Text

Final Written Examination with Answer Key (401-5 Format) (Folder 3)
	ML110630064
Person / Time
Site:	Peach Bottom
Issue date:	01/11/2011
From:	D'Antonio J; Operations Branch I
To:	Wasong A; Exelon Generation Co
	Hansell S
Shared Package
ML102220201	List: IR 05000277/2011301; ML102880050; ML102880462; ML110210429; ML110620164; ML110620239; ML110620274; ML110630047; ML110630064; ML110630070; ML110630120; ML111580037; ML111661880;
References
	TAC U01769
	Download: ML110630064 (221)
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Peach Bottom Initial Reactor Operator NRC Examination January 2011

1. The 2A Drywell Equipment Drain pump was running during automatic pump down of the Drywell Equipment Drain Sump when the following alarms were received:

GROUP II1In INBOARD ISOL. RELAYS NOT RESET (214 D-I)

DRYWELL SUMP VALVES 94 OR 95 CLOSED (215 E-4)

Which one of the following shows the design response of the Drywell Equipment Drain (DWED) pumps and valves?

A. Inboard Isolation AO-094 Closed Outboard Isolation AO-095 Open 2ADWEDPump Running 2B DWEDPump Tripped B. Inboard Isolation AO-094 Closed Outboard Isolation AO-095 Open 2ADWEDPump Tripped 2B DWEDPump Running C. Inboard Isolation AO-094 Closed Outboard Isolation AO-095 Open 2ADWEDPump Tripped 2B DWED Pump Tripped D. Inboard Isolation AO-094 Closed Outboard Isolation AO-095 Closed 2ADWEDPump Tripped 2BDWEDPump Tripped 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II Answer L!

(). #1RO Choice Basis or Justification Correct: C The given co nditions indicate an inboard half isolation, causing OWED sump inboard isolation valve AO-094 to close. Nothing in the stem would indicate that outboard valve AO-095 has closed. If either the inboard or i .~tboard isol ation valy'~~Qses, _~oth pWE[).J>-':l!!l~'ll\filL!!ip~_. ____ ~ ___

I Oistractors: A I Plausible if the applicant does not recall that the 2A OWED pump will trip if

. either isolation valve closes.

i

_1

.. --~-~~~-~'----~~-------.~--.---

B Plausible if th e applicant believes the 2B OWED pump will start on trip of

__::~I~~~~~I~.

during automatic pump down, and does not recall that either

.. e_ cIQ~Lr19 wUUrip both Qumr:>.~_____...._ ..... _ .... __

0 I Plausible if th e applicant believes the conditions indicate a full (inboard and outboard) isolation of the OWED sump.

Psychometrics Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item Other Exam Bank: 0 COL Learning PLOT-5007G-1 m Objective:

KIA System: 223002 - Primary Containment Isolation Importance: RO / SRO System/Nuclear Steam Supply Shut-off 2.8/3.1

~---------~-------------~-------- ------.~ .... ---------

KIA Statement:

K1.14 - Knowledge of the physical connections and/or cause-effect relationship between the Primary Containment Isolation System/Nuclear Steam Supply Shut-off and the following: Containment drainage REQUIRED MATERIALS: ----..---~-t_------- ... --

Notes and Comments:

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

2. Given the following:

Unit 2 is operating at 100% power

The 2A RPS Bus is being supplied by the Alternate Power Supply, 20Y050

The 2A RPS MG Set is in service, ready to power the 2A RPS Bus

The CRS directed you to transfer RPS Power Supplies Which one of the following Primary Containment Isolation System (PC IS) Groups will need to be RESET following the power transfer?

A. Group II Inboard Half Isolation B. Group III Inboard Half Isolation C. Group II Outboard Half Isolation D. Group III Outboard Ha]f Isolation 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Basis or Justification B from the Alternate Power Supply (20Y050 UPS) results in a n"lI"\n"I<::>n1'<>,n/loss of power to RPS 'A' and PCIS Div I logic. This results in a III Inboard isolatkm.

Distractors: A Group II Inboard isolation occurs on loss of vital AC power to 20Y033.

C Group II Outboard isolation occurs on loss of vital AC power to 20Y034.

o Group III Outboard isolation occurs on loss of power to RPS Bus "B'.

Psychometrics DiffiglJl~~~ ___lIime Allowance {Ill inuJes)

I LeveLof KnowledgE}~~~ ~- - .. RO

~

MEMORY 2 3 I 10CFR55.41 (b)(6)

Source Documentation Source: D New Exam Item [XI Previous NRC Exam: (PB 2002)

[XI Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank Learning PLOT-5007G-05h Objective:

KIA System: 262002 - Uninterruptable Powl3r RO/SRO (A.C. / D.C.) 2.9/3.1 KIA Statement:

