incorrect - See A above. 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-3-17-1S2. The GP-9
! power reduction would include use of inserting control rods once the core flow limit
._--_..*....... _ . . . .- ....*._is reached Psychometrics Lev.eL()Lt<!l()l,/Vl~d_g~ I ... .Plffic_u!ty_. (--Time Allowance (minutes) l SRO HIGH I 10CFRSS.43(b)(S)
Source Documentation Source: ~ New Exam Item Previous NRC Exam: 0 o Modified Bank Item o Other Exam Bank: 0 o ILT Exam Bank Reference(s L._ .....A8<;;}18 E-1 Learning PLOT -S008 Obj 7i Objective: ........ _--- --- ...... _ - - - - - - - - KIA System: None Importance: SRO 3.1 KIA Statement: 2.3.1S Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
- ~;;~~~~~::~~~~-t~N~~
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 is to (2) A. (1) safe (2) maintain HPCI flow at the current value B. (1) unsafe (2) maintain HPCI flow at the current value C. (1) unsafe (2) lower HPCI turbine speed to return operation to the safe side of the NPSH curve D. (1) unsafe (2) secure HPCI; any further turbine speed reduction will cause system damage u.. 260 r--~.-~.-~.---:-.-":"'.--:.-~,:----:-.--:-,-
* , , I ..
- I J t
~.:---....,
- TORUS PRESSURE
*-f***-~--*--r***~~**--*r--*-'----*~~--*T--**~*-~*-r*~--
240 .... :- . . _.....:.... " . . . ~ . -..... ~ . . -- --~- -- . . ~ .............~ ......... ;-- .... ~ ......... -~"" .....
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- I
- f l CL 2: 220 ~:"!!':'~-'l-:-~-:'-j':~-~"~'~t"~"~"+'-"~'--~~:::
~ **~*****.*****r***--.*-'-* . *********f*****,** __ ow W . -; ... -,-' - -, r - - _. -: *** _. " ,'** _ ** t " ' - . : *** ~* *** 'lOP SI G AND ABavE I- 200 .. :** -. -:- *.*. ~ .. _. ~.. . . 6 TO 9.99 PSI G
(/) 180 3 TO 5.98 PSI G
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0:: 1SO L----L---1_-'----L._..l...---..!.------_'__---'-_"'--_ 0 T0 2.99 PSI G o o 4.000 I 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. - - -- D 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 ofthe 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 L~~~I . . 9flS.r1()~~~ge__ ____ l2lfficulty ...... ------ Time Allowance (minutes) SRO HIGH I 10CFR55.43(b )(5) Source Documentation Source: [gJ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item Other Exam Bank: 0 ILT Exam Bank _.. _-------_ .... Reference(s): T-117 "Level Power Control" step LQ-18 and LQ-19 1-102 , "PrimaryJ:~_(mtainm.E3nt CQr1!,"-ol"§heet 3 Learning PLOT-2117 Obj 5 Objective: KJA System: None Importance: SRO 4.0 KIA Statement: 2.1.32 Ability to explain and apply all system limits and precautions. REQUIRED MATERIALS: NE Notes and Comments:
Peach Bottom Initial Senior Reactor Operator License NRC Examination April 2013 100. During the performance of ST-O-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-O-032-301-3, the 3C HPSW pump _ ....... (1"-'-)_ and status is tracked by _~,---_. 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 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 in the ALERT range, the pump remains operable. Initiate an Issue to place the pump~!!lm~r~~~~c:lJest fr~~~eJ1cy ""_~~_""""~~_~ 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 that are in_op: "~~~ _____~ __"""~__~""" ____"________ B Incorrect - Pump is not declared inoperable until performance reaches the Action Range. A SOTC entry is not required per the ST. Only for Systems, Structures, and Components that are inoperable. T.S. 3.7.1 only entered if pl!!!1r:~J~ inoperable. ____~ __ ~ 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
~~,,~Lof ~~(.l\l\lI~c:I_9e.
1 ~ ____ J?JffJgLJlty ___ ~""" "r Tirn_~AII()\I\I~nc::~" (mif1u~~st SRO MEMORY 10CFR55.43(b )(2) Source Documentation Source: rEJ New Exam Item Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: 0
---.---~-
I LT Exam Bank ---- Ref~r~_rlC::~(~)~~ __ i ST-O-0~2-301 "HPSW Pumy, Valve, and Flow Functional and In-service Test" Learning ObIE3~Qv_~~_ KIA System: """"- .None tI PLOT 5032 Obj 9 Importance: SRO 4.3 KIA Statement: G 2.2.14 - Knowledge of the process for controlling equipment configuration or status.}}