K1.08 - Knowledge of physical connections and/or cause-effect relationships between Uninterruptible Pow~§~lA.C./l?-.:C.) an~the fQlJowing: Containme,.,t iS91ation system. ___ . . . .~___~_. _ . . . __ ~

~~~~t~~~~~~~~;::~~~JN~NE~_~~_~-~~.~=__-~-_~_~=--~~_~~. . =_-~=_--_~_-~~-_-_.-~~_~_. . ~.--= ..~_-_-_ .=~-~.--~_--__ --1 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

3. Select the correct Shutdown Cooling suction valve electrical power supplies.

MO-2-1 0-17 "Shutdown Cooling Suction Outboard Isolation" is powered by a

_(1)_.

MO-2-1 0-18 "Shutdown Cooling Suction Inboard Isolation" is powered by a

_(2t~ __*

A. (1) 250 VDC Safety Related Bus (2) 250 VDC Safety Related Bus B. (1) 250 VDC Safety Related Bus (2) 480V Emergency Bus MCC C. (1) 480V Emergency Bus MCC (2) 250 VDC Safety Related Bus D. (1) 480V Emergency Bus MCC (2) 480V Emergency Bus MCC 2011 NRC RO Written Exam Rev. Ldoc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 I

Answer Key I cC Qt i,#3RO Choice Basis or Justification -

Correct: B 0-17 is powered by 250 VOC Safety Related Bus 20011.

MO-2-10-18 is powered by 480V E Bus supplied MCC E-124-R-C.

Oistractors: A MO-2-1 0-17 is powered by 250 VOC Safety Related Bus 20011.

I

! MO-2-10-18 is powered by 480V E Bus supplied MCC E-124-R-C.

C MO-2-10-17 is powered by 250 VOC Safety Related Bus 20011.

MO-2-10-18 is powered by 480V E Bus supplied MCC E-124-R-C.

I 0 MO-2-1 0-17 is powered by 250 VDC Safety Related Bus 20011.

I I MO-2-10-18 is powered by 480V E Bus supplied MCC E-124-R-C.

Psychometrics

__Jevel ofKnpwl~.Qae..J_.____.PiffLcul!y~ _._J---.TImeAIIQwance (mil}ut~s) I _R

...O MEMORY I 2.5 I 3 : 10CFR55.41(b)(7)

Source Documentation Source: o New Exam Item [8J Previous NRC Exam: (PB 2005) o Modified Bank Item 0 Other Exam Bank: 0 E-26 KIA System: L.V.JV\.'V - Shutdown Cooling System (RHR Importance: RO I SRO Cooling Mode) 2.5/2.7 KIA Statement:

~2. 02 - Know~Qg~.Qflil~.electrical powe..~LJ.QQI~~Jp..lhe f.9l1owing.;},,'Ip!Qr..9pe..ratec! val~es:____.__ .

REQUIRED

~- ..- .. ~ *..-

MATERIALS:

..- ..- .. --~.----

fONE

..- - . . - ' -*.. ~--'--'.-"---~'---'~~'--'-'---'--- ... ~--

-Notes

.----- and

.. Comments:

.~--------~~----~------~--- -----~--.-----~-.-------------~------,--.----- ~~".--- .. --~-- .. ~-.-.--.--- ---.~

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

4. Unit 2 conditions are as follows:

Reactor power is 100%

RPS bus 'A' is aligned to its ALTERNATE power supply, 20Y050

RPS bus 'B' is aligned to its NORMAL RPS MG Set power supply

A loss of 3 SUE occurred, causing a 4KV emergency bus fast transfer

The E-222 breaker did NOT close on the fast transfer No operator actions have been taken. Which one of the following describes the status of RPS one minute later?

A. NO Scram B. HALF Scram on RPS 'A' C. HALF Scram on RPS 'B' D. FULL Scram 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer uestlon #4 RO Choice Basis or Justification Correct: C loss of 3 SUE and failure of E-222 breaker to close on the fast transfer result in a loss of power to the 'B' RPS MG Set. The E-2 diesel will on E-22 bus under**voltage but will not restore emergency bus MCC (13 seconds) before the 'B' RPS MG supply breaker trips on loss of power (8 seconds). A loss of power to the 'B' RPS bus will result in a half scram on RPS 'B'. RPS bus 'A' is powered from 20Y050, which is nn\li/t:lrt:.)n from the E-12 bus and 2 SUE.

Distractors: A A half scram will occur due to loss of power to RPS bus 'B'. Plausible if applicant does not understand 4160V power distribution to RPS MG Sets and/or 20Y050.

B RPS bus 'A' is powered from 20Y050, which is powered from the E-12 bus

, and 2 SUE. There is no effect on RPS 'A' as a result of this event.

I Plausible if applicant does not understand 4160V power distribution to RPS Sets and/or 20Y050.

o half scram will occur due to loss of power to RPS bus 'B'. RPS bus 'N is from 20Y050, which is powered from the E-12 bus and 2 SUE.

is no effect on RPS 'A' as a result of this event. Plausible if applicant does not understand 4160V power distribution to RPS MG Sets and/or 20Y050. ---~-.--~ .. -~-.

Psychometrics Source Documentation Source: i o New Exam Item o Previous NRC Exam: 0 I o Modified Bank Item o Other Exam Bank: 0

.~ ___.. _~_ _ iJilliLT Exam Bank_ _ ____ *..

~ference(sL_ SQ_M.7.A_~ __ ~_.~ ___..~~~ _____..~_.__.. _~_

Learning PLOT-5060F-2b Objective: *

~~---

KIA System: 212000 - Reactor Protection System Importance: RO/SRO 3.2/3.3 .. __ .

KIA Statement:

-~~~~R~~~i::R~:~~ricaj~6~=UP~S to the following RP§ m:to~-ge~e~at~:~s=_~~_._~

0 - - -..- -..- -..- - - -...- - - - - - - - - - -..- - .. -~ .. - ... -~-. . ..... ... .... . .. . . - - -.. ~

Notes and Comments:

L-_._.._ _ _ _..._ _.._ _ .. _~~~~~_~~~~_ _~._, ______ ~_ .._ _ _ _ ~_ .._ _ _ _ _ _.._ . _ . _ _ _ _ _..__ ~ ______ ~

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 201 I

5. A Loss of Coolant Accident occurred on Unit 3. The following plant conditions exist:

Reactor level initially dropped to -180 inches

Reactor level is now -150 inches and is rising slowly

All RHR and Core Spray pumps started automatically and are injecting

Operators have NOT manipulated any RHR or Core Spray pump controls

Reactor pressure is 250 psig and lowering

Drywell pressure is 16 psig and rising With these conditions present a loss of off-site power (LOOP) occurs. All Emergency Diesel Generators (EDGs) start and load their respective busses.

Which statement below describes how Core Spray pumps will be restarted to control Reactor water level for these conditions?

The Core Spray pumps will _ _~__~_ _ _ _ once their respective emergency bus reaches 95% of rated voltage.

A. require a manual restart B. automatically restart immediately C. automatically restart after a six (6) second time delay D. automatically restart sequentially after 13 seconds (A, C) and 23 seconds (B, D) 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

5RO Choice ...

Basis or Justification Correct: their respective emergency bus reaches 95% of rated voltage, all Spray pumps will start after a 6 second time delay.

Distractors: Plausible if the applicant believes the conditions will cause the Core Spray pump breakers to lockout, requiring manual action to restart the pumps.

The pump start logic imposes a 6 second time delay for Core Spray pump starts to allow the bus to recover from the immediate RHR start of RHR This choice reflects the sequential loading of the busses that occur if the emergency busses are powered by an off-site source rather than the Psychometrics Level Qf Knowl~~~ __ . _.. __~Difficult~_. ._ r-_Time AIIQ.~anc~J!t:linut~~L+-_-J3Q...--~-

HIGH 3.0 4 10CFR55.41 (b)(7)

Source Documentation Source:

o New Exam Item [8J Previous NRC Exam: (PB 2005) o Modified Bank Item o Other Exam Bank: ()

[8J 1LT Exam Bank Rdcl tmce(s) SO 14.7.A-2 ... ---~-- ._._

Learning PLOT-5014-11 Objective:

KiA System: 262001 - A.C. Electrical Distribution Importance: RO/SRO 3.5/3.7 KiA Statement:

K3.01 - Knowledge of the effect that a loss or malfunction of A.C. Electrical Distribution will have on the following: Major systel'l1Joad~ ____~__ . --~~--"-- .... .. ---"-~~--~---- .... ------~ .... ~-~---- . .

REQUIRED

....- - - - -MATERIALS:

... - .... --~--~-- ...- - - - - - fONE

...- _._ ..... ... -~~ ... ---.-~ .... ....- - - - . - -..-

.. ..... ~-

cl!ote§i andJ~9ml'!}ent§i~____.. . _ __ . . ... ----.-.----~ ..... ----.----.... ---------~--~---.--- .. ---~---.------

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

6. Unit 2 is operating at full power when the Standby Liquid Control injection sparger becomes clogged. This results in the pressure input from this line being 10 psig lower than actual.

Evaluate this condition to determine the impact, if any, on indicated Core Plate Flow as read on the Control Room Flow Recorder (FR-095).

Indicated Core Plate Flow on FR-095 will be - - - - - - -

A. higher than actual B. lower than actual C. reading zero D. unaffected 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer QUE!stt()n if. a RO Choice Basis or Justification Correct: B The SBLC injection line is the below core tap for core plate dIp and flow.

I With the below core plate tap having a lower pressure signal, the indicated

will be in a lower than actual core flow Indication.

Distractors: A If the candidate believes that SBLC injects through the above core plate line, then he would determine that indicated core plate flow would be high.

C candidate may believe that the given conditions will result in a sensed of 0 psid, resulting in the flow recorder reading zero.

i D J.jlf the candidate believes that the core plate flow indication uses taps I . unrelated to the SBLC injection line as do Control Rod Drive dIp and core

I spray line break detection, he. would believe hat the indication will be t.

L_~____.._ ____ urlaffecte~ ___...._ . .__ .__.__..._ . __... .__.. __.____._._

Psychometrics I-~~vel ()f Kn()wled.9~_J______ Difficulty. ___ -1 Tirl!.~ AI!Q.warl.9.tlrninut~~ RO . --

HIGH I 3 I 2 10CFR55.41 (b)(6)

Source Documentation Source: o New Exam Item ~ Previous NRC Exam: (PB 2005) o Modified Bank Item o Other Exam Bank: 0

~ ILT Exam Bank

_R~~!ence~t__~-,-______~M:3~~__~ ______. "-"'-'-"~'--"-'-"---"-'---""~-~"--'~"-'-"---' ----

Learning I PLOT-5011-3c Objective: i KiA Systen1~:~----r211000 - Standby Liquid Control System !Importance: RO/SRO i 2.6/2.7 KIA Statement:

K3.03 - Knowledge of the effect that a loss or malfunction of the Standby Liquid Control System will

. ha~ onl~e followina:...QQre pl~te differen!ial pressure indicatio!!-....___._......_.._ .... _ ._-_.

~:t~;~:~~:~~:~:~S:=f-~!lE=-- --"'-"'--'~--'---"---"--'-----'---~-'----'

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

7. Unit 3 is at 100% power with the 3A RHR pump in full flow test in accordance with ST-O-01O-301-3 "A RHR Loop Pump, Valve, Flow and Unit Cooler Functional and In-Service Test".

During the test, a steam leak in the drywell results in the following conditions:

The Reactor is scrammed; all control rods are inserted

RPV level reached -100 inches and is rising due to HPCI injection

RPV pressure is 400 psig and lowering

Drywell pressure is 10 psig and rising

Offsite power remains available to the 4KV buses Based on these conditions, which one of the following describes the status of the Unit 3 RHR System?

A. ALL RHR pumps are injecting.

B. ALL RHR pumps are operating on minimum flow.

C. 3A RHR pump remains in full flow test; ALL other RHR pumps remain shutdown.

D. 3A RHR pump remains in full flow test; ALL other RHR pumps are operating on minimum flow.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer

7RO Choice Basis or Justification Correct: B The 3A RHR pump full flow test valves (MO-3-10-034A and -039A) receive a close signal when RPV pressure is < 450 psig and an initiation signal (RPV level < -160 inches OR drywell pressure> 2 psig AND RPV pressure

< 450 psig) is present. The other RHR pumps receive an auto-start signal (based on drywell pressure> 2 psig and RPV pressure < 450 psig). Since RPV pressure is above the RHR pump shutoff head (-305 psig), the pumps but are all on minimum flow.

Distractors: A . Since RPV pressure is above the RHR pump shutoff head (-305 psig), the pumps are not injecting but are all running on minimum flow. Plausible if applicant does not recognize conditions or does not recall the value for RHR shutoff head.

C Plausible if applicant does not recognize that the given conditions result in

~--~~***-~*****-___Uje~iOn)._

D RHR initiation and injection signals (pump and valve alignment for RPV

-.---.'--'~--"'~-""-'-'--'

Plausible if applicant recognizes that a LPCI initiation signal is present but I does not recognize the conditions are met for an injection signal << 450 I L .. _.~_ _ .~. _ _._. _ _ _ _. 1 _.. ___~~d~ubs~ueI11 va!\I~ re-~ignment. _____~ ___~_~ ___ ~ __ ~ .._~ ... ___ .. '

Psychometrics

~\lel ofKnoV!'!~.g.9~_.~L. ____Piffl~LJl!Y~ ____J_Ilm~ A"o\Nan~~ (mtnutes) . RO HIGH 1 2.5 i 3 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 f-- .._ -.... - ..... - __ hOlbIEXarrl Ban~__....._ .._

~:;;~i;gce(S): -1~~~~~~:~t~BD£,-S-~"1les 118-122 __._.._..._.__ . __.__......_._ - . .--~

Objective: i KIA System: ! 203000 - RHR/LPCI: Injection Mode Importance: RO/SRO i 3.1/3.4 .. - .. -.~

KIA Statement:

K4.09 - Knowledge of the RHR/LPCI: Injection Mode design feature(s) and/or interlocks which provide jQr the following: Surveil~nce for a~ opef!:1b~compQnel'1t~ __._.._ ._._.. ~. __ ~

- REQUIRED MATERIALS:

... ... fONE

... ~..."-"'--'--'- *.- ' - ' ..--.--...... --~-.-~--.~---.~~.~-- .. .------_..._.---.

Notes and Comments:

- . -... - -... ---~~ .. -.------.- - -.. --.~- ..- -.. - . - - - . -...- - -... -~~-.-

2011 NRC RO Written Exam Rev.l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

8. Unit 2 is operating at 100% power.

An electrical transient on 2 Aux Bus resulted in a loss of power to the 2B and 2C Drywell Chillers.

Which one of the following describes the impact of this event, ifany, on cooling water to the Instrument Nitrogen compressors?

A. RBCCW cooling to the compressors will be lost; the compressors must be shutdown and nitrogen loads must be aligned to Backup Nitrogen (bottles).

B. RBCCW cooling to the compressors will be lost; the compressors must be shutdown and nitrogen loads must be aligned to Instrument Air.

C. RBCCW cooling to the compressors will be lost; TBCCW will automatically align to cool the compressors.

D. No impact; the compressors will continue to be cooled by RBCCW.

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Basis or- Justification A loss of power to 2 of 3 DW chillers results in an automatic swap of the DWCW supply to RBCCW. This causes non-essential RBCCW loads to be isolated, which includes the Instrument Nitrogen compressors. Per AD 44A.1-2, Instrument Nitrogen will be shutdown and nitrogen loads will be aligned to Instrument Air (via

~_ _ _.. . . . . . . _--+_____----+--'"Q:4230AJB).______ ----- - - _ .............._ - - - - - - -

Distractors: Per AD 44A.1-2, Instrument Nitrogen will be shutdown and aligned to (backed up by) Instrument Air, not "Backup Instrument Nitrogen" (bottles).

C RBCCW will automatically align to cool the Instrument Air compressors on a loss of

TBCCW. However, TBCCW does not provide a backup cooling source for the

. Instrument Nitrogen compressors. Plausible misconception.

'- - - - -___~. . . . __..__.J._~I~~gg~~ll~ffe~~ to DWCW, ;e-;~~~nt Nitrogen co~presso~~m Psychometrics

_Level of Knowledge Difficult}' Time Allowance (minutes) RO HIGH I 2 2 10CFR55.41 (b)( 10)

Source Documentation Source: o New Exam Item ~ Previous NRC Exam: (PB 2009)

Modified Bank Item D Other Exam Bank: 0 1--......- - - - - - - +--..........,... .:ILT Exam Bank

. .:.=..::..-=:.:~::...:=.:=:...:...--=--.-. -.--.-- .....

Reference( s): AO 44A.1 SO 16.2.A-2 Learning PLOT-5035-4c Objective:

KIA System: 300000 - Instrument Air System (lAS) RO/SRO 2.8/2.8 KIA Statement:

K4.03 - Knowledge of the Instrument Air System design feature(s) and/or interlocks which provide for the f()ll()wing: Securing_()f IASljp()l1loss of coolin~ater.___ . . . . . . . .______.._._ . ____________ _

REQUIRED MATERIALS: NONE ----- ... _ - - - - - - - - _ .... -._._. __ ................-----

Notes and Comments: This question addresses "loss of cooling to instrument nitrogen compressors", which is the pneumatic supply to components inside the drywell. Instrument Nitrogen is not a separate system in the KIA Catalog but typically shows up as "Instrument air/nitrogen".

L. _ _ _. _ _ _ _ _ _ _ _ _ _ _ ~_ _ _ ..J_Ih~r~f()..!:~Jhis_question me~!~ the KIA.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

9. Which one of the following meets the conditions for Automatic Depressurization System (ADS) actuation?

A Drywell pressure at 4.1 psig Reactor water level at 20" for 10 minutes

'A' and 'D' Core Spray pumps operating B. Drywell pressure at 5.0 psig Reactor water level at -165" for 5 minutes

'A' and 'B' Core Spray pumps operating C. Drywell pressure at 1.2 psig Reactor water level at 65" for 5 minutes

'B' RHR pump operating D. Drywell pressure at 2.7 psig Reactor water level at -165" for 3 minutes

'D' RHR pump operating 2011 NRC RO Written Exam Rev. 1.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 I I AnsWE~r QUE!!stl<m # 9 RO e~---****-------~*--**--.-~***~-~-***--*--~**--~~-----**--------- ..- - -....- - ....--~- ..

Choice Basis or Justification D is the only set of conditions shown that satisfies all of the ADS logic level below -160 inches, drywell pressure above 2 psig, at least one RHR pump (or the correct combination of Core Spray pumps)

-,--~~~--------

the 10S second timer timed out.

Distractors: A B The right combination of ECCS pumps is not available; must have at least 1 RHR pump, OR the right combination of Core Spray pumps: A or B and C D. . .-..- - -...

---.----.-.--+-...~.--..- ..~+:::.:..-,=:...:..-~ ---~- ..- -..- - - . - - - - ...- - -..-... -~-- .... -.--..'-.. -~.--~-I pressure is below 2 psig and the 9.S minute high drywell pressure is not present.

Psychometries Source Documentation Source: D New Exam Item rg] Previous NRC Exam: (PB 2007)

D Other Exam Bank: 0 ILT Exam Bank ARC-227 .....

D-4...

~- -~ ... -

G-S KIA System: 218000 - Automatic Depressuri:zation Importance: RO I SRO System 3.8 I _3c_...::..8.... _ _ _ _ _...._ .... ~._1 KIA Statement:

KS.01 - Knowledge of the operational implications of the following concepts as it applies to the

_~~~~;~~~:;:;:~~.:~o.:n S.-~.~t~~6.-~. .~~ I~.i~()pera!iOn~__=_.~-. . _ -..- -.. -..-. _.. .

Notesi!'1d Comm~nts: _____ =-1_.. ___.._. ._. ___.. ~____.. ____~ _.. . . _.__._____...

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

10. Given the following:

Unit 2 is operating at 100% power

FEED WATER FIELD INSTRUMENT TROUBLE (201 H-I) alanns

'A' Steam Line Flow Transmitter (DPT 2-6-51 A) indicates downscale Based on these conditions, the Digital Feedwater System is in and Reactor Feedwater pump turbine SPEED will _ _(2)_ _.

A. (1) single-element control (2) remain steady B. (I) three-element control (2) remain steady C. (1) single-element control (2) lower until level stabilizes at a new lower level D. (1) three-element control (2) lower until level stabilizes at a new lower level 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer

10RO Choice Basis or Justification

~ ~ _ _ _ _ _ _ _ _ _ _ _ _ _J Correct: A The Digital Feedwater Fault-Tolerance Override Logic will not permit the system to remain in three-element control with a failed steam flow detector.

The system would permit three-element control with a failed level detector but not a failed steam flow or feed flow detector.

Distractors: B The candidate may expeict the system to remain in three-element, because it does remain in three-element control for some failures (e.g., level detector). The system conditions would not change if the candidate believes the failed instrument is bypassed by the system when it fails C The candidate may understand that the DFCS will not remain in ooI_""I""I'r

control but may believe that with a failed steam flow instrument the "'\le't""'.....

would be biased to allow level to lower to a new lower level causing RFP turbine to lower.

D The candidate may expect the system to remain in three-element, because it does remain in three-element control for some failures (e.g., level detector). The candidate may believe that with a failed steam flow i instrument the system would be biased to allow level to lower to a new L .._ _ _._ _ _ _ _ _ _ : _____ -'.J9wer_I~"'~t cc:i~sif'I9...RFP turbine spe~J9-'.o~_~r.:~___ __... _________ ~_.~_._

Psychometrics Source Documentation Source: o New Exam Item k8J Previous NRC Exam: (PB 2005) k8J Modified Bank Item 0 Other Exam Bank: 0

-4r KIA System: uCl~'VL - Reactor Water Level Control ROISRO 3.1/3.1 KIA Statement:

K5.01 - Knowledge of the operational implications of the following concepts as they apply to the

~"i.'i'R~::~;~:I~~:1 sJ~~N~E~~C/F()XbQ"'-~iLe~ cOl1troJl~,-op,,!ation.

Notes and Comments:

~ .. --~ .. --.~ .. --~.-- ... ~-~ ... --~ --.~ - - . - -_._.--.-"---"-" - -.. ~--.---- .....----.~

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

11. Unit 2 is in MODE 2 with a reactor startup in progress.

The reactor is subcritical

Control rod withdrawal has begun

2B 24/48 VDC Distribution Panel (200045) is lost What effect will this have on the Wide Range Neutron Monitoring (WRNM) System?

A. Rod Block ONLY B. RPS Channel 'B' Half Scram ONLY C. Rod Block and RPS Channel 'B' Half Scram D. Loss ofB, D, F, and H WRNM ODAs on panel20C005 2011 NRC RO Written Exam Rev. Ldoc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer

11 flO Choice Basis or Justification Correct: C A loss of power to 20D045 will result in a half scram and rod block due to a WRNM INOP condition.

Distractors: A IClU;:'IUIC if applicant does not understand the effects of a power loss on WRNM system and believes the given conditions will only result in a rod i i Plausible if applicant does not understand the effects of a power loss on

, I the WRNM system and believes the given conditions will only result in a I .~_ _ ~ __..~~. ~;__ ~~ half s(~.ram. __ ~~_.. ~~~~~_~__.._~ . _.. _._~. . ~_. __.. _~~_. .~ .._ ... ~~~~~~~~__ .~~

I I D i 20D045 does not power the ODAs; they are powered by 20Y050. I l~ __.. __ ~___:__ ~.~_I~:~~~~i~~et~ ::~~~~~s~~t~~~ecall the functio~_O_f_th~~d_iff_ere_nt_~o.~w~~e_r~~. ~~_

Psychometrics

. . _~~el oL~nowleQge ~l I

__~iffiCI:!~~~_._j Tim~AIl0V'!.cmce (rninut~~) RO -~

MEMORY* 3 I 3 10CFR55.41 (b)(6)

Source Documentation Source: D New Exam Item ~ Previous NRC Exam: (PB 2005)

~ Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank o H-3 --.~~ ~-.-~~-~~-~-~~~-~-~- .... -~-~ ---~.-~--~-.- ... ~-~~- .. ~

5060C-2c, -4a, -4b KlA System: 215003 - Intermediate Range Monitor RO/SRO System (WRNM at Peach Bottom) 3.6/3.8

--~~~~.... ~ .. -~

KlA Statement:

K6.02 - Knowledge of the effect that a loss or malfunction of the following will have on the Intermediate Range Monitor (YVRNMLSys1t?m: 24[48 vol1Q.C: pov~~r._____ ~~.~.__ ~_. . _ ._~ ____..~_~ --~~-. ---~~~.-.l

-~~t~;~~~~!::_:;LS~

~~ ...___ ~_ ...__ ~_~_ ... ___ ~ _ _ _... _~~

-f0N~ --~-...

_ _ _ _. _ ..._ _ _ ~~

..--~-----~~-~---"'-""-"'-'~-'-.'

..._ _ _ _..._ _ _ _ _.._ _.._ _ ~_. _ _ _... _ _ _ _.______M_.__.__

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 201 ]

12. Given the following:

Unit 2 is operating at 100% power

SYSTEM II CORE SPRAY LOGIC POWER FAIL (226 B-1) alarm is received

Investigation reveals a blovvn fuse on 125 VDC power distribution panel 20D022 Subsequently, a LOCA occurs, resulting in the following plant conditions:

RPV level is -75 inches

RPV pressure is 420 psig

Drywell pressure is 4.5 psig Which one of the following describes the status of the Core Spray System?

'A' Core SQrav LooQ 'B' Core SQray LooQ A. Both pumps ON; injection valve is Both pumps ON; injection valve is OPEN OPEN B. Both pumps ON; injection valve is Both pumps OFF; injection valve is CLOSED CLOSED C. Both pumps ON; injection valve is Both pumps OFF; injection valve is OPEN CLOSED D. Both pumps ON; injection valve is Both pumps ON; injection valve is CLOSED CLOSED 2011 NRC RO Written Exam Rev. J .doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question # 12RO Choice Basis or Justification Correct: Plant conditions call for Core Spray initiation and injection (below 450 psig reactor pressure). A loss of 125 VDC power panel 200022 causes a loss of power to Core Spray loop 'B' logic, which will prevent it from actuating a Core 'A' is not affected of 20D022.

Distractors: 'B' Core Spray logic will not actuate. Plausible since each RHR logic

! channel provides pump start signals to BOTH loops of RHR; applicant may i 9i~:~;: ~~;:y;~;~ :~i~-;;:i~ct~~~ 10;li:~~ible sin; each RHR logic channel provides pump start signals to BOTH loops of RHR, but does not i provide redundant valve control interlocks; applicant may confuse Core

~__...._ _ _ _ _-+-~~____ SP@i'.le>gic with ~HR lo~: . . ._~_.... ... .__. __.. .____ ..-...------c D I -::' loop is injecting with both pumps; 'B' loop pumps are not running.

_~_ .... _ _ _.... ~i_~ ~;:~:~~~ :e~:~~~:\~~ti~IZ~tii~~~nfus~d a~~~~does

___ ::t recal~RPV Psychometrics

_ Level of Kl1owledg~~ __ - _... _. Difficulty .._- ~ime Allo\i\lance (minutes) i RO

-~.-- ...

HIGH !

10CFR55.41 (b)(7)

Source Documentation Source: ~ New Exam Item Previous NRC Exam: 0 o Modified Bank Item o other Exam Bank: 0 ILT Exam Bank KIA System: - Low Pressure Core Spray RO/SRO 2.8/2.9 KIA Statement:

K6.04 - Knowledge of the effect that a loss or malfunction of the following will have on the Low Pre~surt3 Core Spray §y"s>~em: D. g. poweL. .~__.. . __ ~_ ... __ ._. . _ . ____

~:~~~~~~:::::~~ ~ONE_ _ . _.. . __ ~_ . ___._~_. _____.. . _ _.. _________._.. _

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

13. Given the following conditions on Unit 2:

Battery Charger 2AD003 is placed in the Equalize Mode in accordance with SO 578.1-2 "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 by the diesel generator Which one of the following describes the status of the 2A Battery Charger one minute after the E-12 bus is reenergized?

The 2A Battery Charger _ _ _ _ _ _ _ __

A. automatically returns to the "float" c:harge mode

8. automatically returns to the "equalize" charge mode C. is deenergized and must be manually returned to service D. is energized but the DC output switch must be manually closed 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Key Qu~#13RO

~.--- ....... ~--"-"""""""""----" -..... ********* _ _ _ _ _ _ _ r_*

Choice Basis or Justification Correct: B From Note 2 in SO 57B.1-2: "Upon a loss of AC input power, the batterY charger will return to the same mode it was in once power is restored. IF the battery charger was in the Equalize mode, THEN the timer will pick up where it was interrupt~~ AND time q~ ........ _------------- .. ---~~

Distractors: A The charger will return to the equalize charge mode. Plausible if the applicant remembers the charger will automatically restart but does not I

remember it will return to the tS..§irne mode it was inprior to the~ower 10ss.

C The battery charger will automatically restart 15 seconds after the E 12 bus is restored. Plausible if the applicant does not remember that the charger will automatic~r~tStart._ _ ~ ......... -------.-~ ..- ........ ------~---.----~-

D Plausible since procedure precaution requires waiting 15-20 seconds 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

_~Iosure) following a}()ss of AC~ower toJhe charg~for the ~<:1me reason.

Psychometrics

_L~,,~I of"Snowledge - -

_Qifficulty TimeAliowance (minutes) RO MEMORY 10CFR55.41 (b)(7)

Source Documentation Source: D New Exam Item ~ Previous NRC Exam: (PB 2008)

~ Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank SO 57B.1-2 Learning PLOT-5057 -6a Objective:

KiA System: 263000 - D.C. Electrical Distribution RO/SRO 2.5/2.8 KiA Statement:

A 1.01 - Ability to predict and/or monitor changes in parameters associated with operating the D.C.

Electrical Distribution cOI"I!rols includin~~,~te!."Y.~h<:1Egkl9/discha19ing rate~.. ..~_~__~_~

REQUIRED MATERIALS: NONE

, _ " "_ _

__" _ _ _ . " _ _ m_m_'"~ _ _ """"""" -----""""~"" * * - ,* * * * * * --------1 Notes and Comments: ** _ _ _ " _ m m _ . " " " _ _ _ * * *

_ __

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

14. Unit 3 was manually scrammed following a loss offeedwater. Current plant conditions are as follows:

HPC[ and RCIC automatically started and injected until they tripped on high RPV water level

Reactor water level is currently +46 inches and dropping slowly

Reactor pressure is 925 psig and being controlled automatically by EHC Which one of the following correctly dt~scribes the RCIC System response to these conditions?

The RCIC _ _(1 )_ _ and the RCIC system will automatically re-inject when Reactor water level has lowered to - - -(2)- -.

A. (1) Turbine is tripped (2) +29 inches B. (1) Turbine is tripped (2) -48 inches C. (1) Turbine Supply Valve (MO-131) is closed (2) +29 inches D. (1) Turbine Supply Valve (MO-131) is closed (2) -48 inches 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 AnsWEtr Key Question # 14 RO

,-----------------------------------------1 Choice Basis or Justification

--+----------------------------------------

Correct: D RCIC does not trip on hi'gh level (because it can not auto reset a trip like HPCI); instead the turbine supply valve (MO-131) closes. The valve will not reopen until a -48" initiation signal is received (unlike HPCI which will 1-----------

restart at +29").

+------~-------------------------------j Distractors: A RCIC does not trip on high level and will not re-inject until -48".

+------t-----------------------

B RCIC does not trip on high level.

-j------------- ----------------------------

C The RCIC turbine supply valve (MO-131) is shut, but the system will not re inject until reactor water level drops to -48".

Psychometrics Level of Knowledge Difficul!Y____________ 1 -Time Allowance (minutes) RO HIGH 2.5 4 10CFR55.41 (b)(7)

Source Documentation Source: D New Exam Item ~ Previous NRC Exam: (PB 2005)

D Modified Bank Item D Other Exam Bank: 0

~ ILT Exam Bank -- ._-_.- ---_._------

Reference(s): OT-110, Step 3.6 ---._--------

Learning PLOT -5013-4b Objective:

KIA System: 217000 - Reactor Core Isolation Cooling I I;' portanee RO/SRO System 4.0/4.0 --

KIA Statement:

A 1.03 - Ability to predict and/or monitor changes in parameters associated with operating the Reactor Core Isolation Cooling System controls including~_ Reactor water level. .__ ._-------------

REQUIRED MATERIALS: NONE - . ---- .. ~---------- -". - -------

Notes and Comments: -- - - - --- - -- _._.------------ . - ----

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

15. Unit 2 was operating at 100% power.

The crew is performing a surveillance test for full load testing of the E4 Emergency Diesel Generator (EDG)

The EDG has just been synchronized to the E-42 Bus Shortly after the E4 Diesel output breaker is closed, a loss of off-site power (LOOP) occurs.

Based on these conditions, which one of the following describes (1) the status of the E4 EDG and the E-42 Breaker, and (2) the required procedural actions?

A. (1) E4 EDG is RUNNING; E-42 Bn~aker is OPEN.

(2) The anti-pump lockout must be manually reset using SO 52A.l.B "Diesel Generator Operations" before the E-42 Breaker will close.

B. (1) E4 EDG is RUNNING; E-42 Breaker is CLOSED.

(2) Monitor and control EDG loading during continued operation using SO 52A.l.B "Diesel Generator Operations".

C. (1) E4 EDG is TRIPPED; E-42 Breaker is OPEN.

(2) Restart the EDG using SO 52A. 7 .A.l.B "Diesel Generator Manual Emergency Start." E-42 Breaker must be manually closed after resetting the anti-pump lockout.

D. (1) E4 EDG is TRIPPED; E-42 Breaker is OPEN.

(2) Restart the EDG using SO 52A. 7 .A.I.B "Diesel Generator Manual Emergency Start". E-42 Breaker will automatically close when the EDG is running.

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question #-15

~""~--"-.-- ..

RO Choice Basis or Justification Correct: A A LOOP und er these conditions will cause the E4 output breaker to trip, resulting in a dead E-42 bus. A "dead bus start in test mode" will send a trip signal to t he E-42 breaker but not to the DG. Because E-42 receives simultaneous trip and close signals from the dead bus condition, the

.. __ ~ __~_'- breaker wl!1_nti-pul"l1J>...!Qckout a and must be r~et f!1anu~._......__.. ~. ....j Distractors: B E4 DG will be running but the E-42 breaker will not close due to anti-pump .

lockout. Pia usible if applicant does not understand DG breaker control j' logic.

C E-4 DG with not receive a trip signal so it does not require restart. The anti-pump lockout on the E-42 breaker must be reset. Plausible if applicant .

I~.~______.J__ ~_.

i D.. J. does not

-~

und

.. -~

~stand DG breakercont~ol logic.._... _ -.- . . .- --...

E-4 D.G with not receive a trip signal so it does not require res. tart. The a.nt.i-.

pump lockout on the E-42 breaker must be reset. Plausible if applicant doe~ not understand DG breaker controllogic. __~~ . _...~_. __ ~. I

~ . ~ ___

I I

Psychometrics f-. Lev/ell of!5nowledg~_~ I-- _____ -'?iffi~~_~ I Time Allowance (f!1if1ut~~) i RO ._-_.

HIGH I 10CFR55.41 (b)(8)

Source Documentation Source: D New Exam Item ~ Previous NRC Exam: (PB 2009)

~ Modified Bank Item D Other Exam Bank: ()

KIA System: I 264000 - Emergency Diesel Generators Importance: RO 1 SRO 3.5/3.6 KIA Statement:

A2.01 - Ability to (a) predict the impacts of the following on the Emergency Diesel Generators; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions ()LC!Q.~rations: Parallel oR.eration of em/elrgel1..9'..gen~~tof: __ ~ .. ____~ . __.__ . ~ __

~:t~~~~~j~::::~~s~_=_ r~E=-. -.-~-.- --~ -------~- -~~.~

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

16. Unit 2 was manually scrammed due to a leak in the Torus.

HPCI is being used for RPV pressure control

RCIC is being used for RPV level control

Torus level is 9.5 feet and lowering Which one of the following describes the required action and the reason for this action?

A. RCIC must be shutdown using RRC 13.1-2 "RCIC Operation During a Plant Event" to prevent exceeding the vortex limit.

B. HPCI must be shutdown using RRC 23.1-2 "HPCI Operation During a Plant Event" to prevent exceeding the vortex limit.

C. RCIC must be shutdown using RRC 13.1-2 "RCIC Operation During a Plant Event" to prevent direct pressurization of the Torus air space.

D. HPCI must be shutdown using RRC 23.1-2 "HPCI Operation During a Plant Event" to prevent direct pressurization of the Torus air space.

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer 16RO Choice Basis or Justification Correct: T-102 directs HPCI shutdown if torus level cannot be maintained above 9.5 feet as this is where the exhaust line is uncovered. Operation of HPCI, with its relatively high exhaust pressure, could result in direct pressurization of the torus air Distractors: Normal alignment for RCIC is with suction from the CST. Therefore there is no concern for the RCIC vortex limit (which is 6 feet). Plausible if the does not recall the normal RCIC lineup and/or is confused on the limits associated with torus level.

alignment for HPCI is with suction from the CST. Therefore there is concern for the HPCI vortex limit (which is below 9.5 feet). Plausible if applicant does not re,call the normal RCIC lineup and/or is confused on the various limits associated with torus level.

C RCIC turbine exhaust pressure is insufficient to cause pressurization of the torus. In addition, RCIC is likely to trip on high exhaust pressure if torus pressure became elevated. Plausible if the applicant confuses the HPCI and RCIC limits associated with torus level.

Psychometrics Source Documentation Source: o New Exam Item ~ Previous NRC Exam: (LGS 2006)

~ Modified Bank Item 0 Other Exam Bank: 0 ILT KIA System: £..V\JV\.,'V - High Pressure Coolant Injection RO/SRO 3.4/3.6 KIA Statement:

A2.07 - Ability to (a) predict the impacts of the following on the High Pressure Coolant Injection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the c~nsequences of thos~~bnormal conditioQs or ~erations: J-ow sum:>res~ion J2QQLlevel~ ___ ~~ ______

REQUIRED MATERI~LS2 I NONL - - ~- -- ~-~---- - - -~- ~-- ----~-. .

Notes

~- --- and

- Comments:

- - - - - - -- - - - ~----- -.~-- ..--------...- - - - .

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

17. Unit 3 is operating at 100% power when the "3A' Reactor Protection System (RPS) bus is manually transferred to its alternate power source.

Based on this event, what is the automatic response of the Standby Gas Treatment (SGTS)?

SGTS will start and the ___,,_ Filter inlet/outlet dampers will OPEN.

A. (1) "B' Fan (2) 'A' Train B. (1) 'C' Fan (2) '8' Train C. (1) 'B' Fan (2) 'B' Train D. (1) 'C' Fan (2) 'A' Train 2011 NRC RO Written Exam Rev. J.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question # 17 RO Choice Basis or Justification Correct: SBGT consists of three fans (A, B, C) and two trains (A, B). Since it is common to both units, e;ach PCIS power supply failure has a unique effect on system alignment: loss of 3A RPS bus starts the 'C' fan and aligns the

'A' filter train dampers; loss of 2A RPS bus starts the 'A' fan and aligns the

'A' filter train; loss of 2B or 3B RPS starts the 'B' fan and aligns the 'B' loss of 20Y033 or 20Y034 will align the 'A' or 'B' filter train not result in starts.

Distractors: A nrrlIT;:>rt fan. Plausible if applicant does not recall or understand the

>TT.

~rTe> of a loss of RPS bus power on the SBGT System.

nf'(,lrrQrt train. Plausible! if applicant does not recall or understand the QTT.<:.f'Te> of a loss of RPS bus power on the SBGT System.

C for loss of 'B' RPS bus. Plausible if applicant does not recall or

.nrlor",t-<3r,rI the effects of a loss of RPS bus power on the SBGT System.

Psychometrics

. i

_~Leyel of Knowl~dge ___ . .~ . . .. Difficljttym .trime Allo,#anc~ (mi!1ytes) m. RO mm __ m MEMORY 2.5 3 10CFR55.41 (b)(7)

Source Documentation Source: D New Exam Item [Zl Previous NRC Exam: (PB 2002)

D Modified Bank Item Other Exam Bank: 0 KIA System: 000 - Standby Gas Treatme*nt System Importance: RO / SRO 3.2/3.1 KIA Statement:

A3.02 - Ability to monitor automatic operations of thE! Standby Gas Treatment System including:

start.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

18. The following conditions exist on Unit 2 following a loss of offsite power (LOOP):

Reactor level lowered to -180 inches and is currently +20 inches.

The EDGs are supplying their respective 4 KV busses.

Offsite power has been restored to the 2 SUE bus.

The DIG AUTO START BYPASS pushbutton has been depressed.

The E-l EDG has been paralleled with the 2 SUE bus.

Which one of the following will occur if the E-I EDG is still operating in parallel with the 2 SUE bus when the E-l EDG Auto Start Bypass Timer times out?

E-l EDG load sharing will _ _ _ _ _ _ _ __

A. remain in the DROOP mode B. remain in the ISOCHRONOUS mode C. automatically swap to the DROOP mode D. automatically swap to the ISOCHRONOUS mode 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 AnSWE!r Key Question # 18 RO Choice Basis or Justification Correct: D Per NOTE 4.3 of SO 52B.2.A, depressing the DIG AUTO START BYPASS pushbutton initiates a 3-minute timer which places the diesel generator in the DROOP mode. CAUTION 4.3 states that diesel generator load sharing will automatically swap back to the ISOCHRONOUS mode when the DIG AUTO START BYPASS timer times out.

"--'-'---'--'-~'--'-'-'--'-"-'-~'-~~'--'-'---'~'-"'-'.

Distractors: A Per SO 52B.2.A, diesel generator load sharing will automatically swap back !

to the ISOCHRONOUS mode when the DIG AUTO START BYPASS timer times out.

B Per SO 52B.2.A, diesel generator load sharing will automatically swap back to the ISOCHRONOUS mode when the DIG AUTO START BYPASS timer I

~--"'---"-"--~"-4 .-

out.

.... -"-"'~~=~~~:~'--'-

C Per SO 52B.2.A, diesel generator load sharing will automatically swap back to the ISOCHRONOUS mode when the DIG AUTO START BYPASS timer times out.

Psychometrics

. ....h~velof Kno~!edgt3~_ __._.__._giffiCUltY__'___'I~Time Allowance (minuteS)~___~_._RQ. __ *.. _ ...

HIGH : 10CFR55.41(b)(8)

Source Documentation Source: o New Exam Item o Previous NRC Exam: 0 o Modified Bank Item ~ Other Exam Bank: (LORT)

._.._ . _ . _ . _.~D !LT Exa~(mk ..*

~-

-~::~~~~;e~L---T~~T~i6~i4-"--"-'-"--"-"---'---'---' --~-.---.--

Objective: !

KIA System: 264000 - Emergency Diesel Generators Importance: ROISRO i 3.4/3.4 .._

KIA Statement:

- . - -..- ....- .. - .*..- ...- -... -.~-- .-.---.~.-

A3.03 - Ability to monitor automatic operations of the Emergency Diesel Generators including:

~::;:E~:~Ti:~id reco0~~'E-- .~.

Notes and Comments: -_."----",-- -_..._-_._--._._.---_..- .

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

19. T-116 "RPV Flooding" is being executed following loss of all RPV level indication.

All ADS valves are open.

Which one of the following conditions indicates that the RPV is flooded to the Main Steam Lines?

Torus Pressure SRV Tailpipe Temperature A. 10 PSIG 240 OF B. 10 PSIG 260 OF

c. 20 PSIG D. 20 PSIG 270 OF Saturated Steam: Temperature Table (From 1967 ASME Steam Tables)

Temp (OF) Press (PSIA) Temp (OF) Press (PSIA) 176 6.869 240 24.968 180 7.511 244 26.968 184 8.203 248 28.796 188 8.947 252 30.883 192 9.747 256 33.091 196 10.605 260 j).4LI 200 11.526 264 37.894 204 268 40.500 208 13.568 272 43.249 212 14.696 276 46.147 216 15.901 280 49.200 220 17.186 284 52.414 224 18.556 288 55.795 228 20.015 292 59.350 232 21.567 296 63.084 236 23.216 300 67.005 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Basis or Justification Distractors: A Psychometrics

-~ le~1 of Knowled~ I [)Jffi(;ul~_~_~ Time Allgwanc~(mjl}ytes) RO HIGH 10CFR55.41 (b)( 10)

Source Documentation Source: [8J New Exam Item Previous NRC Exam: 0 o Modified Bank Item o Other Exam Bank: 0 o IlT Exam Bank ReferenceJ~ T-116 Bases Learning PLOT-1560-9 Objective:

-- -- ---_ .. '- --~---------~----

...* ~-

KIA System: 2~Q002 - Relief/Safety Valves I Importance: RO/SRO 3.6/3.7 KIA Statement:

A4.02 - Abilit~ to manu~lI~ oQeTe andlCl'-m.pi'itol"ln the control room: Tail QiQe ~emper~tlJres, __ -~-~~--

REQUIRED MATERIALS:

------ ~--- . -NONE ---~- ........~--

Notes and Comments: ----~ --~~- ~ -------- -~-- - ~~- ----~ ~~-

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 201 ]

20. Unit 2 is operating at 70% power at the end of cycle with the following conditions:

All APRM channels are in-service and operable

APRM' 1' has a total of 13 LPRM detectors bypassed

APRM 'I ' has 7 out of a possible 10 'B' level LPRM detectors bypassed Based on these conditions, which APRM '1' response is correct if an additional 'B' level LPRM to APRM '1' is manually bypassed?

A. NO alarms and NO Rod Block B. An APRM TROUBLE alarm and Rod Block C. An APRM DOWNSCALE alarm and Rod Block D. An APRM INOP alarm, Rod Block and Scram Vote to RPS 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Key'~-'~-'~-"-'--'~----'----~--~ ----.-.

RO Choice Basis or Justification Correct: B APRM trouble alarm and rod block due to "too few inputs" << 3 LPRMs per level). APRM will continue to average the remaining LPRMs.

Distractors: A Total LPRMs are still greater than the "too few inputs" per APRM setpoint

<< 20 LPRMs total), but < 3 LPRMs per level generates an alarm and rod C alarm only and 30 LPRMs remaining at 70% power will not result APRM downscale trip.

D Psychometrics

!__ __ ~Jffi~Ulty___ ~_ tTime ~lowanceJl'l'llnutes)..__ ~~.

r----~l::~v~12LKn_owl~e __ ~~ +1 __ ...:.O-,,---.~._._~_.~_~

,R HIGH I 4.0 I 3 J 10CFR55.41{b)(6)

Source DocUlmentation Source: D New Exam Item [8J Previous NRC Exam: (PB 2002)

D Modified Bank Item D Other Exam Bank: ()

3.3.1.1 Bases KIA System: 5005 - Average Power Range Monitorl : Importance: RO I SRO Power Range Monitor System 3.6/3.8 KIA Statement:

A4.06 - Ability to manually operate and/or monitor in the control room: Verification of proper Notes and Comments:

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

21. Given the following:

An A TWS is in progress on Unit 2

The CRS directed the URO to inject Standby Liquid Control (SBLC)

The URO placed the SBLC control switch to START SYSTEM 'A' Which one of the following indications is correct for these conditions?

A. BOTH "Squib Valve Continuity" lights will extinguish.

B. ONLY the 2A "Squib Valve Continuity" light will extinguish.

C. STANDBY LIQUID SQUIB VALVE LOSS OF CONTINUITY alarm will be received.

D. GROUP IIIIlI INBOARD AND OUTBOARD ISOLATION RELAYS NOT RESET alarms will be received.

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question # 21 RO Choice .......,...._- Basis or Justification Correct: D , These alarms will be received due to RWCU isolation on SBLC initiation.

Distractors: A Both squib valves fire when either pump is started. However, the continuity I lights remain lit and the "loss of continuity" annunciator will not alarm unless or until the pump control switch is placed in OFF. Plausible if the applicant J r--..----------+-.---- --t, ....::.~--'--~~:~.::J_,u=~=..~t:~~:lls~~~~~~I~~i~~u~~~~h~~~~~~Ft~~t~~dn~_acc~rate~.~icat~o~_

B , Both squib valves fire when either pump is started. However, the continuity lights remain lit and the "~oss of continuity" annunciator will not alarm unless '

or until the pump control switch is placed in OFF. Plausible if the applicant believes only the 2A squib valve will fire and does not recall that the continuity lights do not provide accurate indication of squib valve status 1 - -.... - - - - . - . _.. _ _. - t__ ._. __.. _+_fo_lIowin~BLC sys~m initiation:._.. ___.__.___.. _ .. _ . _.._ . ___.. _.. _ . _. __..___.._._

C Both squib valves fire when either pump is started. However, the continuity I lights remain lit and the "loss of continuity" annunciator will not alarm unless '

or until the pump control switch is placed in OFF. Plausible if the applicant I does not recall that the alarm circuit does not provide accurate indication of I L.._ _._ _.__._ _.._ .._ _.. ...-L_.._ _ _..__.L~9..ulP_v_C!lve~t~!l!!:)lollo~jI~~SBLQ.systemjniti~~ion... __ u _ _ ** _ . _ ** _ _. _*

_ _. _ . 1 Psychometrics

~Lev~1 of Knowled~_1 ____[)iffi2LJJ!y._.. __J .....Iime_Allo.wan_~.Jmi~lJ!~~Lt_ RO

MEMORY i : I 10CFRSS.41(b)(6)

Source Documentation Source: D New Exam Item [g] Previous NRC Exam: (PB 2007)

[g] Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank

. Reference~L __L.SQ._l1J.: B-~ A~~:l11Ji-3~6J~j:~...?14 D:-1; ARC-21 £~-1____._.. __..__ ... _. __ ._._.. __

i Learning I PLOT-S011-4d Objective:

KIA System: J.:~~~~~StandbY Liquid Control System Importance: RO 1 SRO 1------- --_.------.

4.2/4.2 KIA Statement:

G2.4.46 - to that the alarms are consistent with the conditions.

REQUIRED MATERIALS:

Notes and Comments:

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II

22. Which one of the following is correct regarding SRV operation from the Alternative Shutdown Panel in the Recirc MG Set Room?

The ) SRVs can be operated from this location and SRV position indication comes from the SRV __(2)__.

A. (1) A, B, and K (2) acoustic monitoring B. (1) A, B, and K (2) control switch position

c. (1) H, E, and L (2) acoustic monitoring D. (1) H, E, and L (2) control switch position 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer

22RO Choice Basis or Justification Correct: B The A, B, and K SRVs can be operated from the Alternative Control Station. Per SE-10 Caution #104 and Bases, "SRV position indicating lights do not indicate actual valve position, but rather the lights are simply a reflection of the switch position. Acoustic valve position is not part of the nrn,t&>l"t""t1 ASO "

Oistractors: A since position indication for SRVs operated from the Remote Panel is from the acoustic monitors, not SRV control switch C since the H, E, and L SRVs are operated from the Remote Panel, not the Alternative Shutdown Panel.

o since the H, E, and L SRVs are operated from the Remote 11'I'1r\\A/nPanel, not the Alternative Shutdown Panel.

Psychometrics RO ._

10CFRSS.41 (b)(7)

Source Documentation I

Source: D New Exam Item [g] Previous NRC Exam: (PB 2009)

I D Modified Bank Item D Other Exam Bank: 0

_. I [g] ILT Exam Bank ..

.-1 SE-10and Bases; SE-:1

- ~

Reference(s): *__ " h ~ . _ _*_ _ _ _ _ _

_ _ _

_ _ * * ~ _ *

_ *

__ * * ~ *

__

_ ~

Learning I PLOT-S001A-Sd, -Sf Objective:

KIA System: ..1."14111 ..I - Relief/Safety Valves Importance: RO/SRO 3.8/4.0 KIA Statement:

rG2.1.32 - ~bility ~~xplain and aM system limit~i!!lQJ:!recautions. ____ ~_ ~ ~ -- --_._---- _.

REQUIRED MATERIALS: =i=0NE .. _ .......

No.!e~_.§!r1..d_C"O_f!lI"l'l~~~ ___~~_ _ _ _ _ _ ~. _____~ _____~ __~______ ---~

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

23. The Instrument Air System is in a normal lineup when the following occur:

INSTRUMENT AIR DRYER TROUBLE (216 C-4) alarm is received

B INSTRUMENT AIR HEADER LO PRESS (216 D-4) alarm is received

'B' Instrument Air Header Pressure (PI-2425B) on Panel20C012 is lowering

'B' Instrument Air Receiver Pressure (PI-2429B) on Panel20C012 is steady at 110 psig

The TBEO reports there is a valve malfunction on the 'B' Instrument Air Dryer and that neither the 'C' or 'D' drying tower is in service Which one of the following describes the correct action to mitigate this event?

A. Isolate the 'B' Instrument Air Dryer B. Bypass the 'B' Instrument Air Dryer C. Cross-tie 'A' and 'B' instrument air headers D. Cross-tie Unit 2 and Unit 3 'B' instrument air headers 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question '# 23 RO Choice Basis or Justification Correct: D The given conditions indicate both towers for the 'B' air dryer are isolated, which means there is no flow to the 'B' instrument air header from the 'B' compressor/receiver. Therefore, 'B' instrument air header pressure will continue to lower. Per ON-119, the correct action to take for this condition is to cross-tie the Unit 2 and Unit 3 'B' instrument air headers.

Distractors: A This alone will not mitigate this event; the 'B' instrument air dryer is effectively already isolated from the 'B' instrument air header. In order to restore 'B' instrument air header pressure, the Unit 2 and 3 instrument air headers must be cross-tied.

There is no provision for bypassing a malfunctioning dryer in ON-119 or ARC-216 C-4. Both references direct cross-tying the Unit 2 and Unit 3 instrument air headers.

C Cross-tying the 'A' and 'B' instrument air headers will not be effective in restoring 'B' instrument air header pressure since the 'A' supply must pass L .. _~ ____ ~_~ ______ ~ _____ .~L_~ __ ~ __ the 'B' air

~ .*. _ __ in order to the 'B' header. _~~ ____ ~ .....*. _ . _ _ ...*.*. ~

Psychometrics f-~_Level of KnowledgE3__ ~ ______Qiffi~l!I!l _ _~_+ _Tirl1~~lIo\yance (rl1lnut~s.w ____R~ ______

HIGH i 3.0 I 2 I 10CFR55.41Jb)(10t Source Documentation Source: o New Exam Item ~ Previous NRC Exam: (PB 2007)

[gI Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank ON-119 Learning PLOT-5036-6b Objective:

KIA System: .... VI.JV ...'V -Instrument Air System (lAS) RO/SRO 2.9/3.0 KIA Statement:

K3.03 - Knowledge of the effect that a loss or malfunction of the Instrument Air System will have on the following: Cross-tied units.

R~qLJlriEDj!~"ERIA_~s==L~oi!E-=~-=_==-===-=====_===_-==.=--===:-.

Notes and Comments:

20 I I NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

24. Unit 2 was operating at 100% power when a feedwater malfunction resulted in a reactor scram.

HPCl and RCIC initiated on low RPV level

Reactor level is now +20 inches and rising

Subsequently, LT-2-02-3-0nC, one of the two Wide Range Reactor Water Level inputs to HPCI logic fails downscale

All other RPV level instruments remain operable Assuming no operator action is taken, which one of the following describes HPCI System response as RPV level rises?

The HPCI System will __________ at +46 inches RPV level.

A. trip AND isolate B. trip but NOT isolate C. isolate but NOT trip D. NOT trip and NOT isolate 2011 NRC RO Written Exam Rev.l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer QUE!!SUC:m #. 24 RO Choice Basis or Justification Correct: RPV high level trip requires input from both LT-72C and LT-72D (2 out-of-2 logic). With one transmitter downscale, the HPCI system will not trip on high level. HPCI does not isolate on high RPV level (isolations not the other Distractors: A The HPCI trip logic is 2-out-of-2, requiring both inputs to cause a trip.

does not isolate on high RPV level (isolations cause a turbine trip, not the B trip logic is 2-c)ut-of-2, requiring both inputs to cause a trip.

C not isolate on high RPV level (isolations cause a turbine trip, not way around).

Psychometrics Level of Knowleqge_ Difficulty_ I TLm~Allowance (r1}JI'..'Il::Ites) I RO

~---~~ ....... --.. ~

HIGH 3.5 4 10CFR55.41 (b)(7)

Source Documentation Source: D New Exam Item [8J Previous NRC Exam: (PB 2007)

I D Modified Bank Item D Other Exam Bank: 0 i

[8J ILT Exam Bank Reference( s) I SO 23.7.C-2; ARC-221 B-1 Learning PLOT-5023-4c Objective:

KIA System: --- i206000-::'-Hi9h Pressure cooi~-nt injection I II - - . . . . -----------..

Importance: RO I SRO System .____._____~.:.4{.3* ~___._____ . _________

--- -- ~ ---

KIA Statement:

K 1.12 - Knowledge of the physical connections and/l::>r cause-effect relationships between the High rPressure Coolantlnjection System ang the following: Nuclear boilE!!r instrumentation.:______

REQUIRED MATERIALS: NONE Notes and Comments: An event occurred at PB in 2004 (IR #252501) that resulted in loss of 1 input to the high level trip logic due to a lifted lead. This condition

' - - - ........... - _....... went unnoticed for several weeks.

201 t NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

25. Given the following:

Unit 2 is operating at 25% power

#2 APRM fails downscale (not INOP)

This condition will generate an _ _ _ _ _ __

A. alarm ONLY B. alarm and rod block ONLY C. alarm and half scram ONLY D. alarm, rod block and half scram 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer

25RO Choice Basis or Justification Correct: B . APRM downscale (::: 3.2 %) in MODE 1 will generate a control rod 'AIITt"lrlr'<=I'A/

block and downscale alarm 211 C-2 only.

Distractors: A APRM downscale (::: 3.2 %) in MODE 1 will generate a control rod withdrawJ block and downscale alarm 211 C-2.

I----..-.~----*---*--....f-**-----+*-**--***-~*--**--*****-..- - - . - - . - -..- ... - .....- ....- - ' - . - " - ~""--'-~-- - - -....- - - ..

C I A scram vote signal is only generated for: APRM INOP Trip, High Neutron

Flux, and Simulated Thermal Power High. In addition, APRM downscale

.\ will a control rod withdraw 1 o A scram vote signal is only generated for: APRM INOP Trip, High Neutron Flux, and Simulated Thermal Power High.

PS'ychometrics h-. Lev~J Kr1owledg~J_. ________QlffiClJlty . ________ _Iime_t.llg-""l:!n~~(r:nlrJ.l:Ite~) RO I MEMORY I 10CFRSS.41 (b )(7)

Source Documentation Source: D New Exam Item k8J Previous NRC Exam: (PB 2007)

D Modified Bank Item D Other Exam Bank: ()

k8J ILT Exam Bank

$efere~c~:~IAR~,?J 1-C-2=~ ___._._ . _._._._._._.__ .__~_-==-~~

Learning I PLOT-S060-3a Objective: I rt<lA Syste;-- 1215005 -Average-p;;'er Ra~g-;' M~';;tor/-rl~-~ort;n~;-Ro ISRO ----- ---

Local Power Range Monitor System 3.7 I 3.6 1-------- --------------------------- - - - ---- - -- -- -- - - - -----

KIA Statement:

A3.08 - Ability to monitor automatic operations of the Average Power Range Monitor/Local Power Rail9..e_l\,1oni1oI§l'..~tef!lll1£I.l:Iclll!9: ggntr()lI()d bloc~st.~u_~.: __._.._.___._.._..__..____..___._ . _____._._.___.__._. ___._.

==[~ONE REQUIRED MATERIALS:

'. Notes an~co~m~nt~: --==~~_=_=-=======~====~==~==: ________.:_=

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 I 1

26. Unit 2 is initially operating at 100% power when the following events occur (all times are in minutes):

T=O Drywell pressure is 2.1 psig

the E32 BUS DIFFERENTIAL OR OVERCURRENT RELAYS (004 C-l) alarm is received

T=5 -the 'A' EMERG SERVICE WATER PUMP TRIP (002 B-5) alarm is received Assuming no operator actions, what is the status of the 'B' ESW pump and the ECW pump two minutes later?

The 'B' ESW pump is )__ and the ECW pump is __(2)__.

A. (1) running (2) running B. (1) NOT running (2) running

c. (1) running (2) NOT running D. (1) NOT running (2) NOT running 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer

26RO Choice Basis or Justification Correct: B When drywell pressure rises above 2 psig the 'A' and 'B' ESW pumps and the ECW pump will automatically start. With ESW header pressure> 30 psig, the ECW pump will automatically shutdown 45 seconds after it first started. Since the 'B' ESW pump is powered from the E32 bus, it will trip as a result of the E32 bus differential/overcurrent condition. When the 'A'

. ESW pump trips, the ECW pump will automatically restart when ESW i~. ._..__~. _ _.. ~. ._~i_. _ _ ~~ead~~ pr~ssur~droQ~~~12w~~O psi~_.~_ . _ _ ..~. _.__..__.. ___.._.~_. .~. __

Distractors: I A First part is incorrect - the 'A' ESW pump is not running due to the E32 bus differentiallovercurrent condition. Plausible if the applicant does not know r__.___...___..____ +I_._~.~__ ~+-~m p_power_supplie~_9L.~ec9fll1iz~theE:?.£pusj~<2.ked _Ql.IL____~_~~.__

~ C Both parts are incorrect -- the 'A' ESW pump is not running due to the E32 i bus differential/overcurrent condition. The ECW pump is running due to the I I loss of both ESW pumps (low ESW header pressure). Plausible if the ,

1, -'.-~ll.L. :~~~~~~~:sd~~~k.~~t~I.~~;;~~.C;:~~~ri~~~~_;[a~~~~~i~~~~~._E32.~.us i~__

. I 0 Second part is incorrect - the ECW pump is running due to the loss of both 1_.. __.. ___._L__ ~~__;~V::~':~~~o;u~~V:;u~~~~:~P~~~~~~:~~. Plausible if the applicant does Psychometrics

'- Le"el of I<nowl~dg~ __ l __ u _ _ Piff~~J!y_____ Til!1e AII()wance (mir1~tes)_..;.I,. _ . ._.__._c.:R:..:::.O._.____._.

HIGH I I 10CFR55.41(b)(7)

Source Documentation Source: o New Exam Item [gJ Previous NRC Exam: (PB 2009)

I [gJ Modified Bank Item 0 Other Exam Bank: 0

_.~__.__.. _-t-I:8JJLT Exam Ban~_._._.__._._____._.__.._.

Ref~rence(s): __ I SQ §2A.1:B; AI3C-OQ~ C-LARg:9Q2J:!-§. _____ _

Learning tPLOT-5033-4a Objective:

~ .. - - - - - - --- . _ . _ - - - . - - -- -_. -------- -~.,- -~ --------- ---~.-

KIA System: .400000 - Component Cooling Water \ Importance: RO I SRO i System (CCWS) 3.0~. ~/_3_..0"'--__"'_"~'_"' ____i KIA Statement:

.A . . ._.ati~.s. . . tart9f. ~tan.dl?l'. JL.UElP ._. _..__.. ___. . . _. _.__.._.

K4.01 - Knowledge of CCWS design feature(s) andlclr interlocks which provide for the following:

. ._ut~m ~_ ~ . .-- . . . --_..

_._.--~.-.- -~._~ _.~--

RE~UIR~D M~TE~IAh'§-=-__

Notes and Comments:

~-----"'---"'---"'-'-'~-""-"'-'- f . NONL___.~_. ___.. __~_~ ....___ . _.~ ___. _.. ~~~~~___.._. __ ~._. . __

..- .. - ..- -... -.~.-.~ ... ~---- .. -. .-...

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

27. Given the following:

Unit 2 is operating at 85% power

The 2A Reactor Feed Pump tripped

Reactor level dropped to + 15 inches before turning and beginning to rise Based on these conditions, what is the most limiting Recirculation System response and the reason for that response?

The Recirculation pumps will runback to __________

A. 30% to ensure adequate Reactor Feedwater Flow is available B. 30% to ensure adequate Recire Pump Net Positive Suction Head C. 45% to ensure adequate Reactor Feedwater Flow is available D. 45% to ensure adequate Recirc Pump Net Positive Suction Head 2011 NRC RO Written Exam Rev. Ldoc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question # 27 RO ----~~~~~~

Choice Basis or Justification

-,~.-- . +~.~------~---- ....... -----~-----------~---~-----------~--~-----i Correct: With a reactor water level < 17 inches and individual RFP flow < 20%, a 45% runback will occur to ensure adequate feedwater flow is available.

Distractors: A Runback to 30% is incorrect-reactor water level < 17 inches concurrent with a reactor scram will cause a 30% runback. The reason is also incorrect-this is the reason for the 45% run back.

B Runback to 30% is incorrect-reactor water level < 17 inches concurrent with a reactor scram will cause a 30% runback. The reason is correct

_____________ +-~ ____; _ _ .L...... ~ _____ .~ __ ~~~~

isc......._

.. the reason

__ ___ ~

for the_30%.

_ _...._ _ ~_ __~ _________ ~

D Runback to 45% is correct. The reason is also incorrect-this is the reason for the 30% runback.

Psychometrics r- Lev~lof Knowledge r-~~~** .. Difficulty__ .~ I Time Allowance . .

RO MEMORY

I 10CFR55. 41 (b)(6)

Source Documentation Source: New Exam Item [gI Previous NRC Exam: (PB 2009)

[gI Modified Bank Item D Other Exam Bank: 0 I*~ .

[gil LT Exam Bank .-~~-,,--~, ."-.... -----~

Reference(s): - .-

OT-100; UFSAR Chapter 7.9...-and

~--.... . ... ~ -- 7.10

-.......... - ~~---.-------

Learning PLOT -5002-4b Objective:

-_._------ .~--- .... ..............- _.

KIA System: 259001 - Reactor Feedwater System Importance: RO/SRO

-~.-.-------- .. ---------------------"~----.--~ ...... I 3.1

..........._... /3.1 KIA Statement:

K 1.16 - Knowledge of the physical connections and/or cause-effect relationships between the Reactor

£eedVv'(:lter System and the following: Recirculation: ~~~~.-~----

REQUIRED MATERIALS:

~--------~-

NONE ~~----~- .. .~.

.~-

...... -----~------

Notes and Comments: .

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

28. Given the following:

Unit 2 was initially operating at 100% power

7 Drywell Cooler Fans have their control switches in RUN

7 Drywell Cooler Fans have their control switches in AUTO

A loss of offsite power (LOOP) occurs

All 4 EDGs start and re-energize their associated busses

RPV level is being maintained by HPCI and RCIC

Drywell pressure is 0.9 psig With no operator actions, what is the status ofthe Drywell Cooler Fans?

A. ALL fans are tripped B. ALL fans are running C. ONLY the 7 fans in RUN are running D. ONLY the 7 fans in AUTO are running 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II Answer Question # 28 RO Choice Basis or Justification Correct: C power is lost to the 480 V emergency MCCs, the running fans will stop.

power is restored, the fans with control switches in RUN will restart; fans with control switches in AUTO will remain off.

Distractors: A power is restored, the fans with control switches in RUN will restart.

fans would be tripped if drywell pressure was above 2 psig.

B Fans with their control switch in AUTO only start on a low flow condition following a 10-second time delay.

D Fans with their control switch in AUTO only start on a low flow condition following a 10-second time delay.

Psychometrics r-~L~~E9Lof !5no,,!le~ J--~ __ Difficul!Y______ ~-lim~ Altow§lnc~_ (mil"1l.JtesLL__ ~_~9~.____.~.~

MEMORY I j i 10CFR55.41(b)(9)

Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0

,~_. __.__.~_.___ .__+ ___rg]",=_I~L.T ____ Ex_~a~m._B__a_.._n_~k.~. ___~. __~. _~.~. ___.._~. ~~_ . _____ ~ __.___.~~. __..____.____ .____ ._~_____ _

KJA-sy;;;;m:**

- -.. ~~~.-~--~---

t223001 .:

Auxiliaries

.. -----~ ..---..-

PrimarY Conlai';';'e~t

..- , , -.. --~ ..----..- - - - . - - . - -

s~ste";-;'ndr~po.rt;n~;~--RO/SRO

~ - --_..

2.7/2.9 KIA Statement:

JS2.09-=l<nowledge of the electrical pow.E9.[_s~lie_s to theJQ.IIQ,,!i!!g.:J)~ell cQ.olil'!9. fan~:_~__ ~~ __ ~_

-~!~:~-'6::::~~:~:- p~o~~_ ---------------- -~ -- --- - ~- - - - --.*~.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

29. Given the following:

A reactor startup is in progress on Unit 2

The following control rod drive conditions exist:

1. Control rod 18-39 is at position '04' and is stuck

2. Control rod 38-19 is at position '36' and is isolated

3. Control rod 42-43 is at position '48' and has a slow scram time Which of these conditions has a negativ(~ effect on Shutdown Margin?

A. 1 and 2 combined B. 1 and 3 combined C. 2 and 3 combined D.20NLY 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question'" 29 RO Choice Basis or Justification Correct: A A control rod that is stuck at position '04' along with a control rod that is isolated from its HCU at position '36' challenges Tech Spec shutdown margin criteria. Reactor Engineering determination is required under these conditions to determine if SDM criteria are met.

Distractors: B A single control rod with a slow scram time does not challenge shutdown margin criteria. Tech Spec 3.1.4 allows up to 13 (depending on location),

I but the bases for limiting the number of slow control rods is to ensure valid

! and accident not for shutdown C A single control rod with a slow scram time does not challenge shutdown margin criteria.

D A single isolated control rod does not challenge shutdown margin criteria.

Psychometrics

~,=-ev~1 of ~KnqwleSifL~J~_~~~_QIfficulty____ .~i _ Tif!1e Allow!3n~j.'!ljnutes 11----..- ~O ..~

MEMORY I 10CFR55.41(b)(2)

Source Documentation Source: [?3J New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank Reference( s) :-~1j)Q-~JJ~:-2; T ech~~ecdefirlitiQrl~;_T~ch _§pe~1-.t.T*:10J,_[\Jot~ 24__~__ ~. _ ..

Learning

PLOT-1800-3; PLOT-PBIG-2101-T101-4, -6 Objective: I KIA System: Importance: RO I SRO

_____ ~.L _ _ _ _ _ _ _ __

~ ~

3.2/3.8 KIA Statement:

K3.03 - Knowledge of the effect that a loss or malfunction of the Control Rod and Drive Mechanism will have on the Shutdown REQUIRED MATERIALS:

Notes and Comments:

20 II NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 J 1

30. Per T-103 "Secondary Containment Control" Bases, which statement below describes the reason for the following step?

~---------

---------1 I IF REACTOR BLDG OR REFUEL FLOOR I I --- VENTILATION EXHAUST RADIATION B LEVELS EXCEED _ MR/HR, I I THEN VERI FY THE FOLLOWI NG: I I ---- -REACTOR BLDG AND REFUEL I I FLOOR VENTILATION ISOLATION I L!;; ________ -. _________ I

SBGTS INITIATION L- SCC-2 I I

A. Prevent an offsite radioactive release.

B. Provide for a filtered and elevated release.

C. Minimize the radiation exposure to station personnel.

D. Route release path through hardened ducts to prevent ductwork failure.

20 II NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 I I Answer Question" 30 110 Choice Basis or Justification Correct: A high radiation condition, as sensed by RB and RF Floor Ventilation process radiation monitors, will automatically isolate RB and RF Floor Ventilation dampers to contain potentially contaminated air. At the same i time, the SBGT System is initiated. Per T-103 Bases, the purpose of this i action is to provide for a filtered, monitored, and elevated release through the main vice vent stacks.

Distractors: The isolation does not prevent a release, but does filter and elevate the release. Plausible since the purpose of most automatic isolations is to a release.

Radiation dose to station personnel is not changed by the isolation.

However, the severity of the release to the public is minimized.

This distractor is based on a procedural precaution about manual alignment I. of SBGT; it is not the basis for automatic isolations.

Psychometrics Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item [2J Other Exam Bank: (LORT)

ILT Exam Bank Reference(s)_:_. T -19~~ases; SO 9!-~tL _____________

Learning PLOT-5040B-4b Objective:

KIA System: i 272000 - Radiation Monitoring System Importance: RO / SRO 3.7/4.1 KIA Statement:

K4.02 - Knowledge of Radiation Monitoring System design feature(s) and/or interlocks which provide for the following: Automatic actions to contain the radioactive release in the event that predetermined release rates are exceeded.

MATERIALS:

Notes and Comments:

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

31. Given the following:

Unit 2 is operating at 100% power

The' A' RPS bus was transferred to its alternate power supply

Reactor Building Ventilation is being restored in accordance with SO 40B.l.A-2 "Reactor Building Ventilation System Startup and Normal Operation"

After placing all system fans in their normal lineup, Reactor Building and Refuel Floor differential pressures on Panel 20CO12 indicate +0.1 inches H2 0 Which one of the following actions is correct in accordance with SO 40B.l.A-2?

A. remove one RB Exhaust Fan from service B. place one additional RB Supply Fan in service C. remove one RF Floor Supply Fan from service D. place one additional Equipment Cell Exhaust Fan in service 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question # 31 RO -

~ ..

Choice Basis or Justification Correct: C Per SO 40B.1.A-2, the normal ventilation system lineup is 1 Equipment Cell exhaust fan, 2 RB supply fans, 2 RB exhaust fans, 2 RF supply fans and 2 RF exhaust fans. Per the SO, if the normal ventilation system lineup does not establish normal differential pressure (-0.1 to -0.4 inches H2 0), the procedure directs removing one RF supply OR exhaust fan from service. In this case, a RF supply fan must be remo~~cL!rom servi.ge toesta~bll~hJhe 'pf"QP~r negati."..~sJ{Q:__~

. Since normal differential prE~ssure is -0.1 to -0.4 inches H20, removing one RB exhaust fan from service would cause differential pressures to become more t

positive. Plausible if applicant does not know required dIp range andlor does not

~------"~B~- ~~::~rs~:~!;~~~:::~~~:~::~~::;~~.1-t~~o-.4-in-ches H 0, placing one additional i

2 I RB supply fan in service would cause differential pressures to become more I i positive. In addition, SO 40B. 1.A-2 gives direction to avoid running 3 RB supply or

. _________ J_.~._~_:_~!:~~tnf~/~~ ~~:se;~~~~~~~:a:es~~:t:~~~;:~~~~~~~t~:~:reqU~~d dIp 1 o This would make dIp negative, but SO 40B.1.A-2, Caution 4.5.1-1, prohibits running both Equipment CeJ!l exhaust fans. Plausible if applicant does not recall SO 40B.1.A-2 cautions and limitations.

Psychometrics

__ ~vel ()f Krl()wl~dgeJ._ . __ Qiffi~tillY . _ . _._....~I._IiD:I~~II()'v\Iil~c;~J'!lin~tE!SL L~ . ____ J~O_~ .~~ ...

HIGH ! 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 ILT Exam Bank I---~---~""~""---""~-

SO 40B.1.A-2 Learning PLOT-5040B-5b Objective:

KIA System: L.V\JV\.'V - Plant Ventilation Systems RO/SRO 3.2/3.4 KIA Statement:

K5.02 - Knowledge of the operational implications of the following concepts as they apply to Plant Ventilation Systems: Differentiilressure control.

REQt.JIRED MATERIA~~ ~-. NO-NE__=-~~-=~===~=-=~'-=-====~--~------

Notes and Comments:

'---.-"'-"--"~-"~'-~"--"- "-.-"--.~-.~,, .. ~.~.---.-----~---.,,~-.-,,.-.~,,--.-- ... --~ .. -~ - ...--~--.----~ ... -

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

32. ON-114 "Actual Fire Reported in the Power Block, Diesel Generator Building, Emergency Pump, Inner Screen or Emergency Cooling Tower Structures" contains the following note:

NOTE IF power is lost to the Motor Driven Fire Pump (OOP064) controller for more than 8 seconds, THEN the Motor Driven Fire Pump automatic start feature is defeated.

This interlock does NOT affect the ability to manually start the pump. Guidance for resetting the auto start logic can be found in SO 37B.l.A "Common Plant Fire Water System Lineup for Automatic Operation".

The basis for defeating the automatic start feature is to prevent _ _ _ _ _ _ __

A the pump from automatically starting with reduced bus voltage B. overloading the diesel generators during a loss of off-site power C. a simultaneous start with the Diesel Driven Fire Pump and a water hammer D. a spurious start due to loss of power to the fire header pressure instrumentation 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question # 32 HO Choice Basis or Justification Correct: B 14 Bases for the note says that the defeat of the auto start feature after an 8 second loss of power to prevent an auto start during a event which could cause an EDG to exceed its 200 hour0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months Distractors: A The candidate could believe that this interlock is to protect the Fire Pump from damage due to low bus voltage.

candidates are trained to have concern for situations that may cause hammers and resultant equipment damage and possible personnel nnlcA/r;:.I,/r;:.r this is not the concern in this situation.

D of instrumentation is a plausible concern for causing an undesired system operation.

I Psychometrics

_J~veLQLKnow~QR~~L_~~ . __Diffi~ult}'__ ~~ . ._~ -.Time Aliowanc~Jmin..!lt~~ RO I MEMORY 3.0 3 1OCRF55.41 (b)(8) I Source Documentation Source: New Exam Item [8J Previous NRC Exam: (PB 2005) o Modified Bank Item 0 Other Exam Bank: 0

_.. .~. ~. _ . _~~JLT Exam Bank __.._._._~ .._ _ ~ ___~_. __ ~_. _~~

RefereQce( s)~_ ~Q!,!-J1~Base~ __.. _.._ ... _~_~___ .

Learning 550-4 Objective:

KIA System: L\JI,JV'-"V - Fire Protection System Importance: RO I SRO

--~- .. --.~~~.--~-

3.1 13.1 KIA Statement:

K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the Fire P....rote~ion _~JE9m:...f\______C..:~E9leptrical distr~~uti(>n. _ _.._.~_~~. ___._..__ .~. _ _. ~

REQUIRED MATERIALS: ]NONE N()te~<iind S~omment~~ ___L __._.._._~ ______._.._~~__.._____ ~__ ~.~_._~____~_~__..____.._.~ _______

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II

33. Given the following:

Control Room Ventilation was initially in a normal lineup

The PRO performed SO 400.7.B "Place Control Room Emergency Ventilation In Service From The Control Room" Which one of the following describes th,e effect of this action on Control Room pressure relative to Turbine Building pressure?

Control Room pressure will _ _ _ _ _ _ _ __

A. remain approximately the same since it is controlled by modulating dampers B. remain approximately the same since: it is based on supply & exhaust fan capacity C. become more positive since more air is being supplied to the Control Room O. become more negative since more air is being exhausted from the Control Room 20 II NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question # 33RO Choice Basis or Justification

-~-----~-----------~----------------------.------

Correct: B During normal operation, the CR ventilation system maintains the control room at a positive pressure by virtue of the capacities of the supply, fresh air and return fans-20,600 CFM is delivered by the supply and fresh air fans; 18,600 CFM is exhausted by the return and exhaust fans. The remaining 2000 CFM pressurizes the control room and is exfiltrated to the turbine building. CREV operation does not change control room pressure since 3000 CFM is supplied by the CREV supply fans and 1000 CFM is still exhausted (to the TB roof) via the toilet and utility room exhaust fan. The

. remaining 2000 CFM pressurizes the control room and is exfiltrated to the turbine the normal .........."...""t Distractors: A The control room ventilation system uses modulating dampers to control temperature and humidity, not for controlling control room pressure.

C Operating CREV does not change the amount of flow brought in from the outside or the amount that is exhausted.

o Operating CREV does not change the amount of flow brought in from the outside or the amount that is exhausted.

Psychometrics c--__~-~I Of~f!.QWI~d~~J ____ ~_pjfficultY ___ . _.JJime_Allo'.'Vanctimi~utes Ll______B~ ____ _

I MEMORY I I I 10CFR55.41 (b)(7)

Source DocLilmentation Source: I2SI New Exam Item o Previous NRC Exam: ()

o Modified Bank Item Other Exam Bank: ()

o ILT Exam Bank f-~eferenc~}:___ i f!'1-3.§~$O 40D.5.A; SQ40QZ~____ ~ _ _ . ____._.____.. ____.. __._______.___ - -.. ~-.-

Learning I PLOT-5040D-5b Objective:

KIA System: 290003 - Control Room HVAC l,mportance: ROISRO 2.5/2.8 KIA Statement:

A 1.04 - Ability to predict and/or monitor changes in parameters associated with operating the Control Room HVAC controls including: fntrol room pressure.

~---------------------------- ----------------.-----~------

~ ___

REQUIRED MATERIALS:

c------------------------------------ NONE

~-------- ..--------------------. '-

-Notes

and

- - - Comments:

- - - , .__ .._--_.. _---_.._ . . .~---.--.- .. ~-.---------.-"'-------------.-------- .. ----~ .. -~--~-.---.---~.-- .." . - -..---.----.. ~ -- .. --~

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

34. An electric ATWS exists on Unit 2.

The Reactor Operator is directed to perform T-220-2 "Driving Control Rods During a Failure to Scram".

Prior to implementing this procedure, the Rod Worth Minimizer (RWM) loses power.

Which one of the following describes (1) the impact of this power loss on control rod insertion and (2) the action required by T -220-2 to insert control rods?

A. (1) Control rod insertion is prevented (2) Bypass the RWM AND place the Rod Control switch (3A-S2) in the "IN" position B. (1) Control rod insertion is prevented (2) Bypass the RWM AND place the Emergency In / Notch Override switch (3A S3) in the "EMERG ROD IN" position C. (1) Control rod insertion is NOT be prevented (2) Place the Rod Control switch (3A-S2) in the "IN" position D. (1) Control rod insertion is NOT be prevented (2) Place the Emergency In / Notch Override switch (3A-S3) in the "EMERG ROD IN" position 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Key Question#. 34 RO r----------~~_,-~--~-~~--.--.~~----~-. --~--.~.-----.--------.----.--.-- .. --------------~--,

[-----

Choice Basis or Justification Correct: B A loss of power to the RWM (Le., hardware/software failure) will result in all rod blocks becoming active, unless the RWM is bypassed. T -220 directs bypassing the RWM (regardless of specific plant conditions) and inserting control rags using the "Emergency In/NotchQYE:lrride" control switch'--___

Distractors: A T-220 directs inserting control rods using the "Emergency In/Notch Override" control switch.

C A loss of power to the RWM will result in all rod blocks becoming active, unless the RWM is bypassed. T-220 directs inserting control rods using the r--_ _ _ _ _ _---1_________u+.'E-rnergellcy In/Notch Override" control swi!9h.: ___ ._____ u__ ._._. ______ .______ . __ _

D A loss of power to the RWM will result in all rod blocks becoming active, unless the RWM is bypassed. T-220 directs bypassing the RWM

'---__ . . . . . . . . . . . . . . . . _ _----"_____---L(regarQless o~ecific plant conditionSl):

Ps chometrics Level of Knowled e ____Difficultyu.. . . . . . . . . . . .__ ~~~Lm.il'lut_e--'s)'---+-_ _ _R_Q __

HIGH 10CFR55.41 b (6)

Source Documentation Source: I C8J New Exam Item Previous NRC Exam: 0 I 0 Modified Bank Item Other Exam Bank: 0

.......L_D.l!:I Exam Bank .........* --.--~-

ReferecrRC-21lF-S; SO 62.7.A-2; T-220;. M-1-S-20 Sheets 9, 12 -- ....... __u___

Learning PLOT -5062A~6a Objective:

KIA System: I 201006 - Rod Worth Minimizer (RWM) I Importance: RO/SRO

~ ......... ---. 2.5/2.8 u -.---.--~--

KIA Statement:

A2.01 - Ability to (a) predict the impact of the following on the Rod Worth Minimizer (RWM); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Power sl:Ipply los~_....... -- ---------- -_._

REQUIRED MATERIALS:

Notes and Comments:

~.

_ _ u ._ _

INO~ ~-~--~

-~--

-_. _u..._

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

35. The following conditions exist on Unit 2:

An ATWS is in progress

SLC has NOT been initiated

Reactor pressure is being controlled with RWeU in the Recirc Mode

T-227-2 "Defeating RWCU Isolation Interlocks" has been completed

A pipe break occurs in the suction line of the operating R WCU pump, causing RPV level to lower Based on these conditions, the R weu System will _ _ _ _ _ _ _ __

A. isolate on low RPV level B. isolate on high system flow C. remain in service until T-227 is restored D. remain in service unless SBLC is initiated 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Basis or Justification B T-227-2 ONLY defeats RPV low level and SBlC initiation isolation. All other RWCU isolations, such as high flow, are still in effect.

A defeats RPV low level isolation.

ONLY defeats RPV low level and SBlC initiation isolation. All RWCU isolations, such as high flow, are still in effect.

ONLY defeats RPV low level and SBlC initiation isolation. All RWCU isolations, :such as high flow, are still in effect.

Psychometrics f- l~velof KnowlE:!~_+_ . . .__ DifficultY____ i_Time AIICI'.Nanc;ElJ.!:!1iElutel:i)_L___ 139__

HIGH . 10CFR55.41(b)(10)

Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0

~ IlT Exam Bank f--------.--+-~=----------------------.-------- --.

Reference(sL T__~7-?____________

learning Objective:

I-t: PLOT -50 12-4d KIA System: ! 204000 - ReactorW-ate;-C-le-a~-p-S-y-st-e-m--'-li;"portanC;:ROISRO---- . . -..

j 3.6/3.6 KIA Statement:

A3.03 - Ability to monitor automatic operations of the Reactor Water Cleanup System including:

J3Elsponse tClsyste~l~Cllation~~ .-_______ . _________. ___ . .__.....___ .... _._____.. _____.. . .__________ . . . ___~

_R_Eg_UI~§_D_M!'_T_E~I~~§:_______ ...J~--=N..::.O::::-...:N~E=-----. . - - -.... _____.__.____.________ . . . _____.. .___._______ 1 Notes and Comments:

2011 NRC RO Written Exam Rev. l.doe

Peach Bottom Initial Reactor Operator NRC Examination January 2011

36. A LOCA has occurred on Unit 2. RPV water level is now reading -150 inches on LR-I10A blue pen at the 20C004C RCIC Panel.

Based on the above conditions, which one of the following process parameters is providing this recorder level indication?

Level is sensed by the _ _ _ _ _ _ _ __

A. LT-72C, Wide Range level transmitter B. L T -73C, Fuel Zone level transmitter C. LT-112, Wide Range level transmitter D. LT -113, Fuel Zone level transmitter 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II Answer

36RO

~.----~---.-~-------~--~-----~-------~-----~- -------.--------~--

Choice Basis or Justification Correct: B I LR-110A Blue pen comes from either LT-72C or 73C. When reactor level

. (WR) lowers below minus (-) 100 inches, the recorder swaps from Wide Distractors: A Blue pen has swapped from Wide Range LT-72C to Fuel Zone LT-73C.

The green pen always reads Wide Range, not the blue pen.

C I LT -112 is the uncompensated Wide Range input to the HPCI Alternate

. Control Station.

D 13 is always the Fuel Zone input to the L\-113 indicator on the 'A' RHR Psychometrics

_l:~~Lof J$n~wlec:i~__L_____ j:>iffiCI:!I~___ ___ 1Time Allowance (min_l:ut_es_;_:),_+1____ .__R_.O.____.__. __ _

MEMORY I 2.0 i 3 i 10CFR55.41(b)(7)

Source Documentation Source: o New Exam Item ~ Previous NRC Exam: (PB 2002) o Modified Bank Item o Other Exam Bank: 0

~ ILT Exam Bank Reference(s): ST-O-098-01 N-2 ---

Learning PLOT-5002B-5a Objective:

i 216000 -

I KIA System: Nuclear Boiler Instrumentation I Importance: RO/SRO J 3.3/3.1 --_.-.

KIA Statement:

rONE .-----

c.A4.01 - Abilitl'..!2 manu~IIYi>~..!at~ andlorl!l..QnJ!ollnJt1..e~(:mtrol room: Recorders.

- ~

-~~;;~~~!lc:~:;~;LS: ------~-~

~.- - - - - ----- -----~-----~-~-- -- ~- ~ ~ ~~ ~ - - -- ---~ -,---~.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

37. Which one of the following is the reason ON -100 "Failure of a Jet Pump" directs a plant shutdown if a jet pump failure has been confirmed.

A. Invalid heat balance due to inaccurate Recirc flow measurement B. Invalid LOCA analysis due to potential for a displaced jet pump mixer C. Unknown effect on core power distribution due to Recirc loop flow mismatch D. Potential for violating thermal limits due to inaccurate Recirc flow measurement 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question # 3IRO Choice Basis or Justification Correct: This is the reason given in ON-100 Bases-operation with a displaced pump mixer is not part of'the licensing basis. Per Tech Spec 3.4 Oi::t~)I::::::;,

pump operability is an implicit assumption in the design basis loss of coolant Distractors: This is not the reason given in ON-100 Bases. Plausible since various mass flow rates are part of the heat balance equation (although recirc is not This is not the reason given in ON-100 Bases. Plausible since recirc flow mismatch is an ON-100 concern (requiring reference to Tech Spec which is also based on LOCA This is not the reason given in ON-100 Bases. Plausible since inaccurate measurement could negatively impact thermal limit calculations.

Psychometrics J::~vel of Knowledge MEMORY Source Documentation Source: ~ New Exam Item Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 3.4 Bases KIA System: - Recirculation System RO/SRO 3.9/4.0 KIA Statement:

G2.1.27 - Knowledge of s stem ur ose and/or function.

REQUIRED MATERIALS: NONE Notes and Comments: This question meets the KIA since it requires knowledge of the purpose/function of Recirculation System jet pumps from an accident

-~""~--~~-

arlalysis/plant~~l~typerspective~

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

38. Given the following:

Fuel loading is in progress on Unit 2

The first 3 fuel assemblies of a fuel cell are fully seated in the correct core locations and are in the correct orientation

The 4th fuel assembly loaded in this cell is fully seated into its correct core location, but is inadvertently oriented 180 degrees from its correct position Which one of the following could occur due to the misoriented fuel assembly?

A. Reduced core flow through the fuel assembly B. Unmonitored violations of core thermal limits C. Inaccurate calibration of LPRMs using the TIP System D. Improper installation of the Stearn Separator Assembly 20 II NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Basis or Justification Correct: UFSAR, section '14.5.3.5, "a misoriented fuel bundle during power

""Qr""tir,n potentially could lead to unmonitored violations of core thermal Distractors: A misoriented fuel assembly will not impact flow through a fuel assembly, which is primarily based on fuel assembly mechanical design, core orificing, I~i~-***~-~-~*rc~:f~;~:~~~s~f~E~~::s;;E~:e::~~~~~i~~;~!~~;:~~~Si~~

~ . ____ __ rpPlicant does not understand/recall System function and/or opeEation._

h ' gamma flux and will therefore not effect LPRM calibration. Plausible if TIP

D A misoriented fuel assembly will not impact installation of the steam L _ _.. .____ J~_____ ~:~~~~t~~:::L:~~~t:~~~~~:~rg~~Plica:~.does_~ot un~~rstan_~/rec~I~____J Psychometrics

~el of Knowled~_~_____.. DifficuIL_____r Time Allowance (Illlnutes)-+.-_.. ._. RQ_____ _

MEMORY * ' i 10CFR55.41(b)(2)

Source Documentation Source: o New Exam Item [gI Previous NRC Exam: (DR 1997)

[gI Modified Bank Item Other Exam Bank: 0

[gIILT Exam Bank .. _ - - - .~-

mRefer~nce(s): UFSAR,mChapte!.11____.~_. ___.. ._._...____ _ ~ ___._..____.. . ~ *

m ..m Learning NLSRO-0763-1 Objective:

i KIA System: ~34000 - Fuel Handling EquipmE~nt jmportance: RO/SRO

~ ... ... ~ .. ---~-- ... --'~ *....- - . - . - - - -... ~~-.-.... ---..

3.0/3.7

~---.--------~~-.-~--.----...---~.-~---"--~~ ..

KIA Statement:

K5.05 - Knowledge of the operational implications of the following concepts as they apply to Fuel Handling Equipment EueL orienta~on_.___.. . _.~.... ...~__..____~. ._______.. ._ . __ . ----~

REQUIRED

.... - - - -MATERIALS:

-~-~- - ---

... ... ... ---~,

~NE

.. ... .. .... -~-,,-

. .N

__otes and Comments:

~.. ~_m __._ _ _...___.. ._ . . __ . _~._._. ___.. . _._.. __ ~ . . __..__.. .____.. ~. ___

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

39. Unit 2 was initially operating at 100% power when the 2A Reactor Recirculation pump tripped. Current plant conditions are as follows:

APRM power level is oscillating 58% to 69%

OPRM Pretrip Condition alarms (21 t B-5) are being received repetitively Which one of the following actions is required?

A. Insert GP-9-2 Appendix 1 rods ONLY.

B. Manually scram the reactor and enter the T-100 "Scram".

C. Raise the speed of the operating Recirculation pump to suppress the APRM oscillations.

D. Insert ALL GP-9-2 rods, followed by NF-AB-720-1 approved sequence rods, as required to stop the power oscillations.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answelr Key Question # 39 RO -~-------------------------------------------

Choice Basis or Justification Correct: D OT-112 step 2.2 (Immediate Action) requires insertion of all GP-9-2 control rods on a Recirc pump tnip. Step 2.5 (also Immediate Action) directs monitoring for THI and inserting GP-9-2 rods followed by NF-AB-720-1 rods until THI no longer exists. At this stage of the transient, the operator cannot determine if GP-9-2 rods alone will be sufficient.

Distractors: A This action is required when a Recirc pump trips, but it is not the only action required in this case since APRM flux oscillations exceed 10% peak-to peak, indicating THI.

B This action is required when NO Recirc pumps are operating, or when APRM flux oscillations exceed 15% peak-to-peak.

--- ---+-------------------------------------------------------------

C Raising Recirc pump speed is one of the options for exiting Region 2 of the I P-F Map, but NOT a THI suppression action. J Psychometrics Level of Knowledge Difficulty - -

Time Allowance (minutes) RO --~-

HIGH 10CFR55.41 (b)(1 0)

Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 r8J Modified Bank Item D Other Exam Bank: 0 r8J ILT Exam Bank - ---- --

Reference( s): __ QI::112; GP-5-1; AR_Q_-211 B-5 _.----- -- _ _ _ _ _ _ 0. _._.--

Learning PLOT-PBIG-1540-1, -3, -4 Objective:

.-~ -

KIA System: 295001 - Partial or Complete Loss of ~ance: RO/SRO Forced Core Flow Circulation 2.5/3.3 -

__ 0 -

KIA Statement:

AK1.04 - Knowleqge of the operational implications of the following concepts as they apply to Partial or Complete Loss of Forced Core Flow Circulation: Limiting cycle oscillation. -------

REQUIRED MATERIALS: "---~

I NONE "--- ------ --------

Notes and Comments:

~-- ------- I --_. - -----------

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

40. Unit 2 is operating in Mode 4 when a sustained loss of shutdown cooling occurs.

Which of the following is the reason ON-125 "Loss of Shutdown Cooling" directs raising RPV level to above +50 inches?

A. Provides sufficient NPSH for placing a Recirculation pump in service.

B. Satisfies Technical Specification requirements for reactor coolant circulation.

C. Promotes natural circulation and helps prevent stagnation of coolant in the core.

D. Establishes a longer "time to boil" while aligning alternate decay heat removal systems.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer QU6$tion # 40 RO Choice Basis or Justification Per ON-125 Bases, raising level to above the separators (> +50 inches) promotes natural circulation, which will prevent stagnation (thermal of reactor coolant.

Distractors: Not the basis per ON-125. In addition, +50 inches is more than is required for recirculation pump NPSH.

B Per ON-125, raising RPV level to +50 inches does NOT satisfy Tech Spec requirements for reactor coolant circulation. This can only be satisfied by

_restoring.. fQ.rc~d_ circulation (i.e:'m~mRecirc or RttB:..PJJ..'!!Q_ must be in service.

D Not the basis per ON-12ti. In addition, adding inventory has only a minor impact on time to boil, which is primarily based on decay heat load.

I Psychometrics r-~ Level of Knowledge I DifficultL I Time..f\llowance (minute~l RO MEMORY 2.5 ! 3 10CFR55.41 (b)( 10)

Source Documentation Source: o New Exam Item [gj Previous NRC Exam: (PB 2002)

[gj Modified Bank Item Other Exam Bank: 0

[gj ILT Exam :.. Bank

~~~~~. _____ __.__

~ ~_~~~_m_~ ___ ~

Reference( s): ON-125 and Bases Learning PLOT-1550-28c Objective:

KIA System: 295021 - Loss of Shutdown Cooling

Importance: RO 1 SRO I * * *__~~~ ______I~ ___ ~~m __ ~~_~_ _~~_ _~~~~' _ _~"""""""

! __ ~'"

3.3/3.4

- - - - ---~~--. . . . --~-

KIA Statement:

AK1.02 - Knowledge of the operational implications of the following concepts as they apply to Loss of Shutdown Cooling: Thermal stratification.

REQUIRED MATERIALS;-~!NONE

~~~~d Comments: m _ _ _ mmmmlm ____ m

~ ______.___~ __

_ _ ***** _ _ _- __ ~_ .~_- _

.mm_

- - - - ===-_ _._ _ _ . _ ~ m **********

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

41. Unit 2 is operating with the following conditions:

Main Generator volts: 22.0 KV

Main Generator MW: 1100 MWe

Main Generator VARS: 120 MV ARS

Hydrogen pressure: 60 psig A grid disturbance results in steadily lowering grid voltage.

The Main Generator voltage regulator responds as designed by attempting to raise Main Generator terminal voltage.

With no operator action, this transient could result in ------------------

Figure 1 of AO 50.7-2 "Generator Capability Curve" is PROVIDED SEPARATELY.

A. overheating the Main Generator rotor windings B. overheating the Main Generator stator windings C. exceeding the Generator Under Excitation Limit D. Generator Lockout due to reverse power relay trip 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Choice Basis or Justification Correct: B The given conditions (lowering grid voltage) will cause the generator automatic voltage regulator to attempt to raise grid voltage, causing the generator to pick up additional VARS (Le., move up on the Generator Capability Curve). Without operator action, this would result in exceeding i the generator capability curve (B-C) for 60 psig hydrogen pressure. Per !

i _ _~ ...~._~_~__ ~ ...--+..:.F....:,igOLlu:::.:.r.=.e~1....:0::.:..:fAO 50}-2, curve B-C i~f!1ited by arm_~ure (st~!or)heatLnji_~_i Distractors: A This would be true if curve A-B was the limiting factor. Plausible if one was confused between field, c;)rmature, rotor and stator.

C This would be true If grid voltage was rising, resulting In lowering VARS on the main generator (Le., move down on the Generator Capability Curve).

- - - - r* D---IA reverse power trip occ~~~ when real load (MW)i;~~duced to the p~int~l

where the grid supplies the generator. The given conditions would not i

___.~ __j=~~~I~;;~~~dM~'r:~F_~~'::~~~:'~~:r~~eb:;::'~~;~':~~~~Ie ~ j Psychometrics f-J::evel()f KnOwledg~~l_ _ Diffic;!Jlt~ ._Tif!1~.~!lowanc~(rT1inutes) ~ .. --~.--.-

RO ---~ --~---

HIGH I 10CFR55.41 (b)(5)

Source Documentation Source: D New Exam Item [g] Previous NRC Exam: (PB 2008)

[g] Modified Bank Item Other Exam Bank: 0 ILT Exam Bank I-~--~--~+

AO 50.7-2 Learning PLOT-5050-6f Objective:

KIA System: UUl'Ull - Generator Voltage and Electric Importance: RO / SRO Disturbances 3.3/3.4 KIA Statement:

AK1.02 - Knowledge of the operational implications of the following concept as it applies to Generator and Electric Grid Disturbances: ... Over-excitation.

I*~='::::~'::':'::'~:=**:**=**'::'::':':::':=:C:_*=--===:'::':::'~=:C::=.= =.':''::'~C .....=..:.. :=:,=.=.~.-- ....- ~-- .............. --.~-~.~ .......- .... - ...--.-.- ..- - -... --.

of AO 50.7-2 Notes and Comments:

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

42. Refueling movements are in progress on Unit 2.

Which one of the following unanticipated conditions is symptomatic of a refueling event requiring action in accordance with ON-124 "Fuel Floor and Fuel Handling Problems"?

A. FUEL STORAGE POOL HI LEVEL alarms on local Pane120C075.

B. 2 UNIT REFUELING FLOOR AREA HI RADIATION (003 B-4) alarms.

C. An irradiated LPRM detector is dropped in the ISFSI Cask Handling Area.

D. REFUELING FLOOR VENT EXHAUST HI RADIATION (218 A-I) alarms.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer

42RO Choice Basis or Justification Correct: B ON-124 requires entry and action for any Fuel Floor ARM alarm.

Distractors: A alarm does not require entry into ON-124. Note that receiving a "Fuel

""Tr.r~r";" Pool Hi Radiation" alarm, or indications associated with a loss of Fuel Pool water inventory (e.g., lowering level), do require entry into i~~~-~~--~--~~~~--- -+ 24.

--~. -~~-~.+-~~~~~~~~~~~-.--~---~.~~-.. ~~~~--,-~~-~-.-~~,-.-~~~-,-~~~-- .. -~~,---~~--~ -~~- -~~~--. ~~-~~-~---i C 24 entry is required for a fuel assembly or single fuel rod dropped or n::unl:.\"""n but not for an LPRM detector.

D Although this condition obviously requires action, it is not an entry condition into ON-124.

PSy'chometrics Source Documentation Source: o New Exam Item [gj Previous NRC Exam: (PB 2005)

[8J Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank 24 KIA System: - Refueling Accidents Importance: RO 1 SRO 3.4/3.6 KIA Statement:

AK2.03 - Knowledge of the interrelations between Refueling Accidents and the following: Radiation Monitoring_Equ~ment. __ .__ ______ __ _____ ~__________ ~ __.~~~_~__ .___~~_~~_~.~_

REQUIREQJIIIATERIAL~_ -INONE _ _ _ _ _ _ _ _ _ _ _ ~ ~_~~___ ._____ ~~~_

N~tes ,and yomrr'lents: ~ __ _ ___'~_______ ~ ________ ~ ___________~~__ ~ __~~__ ~,~

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

43. Unit 2 is operating at 100% power when a Turbine Trip occurs. Two minutes later the following conditions are observed:

Reactor power is 2E-02%

Rod 22-51 is at position "02"

Rod 50-31 is at position "48"

All other control rods are at "00" Which one of the following statements is correct for these conditions?

The reactor shutdown; an A TWS _ _(2)_ _ in progress.

A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

~ ... ~-.- ..

Answelr Key Question # 43 RO ................ .....

~--

Choice Basis or Justification Correct: A Per T-101 NOTE 23, the reactor is shutdown if it is know to be subcritical with reactor power below 1%. Per T-101 NOTE 24, ATWS criteria stipulate that if any single rod is withdrawn (at '48' in this case), all other rods must be fully inserted. Having one rod at '48' and another rod at '02' does not meet these criteria.

Distractors: B An ATWS is in progress.

C The reactor is shutdown.

D The reactor is shutdown; an ATWS is in progress.

Psychometrics Level of Knowledge ._- ~- .......

Difficulty' I Til"lle Allowance I

(minulf!~LLmm RO MEMORY 3 10CFR55.41 (b)(2)

Source Documentation Source: lZJ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 D ILT Exam Bank Referenc;e(s): T -101 (Notes 23 and 24) ~~-~~~~-,-, ..-. ~.-- ,.---"~-~------ ......

Learning PLOT-PBIG-2100-T101-4, -6 Objective:

~- .... ~-~~

KJA System: 295006 - Scram I Importance: RO/SRO

~-

4.2/4.3 -- - ----

KJA Statement:

AK2.06 - Knowledge of ttl~irlterrelationshJR betvl.t~~n Scram and the foIIQ!,!ing:J3eactor power.

REQUIRED MATERIALS: ___ NONE

~otes and COrTlmen~!:):_---=mm 1 _____ ~_.....-.._.~

- - - - _ _ _ _* _ _.W""* -~

~

---.--.--~- ...... m~

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

44. Unit 2 was operating at 100% power when a large break occurred in the Torus.

As Torus level lowers, which one of the following describes when the LOCA downcomer vents and the SRV tailpipes will become uncovered?

LOCA Downcomer Vents SRV Tailpipes A. 12.5 feet 7 feet B. 12.5 feet 6 feet C. 10.5 feet 7 feet D. 10.5 feet 6 feet 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 I

Answer- Key.. ~- ..- - - . - . -.. --~-.--~.-------~.---~ ...- -

i Ot ,,#44RO I Choice Basis or Justification Correct: C i Per T-102 Bases, the LOCA downcomer vents are uncovered at 10.5 feet.

I Per T -112 Bases, the SRV tailpipes are uncovered at 7 feet.

! I .

Distractors: I A 12.5 feet is when T-102 requires a reactor scram and RPV depressurization due to low torus level.

i I

B feet is when T -102 requires a reactor scram and RPV depressurization I

to low torus level.

feet is the RCIC vortex limit.

Psychometrics 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): I T-102 Bases; T-112

---'--'-'-'-'I.~----'--'---

Bases

--- ---~--- ---- - - - -- ~--- -- ~-- ~~-- -

Learning . PLOT-1560-9 Objective:

KIA System: 295030 - Low Suppression Pool Water RO/SRO Level 3.5/3.8 KIA Statement:

EK2.08 - Knowledge of the interrelationships between Low Suppression Pool Water Level and the i~:~~£:~;:~~~~Ubmr:~:~_-_-==_--=_~~==~= ____

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

45. The following conditions exist on Unit 3:

ATWS

Group I isolation

Reactor power is 40%

Torus Cooling is NOT available Which one of the following limits will be challenged first by these conditions?

A. Drywell Spray Initiation Limit B. Heat Capacity Temperature Limit C. Pressure Suppression Pressure Limit D. Primary Containment Pressure Limit 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Choice Basis or Justification Correct: B The given conditions indicate SRV discharge into the Torus. Without torus cooling, HCTL will be challenged first.

Distractors: A is not an initial concern because there are no given conditions of Containment high pressure or temperature.

B is not an initial concern since there are no given conditions that i indicate the Primary Containment is not functioning properly.

limit is not an initial concern because there is no given condition of Containment high pressure.

Psychometrics

~ev~~Q~nowl~~~_, .._ . ___ Di(fic;ulty....__J_.Iimet.llow~rlce LI111nut~~LL. ___ .B.Q.. ____

HIGH 10CFR55.41 (b)(1 0)

Source Documentation Source: I o New Exam Item r8J Previous NRC Exam: (PB 2009) o Modified Bank Item Other Exam Bank: 0 r8J ILT Exam Bank --_.

o .* &.

"CICI Cllv,,?!

'(5) TRIP Bases .

Learning PLOT-21 02-6 Objective:

KIA System: i 295026 - Suppression Pool High Water II I 'I"'U' LCU IvC. RO/SRO Temperature 3.9/4.0 --

KIA Statement:

EK3.02 - Knowledge of the reasons for the following responses as they apply to Suppression Pool J::Iigh Water TemperatlJ~ SUPPTSion !"lol cooling. ____.______.. __.._.. _ _..___..__..___.____

REQUIRED f-------..- ..-

MATERIALS:

NONE ..._ - . -...- - - ..- - - . -...- . - -... - - -...- - . _..- ...... - - . - -.....- - -........----...-.

Notes and

- ".. ---~-- ... ~~--

Comments:

... ~~-- .. -~.~-~- ..- - - - - . . . - ... ---~-.- .. ~----~ ..- - - - - -...-.-----...- - - -.. ------~--.--~

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II

46. Given the following:

A startup is in progress on Unit 3 with reactor power at 5%

Panel 30Y033 is inadvertently de-em-:rgized, resulting in a loss of power to portions of PCIS logic Which of the following RWCU System containment isolation valves close as a result of this event?

1. MO-3-12-15, Cleanup Inlet Isolation-Inboard

2. MO-3-12-18, Cleanup Inlet Isolation-Outboard

3. MO-3-12-68, Cleanup Outlet Isolation A.IONLY B. 20NLY C. 2 and 3 ONLY D. 1,2, and 3 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Key Question #46 RO Choice Basis or Justification Correct: D of Panel30Y033 causes a loss of power to PCIS inboard isolation valve logic. This results iin closure of associated inboard containment isolation valves, includin~J RWCU valve MO-3-12-15. Loss of 30Y033 also results in closure of RWCU outboard containment isolation valves MO-3 12-18 and MO-3-12-68. This is due to loss of power to the NRHX high outlet temperature relay, which feeds both the inboard and outboard isolation valve logic. Note #2 in GP-8.C describes the RWCU response to a loss of 30Y033.

Distractors: A MO-3-12-18 and MO-3-12-68 also close on a loss of 30Y33.

B MO-3-12-15 and MO-3-12-68 also close on a loss of 30Y33.

C MO-3-12-15 also closes on a loss of 30Y33 .

Psychometrics bt?yel of Knowlt?99...e I Difficulty ~Iime AIIQ~ance (minutes) RO

~--~~-~- .......

HIGH 10CFR55.41 (b)(7)

Source Documentation Source: D New Exam Item [8J Previous NRC Exam: (PB 2007)

[8J Modified Bank Item D Other Exam Bank: 0 ILT Exam Bank Referenct?(~~ AO 58A.2-3 Learning PLOT-5012-6F Objective:

KIA System: 295003 - Partial or Complete Loss RO/SRO Power 3.7/3.7 KIA Statement:

AK3.06 - Knowledge of the reasons for the following responses as they apply to Partial or Complete Loss of A.C. Power: Containment isolation.

RED MATERIALS: l.t!9NE N()it?~and Comments:---~=_~~~J . . _._ _______ __..______~___..................

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

47. According to the UFSAR, which one of the following statements describes the reason for disabling control room controls lAW SE-IO "Plant Shutdown from the Alternative Shutdown Panels" after abandoning the eontrol room?

A. To prevent High Pressure Coolant Injection (HPCI) System automatic operation.

B. To ensure interlocks associated with operation of safe shutdown equipment are defeated.

C. To prevent simultaneous operation from the control room and the Alternative Shutdown Panels.

D. To ensure fire-induced circuit faults will NOT prevent operation, or cause spurious operation, of safe shutdown equipment.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answelr Key Question # 47 RO - - - - - - - - - - - - - - - - _. . . __. _ . _ - - - - - - - _ ._ _ ..

Choice ._.

Basis or Justification Correct: D From the UFSAR: "transfer/isolation switches provide electric circuit isolation between alternative shutdown circuits and circuits that could be affected by a fire in one of the four areas of concern (control room, cable spreading room, computer room, emergency shutdown panel area)." From 10CFR50, Appendix R: "The safe shutdown equipment and systems for each fire area shall be known to be isolated from associated non-safety circuits in the fire area so that hot shorts, open circuits, or shorts to ground in the associated circuits will not prevent operation of the safe shutdown

.. _- r--- equipment." - - - - - - - - - - - - - - - - - - - - - _ ..

Distractors: A HPCI system automatic system operation (trips, isolations and automatic start) is defeated when control is transferred to the Alternative Shutdown Panel, but this is not the design basis reason for disabling Control Room controls .

- .. _ . _ - _ . - - + - - - _ . _ - - - - - - - _ .._._._ .. _ ...__......._ ... _.__._._ .. _._....._._---_.._ - - -

B Interlocks are defeated when operation is transferred to the Alternative Shutdown Panels, but this is not the design basis reason for disabling Control Room controls.

C This could be (and is) accomplished procedurally; it is not the design basis reason for disabling Control Room controls.

Psychometrics Level of Knowledge Difficulty Time Allowance (minutes) RO ~----

MEMORY 1OCRF55.41 (b)(7)

Source Documentation Source: D New Exam Item rgJ Previous NRC Exam: (PB 2008)

D Modified Bank Item D Other Exam Bank: 0 f---------+--rgJ ILT Exam Bank __ __ ..__________...._. __

f-R_e_f_er_e_n_ce-,(-,s),-:_--+_SE-10; UFSAR FPP, Ch.~__. _ _ ._____._...._....._._____._ ....__._

Learning PLOT-1555-9 Objective:

f - - - - - - - - . - I - - . - - . - - - - - - - - - - - - - - - . - - - - - - , - - - - - - - - . - - - ..- - . - - - - - - .

KIA System:

600000 - Plant Fire on Site I Importa.n__c_e_:_R_O...::....../..::.S._R_O_____._.

f--------...J....--.--------- --------.-.

I 2.8/3.4 KIA Statement:

AK3.04 - Knowledge of the reasons for the following responses as they apply to Plant Fire on Site:

Actions contained in the abnormal procedure for plant fire on site. .. _ _ _ ..____ ____ _

f-REQUIRED MATERIALS: I NONE . ______. ..____.... __

. Notes ~nd Comments: . I ... _____.. ____._ _ _. ____.._ ...__.

2011 NRC RO Written Exam Rev. 1.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

48. A fire occurred in the Main Control Room requiring evacuation. The following conditions exist on Unit 2:

The crew is executing SE-l 0 "Plant Shutdown from the Alternative Shutdown Panels"

The URO is performing SE-IO Sheet 2 to establish control at the HPCI Alternative Shutdown Panel After placing the transfer isolation switch for the HPCI Aux Oil Pump to EMERGENCY, the URO observes the white power supply indicating light for the HPCI Aux Oil Pump is lit.

Which statement below is correct for these conditions?

A portion of SE-I 0 Sheet 2 is PROVIDED ON THE NEXT PAGE.

The white power supply indicating light shows that the source of DC power is available to the HPCI Aux Oil Pump.

A. (1) normal (2) control B. (I) normal (2) operating

c. (1) alternate (2) control D. (1) alternate (2) operating 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

SE-10 ASD

>--~

SHEET.

ASD/Ii-4 I

t OPEN THE ASO KEYBOX AND USE THE KEYS AS REQUIRED L. ASD/R-S I BLISH COMMUNICATIONS WITH THE MCR 1IIII<_1IIII<.1IIi _ _ _ _ _ . _ . . . . .

III F REACTOR SCRAM OR MSIV CLOSURE WAS III NOT VERIFIED PRIOR TO CONTROL ROOM III ffiCUATION. I III THEN OIRECT AN OPERATOR TO DE NERGIZE III

- - RPS lOGI C USI NG: I I

o SE-10 ATTACHMENT 14 - UNIT 2 I o SE-10 ATTACHMENT 15 UNIT 3 II L'-::::.::-::.:::-:.:::;"=.':!:_:":::_::-::_:::-:_:::"'_=_~IIII-=_=_:-::_:::-:.=,*=.-=IIiIt=IIiIt=_::"':_=-.=.-=.:-::.:-:.:_:::-:_=-_=_~."'!::.=.::-::+/-.III L. ASOfR-7 IF MORE THAN ONE HPSW PUMP PER UNI TIS RUNNI NG, THEN DIRECT AN OPERATOR TO MONITOR HPSW PUMP ROOM TEMPERATURE LOCALLY USING SE-l0 ATTACHMENT is L. ASOfR-6 I OPEN THE HPCI ASO PANEL AND PLACE ALL TRA NSFER ISOLA TI ON sin TCHES I N EMERGENCY L. A THE WHITE POWER SUPPLY INDICATING LIGHTS ARE NOT LIT.

THEN CLOSE ALTERNATIVE AC AND DC POWER

- SUPPl Y BREAKERS

_ . _ . _____ *

_ *

_ - _ . . . . . . . . . _ - - _ . . . . . . .1 L. ASOIR-lO

..1.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answel' Basis or Justification Per SE-10 Bases Step ASD/R-10; normal (DC) control power is not available if the white power supply is not lit, and closing the alternate (DC)

. supply breaker establishes a separate and independent source of control

' power for the labeled component. As long as a (DC) power supply is present as indicated bY.. the white. l.i9ht, the .S..Witch is n.o.t operated to prevent

]

. .__ . ~ ~ross-tying n9rmal andalternate(DC) pow~r supplie~.~~. .._~~ .....___

Distractors: B The white light provides indication of control power status; not operating

. _.._~~~ .~_I ~~;n:~~~s~~~~t~n'::'~=~\~~:~~r~e~~!~~f~e_riso_l~atio~_swi_tc_he~_alig~_ ~

C The white light indicates that normal control power is available and there is no need to close the alternate power supply breaker. Plausible if applicant

_ ..... _ .... _ - . . ~._____--+_~ges notu.Ilgerstand ~ha!l~ bei l19 accomplished by these steps of§E-19..:....__

D ! The white light indicates that normal control power is available and there is no need to close the alternate power supply breaker. In addition, the white light provides indication of control power status; not operating power.

Plausible if applicant believes the transfer isolation switches align component operating power vice control power, or does not understand

~ ___.. . . ____ . ___ ~ ___ Lwh~t is beiniLacc()!!lpli~t1.edl:l.YJ!:l~se s~eps oJ§E:-10. --.J Psychometrics Source Documentation Source: [gI New Exam Item D Previous NRC Exam: 0 D Other Exam Bank: 0 ILT Exam Bank KIA System: 6 - Control Room Abandonment RO/SRO 2.8/2.9 KIA Statement:

AA1.05 - Ability to operate and/or monitor the following as it applies to Control Room Abandonment:

D.C. electrical distribution. _~ ~ __ . _~ __.. . . _ _.. . _ _.. . _ _- _.. . . _

REQUIRED MATERIAi:"S:.- I NQ.NE__ .. -

Notes and Commen!~~_._ _~_ ~ ~ ___~.__ ...~. __________________ . . . _~~~_..____

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II

49. Given the following:

A radiological accident condition has occurred at Peach Bottom

2 VENT EXH STACK RAD MONITOR HI/TROUBLE A (218 B-5) alarms

2 VENT EXH STACK RAD MONITOR HI/TROUBLE B (218 C-5) alarms

Unit 2 Vent Stack Radiation is reading 2 x 10-5 IlCi/cc and rising Which one of the following could be the source of the radiation release?

A. Standby Gas Treatment Exhaust B. PEARL Building Ventilation Exhaust C. Radwaste Building Ventilation Exhaust D. Recombiner Building Ventilation Exhaust 20 II NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answel" Basis or Justification Correct: Kaawasle Building Ventilation exhausts to the Unit 2 vent stack.

to the main stack.

Buildin~1 exhausts to the Unit 3 vent stack.

Building exhausts to the Unit 3 vent stack.

Psychometrics Level of Knowle~ I Diffic~~ Time Allowance (minutes) RO - ".

MEMORY I 10CFR55.41 (b )(9)

Source Documentation Source: New Exam Item D Previous NRC Exam: 0

[8J Modified Bank Item o Other Exam Bank: 0

-- ,_ [8J ILT Exam Bank --

Reference( s): ON-104 and Bases --

Learning PLOT-1550-9a Objective:

KIA System: 295038 - High Offsite Release Rate RO/SRO 3.9/4.2 KIA Statement:

EA1.01 - Ability to operate and/or monitor the following as it applies to High Offsite Release Rate:

~~~~~::;~~;~~~~~-"'JN~-

,-- ..- -_.- ,-~

,~~ ,.,-- .,,-~

Notes..............................................................

and Comments: ---~ ~~~. - ------~~-------

20 II NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

50. During an ATWS condition, the URO initiated the 'B' Standby Liquid Control (SBLC) System using RRC 11.1-2 "SBLC System Initiation During a Plant Event".

The following conditions exist:

RPV pressure is 1020 psig

SBLC discharge pressure is 1400 psig Which statement below is correct for these conditions?

A. SBLC is injecting at full f1ow.

B. SBLC is injecting at reduced flow.

C. SBLC is NOT injecting; initiate System 'A'.

D. SBLC is NOT injecting; do NOT initiate System' A'.

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Basis or Justification Correct: C Based on the given conditions (1400 psig pump discharge pressure),

is not injecting. Per RRC 11.1-2 and the supporting system operating I* procedure if (SO 11.1.B-2), the operator is directed to verify SBLC is injecting to start the other SBLC A SBLC is not injecting. Plausible if applicant does not recognize 1400 psig pump discharge pressure as abnormal.

B SBLC is not injecting. Plausible if applicant recognizes 1400 psig pump discharge pressure as abnormal, but does not understand SBLC system believes the reduced flow.

SBLC is not injecting but the 'A' SBLC system should be started to comply with the procedure. Plausible if applicant is not familiar with procedure direction and/or believes whatever failure exists will also prevent System 'N from Psychometrics Level of Kn9wledgEl_J___~m __l?ifficuliY m-t_J"irl'le AlloVllance (f'ninutes) RO HIGH 2.75

3 10CFR55.4'1 (b)(6)

Source Documentation Source: D New Exam Item IZI Previous NRC Exam: 0 IZI Modified Bank Item D Other Exam Bank: 0

. ..- -... -~-~----

IZIILT Exam ..

Bank...

-.~ .... ~.- ....

. .____".. ~_~~_~_ . . __ ~_. ___.. ._ _m -.~ .

Reference(

r---.--- . . s):

~---....

RRC 11.1-2;

. - . ---.-. . . SO 11.1.

-~- .. B-2 . . - - -.. -~.---~.--~.--.-~ .... ----~ .... -~~-- .

Learning ! PLOT-5011-4d, -5c Objective:

KIA System: I 295037 - Scram Condition Present and i Importance: RO/SRO Reactor Power Above APRM Downscale or 4.5/4.5

Unknown i KIA Statement:

EA1.04 - Ability to operate and/or monitor the following as it applies to Scram Condition Present and Reactor Power Above APRM Downscale or Unknown: SBLC.

REQUIRED MATERIALS: NONE Notes and Comments: _

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

51. Unit 2 was operating at 100% power when a Loss of Coolant Accident occurred. The following conditions currently exist:

Drywell pressure is 6 psig

Drywell sprays are not available

Drywell bulk average temperature is 275 degrees F

Drywell coolers are being supplied by Drywell Chilled Water (DWCW)

DWCW return header pressure is 28 psig

The Reactor and Radwaste Buildings are not accessible

T-223-2 "Drywell Cooler Fan Bypass" is being implemented T-223-2 Figure 1 is PROVIDED ON THE NEXT PAGE.

Based on these conditions, the Drywell Cooler Fans _ _ _ _ _ _ _ __

A. may be restarted in "Slow" speed ONLY B. may be restarted in "Slow" or "Fast" speed C. cannot be restarted until Reactor Building access is restored D. cannot be restarted until an Engineering evaluation is obtained 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 T-223-2 Rev. 6 Page 6 of (5 FIGURE 1 DRYWELL CHILLED WATER (DWCW) SATURATION CURVE 350 I ! I

-c UNSAFE I,!

325

~

~~

300 ....

C

~

'£.

~

/

V 275 SAFE

/

,~

.r

~

> 250

/

':.L V

C!l Z2S /

V

'* II 200 o 10 20 30 4 a 50 60 70 60 DWCW RETURN HEADER PRESSlJRE ON PI-20262 (PSIG)

IF TI 80146 is out of service, THEN use RT-O-40C-530-2 to determine DW Bulk Average Temperature.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Que$lon f #51 rtU

............- -...... .......... ~-~.- ....... ------~-

Choice Basis or Justification Correct: D Drywell cooler fan units cannot be restarted since operation of the coolers plots on the UNSAFE side of the Drywell Chilled Water Saturation Curve (T-223 Figure 1). Per step 4.1 of T -223, operation must be verified to be on the safe side of the curve, or an engineering evaluation must be obtained, priorJ9 starting~restartl!!9U1Jly drywe~oole!Jan unit. ._--_.............._-

Distractors: Plausible since T-223 states the drywell cooler fans should be started in SLOW speed if drywell pressure is above 0.75 psig.

B Plausible since T-223 states the drywell cooler fans should be started in SLOW speed If drywell pressure IS above 0.75 pSlg, but allows starting the fans in FAST speed if the local fan speed control switches are not accessible is the caSE~

C Plausible since the local fan speed control switches are not accessible.

I Psychometrics

~y~1 of Knowledge ----~-- ...

Difficulty -_. __

Time Allowance (minutes)

RO ..............

HIGH 10CFR55.41 (b)(1 0)

Source Documentation Source: D New Exam Item D Previous NRC Exam: 0 rgJ Modified Bank Item D Other Exam Bank: 0 rgJ ILT Exam Bank

~~E~cccnc-=-,e=->(~s)L-:_----+_T"----.=2=23=_ ______._ ..____ .__.. _.~ __..- ..------~----.----- . . . . . ~--------l Learning I PLOT-1560-4 Objective:

_ ..........._ .... _ - _ . - - - - ........... --~.--.-~--~.-.-.-.--.-

KIA System: I 295028 - High Drywell Temperature ! Importance: RO / SRO 4.0/4.1 KIA Statement:

EA2.01 - Ability to determine and/or interpret the following as it applies to High Drywell Temperature:

Drywell temperature. ~~ _...__ ._______ _~ ______ _

~~~~~~~~::~:~~LS:+/-:'~E ______ ==-_ ____==-_---=-====~~

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

52. The following conditions exist on Unit 3:

The crew is executing T -111 "Level Restoration"

RPV level is -190 inches and slowly lowering

RPV pressure is 550 psig and slowly lowering

An emergency depressurization is in progress with 5 SRVs open

The 3A Core Spray pump is the only available source of injection Based on these conditions, Adequate Core Cooling (ACC) is _ _ _ _ _ __

A. NOT being maintained B. being maintained by submergence C. being maintained by spray cooling D. being maintained by steam cooling 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer

52RO Choice Basis or Justification Correct: RPV water level cannot be maintained above TAF (-172 inches).

steam flow (co()ling) is established by maintaining RPV water above the Minimum Steam Cooling RPV Water Level (-195 inches),

long as RPV pressure is above the Minimum Steam Cooling Pressure ACC is being maintained by steam cooling. Plausible if applicant does not recognize steam cooling conditions are met.

The core is adequately cooled by submergence when it can be l'1ot.",rrn that RPV level is at or above TAF (-172 inches). Plausible if applicant associates TAF with -195 inches.

Adequate spray cooling is provided when design spray flow requirements are satisfied (at least 6250 gpm from one Core Spray loop) and RPV water

' -__*_ _ _ _ _ _ _ _ _ _ _._..l __. _____......L-'..:ley~l_is at or abovetb~~y~tion of the jet pyl!l2...~y~tions (-226 incb~~"- ____ .

Psychometrics r--Level of Knowlecig~_ - - -

Difficulty_ ITirn~ Allow<:ll1 ce (minutes) RO -----

I HIGH I 10CFR55.41 (b)(1 0)

Source Documentation Source: New Exam Item D Previous NRC Exam: 0

~ Modified Bank Item ~ Other Exam Bank: (LORT)

I D ILT Exam Bank ... _- ._--

_Reference(s):-'-l T-BAS Qrltro): TRIP/SAMP Curves, Tables & Limits B.ases: T-111 and B-",ses ____

I Learning PLOT-5014-3a Objective:

I KIA System: ' 295031 - Reactor Low Water Level-- =rportance: - RO ISRO 4.6/4.8

--~~- -

-- ~- ~ -- -- -------

KIA Statement:

EA2.04 - Ability to determine and/or interpret the following as it applies to Reactor Low Water Level:

Adequate core cooling. . . . . . . ___ ~----

~UI~ED MA~IALS: _ I NON~,___ --

Notes and Comments:............_ - - --~--- .. ~ . ..- ..- - .. .... _ - - - -

2011 NRC RO Written Exam Rev. J .doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

53. Unit 2 is operating at 100% power when an EHC malfunction results in the following events:

Turbine control valves swing partially closed then back open

REACTOR HI PRESS (210 G-2) a1arm is received

Reactor power initially rises then returns to the pre-transient level

Reactor pressure peaks at ~ 1065 psig then returns to the pre-transient level Which one of the following actions is required by OT-102 "Reactor High Pressure" for these conditions?

A Perform GP-4 "Manual Reactor Scram".

B. Place the Mode Switch in SHUTDO\VN.

C. Perform GP-9-2 "Fast Reactor Power Reduction".

D. Maintain reactor pressure 035 psig with EHC Pressure Set.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Key -1

-_Question

.....- 53 RO

...- - - -... -~

Choice '--~~ ~.- ....- - -....- - -... Basis or Justification .~ .....- - - - - - -....-

Correct: D The given conditions indicate reactor pressure exceeded the RPV Hi Press alarm setpoint (1053 psig), but not the RPS scram setpoint (1085 psig).

Since RPV power and pressure returned to pre-transient values, the action required by OT-102 for these conditions is to maintain RPV pressure less

.Jhan or equi:11 to 1035psig uSil"lg EHC p~essure s~:... _ _...._ -~--- ..

Distractors: A GP-4 prerequisite is "Plant conditions require a manual scram and sufficient

time is available to perform pre-scram actions." None of the given plant

_cor)ditio_ns r~.9!!ire a r(9a~or scram. --_ ..

I B This choice is based on the applicant believing there was an RPS failure, in which case the correct action would be to initiate a manual scram using the Mode Switch. Plausible if the applicant does not recall the RPS scram setpoil"lt of 1085'p~~

C This is the immediate operator action of OT-102 "Reactor High Pressure" !E reactor pressure continues to rise.

~ ....

Psychometrics r----Levelof Knowledge_1_____. Difficulty. ,. Time Allow?lnce (minut~~ I RO~_

HIGH* . : 10CFR55.41(b)(10)

Source Documentation Source: 0 New Exam Item ~ Previous NRC Exam: (PB 2009) 1

____ __ __ __!8l_'1-T Exam Bank

~ Modified Bank Item

ARC-210 PLOT-5060F- 'I b OT-102 D Other Exam Bank: 0

_~_~_~~~_ . ___...~. _______.. __ ~_.~. . ___~__..__~ __

KIA System: .... "',"'Vi&..... - High Reactor Pressure RO/SRO 4.3/4.3 KIA Statement:

EA2.01 - Ability to determine and/or interpret the following as it applies to High Reactor Pressure:

Rea~to.r. pres~ure'~-_---=F'" _._ . ._.-.-.. . . _-.~.~... __. ....- ......- ._~.'...-.~. . _ . . .

~EQUIRI::D MATERIALS:_ _ ~ONE __ . . _~_~_~ ____.__ ~ __ ~ __.. . .~ _ _.. . ___ ~~__

Notes and Comment~___ _ _~____ __ ~ _ . __~_ . ___.. .__~_ ..._____ _

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

54. Which one of the following correctly shows the impact ofa loss of power to 125 VDC Power Distribution Panels 20D021 (Unit 2) or 30D021 (Unit 3)?

Loss of power to Panel )__ results in the following equipment being unavailable:

A. 20D021 (Div I) Unit 2 HPCI 2A Core Spray Pump 3A Core Spray Pump B. 20D021 (Div I) 1 Diesel Unit 2 RCIC 2A Core Spray Pump 3A Core Spray Pump C. 30D021 (Div I) Unit 3 HPCI 2A Core Spray Pump 3A Core Spray Pump D. 30D021 (Div I) E-l Diesel Unit 3 RCIC 2A Core Spray Pump 3A Core Spray Pump 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer 54RO Basis or Justification Correct: Per SE-13, a loss of 20D021 results in a loss (inoperability) of Unit 2 RCIC, i the E-1 diesel the 2A and 3A Core Spray pumps (and other loads).

Distractors: Unit 2 HPCI is powered from Division II 125 VDC Pane120D022. The E-1 diesel is also inoperable on a loss of Panel 20D021. Plausible since there is cross-tying of DC power supplies between units for ECCS related

. . ____ equipm~.D!~ ___._._.. . ._._.__.__.. . .__.____.. . .___ ...______.. ._____._.. . __

C Unit 3 HPCI is powered from Division 11125 VDC Pane130D022. The 2A Core Spray pump is not impacted by a loss of 30D021, although the 3A

. Core Spray pump is. Plausible since there is cross-tying of DC power

, ____ : supplies between unit~ fo~ECCS related equipment ___.__.. ._ .._ _.. . .

L D I: Of the items listed, only Unit 3 RCIC is affected by a lOS ..s of Panel 30D021. i

. Plausible since other Unit 3 DC panels (30D023 and 30DD306) do impact

_ . ._ _.. _ . . . _ __..__ i diesel generators and Core Sp ral pumps. ___.. n. _ *** _ _ _ *

_ * *

_ **** _ _ _ _

t___

Psychometrics Leve~~~~~eQge QiffL~~___.__ -.TI1'Tl_E:!~llowan~E:!Ln11I1utest RO 10CFR55.41 (b)(7)

Source Documentation Source: [gJ New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 KIA System: - Partial or Complete Loss of RO/SRO 3.8/3.9 KIA Statement:

.G2.?3 - Knowledge Qf thE?_d_es...,.,i"'r-n___,~pr()cedu@1 and.-2.I~~!~ionaLdifferences bE3tween units. ___________

REQUIRED MATERIALS: NONE 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

55. Given the following:

A AIR COMP TROUBLE (216 B-1) alarm is received

The 'A' Air Compressor (2AKOOl) indicates "tripped" on Panel20C012

Investigation shows the 'A' Air Compressor tripped on high receiver pressure

'A' Instrument Air Receiver (2AT006) pressure is currently 105 psig Which one of the following is correct regarding reset of the 'A' Air Compressor trip?

A. Air receiver pressure does NOT allow the air compressor trip to be reset.

B. Air receiver pressure allows the air compressor trip to automatically reset.

C. The trip can only reset by depressing the "Reset-Start" button locally at the compressor.

D. The trip can only be reset by placing the compressor control switch to STOP at Pane120C012.

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer'

-S5RO Basis or Justification

Per ARC-216 B-1, a high receiver pressure trip occurs at 120 psig.

' normal system pressure is 100 to 115 pSig, current receiver pressure (1 I psig) allows compressor reset. Per the NOTE in ARC-216 B-1, a compressor trip can only be reset locally by depressing the "Reset-Start" located at the trlnlr'.;::>ri Distractors: ARC-216 B-1, a high receiver pressure trip occurs at 120 psig. Since n .....'n"I'3 system pressure is 100 to 115 psig, current receiver pressure {105 reset.

the NOTE in ARC-216 B-1, a compressor trip can only be reset locally depressing the "Reset-Start" button located at the tripped compressor.

Per the NOTE in ARC-216 B-1, a compressor trip can only be reset locally by depressing the "Reset-Start" button located at the tripped compressor.

Psychometrics Level ofi<nowledge _ " _ ~[ff~_~ Time Allow§lnce_{minutes) RO MEMORY 10CFR55.41 (b)(1 0)

Source Documentation Source: ~ New Exam Item Previous NRC Exam: 0 o Modified Bank Item o Other Exam Bank: 0 ILT Exam Bank 6 B-1 KIA System: Importance: RO / SRO 3.0/3.0 KIA Statement:

G2.4.50 - Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

REQUIRED NONE Notes and Comments:

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

56. Unit 2 was operating at 100% power when Drywell pressure began to rise. The crew entered OT-IOl "High Drywell Pressure"'.

At 1.2 psig Drywell pressure the crew performed GP-4 "Manual Reactor Scram".

Which one of the following identifies how OT-101 is required to be used in conjunction with T-l 01 "RPV Control" and/or T-102 "Primary Containment Control"?

At 2 psig Drywell pressure the crew must _ _ _ _ _ _ _ __

A. Exit OT-IOI and enter T-I02 ONLY B. Exit OT-lOI and enter T-IOI and T- 02 C. Enter T -1 02 ONLY and execute concurrently with OT -101 D. Enter T-lOl and T-102 and execute concurrently with OT-lOI 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer' RO i Choice Basis or Justification- ...... ---~---- ...

! Correct: D

Per OT-101, "Follow-up Action" step 3.3, if drywell pressure reaches 2.0 I psig, enter and execute concurrently T-101 and T-102. Per OT-101 Bases, this is because OT-101 provides further direction for mitigating the 1_ _ _ _... ~_.~_. __ ~j...---.--_+--=c:..=0..:..:n.=...se=..:q:c:u~.nces-9f the higb_i!~ell pressure (";ondition_._ . _.._~ ~ __..__ _

Distractors: A OT-101 must be executed concurrently with T-101 and T-102.

i B ! OT-101 must be executed concurrently with T-101 and T-102.

C T-101 must also be entered and executed concurrently with OT-101.

Psychometrics r-Level of Knowl~~~_+__.. Difficl:J!!y Time Allowance (rninutes) I RO MEMORY I 10CFR55.41 (b)(1 0)

Source Documentation Source: I2J New Exam Item Previous NRC Exam: 0 o Modified Bank Item o Other Exam Bank: 0 ILT Exam Bank OT-101 and Bases 02-3 KIA System: - High Drywell Pressure RO/SRO

~--- ~--- ----~.~--.------------~--~~--~-~-~-.-.-- ...*

4.2/4.0 KIA Statement:

G2.4.8 - KJ')9wledge of hoVllabnormalgperatiJ')g procedures areLJ~ed in conjur'lction with~9Ps.~_~~

-~:~~~:~:::~~~~§-: __ I NONE ----==~- . -------.-... __

.._~ ~==~=~-=::

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

57. Given the following:

Unit 3 is operating at 100% power

A feedwater controller failure occurs

RPV water level rises at a rate of 2 inches per second Assuming no operator actions, one minute later the reactor recirculation pumps will be - - - - - - - - -

A. tripped B. operating at 30% speed C. operating at 45% speed D. operating at the initial speed 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Key Question # 57 RO -----~----------------------------- - -

Choice Basis or Justification

---+----- - - - - - - - - - - - - - - - - - - - - - - - - - -

Correct: A With no operator actions ;and RPV water level rising at 2 inches/second, level will reach the high IE!Vel turbine trip setpoint (+46") in approximately 10 seconds. A turbine trip from full power will result in high reactor pressure and SRV actuations. This transient, as analyzed in Section 14.5.2.2 of the UFSAR, results in a peak reactor pressure of -1250 psig (at the bottom of the vessel; equivalent to '~1200 psig in the steam dome). An ATWS (ARI)

RPT will occur at 11 06 psig ... approximate!yJ_~ seconds into the event.

Distractors: B Recirc pumps trip on highi reactor pressure. This choice is based on the applicant not accounting for Recirc pumps tripping on high RPV pressure and the high level trip of feed pumps causing a 30% runback due to total feedwater flow < 20% along with RPV level < 17 inches (following scram).

C Recirc pumps trip on high reactor pressure. This choice is based on the applicant not accounting jfor Recirc pumps tripping on high RPV pressure and the high level trip of feed pumps causing a 45% run back due to any feed pump flow < 20% al()ng with RPV level < 17 inches (following scraI11J__ _

D Recirc pumps trip on high reactor pressure. This choice is based on applicant not recognizing the conditions will result in any change to Recirc pump status. Plausible if applicant does not understand the sequence of events and plant response to a feedwater contrQ!I~!taj_lu_re_.__________'

Psychometrics Level of Knowledge Difficulty ---

Time Allowance (min~_tes) RO --------

HIGH 3 2 10CFR55.41 (b)(6)

Source Documentation Source: D New Exam Item IZI Previous NRC Exam: (PB 2007)

D Modified Bank Item D Other Exam Bank: 0

j---"=

IZII LT Exam B a n k _ _ _ _ _ __ _

f-R_e_f_er_e_nc_e--,-(s--,)_:_-+-A_R_C-314 C-3; UFSA~ 14.5.2.2, UFSAR Figure 14.5._~______________________ _

Learning PLOT-5006-3j Objective:

-Kl--A-S-ys-t-e-m-:----+-2-95005 - Main T urb;~e Generator Trip Ilm-;;;'rtance: ~uRO I SRO------

3.2 / 3.3 KIA Statement:

AK2.03 - Knowledge of the interrelationships between Main Turbine Generator Trip and the following:

Recirculation system. - - - - - - , - - ---------- - - - - - - - - - - - -----

REQUIRED MATERIALS:

I NONEu~==-~~--=-=_____ --_~-=_=-~-~~

Notes and Comments:

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

58. Given the following conditions:

Unit 3 was initially operating at 100% power

3A TBCCW pump tripped due to a motor fault

3B TBCCW pump is blocked Which one of the following describes the impact on continued power operations?

A. A Reactor power reduction will be fe:quired due to loss of Stator Water Cooling.

B. A Reactor power reduction will be required due to loss of Isophase Bus Cooling.

C. An immediate plant shutdown will be required due to loss of cooling to the CRD pumps.

D. An immediate plant shutdown will be required due to loss of cooling to the Condensate pumps.

201) NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Basis or Justification i Correct B Isolated Phase Bus coolers are not vital TBCCW loads. Therefore, I ' . of TBCCW, they are isolated during the swap to RBCCW. Per

~~__ ~~~_:....,_t.c....h_is_:.-,----=---,,--,--.--,,--=----c.._-,-,----,----,---,,__ c....:....:re=---=d--=-u--=--ctc::..io=-=-n-=---t-=-=o~<_~1-=-:: *.. =-=--=---=-s=ta.:..::.to=-:r..... c:.:..:...:..c..-::.:::..:.IA..:...:.W

...:..:==-_ -::.....:::=-:.__

. Distractors: A Stator Water Cooling System coolers are cooled by Service Water and

. . ~~ 1.----- ,~~~~:~~~ :~i~~~~~~:~~~9;~f~~~h~~;~:v~ PI~USib!e if th~~pplicant does

~.CRD pumps are vital TBCCW loads; cooling swaps to RBCCW on loss of 1-.. . :

C

_~~

D TBCCW. A plant shutdown andlor power reduction is not necessary.

~_ **** _ _*

_ **** _ _ ~.a ** __

. Although a loss of TBCCW does result in a loss of cooling to the

_ ___ ~_~_,_~ __ **** _ _ _ _ _ _ ._~_ ***** _ _ _ _ _ _ _ _ _ _ _

l I Condensate pumps, ON-'118 does not direct an immediate plant shutdown.

Ins.t.ead.' .mon.it.ori.ng . . Of Condensate pump temperatures is. direc.ted... and .i.f.

necessary, the pumps are removed from service, which requires a power

___~. . _ _ _.. . _.. ____~requcti'O.!1_._ .....____ ~_.__.. ___. _______.

Psychometrics Level of Knowledge I Difficulty Time Allowance (minu!es) . RO r-HIGH 2.25 3 i1OCFR55.41(b)(1O)

Source Documentation Source: New Exam Item ~ Previous NRC Exam: (PB 2005)

. D Modified Bank Item D Other Exam Bank: 0

... ~~ILT Exam (;3ank ~ __.. ____ ~_ ... ___...__..._ ... _ _ _ _....

Reference(s): _ 0ti-,11? Bases ,_.m.~ _ _**_ _ _ _ .~ _ _ ***..* _ _ _ _ _ ~ __________

_. _ _ _ _ _ _ _ _ _ _ _ .~_ ***

Learning PLOT-5034-3b Objective:

KIA System: 295018 - Partial or Complete Loss of Importance: RO 1 SRO

n""'I"\,..,.... oln~ Cooling Water 3.5/3.6 KIA Statement:

AK1.01 - Knowledge of the operational implications of the following concepts as they apply to Partial or ComElete Loss of ComE~nent cping Water: Effects on C<l.mponentlsystem operatio",- _

.BEQUIRED MATERIAL~~ NO~E _ _ _ _ _ _ ~_____ .

Notes and Comments:. ___ ~ _________ ~___ ~______ ~_. _~_~~ _______, __

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

59. Unit 2 was operating at 100% power for 6 months when the crew scrammed the reactor due to a loss of main condenser vacuum. The following conditions exist shortly after the scram:

HPCI is unavailable

RPV level is 30 inches

RPV pressure is 930 psig

Vacuum is 6" and degrading Based on these conditions, and with no operator action, reactor pressure is as a result of __(2)__,

A. (1) rising (2) MSIV closure B. (1) rising (2) Bypass Valve closure C. (l) lowering (2) RPV cooldown D. (1) lowering (2) Bypass Valve operation 20 II NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answel' 59RO Basis or Justification Correct: automatically close at T' Hg vacuum, which results in a loss condenser as a heat sink. With no operator action, RPV to Distractors: Automatic MSIV closure does not occur on a loss of main condenser vacuum. Plausible since OT-106 directs manual closure of MSIVs at 5" Hg vacuum.

C Based on 6 months of operation at 100% power, there is sufficient decay heat to cause reactor pressure to rise.

D Bypass valves automatic,i)lIy close at 7" Hg vacuum. Since the given conditions state that vacuum has lowered to 6" Hg, bypass valves are Psychometrics Level otKnowledge ,....~ . . ~- Difficulty_ _ --t~mej\IIowance~)nutes) I RO HIGH 10CFR55.41 (b)(5)

Source Documentation Source: New Exam Item Previous NRC Exam: 0 Modified Bank Item Other Exam Bank: 0 ILT Exam Bank 06 540-5 KJA System: 295002 -- Loss of Main Condenser Vacuum Importance: RO 1 SRO 3.6/3.8 KJA Statement:

AK1.03 -- Knowledge of the operational implications of the following concepts as they apply to Loss of M_a_inCondense..!_Va_c_ul!rr!_:L_o~...:;:o:.:..f~. == ......s-.:.:in. .:.:k.:.:. __.. . ._~. _ _.. . . . __~_~_._~_~_ . . . _~____ ~ __ ~_....._ ..

REQUI REDI\II_ATE R_IA!-§-=--:.. . . ._------+~N--=-OcN..:.~E_~ ________.__.. _.____________.

Notes and Comments:

--~ ....... -----"---~------

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom lnitial Reactor Operator NRC Examination January 2011

60. Unit 2 was operating at 100% power when Drywell pressure began to rise. The crew entered OT-IOI "High Drywell Pressure".

At 1.2 psig Drywell pressure the crew perfonned GP-4 "Manual Reactor Scram".

The following conditions currently exist:

RPV level is -5 inches and is being controlled per T-100 "Scram"

Drywell pressure is 1.5 psig and slowly rising

All PRO and URO scram actions have been completed

No other actions have been perfonned Which one of the following is the pneumatic supply to the ADS valves under these conditions?

A. Backup Instrument Air Supply B. Backup Instrument Nitrogen bottles C. Backup Instrument Nitrogen from CAD D. Instrument Nitrogen Compressors "A" and/or "B" 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Questicm # 60 RO Choice Basis or Justification Correct: A Based on the given conditions, a Group 111111 isolation signal occurred due to low RPV level (-1 inch). This results in an isolation of the N2 compressor

. suction valves and the N2 receiver supply to the A and B drywell headers.

Since all PRO scram actkms are complete, the A and B drywell header isolation valves have been bypassed and reopened per RRC 94.2-2, aligning the N2 receivers to drywell loads. As N2 receiver pressure lowers to 85 psig, the Backup Instrument Air isolation valves will automatically

~~I'lJO re-pressurize thenreceivers and supply__dryweILpneumatic loads.

Distractors: B Backup Instrument Nitrogen from N2 bottles to ADS SRVs is not permitted in T-1DO; only from T-101 "RPV Control" (there are no given T-101 entry conditions) and only if specifically directed to be aligned (not part of the f~---.~...........~--+_.~~-+-lJBO or PRO scram actig~~. ..

C Backup Instrument Nitrogen from CAD is not permitted in T-1 00; only from T-101 "RPV Control". Since drywell pressure has not reached 2 psig and

'~_n~__ . .~--+_.~~-+-~EY level is well abov~-=~~9hes, there are no T-::1()j.~,.,t conditions.

N2 compressors A and B tripped due to the loss of suction generated by the Group III isolation signal (-1 inch) .

.- - - - .............. ---~ .. ~- --_. ._ - - - ............ ~-~- ... --.--~--~ ..

Psychometrics Level of Knowle.d.. . n. e _I....

p HIGH - ~-----~~---i-*

Difficulty

Time Allowance (minutes\ RO L I 10CFR55.41(b)(7)

Source Documentation Source: D Previous NRC Exam: 0 D Other Exam Bank: 0 f--............... ----~--+_--'==.....:..:=-'---.:C=.:..:..::....=.::.:.~-.-- ...............- - - -..--~-----~...~-.. - - - - - - ...........----~... -------~...

Reference RRC 94.:.. =2:.=-2=--________

Learning Objective:

KIA System: 295010 - High Drywell Pressure RO/SRO 2.6/2.8 KIA Statement:

AK2.04 - Knowledge of the interrelationships between High Drywell Pressure and the following:

-~~~~~~~~:=:~- -1 NONE__ .. _ _ ===-===--=--~~~-~-

20] 1 NRC RO Written Exam Rev. l.doc

Peach Bottom rnitial Reactor Operator NRC Examination January 2011

61. A high Torus water level condition exists on Unit 3.

In accordance with T-I02 "Primary Containment Control", Torus level is being lowered in an attempt to maintain below Curve T/L-I "SRV Tail Pipe Limit" (provided on the NEXT PAGE).

The reason Torus level is maintained below the "SRV Tail Pipe Limit" curve is to prevent _ . . _ _ _ _ _ _ __

A. exceeding the Torus level Tech Spec Limiting Condition for Operation B. flooding the Safety Relief Valve solenoids, rendering the SRVs inoperable C. direct pressurization of the Primary Containment without pressure suppression D. covering the highest vent capable of passing all of the decay heat from the reactor 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 It SRV TAIL PIPE LIMIT 17.6 17.2

~ ;::=;:::: \

16.8 ,,UNSAFE

/"""',

t--<

~

16.4

.....l w

>w \

.....l 16.0 C/)

SAFE

\

J 0

::

15.6

\

0 t--< \

15.2

\

~

14.8

\

14.4

\

o 200 400 600 800 1000 1200 RPY PRESSURE (PSIG) 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Choice Basis or Justification Correct: The SRV Tail Pipe Limit Curve is specifically designed to prevent the back pressure during SRV operation from damaging the SRV components, potentially causing direct pressurization of containment without pressure Distractors: The Suppression Pool Water Level LCO is exceeded when level is above 14.9 feet, which is below the values shown in curve T/L-1.

B The concern for rendering the SRVs inoperable due to solenoid flooding is but does not occur until 21 feet Torus level, as opposed to the just foot limit shown on the SRV Tail Limit Curve.

i - concern for exceeding the level of the highest vent that can pass all of decay heat after shutdown is real but it is not a concern until level has l~~________________L_~__ I ~;~~/~~~~I~~~Li:~~~~~~~~ymg~~~!~i~~_~d1dfe!t~:ceeded,Whi~h~:ill_~~~_

Psychometrics r-mLevel ofKno~ledgel __~_l:)iffiCul!Y______ Time Allowan~~(minu~~ RO MEMORY 3.5 3 10CFR55.41 (b)(1 0)

Source Documentation Source: D New Exam Item [g! Previous NRC Exam: (PB 2005)

D Modified Bank Item D Other Exam Bank: 0

[g! ILT Exam Bank Reference(s): T-102 Bases, Step TlL-23; f------

--+--TR_~E/SAMP_CURVES-,_I6.I:3LES, &LlMITS_-= BAS~$-,--$tep 27_ ~ ..- - - - - - - - - - -... - -- ----

Learning

PLOT-1560-03, -09 Objective:

KIA System: 295029 - High Suppression Pool Water i Importance: RO/SRO I Level 3.6/4.0 I --

KIA Statement:

EK3.02 - Knowledge of the reasons for the following response as it applies to High Suppression Pool Water Level: Lowering su~~resTn ~ool water level. ------ .... _----_ __ ..... _-----_ _---_.---------

REQUIRED MATERIALS: -----

NONE --~-- .....-------- - -

.. .. ---~--.----- .-------- .... ----~--.---,,------ .

~

Notes and Comments: ~------ -- -- ~--- -- ---

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom [nitial Reactor Operator NRC Examination January 20 II

62. Given the following:

Unit 2 is operating at 100% power

The 2A CRD pump is blocked for maintenance

The 2B CRD pump tripped on motor overload Which one of the following describes the: impact, ifany, on RWCU System operation?

A. No impact; RWCU operation may continue.

B. RWCU must be shutdown to prevent pump damage due to loss of seal cooling.

C. RWCU must be shutdown to minimize pump motor area contamination due to loss of purge supply.

D. RWCU flow must be maximized in preparation for a reactor scram and trip of both Recirc pumps.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Question #. 62 RO Choice Basis or Justification Correct: RWCU shutdown is required by ON-107 if CRD restoration is not This is to minimize radioactive contamination of the RWCU pump motor

_~ __.. ._ _..

Distractors: TA i~~ea due to the loss of RWCU pump seal purge SUPPIY'---.. ---~l I, RWCU shutdown is required by ON-1 07 ~ CRD restoration is not imminent. I

-~*-*-*-*~i~B --I ~~~~t~:o~~v:~~~;e::~~;,:~a~~i~I~:~n~6~~i;sat~~li~~~~c:~~~;- ",

~!~***-****--***---*-I D_-t@nf::~;~~~~~;~;~~~:~~~~:I~~~~r~~~~!~~=~~~~~st l ... _ _ _ _ _ _ _. _ _ _ _ _ _ _ :

~~_

secured on a sustained loss of which is the case here.

Psychometrics c---Levf::lof Kn~wledge _4__ ~ __[)ifficulty______ Time AJI()wan9~ (minlJl~j-10CFR:s~1(b)-(-10)

MEMORY Source Documentation Source: , 0 New Exam Item 0 Previous NRC Exam: 0 I o Modified Bank Item 0 Other Exam Bank: 0

~ ...- . - - ~------~--~...

i 0 ILT Exam Bank ._.....

c-_gef~ren~!5~):__ I ON...::J07 B§!!5~ ____.. _ _..____.. ____.. ._____.. .____...._____.. . ----.-...... --------

Learning ! PLOT-5014-6g Objective:

KIA System: 295022 - Loss of CRD Pumps RO/SRO 2.5/2.6 KIA Statement:

AA 1.04 - Ability to operate and/or monitor the following as it applies to Loss of CRD Pumps: Reactor

~ter Cleanup Syste,!!... ~_ ____ _____ _________________ ~~ ...____.

~~~:~~~::~I:L~:_- L~9NE -==---=---=~=- =-- =--=- _--=~~==~===~==__ . _.. -=_ .- -

2011 NRC RO Written Exam Rev. 1.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

63. During Refuel Floor operations on Unit 2 the Control Room receives the fol1owing alarms and indications:

REFUELING FLOOR VENT EXHAUST HI RADIATION (218 A-I)

REAC BLDG ZONE VENT EXHAUST HI RADIATION (218 B-1)

REAC BLDG OR REFUELING FLOOR VENT EXH HI RAD TRIP (218 D-4)

Refueling Floor Radiation Trip Units RIS-2-17-458 A and D are both reading above 16 mR/hr Which one of the following describes the ventilation system response to these conditions?

A. Reactor Building Ventilation trips Refuel Floor Ventilation trips SBGT initiates and aligns to the Reactor Building and Refuel Floor B. Reactor Building Ventilation continues to run Refuel Floor Ventilation trips SBGT initiates and aligns to the Refuel Floor C. Reactor Building Ventilation trips Refuel Floor Ventilation continues tiQ run SBGT initiates and aligns to the Reactor Building D. Reactor Building Ventilation continues to run Refuel Floor Ventilation continues to run SBGT remains in standby 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Key

"'" . ., In.#63 ~O

Choice Basis or Justification Correct: A trip of "A" and "D" Refuel Floor radiation monitors will result in a Group III isolation. The Group III isolation will trip both Reactor Building and Refuel Floor Ventilation and align SBGT to the entire Reactor Building and Refuel Floor.

Distractors: Reactor Building will also trip even though the high radiation was on the Refuel Floor. SBGT is aligned to both the Refuel Floor and Reactor Building. Plausible if applicant does not recognize, based on the given

- - ~* * * * *- - - -t* * * - C- -t- :-~; ~:a:;!;-:~;;:~~~n:!:~:~oorV~;;W~tio;;- -w~1 ~

conditions, that the logic requirements are met for a Group III isolation of will trip ;;;;dSBGT i be aligned to both areas. Plausible if applicant does not recognize, based

on the given conditions, that the logic requirements are met for a Group III I 1_~ ______""r..._ _ _. __-+--,,-is=-::o~la-:::::ti~:m of both RB Cli"ld RF ventilati ClI1 systems. ____________ J D Both Reactor Building and Refuel Floor Ventilation will trip and SBGT will I be aligned to both areas. Plausible if applicant does not recognize, based on the given conditions, that the logic requirements are met for a Group III isolation ofJ:?P!11 RB aQQ_RF v~ntilation systeIT1~LaI"lSLSBGT !l1iti_Cl!ion_.___ _

Psychometrics Source Documentation Source: D New Exam Item r8J Previous NRC Exam: (PB 2005)

D Other Exam Bank: 0 ILT Exam Bank B-1 and D-4 c

KIA System: 295034 - Secondary Containment Importance: RO I SRO Ventilation High Radiation 3.8/4.2

.---------------------~-------- .........

KIA Statement:

! EA2.01 - Ability to determine and/or interpret the following as they apply to Secondary Containment

,{entilation High Radia!L9n: Ventilation radlCl!ion level~_____________.____..__

~::~~:~~::::~A~S: -~ ItlON~ - ------__ . . _ ._._.-~--_---.-

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

64. Unit 2 is operating at 100% power when an inadvertent Group IIIIlI isolation occurs due to a spurious low RPV level (+ 1 inch) signal.

Which one of the following describes the status of Drywell Chilled Water flow to the Drywell cooling units and Recirc pump motor coolers one minute later?

Drywell Cooling Recirc PumQ Motor Cooling A. Isolated Isolated B. Isolated In-service C. In-service Isolated D. In-service In-service 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer QuestiOI't# 64 RO Choice Basis or Justification Correct: D Both are in service. Drywell Chilled Water (DWCW) to the drywell units and Recirc pump motor coolers does NOT automatically Group 111111 isolation signcll. Per GP-S.B, if containment conditions isolating these loads (i.e., drywell pressure is greater than DWCW Distractors: Both are in service. Drywell chilled water to the fan units does the fans trip (-160 inches RPV level OR 2 psig drywell pressure),

would not occur on a +1 inch low RPV level isolation.

Both are in service. Drywell chilled water to the fan units does Isolate the fans trip (-160 inches RPV level OR 2 psig drywell pressure), but this

~~-............... ........................ ---+-.--.~-i would not occur on a +1 inch low RPV level isolation. - ... - . -...- ... ~--~-~ ..... -.~ ...... ~.-.

Both are in service.

Psychometrics Level of Knowl~~ DifficultY' . Time Allowarlce (minutes) I RO HIGH I 10CFR55.41 (b)(7)

Source Documentation Source: New Exam Item IZl Previous NRC Exam: (LGS 2006)

D Modified Bank Item Other Exam Bank: 0 IZlILT Exam Bank ~ .. ------~-~ ... ~.

Reference( s): GP-S.B

............. c---. ._.-. ------.

Learning PLOT -5007G-1, -3 Objective:

KIA System: 295020 -Inadvertent Containment Isolation Importance: RO/SRO

_....... ~---.

....... ~.-.-- ........- ................. -------~-

3.2 / 3.2

.~---

KIA Statement:

AA1.02 - Ability to operate and/or monitor the following as it applies to Inadvertent Containment Isolation: Orywell ventilation/COOllin g system. ---_.

REQUIRED MATERIALS: -

NONE - - _ ................... -~-- ....... . ...................-----.-- - - -- ----------- .. ~.-.

Notes and Comments:

_---------------------------------- ---------............ ~---

................... _-- ........................... ~~-- ~-.--

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

65. Given the following:

Unit 3 is operating at 100% power

A loss of feedwater heating occurs

The crew enters OT-104 "Positive Reactivity Insertion"

OT-I04 directs lowering reactor power using GP-9-3 to at least 10% below the pre-transient power level To comply with this step the operator must first _ _(1 )__. The reason for lowering power at least 10% is to __(2)__.

A. (1) insert control rods (2) provide margin to the full power thermal limits B. (1) insert control rods (2) avoid reaching an APRM rod block or scram setpoint

c. (1) reduce Recirc flow (2) provide margin to the full power thermal limits D. (1) reduce Recirc flow (2) avoid reaching an APRM rod block or scram setpoint 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Exam ination January 2011 Answer' Questi~n#65 RO Choice Basis or Justification

~--~---+---~------~------~~-~--------------.--~-~--~-~--

Correct: C i Based on the initial conditions of 100% power, GP-9-3 will direct lowering Recirc flow first until either flow reaches -61.5 Mlbm/hr, or an APRM HIGH alarm occurs. Per OT-104 Bases, the reason for lowering reactor power by at least 10% is to provide additional core thermal margin under potentially

, Ai~;~::~t;,;~":!:~t~f:~~:!c:~::~::~,

i Distractors:

GP-9-3 directs 1~':'eringReCirC--

flow first until either flow reaches -61.5 Mlbm/hr, or an APRM HIGH alarm i B Incorrect first action - for these conditions, GP-9-3 directs lowering Recirc flow first until either flow reaches -61.5 Mlbm/hr, or an APRM HIGH alarm occurs. Incorrect bases - avoiding the APRM rod block and scram I-~- - -I F:~~~:= ~a::~-";~;:~;~~:::~:i~t~~:~e:~d scramsetpoint;;s not i ' t h e bases for performing this step.

~.~..... . ~-.-.. - - .....- -....- - -....-~---....- - . - -....- - . - - -....--~-....----~.....~----....- - - .. - -...

Psychometrics Level of Kn9wledg~-+-_~._-=D-=ift'iCulty__. ___1 Time Allowance (minutes) i ..._ RQ__ .__.. _

HIGH I I 10CFR55.41(b)(10)

Source Documentation Source: [;8J New Exam Item D Previous NRC Exam: 0 i D Modified Bank Item D Other Exam Bank: 0

_... ..._t-OJLT Exal11 Bank_.___.. .~__.__.. .___~. . ___~..... _____.. ____...__ ..____

~efeIen<::e(s):_ +~-!.::104.and Bas_es;GP-9-~ ____.. ...______~ _ _.. .____ ~. . .________ .

Learning , PLOT-PBIG-1550-3 Objective:

KIA System: 4 -Inadvertent Reactivity Addition RO/SRO 4.3/4.4 KIA Statement:

G2.1.23 - Ability to perform specific system and integrated plant procedures during all modes of plant REQUIRED MATERIALS:

Notes and Comments:

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

66. Given the following:

Both Units are operating at full power.

The entire shift team was present at a 0730 Shift Turnover Meeting conducted by the Control Room Supervisor (CRS) and the Shift Manager (SM).

At 1100, the Fourth Reactor Operator (RO) enters the Unit 2 Controls area to relieve the Unit 2 RO for lunch.

The Unit 2 RO will be eating in the Control Room Lunchroom.

What is the MINIMUM TURNOVER ACTIVITY and the MAXIMUM DURATION of this mid-shift turnover in accordance with OP-AA-112-1 0 1 "Shift Turnover and Relief'?

A. Tour the Main Control Boards with the off-going RO; relief duration shall be

< 30 minutes.

B. Tour the Main Control Boards with the off-going RO; relief duration shall be

< 60 minutes.

C. Review the Shift Turnover Checklist including any deviations; relief duration shall be < 30 minutes.

D. Review the Shift Turnover Checklist including any deviations; relief duration shall be < 60 minutes.

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 I I Answer

66RO Choice Basis or Justification Correct: Section 4.1.7 states that mid-shift relief for "less than one hour" only requires a review of the Shift Turnover Checklist and any deviations from status.

Distracters: "'1'\,.."Trl'\l board tour is not required and the turnover is valid for 60 minutes.

,...... ,..,'T ...... I board tour is not required.

C This mid-shift turnover is valid for 60 minutes.

Ps chornetrics Level of Kn9~ledge__ L Difficult.L..Y_ __ Time Allowance (minutes MEMORY 2.5 3 Source Documentation Source: D New Exam Item ~ Previous NRC Exam: (PB 2005) o Modified Bank Item o Other Exam Bank: 0

[g] ILT Exam Bank Reference(s): OP-AA-112-101 ~~~~~-

Learning PLOT-1570-17 Objective:

KIA System: G2.1 - Conduct of Operations I Importance: RO/SRO 3.7/3.9 .~-~~

KIA Statement:

G2.1.3 - Knowledge of shift or short-term relief turnover ~actices. - ------~.--------- ~ ~~~--

REQUIRED MATE~: I NON~"- ~~~- - ------ -- _......._.

Notes and Comments:

--~---

.. ~

............ ~-~--

20 II NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

67. Unit 3 is in MODE 5 with refueling activities in progress.

Which one of the following conditions requires the Reactor Operator to notify the Fuel Handling Director to suspend core alterations, in accordance with FH -6C "Core Component Movement Core Transfers"?

A. A FUEL POOL SERV WATER BOOSTER PUMP OVERCURRENT (216 C-5) alarm.

B. Shutdown Cooling (SOC) has been removed from service to complete a swap of SDC loops.

C. Wide Range neutron count rate doubles when a fifth fuel bundle is seated around the' A' WRNM detector.

D. The white rod pennissive light on Panel20C005 is NOT lit when the refuel platfonn is over the core with fuel loaded on the main hoist.

20 II NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examiination January 2011

~~----------------.----

Answer Key Question # 67 RO r------------ - - - - - - - - - - - - - - - - - --------

Choice Basis or Justification Correct: C FH-6C requires notifying the FHD to secure fuel handling when the WRNM count rate doubles after tile fourth fuel bundle is placed around the detector (step 10.2.14). ON-124 entl}' is required if count rate doubles twice.

Distracters: A Loss of a Fuel Pool Cooling Service Water pump does not require securing fuel handling .

B Swapping SDC loops dOE!S not require securing fuel handling.

- + - - - - - - - - - - - - - - - - - - - - - - - - - - ----------

D Under the conditions described, the white light should be extinguished.

Psychometrics I

Level of Knowle~i Difficulty --.- .. -~

__Ti,!!~~lowance (minutes) RO -

MEMORY 3.0 3 10CFR55.41 (b){1 0)

Source Documentation Source: D New Exam Item [8J Previous NRC Exam: (PB 2005)

D Modified Bank Item D Other Exam Bank: 0

[8J ILT Exam Bank -

Reference(s): FH-6C; ON-124 ._--_._- - --- .._---.--

Learning NLSRO-0763-6 Objective:

~--

KIA System: G 2.1 - Conduct of Operations I Importance- RO/SRO 3.9/3.8 - - - - -

KIA Statement:

G2.1.44 - Knowleqge of RO duties in the control room during fuel handling such as responding to alarms from fuel handling area, communication with tlhe fuel storage facility, systems operated from the control room in support of fuelin~ operations, and supporting instrum_entation. --- -.

REQUIRED MATERIALS: NONE .-~-.--

Notes and Comments: -_.- -_. -~------------

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

68. In order to operate at the Maximum Power Level stated in the facility license, which of the below conditions must be met?

1. Three Leading Edge Flow Meters must meet TRMS 3.20 "LEADING EDGE FLOW METER (LEFM) SYSTEM" requirements

2. Core Thermal Power (CTP) calculation must be available

3. Shift Average CTP must NOT exceed 3514 megawatts thermal at any time A. 1 ONLY B. 2 ONLY C. 1 AND 2 ONLY D. 1,2 AND 3 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II Answer

68 RO Choice Basis or Justification

-T---~-***~***~~~**---~~----~~--~~--------****~~~~-----~ .............~

Correct: C GP-5, Note 2 on page 42 states that MAPl is determined lAW TRM which requires three OPERABLE lEFMs (for Reactor Operators, this is "above the line" TRMS knowledge). GP-5, Step 6.1.6 describes the required actions to take if CTP calculation is NOT available - reduce generator output by 5 MWe. GP-5, Note 7 on page 44 describes how to maintain Shift Average CTP below MAPl for the shift, allowing shift MAPl is not exceeded at end of shift.

Distractors: A Plausible as this is a TRM requirement. Candidate may be aware of this limit but not know CTP requirement.

B Plausible as this is a GP-:5 requirement. Candidate may be aware of this limit but not know TRM requirement invoked by GP-5.

o Plausible if candidate believes Shift Average CTP must be maintained below MAPl in addition to knowing the TRM and CTP requirements. Note that the Facility Operating License authorizes Peach Bottom to operate "at steady state reactor core power levels not in excess of 3514 megawatts GP-5 is meant Psychometrics level of Knowledg~__ ! _____ D_iffiCUI!y_

Tim§l!\lIowance (minute:..=sCL)~_ _----'R--=-O=----_______

MEMORY i 10CRF55.41(b)(10)

Source Documentation Source: ['8'J New Exam Item Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 IlT Exam Bank Befer§ln_c_e-,--(s--'--):~_~i-F_~9i1ity OperatinA!:i9~nse; Gf>:::§;~T=--:..R-=-Mc..:....::..3=.2,--,,-O_ _ __

learning PlOT-DBIG-1530-3 Objective:

KIA System: G2.2 - Equipment Control RO/SRO 3.6/4.5 KIA Statement Notes and Comments:___ _

I G2.2.38 - Knowledge of conditions and limitations ill tl}e facility license_._____ __

REQUIRED MATERIAls: NOiliEmn

~ ___ ~ _________ ~ _ _

- un ------------~-

~_._

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

69. 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-l) alarms

- A RFPT TRIP (201 G-4) alarms

- B RFPT TRIP (201 H-4) alanns

- 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 0 inches

- Reactor pressure is 1028 psig

- Reactor power is 100%

What actions are required for these conditions?

A. Perform GP-4 "Manual Reactor Scram".

B. Trip the Main Turbine and enter T-100 "Scram".

C. Scram the Reactor and enter T -100 "Scram".

D. Scram the Reactor and enter T-l 01 "RPV Control".

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 I 1 Answer Basis or Justification given conditions indicate the main turbine did not trip on high reactor level eX!Jected (which would have caused a reactor scram). Since the feedwater and RPV level has lowered below the scram setpoint of +1 inch, an has occurred. This is an entry condition for T-101: "scram condition above 4% or unknown".

A 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 thesl~ conditions. In addition, since a scram should have I occurred, the operator is required to manually scram the reactor (place the mode

. . switch in shutdowDl .m.. ._.m........... _ _ _ _._ _ _ _........... __ .

B This would rely on the Rector Protection System to scram the reactor, which

, violates the "Reactivity Management" Operations Fundamental (do not rely on the reactor protection system to protect the reactor during reactivity events). Since a scram should have occurrecl, the operator must manually scram the reactor (place the mode switch in shutdown). Plausible since the main turbine should have tri ped on a h!.9.tJ.reactor water leY.§l~ __.m _ _ _ _.

C 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

,tnrn",ti,..",lhl scram as This overrides OT -110 direction.

Psychometrics

_l:evel of Knowledge .- -----~

DifficllltY_m ._m __ ~ TilTlt:?Allowance (I'l'linutes) ...........

RO

--~~~-~-~-- ._--- ._ --~-----------

HIGH 3.0 3 I 10CFR55.41 (b){10)

Source Documentation Source: o New Exam Item o Previous NRC Exam: 0

~ Modified Bank Item o Other Exam Bank: 0

~ ILT Exam Bank ~--~-- .. -.- .. _------- --_._--......._-----

_Reference{s}: ~ARQ.-2Q§C-1; OT-11 0;.GP-4; T -100 ~ .. ~---- .... -_........

Learning PLOT-1529-2 Objective:

KIA System: G2.2 - Equipment Control I Importance: RO/SRO 4.2/4.4 KIA Statement:

G2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and dir~_qi~es aff~ct Rlant and system cQ.ndJ!ions. ......... ------~~------

REQUIRED MATERIALlj:_ ] NO~~_ __ _ ___ ....... --~~-

Notes

- and Comments:

- -- ---- - ~ - . - --- .. ~--

..... -~,--". ---

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

70. Both units are operating at 100% power with the following conditions present:

RIS-0760D "Main Control Room Ventilation Radiation Monitor" is failed with a trip inserted per GP-25 Appendix 14 "MCR Ventilation Isolation, Division II"

CONTROL ROOM RAD MONITOR DIV II INITIATED (003 A-3) is lit due to the GP-25 trip One hour later, an annunciator is received and the PRO observes:

CONTROL ROOM VENT SUPPLY FAN HI-LO (003 A-I) is in alarm

CONTROL ROOM VENT SUPPLY LO FLOW CREV START (003 A-5) is in alarm

CONTROL ROOM RAD MONITOR DIV I INITIATED (003 A-2) is in alarm

Flow Recorder FR-0765 indicates 200 scfm and lowering

RIS-0760C "Main Control Room Ventilation Radiation Monitor" is failed upscale Based on these conditions, the Control Room Emergency Ventilation System has A. started due to the low flow condition B. NOT started as indicated by the low flow condition C. started because the Rad Monitor initiation logic is satisfied D. NOT started because the Rad Monitor initiation logic is NOT satisfied 2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 LnJ#~!iil.rll'ln # 70 RO Choice Basis or Justification A The CREV system is in sE~rvice as indicated by 003 A-5, and was initiated by Low Flow. The Rad Monitor combination would NOT result in CREV initiation Monitor is "A B B The low flow signal is actually from normal Control Room Ventilation and normal during a CREV initiation.

C Plausible because the alarms indicate Div I and Div " initiated, even though the logic for CREV initiation due to Rad Monitors is NOT satisfied (Rad i Monitor logic is "A or BAND C or D").

D Plausible because CREV has NOT started due to Rad Monitor logic, it has started due to LOW FLOW condition.

Psychometrics

, I c-~Level ofKnowledge_~_

~. .~-- !li:i~~ltY_~ ~_ 1--.l1 me

- Allowa~ce (rninuteS)i-oCFR5~~1(bi11)

HIGH Source Documentation Source: o New Exam Item o Previous NRC Exam: (PB 2009) o Modified Bank Item o Other Exam Bank: 0 o ILT Exam Bank .. -.

mReferel1ce(sLm_~iGP-=~5 ~pendi)(~4; SQ 40D.1.~, ARC-0_93 A-1, A~C . .Q03 ~A-2, AR9:Q03 A-tL __ . __

Learning Objective:

I:OT-50400-4a c---'" ... i *~**"-*--*----~--------~--~***~~~-----~---****-T---- - ~---.-----.-

KIA System: . G2.3 - Radiation Control Importance: RO/SRO 2.9/2.9 ... - .

KIA Statement:

G2.3.5 - Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, l..2ortat>le surv~.Ylf'lstrl!rn~Qts, "per~onQ~f'l"lonitofing equipmenh~L. __ ~__ .m_. __ ~_ .... --~--.- .. -~---- ..

REQUIRED MATERIALS: NONE Notes and Comments:

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

71. A transient on Unit 2 resulted in the following conditions:

Containment venting is required for combustible gas control using T-200-2 "Primary Containment Venting"

Chemistry determined that the maximum Containment vent rate that will not exceed the General Emergency release rate is 2,000 sefm

Drywell pressure on PR-2508 is 25 psig

Standby Gas Treatment is available Using Figure 1 ofT-200-2, PROVIDED ON THE NEXT PAGE, determine which one of the following vent paths will most quickly remove the combustible gases without exceeding the General Emergency release rate.

A. 2 inch hard vent to SBGTS B. 6 inch ILRT line C. 16 inch Torus Hardened Vent D. 18 inch vent to SBGTS 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 FIGURE 1 MAXIMUM PRIMARY CONTAINMENT VENT RA'fE FOR VARIOUS VENT PATH SIZES Maximum vent Path Volumetric Flowrate 1,000,000 i"

100.000 l - I-- ..

ft= ....... 18 inch \le'1: to S8GTS an

~

=!!

2-iii

-"-16 Inch Torus S 10,000 III:: 1== Haroel'ed VfIt

~

....2

""11"E "0

J 1.000

~ _ 6 inch lLRT line iii

> 1/

100

.......2 inch Hard Vent toSBGTS 10 I I I I I J o 10 20 30 40 50 60 70 SO 90 100 110 Containment Pressure /P&IOI 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Basis or Justification Correct: Plot Containment pressure of 25 psig and vent rate of 2000 SFCM, the point is ABOVE the 2 in Hard Vent to SBGTS line and BELOW the 6 inch ILRT line; the 2 in Hard Vent to SBGTS is the largest vent path that will NOT exceed the GE releclse rate.

Distracters: B Plot Containment pressure of 25 psig and vent rate of 2000 SFCM, the point is ABOVE the 2 in Hard Vent to SBGTS line and BELOW the 6 inch ILRT line; the 2 in Hard Vent to SBGTS is the largest vent path that will NOT exceed the GE release rate. Plausible if plotted wrong or the c---*__~i;td~:~:a~nQ!:~o;~~:::~:!:;~~h:s~U~ne~ve;rt~~t;-~f 20-00SFCM~h~

point is ABOVE the 2 in Hard Vent to SBGTS Line and BELOW the 6 inch ILRT Line; the 2 in Hard Vent to SBGTS is the largest vent path that will NOT exceed the GE releclse rate. Plausible if plotted wrong or the

candidate does not understand the curve.

D Plot Containment pressure of 25 psig and vent rate of 2000 SFCM, the point is ABOVE the 2 in Hard Vent to SBGTS line and BELOW the 6 inch ILRT Line; the 2 in Hard Vent to SBGTS is the largest vent path that will NOT exceed the GE release rate. Plausible if plotted wrong or the candidate does not understand the curve.

Psychometrics

~vel of Knowledge _.____PlffiClll!Y.______ _Time Allowance (ITlif1.!:1tes J )l. ___BO __, . _____ M. _ _ _ _ _

HIGH i i

10CFR55.41 (b)( 10)

Source Documentation Source: New Exam Item !Xl Previous NRC Exam: (PB 2009)

!Xl Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank 00-3 KIA System: - Radiation Control Importance: RO I SRO 3.8/4.3 KIA Statement:

._G2.3.11 - Abili!LtQ._coQ!!"QI!adia=tr-io~n_:...c.r_e. :.--=le. . :. a-".se..:. .sC-.._. _____._ _._ _.. ._.._.___...____ . ____ _

i REQUIRED MATERIALS: NONE

Notes and Comments:

2011 NRC RO Written Exam Rev. 1.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II

72. For an actual fire reported at Peach Bottom, which one of the following affected areas will REQUIRE entry into ON-114 "Actual Fire Reported in the ... "?

A. Inner Screen Structure B. Water Treatment Plant C. SU-2S Startup Switchgear House D. Low Level Radwaste Storage Facility 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 J I Basis or Justification Correct: A ON-114 entry symptom includes the following areas: Power Block, Diesel

, Generator Building, Emergency Pump Structure, Inner Screen Structure, and the Emergency Cooling Tower Structure. Reactor Operators are required to know entry conditions for ON, OT and TRIP procedures from Distracters:

i

. ON-114 entry symptom includes the following areas: Power Block, Diesel J. Generator Building, Emergency Pump Structure, Inner Screen Structure, and the Emergency Cooling Tower Structure. Plausible because a fire in I~_' .. ____ ~_ ._.__ f--~ __ . ~9~:~e~_~~~~~m:~:;I~~~;~~l~~~I::~dispa~~~_~he Fir~~rigade,but is_ i C ON-114 entry symptom includes the following areas: Power Block, Diesel Generator Building, Emergency Pump Structure, Inner Screen Structure,

. and the Emergency Cooling Tower Structure. Plausible because a fire in

, the SU-25 Switchgear House would involve dispatch of the Fire Brigade, but is NOT an ON-114 Entry Symptom Area.

D ON-114 entry symptom includes the following areas: Power Block, Diesel Generator Building, Emergency Pump Structure, Inner Screen Structure,

! and the Emergency Cooling Tower Structure. Plausible because a fire in

, II the LLRW Facility would involve dispatch of the Fire Brigade, but is NOT an I ON-114 Entry Symptom Area. . _ _ _ _ _ _----'

Psychometrics Level ofKnowledg~_ -~~"- ....

Difficul!Y.

Time Allowance......(minutest I

.~----- ------

RO MEMORY 2 I 3 110CFR55.41(b)(10)

Source Documentation Source:

J.~.00 New Exam Item Modified Bank Item

[8J Previous NRC Exam: (PB 2002)

D Other Exam Bank: 0

... _t8JJLTExam Bank n .. &_(s):

,",CICI CllvC~ ! ON-114 Learning PLOT-PBIG-21 00-3 Objective:

KIA System: - Emergency Procedures/Plan II II ,",VI La I RO/SRO 3.4/3.9 KIA Statement:

G2.4.?7 - Knowl~ci~..2i"fire inJhe plant" pr",..,...,jures.

REQUIRED MATERIALS: NONE Notes and Comments:

20 II NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011

73. A Main Control Room Annunciator Mode Switch is to be placed in manual for greater than one shift.

How is the position of the Annunciator Mode Switch indicated to the Control Room staff?

A. Equipment Deficiency Tag (EDT)

B. Equipment Status Tag (EST)

C. Green Triangle or Sticker D. Red Triangle or Sticker 2011 NRC RO Written Exam Rev. 1.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Key QUE~sti()O # 73 RO Choice Basis or Justification Correct: 03-102, Equipment Status Tags are used to track annunciator mode switch repositioning for more than one shift.

Distracters: A I Plausible because per OP-AA-1 08-1 05-1 001, the use of the Equipment Deficiency tag would be appropriate if the alarm itself were malfunctioning.

In this case the EST is used to track the status of the annunciator mode

."AIIT,.n not the annunciator function.

C Plausible because OP-PB-1 03-1 02-1 002 directs use of GREEN triangles or !

stickers for nuisance alanns expected to remain annunciated for greater I than one shift.

D ! Plausible because OP-PB-1 03-1 02-1 002 directs use of GREEN triangles or stickers for nuisance alarms expected to remain annunciated for greater than one shift, and RED triangles are used for grouping alarms for first-in I I identification.

Psychometrics Lev~1 of Kno'l\lledg~ . . _._._._ !?iffi~llI!Y ILf1}e Allowa~ce (Illinutes) RO MEMORY 10CFR55.41 (b)(1 0)

Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 OP-AA-103-1 OP-AA-108-105-1001 - _......_------_.. _-_.-.

Learning Objective:

KIA System: - Emergency Procedures/Plan Importance: RO / SRO 4.2/4.1 KIA Statement:

_G2.4.31--=l<nowle~e of a.l1_nunciatoLalarl}1s, iF!dica~ions, or res[,onse ['rocedures_._ _.. . ____..........._..__

REQUIRED MATERIALS: NONE Notes and Comments:

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 20 II

74. An Equipment Operator (EO) has been assigned to enter the Moisture Separator Area to investigate a stearn leak. The following information has been provided:

The Equipment Operator has 3280 mRem TEDE annual Exposure

Expected dose for investigation of the stearn leak is 300 mRem In accordance with RP-AA-203 "Exposure Control and Authorization", which one of the following describes the action require:d, if any, to investigate the stearn leak under these conditions?

A. A Planned Special Exposure must be obtained B. A Dose Control Level Extension mus.t be obtained C. An Emergency Exposure Extension must be obtained D. No action required since total exposure will be < 4000 mRem 2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer m::y Question # 74 RO Choice I Basis or Justification Correct: B RP-AA-203 requires dose extension above 2000 mRem TEDE. Dose extensions are granted in 500 mRem increments. The current extension is i good to 3500 mRem. Anc)ther extension is required to get to the 3580 r -.... ~--~-~--.--.....--t~-- ~._-0DRem eX-'pected~xposure___~_. .___..__.__.. _ .._ . .___.. ____._.....

Distractors: A This evolution does not qualify as a Planned Special Exposure, which is separate from and in addition to the annual exposure limits.

C This evolution does not re!quire an Emergency Exposure Extension since the conditions do not rise to the level of "lifesaving or protecting valuable D RP-AA-203 requires dose extension above 2000 mRem TEDE. Above 4000 mRem, Site Vice President approval is required.

Psychometrics

__ Level ()f KnoVv'~~L ___.. Difficulty_----J.lime Allowance (minutesLJ RO MEMORY I 2.5 I 3

10CFR55.41 (b)(12)

Source Documentation Source: o New Exam Item [g] Previous NRC Exam: (PB 2007)

Modified Bank Item Other Exam Bank: 0 ILT Exam Bank KIA System: - Radiation Control RO/SRO

~.~-----.--~.-.-~.-------.---------- ~----~-.--.-------.

3.2/3.7 KIA Statement:

2011 NRC RO Written Exam Rev. J.doc

Peach Bottom Initial Reactor Operator NRC Examination January 201 1

75. Given the following conditions:

A complete loss of off-site power has occurred

RPV water level is -20 inches

Drywell pressure is 6 psig

All four Emergency Diesel Generators (EDGs) are running but do NOT have cooling water available

The Control Room Supervisor directs you to shutdown the running Emergency Diesel Generators In accordance with SE-1 1 "Loss of Off-site Power", which one of the following describes how the EDGs are required to be shutdown?

A. Install jumpers at each local EDG Gauge Panel.

B. Depress the STOP Pushbutton on each local EDG Gauge Panel.

C. Install jumpers in Main Control Room Panels 00C029A, B, C, and D.

D. Place each EDG control switch in "Pull-to-Lock" at Main Control Room Panels 00C026A, B, C, and D.

2011 NRC RO Written Exam Rev. l.doc

Peach Bottom Initial Reactor Operator NRC Examination January 2011 Answer Key I Distracters: A SE-11, Attachment A directs shutting down the EDGs by installing jumpers in the MCR.

i B

SE-11, Attachment A directs shutting down the EDGs by installing jumpers in the MCR.

D The MCA relay is picked up due to a LOCA signal; this precludes shutting

down the EDG from the normal control location.

Psychometrics

,=evel of Knowledge J__ ~_~ Difficul~_. . _~l~ime Allowi:lnce (minutes) I RO~_H HIGH 2.25 I 3 I 10CFR55.41 (b)( 10)

Source Documentation I D D Previous NRC Exam: 0 Source: New Exam Item Modified Bank Item D Other Exam Bank: 0 k8J I LT Exam Bank J=ieference( s):_~ I~SE-11 , Attachment A _.

Learning PLOT-1555-3 Objective:

.. ~

KIA System: G2.1 - Conduct of Operations IJJUlldll\";~, RO/SRO I 4.4/4.0

. _ .- - - - - . ..... -~-,- .. - -......- - . _ ..... -~.~ .

I KIA Statement:

c1~1.30 . -:::- Ability t9 10cat~~f"1d opera!e compcments, inc~.!Jdinglocal c9.ntrol!:)_____ ~_. . _..~~~_.~ __.~. . ._~~~ ___

REQUIRED MATERIALS: NONE Notes and Comments:

2011 NRC RO Written Exam Rev. I.doc

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20 I I

76. Given the following conditions:

Unit 2 is operating at 100% power

2A TBCCW pump is in service

TURB BLDG COOLING WATER SUPPLY LO PRESS (217 C-5) is received

ISO-PHASE BUS TROUBLE (206 F-5) is received

2B TBCCW automatically starts Two minutes later, TBCCW system pressure on Panel20C012 (PI-2229) is 25 psig.

Which one of the following is correct fOJr these conditions?

TBCCW system pressure 1)~; the CRS must direct the crew to A. (1) is low (2) reduce generator load to < 18,000 stator amps using GP-9-2 "Fast Reactor Power Reduction" B. (1) is low (2) perform a plant shutdown using GP-4 "Manual Reactor Scram" and remove Condensate pumps from service C. (1) is low (2) perform a plant shutdown using GP-4 "Manual Reactor Scram" and remove Station Air Compressors from service D. (1) has been restored (2) restore TBCCW pump lineup to normal using SO 34A.l.A-2 "TBCCW System Startup and Normal Operations" 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer Question:# 76 SRO -~-"""-T~-~"- ---"~~---""

Choice Basis or Justification Correct: A TBCCW system pressure on PI-2229 is - 100 psig. The standby TBCCW pump starts at 70 psig and a low pressure condition (217 C-5) is alarmed at 50 psig.

Alarm 206 F-5 is received due to loss of TBCCW flow to the Iso-phase Bus Coolers.

Per ON-118, ifTBCCW cooling cannot be restored (as is the case here) power must be reduced to < 18,000 stator amps lAW GP-9-2). This is done prior to securing both TBCCW pumps and transferring vital loads to RBCCW, since the Isolated Phase Bus

" "~~"""---~~"""-~ "-~ -+-~"

coolers

"~.--~-.- ...

are..------

~---

not vital TBCCW

-,,-... ~. ----_

loads.

Distractors: B

Although a loss of TBCCW does result in a loss of cooling to Condensate pumps, ON I 118 does not direct an immediate plant shutdown unless a unit trip is likely (imminent r per ON-118 bases), which is not the case based on the given conditions. Instead,ON "1118 directs monitoring Condensate pump temperatures and, if preset values are exceeded, removing the pumps from service, which first requires a power reduction LJsing~E~~-2.~ "_~~ ____ "~_" "~_.""~~_" ____ ~".

C There are no direct actions in ON-118 for loss of cooling to the Station Air Compressors. ON-119 "Loss of Instrument Air" directs a rapid plant shutdown using GP-4 only if air header pressure cannot be stabilized above 75 psig, or if equipment critical to continued plant operation begins to malfunction due to low air pressure. For a sustained loss of TBCCW, ON-119 directs cross-tying the Unit 2 instrument air system to Unit 3.

D TBCCW system pressure has not been restored.

Psychometrics r-b~.vel~f Kno\Alleqg~ __ i____ ____ DiffJc;UJ!y_ : Til1)~ AII()\Alal1c~_lrl1j!lLJ!~~ ---~----

SRO HIGH I I 10CFR55.43(b )(5)

Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: ()

o Modified Bank Item Other Exam Bank: 0

,OILT Exam Bank-"""---"

- ~*-"-"-"""~-*""~"----"1--~~"""--"~""---""-*"

Referenc;e(s):_~_V~B_~-g.17 C-:5; A~C-206 .

Learning I' PLOT-5034-3b Objective:

KIA System: rS018- Partial or

. Component Cooling Water Co~plete Loss of I Importance:

I SRO 2.9 KIA Statement:

AA2.05 - Ability to determine and/or interpret the following as it applies to Partial or Complete Loss of

~~~t:~~~:_::_r~:. t";~.:_.~ ~ _..~~_~Sl~~:~~e 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

77. Unit 2 is operating at 100% power with rising Drywell pressure.

The crew began venting the Drywell using SO 7B.3.A-2 "Containment Atmosphere Pressure Control and Nitrogen Makeup"

The Reactor was scrammed at 1.1 psig Drywell pressure

RPV level lowered to -10 inches following the scram; current RPV level is +20 inches

Drywell pressure is now 1.2 psig and rising slowly In order to reestablish Drywell venting, which one of the following is correct for monitoring Drywell radiation levels using the Primary Containment Radiation Gas Sampler (PCRGS)?

The PCRGS A. is isolated; direct resetting the isolation using GP-8.B "PC IS Isolation Group II and III" B. is isolated; direct bypassing the isolation using GP-S.E "Primary Containment Isolation Bypass" C. is NOT isolated; direct manual isolation using GP-8.B "PCIS Isolation Group II and III" if drywell pressure exceeds 2 psig D. is NOT isolated; direct continuous monitoring of drywell radiation levels using SO 7B.3.A-2 "Containment Atmosphere Pressure Control and Nitrogen Makeup" 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

~ ...

Answer Key

- - . -.. ~"-~,-""-,-~ ..

UUE!Sm:m # 77 SRO..""- "-""'"""~

Choice Basis or Justification

--~---""--~ -'"~---~~---

Correct: A OT-101 "High Drywell Pressure" directs venting lAW SO 7B.3.A-2 if drywell radiation levels are within acceptable limits. SO 7B.3.A-2 requires continuous monitoring of drywell radiation levels while venting. PCRGS provides this function but its sample points tap off the CAD H2/02 gas analyzer sample lines, which isolate on a Group III isolation (including RPV level below +1 inch). Since RPV level drops below +1 inch following a reactor scram from 100% power, it is necessary to reset the isolation, once RPV level is restored, to allow continued

~ ___ ...____ ~ ____ "~_._"._._~__+ monitorillg~(j~ellradiatiQfllE!~~ls wj1Jle ventiflg.._~_

Distractors: B PCRGS automatically isolated at +1 inch RPV level. Since RPV level is above the isolation setpoint, bypassing the isolation is not required. Section 6.0 of GP-8,E provides direction for bypassing the CAD gas sample valves isolation, if necessary,

--"~.--~~. "---~- -+--""-.-"-~

to restore

-""-", PCRGS.

~""-

C 1 PCRGS automatically isolated at +1 inch RPV level. It will also automatically

isolate at 2 psig drywell pressure, GP-8, B is the correct procedure to perform a manual ...

~---~+~~~~~-. ~-~

isolation

"--~

if the automatic isolation did not occur as designed.

-"--- ""---" ._---_." - - .~" ,--""~"-"-"--~"-- "----~" "-- -"~

o PCRGS automatically isolated at +1 inch RPV level. SO 7B.3.A-2 is the correct procedure for monitoring drywell radiation levels while venting, but the PCRGS must first be restored (isolation reset) in order to continue monitoring drywell radiation levels.

Psychometrics Level of Knowl~~JL~ Qlfficulty~__ '--1 Time A"OWC3I1.C~jr:njI1U~~Sll SRO -

HIGH i 2.5 , 3

10CFR55.43(b)(5)

Source Documentation Source: o New Exam Item [gI Previous NRC Exam: (PB 2002)

[gI Modified Bank Item 0 Other Exam Bank: 0

_----1._~_". ___ ILT Exam Bank

"~._._

Ref~rencei~L .~Q ?B.3:~-2; GP-f!.~;9T-101* T-102 Learning I', PLOT-5007B-4d Objective:

KIA System: T29-S024 - High Dryw~IIPre~sure i Importance: SRO

_._~~ _ _ ~~ ___ "__~~_~_l_" ___ ~ __,~.__..~_ .. _~._,~" 4.0 KIA Statement:

EA2.08 - Ability to determine and/or interpret the following as it applies to High Drywell Pressure:

QrywelL radiation lev~ls-,-___ ~""_----'-""""~ __~' __ "' ~~ __.~..__"

REQUIRED MATERIALS: NONE --""-"--""- -" -------~~-- .""--~ ..~-

Notes and Comments: Used ol1_~OO~NRC SRO exam for'"" the same KIA {~~§QtO_&A2.03L 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20 II

78. 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 180 degrees F and rising

RHR loop 'A' is in Torus cooling; loop 'B' is unavailable

Torus pressure is 6 psig and slowly rising

Torus level is 16 feet and slowly rising

HPCI is injecting at 5000 gpm Which one of the following describes the required action and the reason for taking the action?

Portions of T-102 "Primary Containment Control" AND T-117 "Level/Power Control" are PROVIDED ON THE NEXT TWO PAGES.

A. Perform Emergency Blowdown per T-1 ] 2 due to inability to maintain RPV level above -195 inches.

B. Reduce RPV pressure to less than 900 psig in order to maintain on the safe side of T/L-l "SRV Tail Pipe Limit".

C. Perform Emergency Blowdo'WTI per T-112 due to being on the unsafe side ofT/T 1 "Heat Capacity Temperature Limit".

D. Reduce RPV pressure to less than 900 psig in order to maintain on the safe side of T/T-1 "Heat Capacity Temperature Limit".

2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 T-I02 "Primary Containment Control" "SRV Tail Pipe Limit" Curve I CURVE T/L-1 SRY TAll PI PE LIMIT I 17.8 I

I

...~*

17.2

,I CONTINUE EFFORTS TO I

I

16.4

~

w 18.8 18.0 RESTORE TORUS LEVEL

w 16.8

0 I.... lIL-21 I 16.2

I

,~ (YES)

I....

...'"0 14.8 I*....

0 200 400 800 800 1000 I 200 RPV PRESSURE (PSIDI 6 (NO) I....

1£ T-101 HAS NOT ALREADY BEEN ENTERED.

1. MANUALLY SCRAM THE REACTOR USING GP-4

2. ENTER T-IOI ANO EXECUTE IT CONCURRENTLY WITH THIS PROCEDURE ~ ____ ._.~ I....

3. PERFORM RPV DEPRESSURIZATION PER T-1 01 I.... '-T/ l . - 2 - - - - - - - - - '" , - - - - - - - - - - - - ' V RC-l I

T-I02 "Primary Containment Control" "Heat Capacity Temperature Limit" Curve I

I CURVE 111-1 I

lIEn CAPACITY TEMP UNn I.,. . .* - - - . . --****--****-~*

SAFE !lEGION IS BELOW THE CURVE 2BO.,.....-,......,..-..,-...,.-,..---.,

%fi0 t: .!.E REGARDLESS OF RX POWER, ENTRY INTO T-116 OR STEAM COOLI NG

} 240 290 RPV PRESSURE o TO 74.9 PSI G I SECTION OF T-111 IS 1H!.I REQUIRED, 220

=:: THEN LOWER RPY PRESS TO NAUTAIN AN -/ v::

OPERATING POINT BELOW CURVE T/1-1 210 w 75 TO 289.8 PSIG EXCEEOING COOLOOWN RATE LIMITS IF NECESSARY ...'"

II<

200 300 TO 488.9 PEID A

L -______________~-------------- /' c r I.... TIT-8 /'

III 180 180 600 TO 898.8 PSI 0 700 TO 899.8 PSIO

!! 900 PSI 8 DR ABO VE

/' 170

/'

/' '"iii! HIO

/' /' e I

/' 160 IS I.... (YES) IOJi 11 IZ 13 14 TORUS lEVEL (FTl 16 17.1

....... ~ I....

PERFORM AN EMERGENCY BLOWDOWN 1...-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-'

USING T-112 V EB-l I.... T/T-IO 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 T-117 "LevellPower Control" (YES) l L..I 8 >-----.. ~:

l - (NO)

~

i t

TERMINATE AND PREVENT RPV INJECTION USI NG T-240 (A TT ~ 1 ~ FI G. S) l - LQ-21 PERFORM AN EMERGENCY SLOWDOWN USING T-112 _.F-DI>l-l- LI-22 V EB - 1 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201 1 Answer Question '# 78 SRO r"--"-"""~-~""~-"~-**-~-"--"-~-"-----," "-""-

Choice Basis or Justification Correct: C For the given conditions of torus level at 16 feet and torus temperature at 180 degrees F, the HCTL curve has been exceeded. This requires an emergency blowdown per T-102, step TIT-9 and TIT-10, which require an blowdown if unable to "maintain" below the curve. i Distractors: A While RPV Level is below -195 inches, it is only 5 inches below band and is I rising due to HPCI injection. The criterion for T-117 LQ-20 is whether or not level can be restored and maintained above -195 inches, which it can.

T -112 is not warranted under these conditions.

B Plausible since Torus level is -1.8 feet away from T/L-1 limit and level is rising slowly. Reducing pressure for the purposes of maintaining this curve is not warranted at this time.

D The HCTL curve has already been exceeded, requiring an emergency blowdown. Plausible if applicant incorrectly plots/interprets the HCTL

'-." ____* __" _ _ _ ~ __ *._ _ . _.. _....L~. _l"_~~~~~~~~~:~~j~~~~e~d_~cin~ . pre~sur~_~~~es~~re~~er~~~n_t~ t~~_sa~_~i~e~of J Psychometrics

,Le'{~Ls~Ll5no~I~_qfl~, I HIGH i Source Documentation Source: D New Exam Item r8J Previous NRC Exam: (PB 2008) r8J Modified Bank Item 0 Other Exam Bank: 0

,t8JJLT E~amBank r--'~----"""~'~----i-T -102 and Bases Learning PLOT-PBIG-21 02-5a Objective:

KIA System: 295037 - Scram Condition Present and SRO Reactor Power Above APRM Downscale or 4.1

, Unknown """--"-""""","'-"""""

K/A Statement:

EA2.04 - Ability to determine and/or interpret the follc,wing as it applies to Scram Condition Present and Reactor Power Above APRM Downscale or Unknown: _?l.!2Qr~si0r1_2()()L!~ITI£er~tl.lre.

REQUIRED MATERIALS:

Notes and Comments:

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20 I ]

79. Unit 2 was operating at 80% 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 Mlbmlhr

Indicated Core Flow (FR-2-2-3-095 black pen) is 51 Mlbmlhr

APRMs are oscillating between 39 and 43% 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" due to indications of Thermal Hydraulic Instability C. (1) Region 2 (2) exit Region 2 by raising '2A' Recire pump speed using SO 2A.l.D-2 "Operation of the Recire 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" 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Senior Reactor Operator NRC Examination January 2011 AO 60A.1 2 Rev. 0 Page 8 of 11 ATTACHMENT 1 PBAPS BAC KUP STABILITY SO LUTION POWER FLOW OPERATION MAP I ',,*'~'~' """'"'"*1 ",,, "''''''''''T~' "*T"""""""''''' ' " ' . " " , " " " ' ' ' ' ""."T'"~'-.'r '~""""*~I''' ""'-~,"~"""""'.'''r 3d, 40: 50: 60 : 70 80 90 100110 120

¢ore Flow (tn Iblhr)

<--- 't"'---------I------- -t -------t----------I- ----"1---------t--

...£ ... - - - I. _ _ _I

__ _L_,, _______ L_____ ___ , ____ _________ L _____ 4 _________

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------ t*

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. fit _

~ ____ ~ ~

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......----+--~=-~~ ..

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~- T ----------

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~---- ---~-- # ---~- '"

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I,\) APRM SC:RAM b.65WD + 63.2 0" , ' /'

3: - ..

- - ~.. - ,., .. - -

_______ L __

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APPX CAVITATION tN-;RlOCKILINE 1

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,~ I ! ! ! I

~--- * ~-----t----- --l------- l- ------ -l q '

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qECIRC, PUMP SUCTIONiC.;\ I-.~TION

__ QI3Qtl_L"lQN_L.lN_E_ _ _ _ L_ ~ _.J _________ .L_ _ _______ .1___ __L I ___ J ___ _

I ",. I ,,1 I I I I

...,." I"",tfo I I I I

.. ... - .r "" TION CAVITAH:~N o .. i ~ , .. I .,

o 50 100 II C lJO Core Flow 2011 NRC SRO Written Exam Rev_ 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer j Question ;# 79 SRO Choice Basis or Justification Correct: The calculation of core flow 51-2(5) =41 Mlbm/hr 1102.5 Mlbm/hr =40%

(alternatively, 4'1 Mlbm/hr can be found on the upper 'x' axis). Plotting 41 Mlbm/hr vs. 39-43% power shows the reactor is operating in the normal

.oQera!il}Q.r~iC)'!..IheJC)iJe>..V\I-:up~p!i()ns of OT-112 ar~r~ctLlir~d~ __

Distractors: A If a core flow calculation error is made, the applicant could believe the reactor is operating in Region 1.

t-*-***----**** -.---... - ..- . - - -...--- ..- - - . - - .

B If a core flow calculation error is made, the applicant could believe the reactor is operating in Re'gion 2. The indications provided do not meet the !

criteria for THI, although inserting GP-9-2 rods would be a correct action if QP~§tin9.jl"l Regle>..n.L. _______ _

C If a core flow calculation I~rror is made, the applicant could believe the reactor is operating in Region 2. Raising recirc pump speed would be a correct action iLQP~r:t!it'l9 !1l..13~gion 2 without indications of THI.

Ps chometrics Level of Knowle.Qye____ _____ Diffip.!Jl!y__..__...__.. __+_ Time Allowance SRO HIGH Source Documentation Source: o New Exam Item [8J Previous NRC Exam: (PB 2008)

[8J Modified Bank Item o Other Exam Bank: 0 J81JLJ...Exalll~Jlk

!3efeIenc~( s L __ OT-11 AO 60A.1-2 Learning PLOT-PBIG-1540-3, -4 Objective:

KIA System: i 295001 - Partial or Complete Loss of Importance: SRO Forced Core Flow Circulation 3.8 KIA Statement:

AA2.01 - Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation: Powerlflow ;'--'..:_':'C... L.:.._. _____.._.

REQUIRED MATERIALS: AO 60A.1-2 "PBAPS Backup Stability Solution Power Flow Map" (whiteout "immediate exit" in u er left corner) __ (im~e_ct..d_~~I)_

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 Qpe~Citjon_M~J!_whic::h is an "immediate C)P~r~!()i.C1~tiQn~.C>LQI-11~. _

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

80. The following conditions exist:

Both units were initially operating at 100% power

A complete Joss of offsite power (LOOP) occurred

SE-ll "Loss of Off-Site Power" is being implemented

Attachment U "Opening Secondary Containment Doors to Support Long Term HPCI I RCIC Operation" is required since HPCI Room Cooling is NOT available For these conditions, which one of the following statements is correct regarding HPCI and RCIC operability per Technical Specification Bases?

A. HPCI is operable without HPCI Room Coolers available.

B. HPCI is NOT operable without HPCI Room Coolers available.

C. HPCI is considered operable once Attachment U is implemented.

D. Implementing Attachment U also causes RCIC to become inoperable.

2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Question # 80 SRO Correct: A . Per Tech Spec 3.7.2 Bases, "ESW provides cooling to the HPCI and RCIC I room coolers; however, Goaling function is not required to support HPCI or I RCIC System operability." Per SE-11 Bases for implementing Attachment Distractors: *~B-~I ~~;.~~~~~::~I;.r~.~r:n~O~.~~~ir:;::s~~;~:~~~~~:~c~~:~~~~t required *1 to support HPCI System operability. Plausible if applicant does not recall C *1 ~:;:~~1c;:~:'*i!r ~;~;~;ntr;;gAttact;ment U, *, . room coolers are not required to support HPCI operability." Plausible if applicant believes

opening RCIC room doors into the HPCI room (Le., provide cooling from D1'~~~~;:;::~!:~:"~;Sr!~~;:r:~~~:~~~;i~;t:::;;~~~=~.

Plausible if applicant believes opening RCIC room doors into the HPCI I room (increasing load on RCIC room coolers) also causes RCIC to become

.__ .~~lD.oper~ble .......

Psychometrics r- Lev~lPiJS..nowl~29.EL*l __***_*_Diffi(;ulty .. __ .... ._Ti'!le AII{>v"anc~Jl'!1inU~~l.f_._. __ ..§J~O....

MEMORY i

10CFR55.43(b)(2)

Source Documentation Source: rgj New Exam Item 0 Previous NRC Exam: ()

o Modified Bank Item 0 Other Exam Bank: 0

. 0 ILT Exam Bank B~f(;}~~.(;e(iL=--~~~E-11*~~~~Tech.§Qe.c 3-L2- E!ases~I.BM 3.11 Bases Learning PLOT-5033-9

~~:::~.-**J*295*003~P~rti;!o~

-_.. ~.----- . . .-

Power

C~~plete Loss of A.C. Importance: SRO 4.2 ......... _ ........ .

KIA Statement:

G2.2.25 - Knowledge of the bases in Technical Specifications for limiting conditions for operations and

~ftllit li'!li!~:___

REQUIRED

-~,.-- ..

MATERIALS:

~----.--~.------.--

_.. "-~l""

NONE

~

~~----- ....- - - -.. - ,

Notes and Comments: .~.. _.. ... __ ........ . .....__.....

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

81. Unit 2 was operating at full power when a small break Loss of Coolant Accident (LOCA) occurred. The following conditions currently exist:

Torus level is 17 feet and rising

Torus pressure is 9.8 psig and rising

Drywell temperature indicated 165 degrees F before TI -80146 "Drywell Bulk Average Temperature Indicator" failed

Based on T-I02 "Primary Containmt:nt Control" NOTE #27 below, the crew attempted to perform a manual calculation of Drywell Bulk Average Temperature using RT-O-40C-530-2 "Drywell Temperature Monitoring" but the calculation was invalid t::::::\#l27 I.E. TI -8014S( 90148) I S OUT OF SERVI CE. lHE1l USE

~ RT-O-40C-530 TO DETERMINE OW BULK AVG~

Evaluate these conditions to determine the appropriate action related to spraying the DrywelL A. Do NOT spray the Drywell since the safe side of the DWSIL curve cannot be verified per RT-O-40C-530-2.

B. Do NOT spray the Drywell since Torus level may rise above the limit ofT-102 "Primary Containment Control" for spraying the Drywell.

C. Spray the Drywell per T-I02 after verifying the safe side of the DWSIL curve using TI-2501, Point 136 plus 19 degrees F.

D. Spray the Drywell per T-I02 after verifying the safe side of the DWSIL Curve using the hottest temperature indicated on TI-2501, Points 119-127.

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20 II

81 SRO Choice Basis or Justification

---~--~ *... ~ ..- -~--~-----~

Correct: A I RT-0-40C-530-2 precaution 4.2.2 states that if the calculation of Drywell

Bulk Average Temperature is invalid, the safe side of the DWSIL curve cannot be verified. DO NOT SPRAY THE DRYWELL.

i Distractors: B

-'...:...=---::~:~~--.---~-

Per T -102, the Torus level limit for spraying the Drywell is 18 feet. If

. Drywell sprays are required and all other conditions are met, Torus level at

_~... ~___ .___ ~ __ ~._-+--._~.. _.____ _ J 7 f~et an.9Jj§;jn~woul£l_n.QLpre'y~nt sPTayinglhePiY'!V~IL_

c TI-2501, Point 136 (plus 10 degrees F) can be used to calculate approximate drywell temperature for entering ON-120 or T-102, but not for spraJlil"lft!bE3~d~ryyv~~:.....

D Using the hottest temperature from TI-2501 points 119-127 is an acceptable method of determining when to initiate RPV blowdown, but it is

____.. ~_i~n-----=otaccE3.Qta tl lE3 for use on the DWSIL curve.

Psychometrics Source Documentation Source: D New Exam Item [gJ Previous NRC Exam: (PB 2008)

Modified Bank Item D Other Exam Bank: 0

". ___._~. ___ ~~.~_t81J,=:r_Exam Bank ___ .__ ~.

BefE3rE3n~~~L ____~1":::1Q.?JNotE3_!27); RT-O-40C-530-2 Learning i PLOT-1560-11 Objective: .

KIA System: 295028 - High Drywell Temperature SRO 4.3 KIA Statement:

G2.4.20 - KnowleC!~()ftbE3_()perational impU.9ati()ns of EOP warnJI'"I9!,. .9autions andl'"l()!es.

~:;;~~~~~~~~~L1!=-_ JN9~E . -_. _- - .___.

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

82. Unit 2 was operating at 100% power when a Group I isolation occurred. The following conditions are present:

All APRM downscale lights are lit

7 control rods remained at position '48' on the scram

RPV water level lowered to - 30 inches and is now inches

Torus temperature is 105 degrees F and rising slowly

Drywell pressure is 0.7 psig and steady What action is required when Torus temperature exceeds 110 degrees F?

T-117 "Level/Power Control" is PROVIDED SEPARATELY.

A. Lower RPV water level to -60 inches B. Lower RPV water level to - 172 inches C. Maintain RPV water level -195 to +35 inches D. Raise RPV water level; maintain +5 to +35 inches 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 f----------~---- -

Answer-----

-~-

Key -~- ---------~---------------

Question # 82 SRO

-~-----------

Choice

r------------------

-~---------- --~

Basis or Justification Correct: C The conditions given show that an ATWS is in progress per T-101, Note 24.

Therefore, RPV level control is directed by T-117. With APRM downscales are lit, reactor power is < 4%. Therefore, T-117 does not require lowering RPV water level. Step LCH8 directs a water level control band of -195 to

+35 inches (based on anslJ'l~jI}R'N9:_at st(3I?_~g-JJl"____________________ _

Distractors: A Based on answering 'YES' at step LQ-6 (power above 4%) and 'NO' at step LQ-10 (are any SRVs open-plausible if applicant does not recognize that SRVs would be open following a Group I isolation, which is indicated by n rising Torus teml?eratureL~__ _ _ unm ___________ n n __

B Based on answering 'YES' at step LQ-6 (power above 4%) and 'YES' at step LQ-10 (are any SRVs open-expected following a Group I isolation

_______9 nd is indicatE~~:IJ~y!~~!1gIoi~!)_!~mJ:>~rC!!lJre).______ ~___________ _

D Plausible since this is thE! level band directed by T-101, RC/L. However, RC/L is exited if an ATWS is in progress (step RC/L-2). Plausible if appJic;C!nt9Q(3~notrecaIIJ-10LRC/L direction. ____ ~ _____ ~ _____________

Psychometrics

__~~~~I_gLlSnowl~_~_ _ ____ Aifflf_Lllty ________ _Ii!!l_~t.llowan~~f!1il}ut~~L SRO HIGH 10CFR55.43(b)(1 0)

Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 f - - - - - - - - - - - - - ____ __D!~T~~~!!lJ!a_nk ____ ____ ________________ - --------------------

J~~l~renc(3ts L_ ____ T-ltL________ _______~___ ~________ ________________________

Learning PLOT-DBIG-2117 -9b Objective:

KIA System:

f--------~----

- ~=:r~~~pp;ess;;;;;p~~1 Higl1 w;;e;-II,,;p~rt~~;;-

~----------~------~---~----- -- ______________ L__ -

-::0 KIA Statement:

G2.4.21 - 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 system integrity, containment conditions, radioactive release control, etc.

~::::~~::~:~ALS~ t-l1 ! .. L~.,:~r~nkOI~~~

__ ~____________________________ _

__~_=--___ ~ ___

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

83. The following conditions exist on Unit 3:

Drywell pressure is 5 psig

Torus pressure is 4 psig

Torus level is 15 feet

Drywell bulk average temperature is 250 degrees F and rising

Initiation of Torus Sprays using RHR per T-204 is complete Which statement below is correct for spraying the drywell under these conditions?

Drywell Spray Initiation Limit (DWSIL) Curve DW/T-2 is PROVIDED ON THE NEXT PAGE.

Spraying the drywell in accordance with T-102 "Primary Containment Control" A. is required and the resulting evaporative cooling pressure drop must be controlled by throttling spray flow B. is required and the resulting convective cooling pressure drop must be controlled by terminating spray flow C. must NOT be performed since it may result in an evaporative cooling pressure drop to below the high drywell pressure scram setpoint D. must NOT be performed sincc it may result in an evaporative cooling pressure drop greater than the capacity of the Reactor Building-to-Torus vacuum breakers 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 CURVE DW/T-2 OW SPRAY I NITI ATI ON LI NI T 800 --~

...... 560 ell! _.

,/

~SOO -r-.- UNSAFE J

.. Q..

2460 /

w 400 -... Il w

t:II

~360

.L w

300 -

J

,I SAFE

~2&O

)

=200 /

!II Q t50 If 100

/

o 2 4 6 B 10 12 14 16 DRYWELL PRESSURE (PSIG) 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201 1 Answer Key Choice I Basis or Jus.t..i.f.~i~c~at~io_.n___ ~~.__.. . ~~~

Correct: C I Drywell temperature and pressure plot on the unsafe side of the DWSIL curve and

therefore drywell sprays must not be initiated. Per TRIP Bases, DWSIL is the highest drywell temperature at which initiation of sprays will not result in an evaporative cooling pressure drop to below the high drywell pressure scram setpoint. If drywell sprays are initiated while on the unsafe side of the DWSIL curve, the evaporative cooling pressure drop will reduce drywell pres:sure below the 2 psig scram setpoint. The 2 psig limit

provides margin to preclude containment failure or de-inertion following initiation of

_~~ _~~+-_._~._ + <!ryy'{eIL!>pra~: _...

Oistractors: A Drywell temperature and pressure plot on the unsafe side of the DWSIL curve. Per T 102 Bases, initiation of drywell sprays can result in a relatively large drop in drywell I pressure (due to evaporative cooling) that may occur at a rate faster than can be compensated by the operator. Spraying only when on the safe side of the DWSIL i curve ensures this will not occur.

--B~-IDrywell tem~~;~ure and pre~~~;~~i~t-on the ~~saf~ side of the DWSIL curve. Per T 102 Bases, convective cooling occurs when water is sprayed into a saturated atmosphere (Le., after evaporative cooling has occurred). Per T-102 Bases, convective I cooling occurs at a much slower rate than evaporative cooling and can be controlled by

~~'_~~__~"_..... I~_~~ ____.+--=,te-=-:r.:..:minatir1g_sPi(iy'~ ______ . . . _~_~_~~~~ . __ ~__ ____~ __ ...... ~~.~.

o Plausible since one of the as.sumptions for determining DWSIL is that the evaporative

, cooling transient is complete before the Torus-to-Drywell vacuum breakers operate (not I the Reactor BUilding-to-Torus vacuum breakers), thereby ensuring sufficient margin to L______~ __~ ___ ~_~, . . .

_~_J ...ill.gid d~-ir1ertin9...th_~pr:il11a'!y con19inmentlJ'.'~~~~P~ClY!3 arei I11tiated:____ ...

Psychometrics

~_~I?"el of Kn_9~~l?ci9~_~~__ Oifficul~y_~__ Time Aliowance(l11jl1_u!l?~l ------_ ..

SRO HIGH I I i 10CFR55.43(b)(5)

Source Documentation Source: [8J New Exam Item D Previous NRC Exam: 0 D Modified Bank Item D Other Exam Bank: 0 D ILT Exam Bank

~~:~;;(st=i ;~6~;~~~;~Rip/SAMP Curves,_Iables & Lin1jt~ases Objective: I KJA~~:te"'~_~F12-H,ghD~ell ~ra~ure _ Importance: SRO 4.1 KIA Statement:

AA2.02 - Ability to determine and/or interpret the foillowing as it applies to High Orywell Temperature:

~!~~:~=:~~f-0NE ----- ___ _

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

84. Unit 3 was operating at 100% power when a feedwater level control malfunction caused RPV level to rise to +90 inches as read on LI-2-2-3-86.

Current plant conditions are as follows:

All control rods are fully inserted

RPV pressure is 1060 psig and rising slowly Which one of the following actions is required for RPV pressure control?

Figure 1 ofOT-110 "Reactor High Level" is PROVIDED ON THE NEXT PAGE.

A. Restore and maintain reactor pressure below 1053 psig using the Bypass Jack per OT-102 "Reactor High Pressure".

B. Restore and maintain reactor pressure below] 053 psig using EHC Pressure Set per OT-l 02 "Reactor High Pressure".

C. Reduce reactor pressure below 1050 psig using a single SRV and prolonged SRV opening per OT-110 "Reactor High Level".

D. Reduce reactor pressure below 1050 psig using multiple SRVs and short-duration SRV openings per OT-IIO "Reactor High Level".

2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 OT 110 PROCEDURE Rev. 7 Page 6 of 6 FIGURE 1 LI-2(3)-2-3-86 INDICATION (INCHES) 8 SAFE 7

o 100 200 300 400 500 600 700 800 900 10001100 REACTOR PRESS (PSIG) 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer Key I Question #84 SRO Choice Basis or Justification Correct: C When RPV pressure reaches 1050 psig, OT-110, which is executed I concurrently with T-101 "RPV Control", directs manual SRV operation using a single§RV {if pos~ible)and prolol'l~<L$~'L()Q~niI1Jl~~~____~_~___ ~

Distractors: A Per OT-110 Figure 1, an indicated level of +90 inches indicates that actual i RPV level may be at or above the main steam lines. OT-110 directs closing the MSIVs if RPV level cannot be maintain below the bottom of the

. MSIVs (+108 inches), thereby taking away the use of BPVs. In addition, I while OT-102 does direct maintaining reactor pressure below 1053 psig,

. since the reactor is scrammed, OT-102 is no longer applicable. OT-110 is

.___~xecu!~c! c()D~urrel1fu'..wi!h.J-=-101:.__________________________ . . __.. . _

B Per OT-110 Figure 1, an jindicated level of +90 inches indicates that actual RPV level may be at or above the main steam lines. OT-110 directs I closing the MSIVs if RPV level cannot be maintain below the bottom of the i MSIVs (+108 inches). In addition, while OT-102 does direct maintaining I reactor pressure below 1053 pSig, since the reactor is scrammed, OT-102 i is no Ion er a licabl . OT-110 is executed concurrentl with T-101.

! I D OT-110 directs prolonged SRVopening using a single SRV (or as few as I I I . possible) in order to minimize SRV tailpipe loading and the number of I

~_______ -.1 ___l.§!3Vs that are effected b1lJ:!l9.her than normCllJoads_._ _.._____. _ ..____

Psychometrics j~vel oflSnowl~Qg~_ r ______Diffic;.!:I'!l. ____-+_Time_Allo\Alanc~_frI'llI'lLJte~.L_~ ___ ~B~ ____

HIGH

i 10CFR55.43(b)(10)

Source Documentation Source:

o New Exam Item r8J Previous NRC Exam: (PB 2007) o Modified Bank Item o Other Exam Bank: 0 I

r8J ILT Exam Bank Refi:;n::;dce(s): OT-102; OT-1l0; T-101 -

Learning PLOT-1540-4 Objective:

KIA System: 295008 - High Reactor Water Level Importance: SRO 4.2 -------- -_ ..

KIA Statement:

~~6~,~;~~O;:;~iA~:bnE-m1To~_~;~O~:~:;d~::~ded-)-**--**-------~.--.~~.-

f - - - -...- -...- -..- -..- - - -..---.-.. . . - . -..- . -.. --~--------- ..- - - -.. ---~.--~- ..- ... -.--.

Notes and Comments:

..- -...- -...- -...- - -... -.~-... ...- ..--...- . - - . - - - - - -..- ..- - . -..- - - -.. --~

~ .. ~

2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

85. Given the following:

Unit 2 is in MODE 5

In vessel fuel moves for Core Shuffle I are complete

Control rod drive mechanism removal and replacement is in progress

Refuel Floor exhaust ventilation radiation monitors RIS-17 -4 5 8C and RIS-17-458D are determined to be inoperable What Technical Specification action, if <my, is required in order to continue the current activities?

Technical Specification 3.3.6.2 "Secondary Containment Isolation Instrumentation" is PROVIDED SEPARATELY.

A. No actions are required; these monitors are NOT required for these conditions.

B. Place both channels in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or take the actions required by Condition C.

C. Place both channels in trip within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or take the actions required by Condition C.

D. Restore isolation capability within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or take the actions required by Condition C.

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer Question 85 SRO Choice Basis or Justification Correct: into Tech Spec 3.3.6.2 Condition A is required for Function 4 on 3.3.6.2-1, which requires placing both channels in trip within 24 i hours. This is because 1 channel in each trip system is inoperable but 1_ . ' ~~i:~~~~~~!I~n~m~~n6a::~~~,aann~ ~e:::~I!n~Vha~~~I:~o~;~t~~~;r Distractors: . . . . i ~A--l ~~~n~~~~:i:e;:::i~~::~~~d~n the applicant beli~~ingthe r;diation - ~- .

I monitors are not required while in Mode 5 and does not correctly apply sub t

~--~-. . ---~....... ~B-* ~~i: ~~~~c:~::II:~~~~~~~~;dt:~ ;~:e~:;;i~:~~0~;;i~~i~9 that 1'~hannelin each trip system is inoperable but isolation capability is maintained, and incorrectly applying Condition A to Function 4 of Table 3.3.6.2-1. The 12

'--_ _ _._.. _----+___... --+..:.-h...:.o...:..:u......

r .....:re:....:.tlJirement applies to Functions 1 and 2 only.

D

This choice is plausible for the reasons stated above, and if the applicant

. believes that 1 channel in each trip system is inoperable and that trip maintained.

Psychometrics I

_~eyel of Knowl~dg~~J_. . . . DifficLJI~_. _ _ L Time~lIo\'yance (f!1inutes) SRO HIGH 10CFR55.43(b)(2)

Source Documentation Source: o New Exam Item 0 Previous NRC Exam: 0 lSI Modified Bank Item Other Exam Bank: 0 ILT Exam Bank ____ .. ___ ._______ _____

~=~~ ~ .o l3eference(Si)_:_ "Le9~ Spec~.:~:(5.".2_~ _ _ ....__ . 0 * * *0 * * * .

Learning

PLOT-5040B-8 Objective:

KIA System: 295033 - High Secondary Containment Importance: SRO Radiation Levels 3.2 KIA Statement:

EA2.02 - Ability to determine and/or interpret the following as it applies to High Secondary Containment Area Radiation Levels:

REQUIRED MATERIALS: 3.3.6.2 for Unit 2 Notes and Comments:

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

86. Unit 2 is operating at 100% power when the following occur:

BLOWDOWN VALVES POWER MONITOR (227 C-5) alarms

The GREEN indicating light for the 'C' and 'D' SRVs are NOT lit

The GREEN indicating lights for ALL other SRVs are lit

Subsequent investigation identified blown fuses associated with the

'c' and 'D' SRV solenoids What is (1) the impact on the ADS System, and (2) the Technical Specification action required for these conditions?

Technical Specifications 3.4.3 and 3.5.1 are PROVIDED SEPARATELY.

A. (l) NO ADS valves are inoperable (2) Be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Mode 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months B. (1) ONE ADS valve is inoperable (2) Restore ADS to Operable status within 14 days

c. (1) TWO ADS valves are inoperable (2) Be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to :S 100 psig in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. (1) TWO ADS valves are inoperable (2) Restore ADS to Operable status within 14 days, or be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to :S 100 psig in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer Choice Basis or Justification Correct: B given conditions show a loss of power to the solenoids for SRV 'C'

'D'. 'C' SRV is an ADS-SRV; 'D' is a non-ADS SRV. Therefore, 'C' is inoperable per Tech Spec 3.5.1, requiring entry into Condition E for 1 ADS valve inoperable. '0' SRV remains operable per Tech Spec 3.4.3 since Tech Spec Bases for SR 3.4.3.2 states: "If the valve fails to actuate due only to the failure of the solenoid, but is capable of opening on

()yerp~~ssur~~~saf~yJunction of the~SR'Lis consid~red OPERA~!J:-,-"__ _

Distractors: A 'I The choice is based on applicant believing two non-ADS SRVs are

, inoperable and require solenoid actuation to be considered operable per 3.4.3.

choice is based on applicant believing two ADS SRVs are inoperable correctly applying the required actions of Tech Spec 3.5.1.

D The choice is based on applicant believing two ADS SRVs are inoperable and incorrectly applying the required actions of Tech Spec 3.5.1.

Psychometrics 1

Source Documentation Source: I [8J New Exam Item D Previous NRC Exam: 0 0 Modified Bank Item 0 Other Exam Bank: ()

D ILT Exam Bank f--Reference(!):___ ~AR~G__ 22?_ C-5;_Iech§p~c;~.~~~ and 3.5.1; _M-1-S-:~ _________

Learning i PLOT-5001A-8, -9 Objective: I i

KIA System: I 218000 - Automatic Depressurization SRO i System 3.6 ---,

KIA Statement:

A2.05 - Ability to (a) predict the impact of the following on the Automatic Depressurization System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of clhQSe liQnormal CO-"diti1opeJations: Loss of A C .. or D.C. power tQ.ADS valves. ___,____"____ ,_

"REQUIRI;D M~TERIAL§: Iech Spe~J.4..!~1*1.c:L~,,5.1 ~ Unit 2 wI NO Surveillance Requirements Notes and Comments:

,--,--,--,--,-,---~,----------,--,-----,---"'-----'---------"--- ,--,

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

87. Given the following:

Unit 2 is operating at 100% power

APRM' 1' is inoperable and bypassed

APRM DOWNSCALE (211 C-2) is received

The ODA for APRM '4' indicates downscale Which one of the following describes (1) how these conditions impact the Reactor Protection System and (2) what action is required?

Technical Specification 3.3.1.1 "RPS Instrumentation" is PROVIDED SEPARATELY.

A. (l) An RPS channel trip will occur (2) Perform AO 60F.2-2 for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months then, if operability is not restored, place ARPM '4' in the tripped condition using GP-25 B. (1) An RPS channel trip will occur (2) Perform AO 60F.2-2 for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months then, if operability is not restored, place ARPM '4' in the tripped condition using GP-25 C. (l) An RPS channel trip will NOT occur (2) Place ARPM '4' in the tripped condition using GP-25 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months D. (1) An RPS channel trip will NOT occur (2) Place ARPM '4' in the tripped condition using GP-25 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months NOTE:

AO 60F.2-2 is titled "Defeat of an RPS Half Scram"

GP-25 is titled "Installation of Tripsllsolations to SatisfY Tech Spec/TRM Requirements for Inoperable Instrumentation" 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer

87SRO Choice Basis or Justification Correct: 0 An RPS trip will not occur since (1) an APRM downscale condition does not cause an RPS trip (rod block only) and (2) the 2-out-of-4 logic (voter) modules must see I two of the same trip input signals to cause an RPS trip. With APRM 1 bypassed, a ny trip signals generated by APRM 1 are removed. Tech Spec Table 3.3.1.1-1 requires 3 APRM channels per trip system in Mode 1. Since only 2 channels per t rip system are available, Condition A applies, which requires placing the channel (APRM.f) in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This is done using GP-2~Appenc:Jix 24.____._

Distractors: A Part 1 is incorrect; part 2 is correct (if a scram were to occur). Plausible if applicant believes the given conditions are sufficient to cause a trip of the 2-out-of-4 logic .

~----.

(' modules, which if were true, would result in a reactor scram.

_ _~C*_* _ _ '

_ _ _* _ _ _ _**_ _ _*_ _ _ _ _' _ _' _ _ _ _ 1 I

B i Part 1 is incorrect; part 2 is incorrect. Plausible if applicant believes the given

.conditions are sufficient to cause a trip of the 2-out-of-4 logic (voter) modules and C

---LtPart he...

trip ....resulted

_- -_ in .

a __

half scram vice a full scram, which is not the case.

1 is correct; part 2 is incorrect. Plausible if applicant believes Tech Spec 3.3.1.1 Condition B applies. This might occur if applicant incorrectly applies note

'c' on Table 3.3.1.1-1 (each APRM channel provides inputs to both trip systems) and misses the note for Condition B, which states the condition does not apply for Functions 2d or 2f.

Psychometrics

~evel Qf KnQwledg!L_ __....__Diffic':I!!L _ _ _ ._Ti~~Allo\}.lanc~ (minu!.~LL _____ ..§.~.9 ______

HIGH I I 10CFR55.43(b)(2)

Source Documentation Source: IZI New Exam Item 0 Previous NRC Exam: 0 Modified Bank Item 0 Other Exam Bank: 0 ILT Exam Bank

~~;: e(s2_:-1-PL-OT_.5~6b~~8Tech--~Q.~£..~.:..3.j.J;..GP..: 25;-t'060F:?-2 ___________._ ... -..---".

Objective: .

KIA System: 212000 - Reactor Protection System SRO 3.7 KIA Statement:

A2.04 - Ability to (a) predict the impact of the following on the Reactor Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those

. . abrl.Qrm~lconditi~r1~Q!"_Qf>.(7rations: ~ucle~LinstrumenL!)ystem fajluf~ ___.__

~:~;~~~~:::;:~~~

' -.. ~ .. -"--"'--"-"'~--~~...

1Tec~

-~

Spec 3.3.1.1jUnit 2_)-

...- - . - - -..- - - . " ' - -.. -~----

. ----~======_=~_~ =

..- - -... ~--.

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

88. Units 2 and 3 are operating at 100% power with the following conditions present:

The 'A' SBGT Filter Train is INOPERABLE due to water intrusion

The '8' SBGT Filter Train is unaffected and OPERABLE

The damaged filter is expected to be returned to service within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months

A Prompt Investigation has been initiated Which of the following are required, if any, for these conditions?

(1) Prompt notification to the NRC due to "Loss of System Safety Function" (2) Notification to the NRC as a "Condition Prohibited by Plant Technical Specifications" (3) Notifications per OP-AA-I06-101 "Significant Event Reporting" A portion of Technical Specification 3.6.4.3 "Standby Gas Treatment (SGT) System" is PROVIDED ON THE NEXT PAGE.

A. (1) and (3) ONLY B. (2) and (3) ONLY C. (3) ONLY D. No reports required 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Std Gas Treatment [SGT) stem LCO 3.6.4.:) Two SGT subsystems be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment, During operations with a 81 for draining the reactor vessel f DPDRVs L REQ:JIRED ACTION A. One SGT 5 A.I Restore SGT subsystem 7 incperable. to OPERABLE status.

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201 1 Answer Question # 88 SRO Choice Basis or Justification Correct: Initiation of a Prompt Investigation requires notifications lAW OP-AA-106 101 (see Attachment 1, Page 1 of 2).

Distractors: A SBGT is common to both units, however only ONE filter train has been impacted. Per Tech Spec Bases, since the filter trains share a common inlet plenum, each unit is allowed to claim the remaining filter train for

!_~_~ ~ __~~~_~~~~~ __ --+_~_~ ____-+~uCJr=-:o,-=s~s of s~tisfyl!!9Jb_e LCO. The safety function_~as ti9T beerl}Qst.

B SBGT is common to both units, however only ONE filter train has been impacted. Per Tech Spec Bases, since the filter trains share a common

. inlet plenum, each unit is allowed to claim the remaining filter train for i purposes of satisfying the LCO. 7 days is allowed for restoration of the

____ ___ ~,------+inoper@l~e slJ~syster.:n_.__~_~~ __~ ____ ~ ___ ~ ____~ __~_~____~~~~_~~ .

D Initiation of a Prompt Investigation requires notifications lAW OP-AA-106

\ 101 (see Attachment 1, Page 1 of 2).

Psychometrics

_Levelo::~'Nle~e__ I______ DiffiC::LJl!L_ --~~~

Till1~AllolNanceJ.1l11nut~~LL ______SRQ______

i 10CFRSS.43(b)(S)

Source Documentation Source: [:gj New Exam Item 0 Previous NRC Exam: 0 I o Modified Bank Item 0 Other Exam Bank: 0 D ILT Exam Bank

-Refer; ncels): _rU-~t2& LJDij 3 T~9h spe~ 3.6:4.3 aQ9 B;~es; O~=AA-t06--101; LS=AA=1020--__

Learning

PLOT-S009A-8, -9 Objective:

KIA System: 000 - Standby Gas Treatment System I Importance: SRO I

i 4.1 KIA Statement:

G2.4.30 - Knowledge of events related to system operation/status that must be reported to internal

..QI9anizations.Q!'extern§ILagerlcies,_~uch as the Stat~, the NRC, ~r the transmissioD_system ope~ator. __ _

REQlIIREDI\IIATE~IA!-~~-TN-ONE ~~_~~~_._.__~~_~_ _~ ~ ~ ~ _ _~~~~ __._. __ ._.. __

Notes and Comments: ~sed oDElant9P~X fro.!!1MC'i:LQf.1Q03 nR~160?84:l_.~~.~._... __._. __

2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

89. Given the following conditions:

Unit 2 is operating at 100% power

The 2A RBCCW pump is blocked for maintenance

B REACT BLDG COOLING WATER PUMP OVLD (216 B-1) alarms

The temperatures for components cooled by RBCCW are beginning to rise Which one of the following describes the actions required by ON-113 "Loss of RBCCW" for the current plant conditions?

A. Shutdown the running R WCU pumps and lower power per GP-9-2 "Fast Reactor Power Reduction" to reduce RBCCW System heat load.

B. Perform GP-4 "Manual Reactor Scram" and trip both Recirc pumps to prevent exceeding Recirc pump motor bearing and/or seal cavity temperature limits.

C. Trip one Recirc pump and execute OT-112 "UnexpectedlUnexplained Change in Core Flow" concurrently to prevent exceeding Recirc pump motor bearing temperature limits.

D. Reduce the speed of both Recirc pumps per GP-9-2, remove one Recirc pump from service and enter single loop operation per GP-5 "Power Operations" to reduce RBCCW System heat load.

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer i Ot ,,'89 SRO Choice Basis or Justification Correct: A i Per ON-113 Step 2.2, if restoration of RBCCW is not imminent, RWCU I pumps are shutdown and reactor power is lowered in accordance with GP i 9-2. Per ON-113 Bases, the reason for these actions is to reduce the heat

load on the RBCCW syst.em and allow more time to diagnose and correct

-~:r:;:~v:~t:~~~~~;!:~:t:~ ispre;at~re andd~-;-notfct(;;;the-Distractors: B I ~uidance given in ON-113. Direction to perform GP-4 is only given if it is necessary to trip both Recirc pumps, which is based on reaching certain i

1 r--------:~-c-- i~:0:~~:~i~s9n:~~~:e~e:~~:~t; ~:;r:r~~~~I:~::r b~~rin~emp~~tur~

exceeds 194 degrees F.

This action is not taken unless a Recirc pump seal temperature exceeds i 200 degrees F, and lowering pump speed does not reduce seal

. to below 1 F.

Psychometrics

~1J_el of~Knowled~_~(_____ DiffiCtillY ____ ~. .~me Allowanc;~lll1jnutes~__. ____ SR.Q_____ ~~

HIGH' I

10CFR55.43(b}(5)

Source Documentation Source: cg] New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item D Other Exam Bank: ()

ILTExam Bank 13 and Bases Learning PLOT-PBIG-1550-3, -18a, -18b Objective:

KJA System: 400000 - Component Cooling Water Importance: SRO 3.0 KJA Statement:

A2.03 - Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal

~0_;Q_~~~~~~.~~_r_:_!;_~_a~_i~~~.~:-i9_h./lN.C.6-:-E-.-.te-m-p~r-a-ture.-_--_._.-=-.=_. ~=_-. _

. -.------.~-=..- . _.._~-~._._..._. ~__--_~_~_-_.-. _. _._.. .---...

Notes and Gomll1ent~.:. _____~____ ___~. _ _.______._ _ _ .... _ . ______ ._.___ ... _.._....~._.

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201 I

90. The following conditions exist:

Both units are operating at 100% power

Offsite electrical sources are in a normal lineup

3SU Transformer Load Tap Changer (LTC) is in MANUAL

The Main Supply Fan for the E-l Diesel Generator is blocked What action, if any, is required for these conditions?

Technical Specification 3.8.1 "AC Sources-Operating" is PROVIDED SEPARATELY.

A. Enter Tech Spec 3.8.1 Condition A for both units.

B. Enter Tech Spec 3.8.1 Condition B for both units.

C. Enter Tech Spec 3.8.1 Condition E for both units.

D. No Technical Specification action is required.

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer Question #90'SRO Choice Basis or Justification Correct: B i I

Based on the given conditions, the E-1 diesel is inoperable. Tech Spec

' 3.8.1 Bases requires the main supply fan to be operable in order to I* consider the diesel generator operable. Although the 3SU transformer is inoperable when its LTC is in manual, per Tech Spec Bases, this is not a i required offsite source. Since there are two qualified offsite sources f-_ _... _ _ _ _.... _ _ -+.... _._.__..-I--a_v..~ailable ,_I~ch SJ)~c 3:..§.:1 C(mditiol'l~t!J~..oGi~().JJerC!QI~L§!Q2Ii~~~.. ~ ......._.

Distractors: ,

I I.

A Based on the given conditions, two qualified offsite sources are available.

Plausible if the applicant (1) does not recall the EDG main supply fan must be operable for its EDG to be operable, and (2) believes the 3SU source J I must be aligned with its LTC in automatic to meet Tech Spec 3.8.1 r - * - - -.... ---I-ciaas~~e~:~;giVen-;;onditi~ns, tw~ qualif,; d off;; e sou;-ces--;;re available-:-- !

~""~-~'---i-"D~-~:~~~~~:~:~i:!!:¥!~~'~!~;~i~~:;:s:~~I:I::~:~;~:~ j

! ' _~tf~~~~~:;;~~I~~~aUthe_ED~ m~insuP~~fan :~st b~.~per~~~_f~~ .i~SJ Psychometrics

... Level of ~nOwlecl.9..~_J~ _ _DiffictillY . . . Tim~_,il.llowCln~~ (fI'liD..Yt~~) SRO HIGH i 10CFR55.43(b)(2)

Source Documentation Source: rgj New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0

0 ILT Exam Bank I - - - . - - -..... - - - ....~... - - - -...-------~.... - -..- - - -....- -...- .. ~-..- - . -....- -.... ... --~ ....~----

Referer1ce(s): . . _---+-I~ch_..§pec 3.8.1Cl...rI.9J?ases~__~:.L ___.. .__.. _.. ____.. _._.,

Learning

PLOT-5051-8, -9; PLOT-5052-8, -9 Objective:

KIA System: - A.C. Electrical Distribut.ion Importance: SRO 4.7. , , - - _ . _ - - -

KIA Statement:

G2.2.22 - KQPwledge of limiting conditionslor oJ)eratl()ns anc:Lsafety !!mits. _______ . _ ..._.._ . .__

R!=QUIRED MATEBlALS,--- -trec!!Spec 3.8.1 for BOTH Units_ .._ _ _ _ ........._____.. _.. __ ....___

J::JotesandComl1J~nts: ______ ___ . ___________ . _________.. _ ___________

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

91. Unit 2 was initially at 100 % power when the following transient occurred:

Feedwater Heater String A3/4/5 isolated

Reactor power rose to 106%

Core flow is 91 %

Thermal LimitIFLLLP values are as follows:

o MFLCPR - 0.979 o MFLPD-0.955 o MAP RAT 0.945 o FLLLP - 1.009 Which one of the following describes (1) the impact of these conditions and (2) the required action to be taken?

GP-5-1 "PBAPS Power Flow Operation Map" is PROVIDED ON THE NEXT PAGE.

A. (1) Core operation is outside the analyzed region of the Power to Flow Map.

(2) Reduce reactor power lAW GP-9-2 "Fast Power Reduction" to less than 90%

per OT-104 "Positive Reactivity Insertion".

B. (1) Core operation is outside the analyzed region ofthe Power to Flow Map.

(2) Reduce reactor power by insertion of GP-9-2 Appendix 1 rods to less than 90% per OT-104 "Positive Reactivity Insertion".

C. (1) A Thermal Limit / FLLLP violation has occurred.

(2) Reduce reactor power using Recirc flow IA W GP-5 "Power Operations" until FLLLP is less than 1.000 per GP-13 "Resolution of Thermal Limit Violations".

D. (l) A Thermal Limit / FLLLP violation has occurred.

(2) Reduce reactor power lAW GP-3 "Normal Plant Shutdown" such that thermal power is less than 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Exhibit GP-5 1 Rev. 6 1 of 2 NHN;nhn PBAPS POWER FLOW OPERATION MAP 130 40: 50: 60: 701*10 : 20

_____ <tore

~ ___ -Flow _ __I(t!lHbfhr)

_________ ~ I ________ I _________1_________ 4 - ________ I ____ _

~ ~

120 I I I I I I J 1 I /IIIIf" _ ..l MIt .. ,. _ f"", _ _ .. ...J I I I... I I I i I -, 1 ! I 110 --------+------- -----~--;-~-~ --~--------~---------~--------

I ~

til ... : . .

I .; fill' I

+ - ..... :

100 ----- . . ---!- ---- -~--- - -- -,..---- - ---..,- - ---- - --T-- ------ -, ..!"- - ---- ~ - -,-!!- ---~:;-,.----....,.---'I""-~

I I 1 I ~ .,. I I ., I ... ~"

I I I I filii *' ,. fill I ... ~ I" '" " ~ ... ,.

________ -l ________ _ .L ___ _ _ _ _ _ _ L. ________ ..J_

~, _ ___ Ji.__ _ _

i __J... __ "'__..__"

~: . ~:~~

_1 _______ .... .::

90 I ., I l,oj" ,," I ,,"

I .. I I I ~.j. I I 'II ' I ill j I _...... I ",,,,r""

p _dIJlf' ~jr ~

-:----- -;-'"-:- ----;.. . .!'t---- ..~'-'~.:.:-- ---i-I I , , I I ... ., ,

S!G 80 :--~~i---.:.:-

-IX

~

0 70 ---i- ___ n -~-u------~-/~-7-"J~-----

63,~.i ",'

, I.. .. I I , .,'

!:. '-.. .,-.:.:~~----.ca~~:::'-r--

I t1.VJ"'*

~. -- I:

I I

__ un_

I..- - - ----'

... S~RAM

!0 Q. 60

i----;----I' APRM

--- -- _ --~-

O,66WO +

__ - - - - __ ..1. - - - _ - - __ -~ - , - I .. - ... -..:":-

",i"

~ - ---- - - -~ _ :--:-: -.._.r_~_

~"

iii A!R!'l!~~B;-O~K_ ~O,!5!D_+54, , " : "'~~'r' E I

4" * * :

!I 50 i ,1",6

,---------r---

I " I I I ** ' - I I I " ,. I I 40 t~--i

~-~~-+---------I---

! I

~------ I

--~-- I I

..... 1 I I I I I

I I I I 30 - --r-~- --~ - --- .... ~ ----- - ---t-- - --- - - -:--- - ----

A;;P:~" .. j~=

I _I 1_ ., .... _

20 T---------r---- : : : : : :

--~---------r--------,---------r---------I-------- ~----

,C

. ----'-i

/ I I f I

~F"X, liNe I !;cE;I~b o!.!"""

0

~I~: --~----F--~--

...... : ,,/ I

J----

... I

,J,;

I a

Q 10 20 30 40 ~ 10 ro 80 90 100 110 120 Core Flow (% of Rated)

FR-2 3) 3- (Black pen) OR B015 (B315) CM-I 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer SRO Basis or Justification Correct: Per the given conditions, the core is operating outside of the analyzed region of the Power to Flow Map. OT-104, step 3.5.1 direction for these conditions is to lower reactor power lAW GP-9-2 as necessary to maintain APRM power within the analyzed region of the Power-to-Flow Map and at least 10% below the pre transient power level. For the given conditions, this will require reducing Recirc flow to <

Distractors: B Part 1 is correct; Part 2 is incorrect. OT-104 directs reducing power lAW GP-9-2, which will involve an initial flow reduction. Plausible if applicant believes insertion of control rods to restore acceptable point on the Power-Flow map is required, as the case if 2.

C Part 1 is correct; Part 2 is incorrect. A FLLLP violation has occurred, however the action in Part 2 is required by GP-13 ONLY IF unable to restore thermal limits or to less than 1.000 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Psychometrics Source Documentation Source: rgj New Exam Item Previous NRC Exam: 0 o Modified Bank Item o Other Exam Bank: 0 ILT Exam Bank KIA System: 256000 - Reactor Condensate System Importance: SRO

.. ---~ ... -~.--- .. --~ ..- -...* --~-..- . - - - - ...- - -.. - .*..-~~-...-.

2.9 KIA Statement:

A2.17 - Ability to (a) predict the impacts of the following on the Reactor Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conclitions Q!'..QQerations: F~edwater_~eater stri!!9~ ___.____.. .____.......__.~.___ _._. _._. _____.. _

REQUIRED MATERI~!:~-=----~ONE___. _ . _ .~_. . _____ ..._ . _ ._______..____._... __._

Notes anclJ~:;omll1e~t~__ ___ ~I_.

2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

92. A Unit 2 startup is in progress with the following plant conditions:

Reactor power is 25%

Generator output is 200 MWe

Annunciator TURBINE STOP V. CLOSURE & CONTROL VL V FAST CLOSURE SCRAM BYPASS (210 A-2) is lit

A relay failure causes the Power-to-Load Unbalance circuit to actuate

The POWER LOAD UNBALANCE TRIP (206 B-1) annunciator alarms Which one of the following describes (1) the automatic plant response and (2) the correct procedural direction for this event?

A. (1) The turbine and generator remain online; the reactor does NOT scram (2) Perform applicable sections of SO IB.2.A-2 "Main Turbine Generator Shutdown" B. (1) Generator lockout, turbine trip and reactor scram (2) Enter T-100 "Scram" C. (1) Generator lockout and turbine trip ONLY (2) Halt GP-2 "Startup" D. (1) Reactor scram ONLY (2) Enter T-100 "Scram" 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer

92SRO Choice Basis or Justification Correct: C If the PLU circuit (part of EHC logic) energizes, a generator lockout and turbine trip will occur. Since reactor power is < 29.5% RTP (turbine 1st stage pressure is < 138.4 psig), 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 operating at 25% power with the turbine-generator off-line. Per Hu-i AA-104-101, "Procedure Use and Adherence", steps 4.1.5 and 4.1.7, halting progress on

_ ~.___.. ~_ ---1----- _~hv~~::X~f~~~~:~~=~i~~~t~~c:e~i~~~~ ~~~~~:~~sWhiCh:oul~~ecessit~~~~~~ ___ ~

__~istractors: -~---J ~~~;~~r;c~~t.~~~p~~~~r;u~~!;::~~~~~k~t an~ turbine trip. _Plausible if applicant does .

I: . I B The reactor does not automatically scram. The Recirc pumps will not trip either since house

~

. . . I.

. .1. loads are not trans~.erred to the unit auxiliary transformer unt.il. 210 A-2 is clear. Plausible if i the applicant does not recognize the turbine trip auto scram is bypassed, believes the relay

__... ~ __________ ~----~-- ~:~~~~~e:~~: ~~~h~c~al!~~fr~~~~~~nywa.~r belie~e~~~cram will occur due to high

.. i o PLU circuit actuation causes a rapid closure of turbine control and intercept valves, which is

. functionally like a turbine trip. Turbine control valve closure results in a reactor scram if st power is above 29.5%, as measured by turbine 1 stage pressure. In this case, the scram is bypassed as indicated by annunciator 210 A-2. Plausible if the applicant does not recognize the turbine trip auto scram is bypassed, believes the relay failure will cause the scram to occur anyway, or believes a scram will occur due to high reactor pressure or high neutron flux.

Psychometrics r--~~veL.c:>fJSr1~wledg~_ ____ ~ifficul!Y ____ ~. Time. Allow~ce (l'llinu~§L SRO ..*.

HIGH 10CFR55.43(b)(5)

Source Documentation Source: D New Exam Item r8J Previous NRC Exam: (PB 2007)

D Modified Bank Item D Other Exam Bank: 0

.---.--..---1--.~J-'=T Ex~m B~nk . __..__......_______ . _...__ . __.__ . ___ ...____ ..__.__..___.__.

ARC-206 B-1' Tech :3.3.1.1 Bases Learning PLOT-5001 B-6a Objective:

KIA System: 000 - ReactorlTurbine Pressure Importance: SRO System 3.8

--.~ ...- . ._-_ .. __ __

KIA Statement:

A2.17 - Ability to (a) predict the impacts of the following on the ReactorlTurbine Pressure Regulating System; and (b) based on those predictions, use procedures to correct, control, or mitigate the

~_onseguen<?es oftho~~iij:).!!()rmal. conQition~qr:J?Peraj:ions:...T~urbine!r~ __..

_~E~l.I~REJ?_M~!ERI~L§~___I NO~E .__ .~ _____.. _______.. __ ....__.. _..... _ . ___..__.._ .....

t'-Jote.! and Comment~:~ ___ J___._.. . . _ . _._.__.. __.___ . ___"___.. _.~_._ . .___ "

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20] 1

93. The following conditions exist on Unit 2:

Reactor is shutdown

Reactor level is -200 inches

Reactor pressure is 100 psig

RHR Loop 'A' is unavailable

RHR Loop 'B' is injecting at 18,000 GPM

Core Spray Loops'A' and 'B' have failed to inject

Drywell temperature is 300 degrees F

Drywell pressure is 8 psig

Torus pressure is 6 psig

Torus level is 17 feet and steady

Containment H2 and 02 concentrations require performing step DW/G-3 of T -102 "Primary Containment Control" Containment Spray must __(1 )_ _ based on __(2)__.

Portions of T-l 02 are PROVIDED ON THE NEXT TWO PAGES.

A. (1) NOT be initiated (2) lack of adequate core cooling B. (1) NOT be initiated (2) Drywell Spray Initiation Limit curve C. (1) be initiated (2) drywell temperature exceeding design limit D. (1) beinitiated (2) potential for loss of Primary Containment integrity 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 L- O'IIT- 1 2 : r:;;;::;...

r---------'---------- ~~;,~ CURVE DWIT :z -

DW SPRAY INITIATION LINIT Ir'~

800 I!!

I TORUS LEVEL IS BELOW 18 FT AND ~6&0

.I

!: 500 UINSAFE I OW BULK AVOTEUP MID OW PRESS I I - - (tiD linT SPRAY)

I ARE BELOW CURVE OW/T-Z. -a-------

I

!Ii w

4&0 /

if I THEN ~sO~af~LF~ AVO TENP I

400

1&0 I

I I w /

I 1. SHUT DOWN RECI RC PUUPS I ~300 - SAFE I

=

2. SHUT DOWN OW COOLl NG FANS

~7&O

3. INITIATE OW SPRAYS USINI): I /

I RHR (T -204) OR I '"

"'200 I HPSI CT -20S) II I

'" 160 100 L I) 2 8 8 10 l2. 14 IS DRY LL PRESSURE PUll) 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20 II OW/G-3 ow 0: LEYEL IS AT LEAST 6S OR UIIKNOIII

.!.l!!I..

OW OR TORUS H2 LEVEL IS AT LEAST 61 OR UIIII:MOIII EIITER T-101 AIID EXECUTE IT COIICURRENTLY WI TH TMU PROCEDURE ----~ VftC-t L.. 01/11-8.1 PERFORM Aft EIlERGEIICY SLOWDOWN USUG T-112 I.---------r----------'

r=m>V '

EB-t L.. 01/11-3.2

..*........ ~- ........

L..(yESI

..-....--....-.-..... -~

f I

I iG;- -C;:;;;;-R;::;;-O:;; -:s;; - - -: '- ~---.--. J I !!!. 1 I !!l!! SUCCESSFUL, t

~I T -201 IS I !.!!£!!. *** ~L EXCEEDllItl OFFUTE RELEASE RATE LIMITS SECURE T-201 AHD DII\I-3.8 IF NECESSARY I PROCEED TO STEP DW/G-S~

VEIIT THE DI DIRECTLY DR INDIRECTLY U;

r------- I

' - D./G-8.~

THROUGII AMY POSSIBLE PATH USI NG *T-200

'- 01lIQ-3.8 EXCEEDING OFFSITE RELEASE RATE LIWITS IF NECESSARY .!.f OW VEltTI ItG IS 1 N PROGRESS.

VENT AND PURGE THe ow WI TN AIR THEN PURGE THE OW OIRECTLY OR INDIRECTLY U8IIItl T-201 -- WITH N! USING T-707

' - OI/II-S.1 CAUTION #10 REDUCING CrllT PRESSURE REDUCES MARGIN TO TIT LeG "PSI! LII/ITS ow CURVE OIIG 1 -

SPRAT IMITIHIOR lIlIIT II REGARDLESS OF Ace

'- UNSAFE (DO !lOT ;PRAYl I TORUS LEVEL IS BELOW 18 FT I

OW BULK Ave TEMP AND DW PRESS I

J SAFE I ---- ARE BELOW CURVE OWIG-l,

1. SHUT Down RECI RC PUMPS

/ 7. SHUT DOWN OW CODLING FANS

3. INITIATE OW SPRAYS USING:

I II RHR (T -204) OR HPU IT-2(5)

I I 0 2 + B B 10 12 1+ II '- 01l/ll-8J1 RYIlELL PRESSURE (PUGl 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer Choice Basis or Justification Correct: D Based on the given conditions, and the guidance of T-102 Step DW/G-3.9, containment sprays are required regardless of ACC. Per T-102 Step DW/G-3.9 Bases, spraying the drywell is performed regardless of ACC because of the potential for deflagration, which could result in a loss of primary containment integrity leading, in turn, to a loss of core cooling Distractors: A T-102 Step DW/T-13 directs spraying only with those pumps not required to assure ACC, and using the only available loop of would jeopardize ACC, T-102 Step DW/G-3.9 directs spraying of ACC.

B Drywell Spray Initiation Limit (DWSIL) curve is NOT exceeded. Plausible if applicant uses torus pressure to plot the DWSIL curve.

C drywell temperature has exceeded the design limit of 281 degrees this is not the reason containment spray is required since this guidance from the DWIT leg of T-102; the reason containment spray is in the T-102.

Psychometrics

__ Level of ~nov.1E.7dge~ __ __ ~_~_ DifficultY: __~~ __L Tim_~!I2w<iii!c;~(min~!es )_L__~SR~_, ,~~__

HIGH i I 10CFR55.43jb)(5)

Source Documentation Source: ~ New Exam Item 0 Previous NRC Exam: 0 o Modified Bank Item 0 Other Exam Bank: 0 Learning 02-5a, -6 Objective:

KJA System: 226001 - RHR/LPCI: Containment Spray Importance: SRO System Mode

_ _ _ _ _ ' _ _ ~ _ _ ___'~~_, ___ ~, _ _ _ ,~ __ ~, _ _ _ _ _ _ ~, __ ~ ______ ~_ _ _ _ _ ~,

46

_ _ _ _~_~ _ _ _,J__ _ _ _ _ _ _ _ _ _ _ ~_ __'., _. _ _ _ _ ~ _ _ _ _ ~~_~. __

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201 ]

94. Unit 2 is in day 2 of a refueling outage with the following conditions present:

The Reactor is in Mode 4

RHR Loop 'B' is in Shutdown Cooling

C RHR PUMP ROOM FLOOD (224 C-5) is in alarm

TORUS WATER LEVEL OUT OF NORMAL RANGE (226 A-4) is in alarm

'C' RHR Pump Room water level is reported to be 18 inches

Torus level is 14.5 feet and lowering What is the highest EAL classification, if any, for these conditions?

T -103 "Water Level-Alarm and Action Levels" table is PROVIDED ON THE NEXT PAGE. EP-AA-I007 Table PBAPS 3-1 "EAL Matrix" is PROVIDED SEPARATEL Y.

A. Unusual Event per FUI B. Unusual Event per HU S C. Alert per HAS D. No EAL classification 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 TABLE SC/L-2 WATER LEVEL-ALARM AND ACTION LEVELS ALARM ACTl ON LEVEL AREA INDICATION STAruS LEVEL urnT 2 UNIT 3 TORUS ROOM BIN. tOO IN. 100 IN. LI -2( 3)919 SUMP ROOM NONE 1 FT 7 IN. 1 FT 4 IN. LOCAL 8I SN OR ReI ~RROOM BIN. Z FT 5 IN. 2 FT 5 IN. LOCAL 818N HPCI ROOM 8 IN. Z FT Z HI. 2 FT 2 IN. LOCAL 8I GN A RHR ROOM 8 IN. 2 FT t 1 IN. 3 FT 5 HI. LOCAL 8I8N OR C RHR ROON B IN. 1 FT 3 IN. 3 FT 5 IN. LOCAL 8I GN B RHR ROOM S IN. 1 FT 5 IN. 3 FT 5 IN. LOCAL 81 GM OR o RHR ROOM B IN. 3 FT 4 IN. 3 FT 5 IN. LOCAL SIGN A CS ROOU SIN. 1 FT 10 IN. 3 FT 3 IN. LOCAL srSN OR C CS ROOM 6 IN. 3 FT S IN. 3 FT 1 IN. LOCAL SIGN B C8 ROON 6 IN. :2 FT I) IN. :2 FT 4 IN. LOCAL SIGH OR o CS ROOM B IN. 2 FT 3 IN. 2 FT 10 IN. LOCAL 81 GN 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20 I I Answer

94SRO Choice Basis or Justification Correct: C threshold value for HA5 is met since the 2C RHR pump room is a Table H3 and water level (at 18 inches) is above the T-103 Action Level of 1 foot, 3 (15 inches).

Distracters: A i choice is based on the applicant believing a Torus leak indicates a loss or

potential loss of the Primary Containment Barrier. However, none of the threshold values for FU1 apply to a Torus water leak. This includes "Emergency Director Judgment". In addition, FU1 only shows up on the EAL Hot Matrix and is therefore only applicable in Modes 1, 2 and 3. Plausible if the applicant makes some i

incQrrec! assu!!![>!i()ns a_n!!;'!Qplies the wrong lHJ:l!lmatr~ forJhe given congition~_'1 B The applicant may consider the HU5 threshold value met based on the given

conditions. However, a single operable loop of RHR (B in this case) meets the Tech Spec requirements for ECCS and Shutdown cooling in Mode 4. Therefore, flooding in the 2C RHR pump room does not meet HU5 criteria since 2C RHR is not needed for the current operating mode. In addition, HU5 is not the highest EAL

classification.

-.-~~ ~-.-i--

D criteria for HA5 are met. Plausible if applicant evaluates the T~103 Action i Level table incorrectly (i.e., wrong pump room and/or wrong unit).

I Psychometrics

__~Level_ of KnowL~~__L __.. ~ifficl!ltL__ .

HIGH :

i--Time~I!QwC3J!~eLlTlinut~~ SRO 10CFR55.43(b)(5)

Source Documentation Source: IZI New Exam Item D Previous NRC Exam: ()

D Modified Bank Item D Other Exam Bank: ()

ILT Exam Bank -~.-~-.~.-.--- .. ~--- ... ~---

EP-AA-1 Table PBAPS 3-1 and Table PBAPS 3-2 G6-8 KIA System: i G2.1 - Conduct of Operations SRO

-~----~------.----.--~---------

4.2 KIA Statement:

G2.1.25 - Ability to interpret reference materials, such a~aphs, curves, tables, etc.

REQUIRED MATERIAL~-: -PP-AM007 Table PBAPS3.1 (EAL MATBIX}--~= ____

t'-Jot~s~nd Comments:____ ____~__________ __ __ _ _ ___ _ ____ _ __~ .....__~_ . __ .

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

95. Both units are operating at 100% power.

A breach is being planned for the Cable Spreading Room Return Ducting that passes through the Control Room Envelope (CRE).

The breach size has been calculated to:

be equivalent to 144 in2 in the 12 inch thick wall

result in an additional 400 cfm of leakage burden This breach will result in MCREV train becoming INOPERABLE, due to exceSSIve __,_

These items are PROVIDED ON THE FOLLOWING PAGES:

1. CRE section of A-487

2. Current CRE Temporary Breach Log

3. CRE Boundary Operability discussion from GP-30 A. (1) A (2) out-leakage B. (l)A (2) in-leakage C. (l)B (2) out-leakage D. (l)B (2) in-leakage 20 II NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 nUTSlD[ AlH C ~~~....J CONTROL ROOM

~,t)nM t:NVCLJPE :\Jr:L:IUf:~: WAI_LS.

FLOOR I CEIl,Nl JF CONTRDL LECLND' iJUT ~LfAXAG~ bAHRI::P TION J ~HIS 0 AW NG LJrlrIL7ERE:: ;fH.. EAKAGE P,AR~j£~ ~AY N~ R LlGIBLE AT A R~~~ ED S Z~

From A~487 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Page ATTACHMENT A "A" MCREV TRAIN CONTROL ROOM ENVELOPE TEMPORARY BREACH LOG (1) (2) (3) (4) TOTAL (5) (6) TOTAL Breach Breach Breach CALCULATED TOTAL INLEAKAGE CALCULATED TOTAL BREACH Number Added Restored INLEAKAGE INLEAKAGE < 500 CFM AREA DATE/TIME DATE/TIME (CFM) (CFM) Y or N? < 4.128 FT2 450 2.122 7\ fT1fTl7\ "-'U1.A'C'l\lfT1 D

.Ll.l ..L r-i\....d ll*1l.:.Jl\l.L .u "B" MCREV TRAIN CONTROL ROOM ENVELOPE TEMPORARY BREACH LOG (1) (2) (3) (4) TOTAL (5) (6) TOTAL Breach Breach Breach CALCULATED TOTAL INLEAKAGE CALCULATED TOTAL BREACH Number Added Restored INLEAKAGE INLEAKAGE < 500 CFM AREA DATE/TIME DATE/TIME (CFM) (CFM) Y or N? < 4.128 FT2 50 3.556 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20 II From GP-30 CONTROL ROOM ENVELOPE BOUNDARY INTEGRITY "The CRE boundary may come inoperable due to inleakage and/or outleakage that lowers the ability of MCREV system to protect MCR personnel from radioact mate I during an accident. Depending on the location of the kage, one or both trains of MCREV may be affected. Generally, the CRE boundary can become inoperab due to any of the following:

o Unfiltered inleakage above 500 cfm either the "A" OR "B" train of MCREV. Work act ties will be assigned a value Calculated kage . . . as shown on A-487, "Barrier Plans Elev 165 0". Work act ies not affecting this equipment may have an assigned value of 0 (zero) for Cal ed I kage.

o Inability to rna in positive ssure within the CRE boundary due to a breach (i.e., a breach the physical CRE boundary other than in equipment described for unfiltered eakage) . .

. The maximum breach area that will maintain pos ive pressure 12" ck portions of the CRE boundary is calculated to be 4.128 Breaches in the CRE boundary will be assigned a value Calculated Breach Area. Work act ies not affecting this equipment may have an assigned ue of 0 (zero) for Calculat Breach Area."

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201 1 Answer Key

~------------------------------------

Question # 95 SRO i Choice Basis or Justification Correct: C i As indicated on A487, the Cable Spread Room Return ducting forms part of

. the OUT-LEAKAGE Barrier - the integrity of the OUT-LEAKAGE Barrier is required to ensure the ability to maintain a positive pressure in the MCR.

Provided this barrier brealch area is less than 4.128 sq ft, operability is maintained. If greater than this area, operability is NOT maintained, as discussed in the portion of GP-30 provided. The candidate must correctly

identify the barrier on the A-487 print and then translate this correctly to the L . ._~. ._;___I ~:;~r~~I~':~~':~~~i~~~e conclUSiontha~h~B~ OUT~EAKAGE

!___Distracters: .. L~_.~~:~:e!;:yo~e~~~:~~~:~;~~~e::e~o~~~=e~IY identify the

B 1 i Candidate may select this option if they do not correctly identify the boundary type or misappl the leaka e t e.

D I Candidate may select this option if they do not correctly identify the

, boundary type or misapply the leakage type.

Psychometrics c-LeY~lgLKn_O\Nle~~_~_. _____DiffiCl:l!!Y _ _ ~-l Time_AIIOW~rl~.l.f!1ir1~LJtest~I~__ SRQ _

HIGH! i

10CFR55.43(b)(5)

Source Documentation Source: New Exam Item Previous NRC Exam: 0 Modified Bank Item D Other Exam Bank: 0 KIA System: G2.2 - Equipment Control KIA Statement:

G2.2.15 - Ability to determine the expected plant configuration using design and configuration control

~;~~:~~:~~~:a~~l~~s't~_::e~~_.=_~------.---.-------.

20 II NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

96. While operating at power, aT-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"

Drywell radiation suddenly spikes to 2.5 x 10-1 JlCi/cc and continues to rise Based on the above conditions, what action must be taken and why?

A Re-align the vent path to the Torus to "scrub" the release.

B. Terminate venting to ensure ODCM release limits are not exceeded.

C. Perform a GP-15 "Local Evacuation" of the Radwaste Building.

D. Direct Rad Protection to perform dose calculations from Main Stack data.

201) NRC SRO Written Exam Rev. )

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201}

~- ... ..

~ ..

~.-

Answer Key .-~.---- .. ~- .. - ..... --~ ~~ .. - .. ..- ...-

~

..u 01 ,,#96SRO Choice Basis or Justification Correct: Per OT-1 01 Bases, venting is required to be terminated to ensure the ODCM release limits are not exceeded.

Distractors: I This is not in accordance with the direction in OT-101, but maybe

with the guidance for venting in T-200.

I Evacuation of the Radwaste Building is not required for th is venting operation. This would only be true for T-200 venting.

D Offsite dose calculations are not required since terminating the venting I'In~~r~I'll'In ensures ODCM (and Tech Spec) limits are not exceeded.

PS"ychometrics J~E:vel ofJS.nowlei!9.~__ J~_. _ QifficultL __ ~1~irlJe AllQwa~~lmir1utE:~~ ___ ~~§B~___ .'~

MEMORY I 3.5 3 I 10CFR55.43(b)(4)

Source Documentation Source: I 0 New Exam Item L8J Previous NRC Exam: (PB 2002)

. [8J Modified Bank Item 0 Other Exam Bank: 0

_~ . _~. . _.~~~~[8JJLT Ex:am l:Jan~_~ __~~ _ _ . ____ . .~_

. OT-101' SO 7B.3.A-2 .~~------~.~------~

PLOT-1540-04 KIA System: - Radiation Control \ Importance: SRO 3.8 KIA Statement:

G2.3.14 - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or

~_~~~_e~~;~_c:!:t_:~_~~~~~vi~~f~_E ~_~'~'~.~-~===:::..-==~=-~~_-~~_~ ~_. ~_ _.~ .

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201 I

97. The following conditions exist on Unit 2 following fuel failure with a Primary System breach in the Reactor Building:

Reactor power is 40% and lowering

Control rods are being inserted per GP-9-2 "Fast Reactor Power Reduction"

2 VENT EXH STACK RAD MONITOR HI-HI A (218 B-4) is in alarm

2 VENT EXH STACK RAD MONITOR HI-HI B (218 C-4) is in alarm

Vent Stack radiation on RI-2979A(B) is 3.63 E+04 flCi/sec and rising

MAIN STACK RADIATION HIGH-HIGH (003 D-l) is in alarm

Main Stack radiation on RI-050A(8) is 4.17 E+07 flCi/sec and rising

The Primary System breach has NOT been isolated Which one of the following describes the actions required by T -104 "Radioactivity Release" for these conditions?

_(1 based on _ _(2)_ _.

A. (1) Manually scram and depressurize per T-I0l "RPV Control" (2) Main Stack effluent B. (1) Manually scram and depressurize per T -101 "RPV Control" (2) Vent Stack effluent C. (1) Perform T-112 "Emergency Blowdown" (2) Main Stack effluent D. (1) Perform T-112 "Emergency Blowdown" (2) Vent Stack effluent EP-AA-I007, Table Rl Table R1 - Effluent Monitor Thresholds Release Path General Emergency Site Are,l Emergency Alert Unusual Event Main Stack I (RI-O-11-050,AJ8 5.57 E+09 ~ICilsec 5.51 E+08 ~ICilsec 6.36 E+01 ~tCiisec 6.36 E+OS ~Cilsec Common)

.1 Vent Stack (RI-2979AlB Unit 2 or 3.36 E+{}S pCI/sec 3.36 E+07 "CI!sec 3.83 E+06 3.83 E+04 pCi!sec RI-3979AlB Unit 3j 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer QUEtstfCm '# 97 SRO Choice Basis or Justification Correct: A Main Stack effluent is approaching the Alert level; Vent Stack effluent is above the Unusual Event level but well below the Alert level. For these step RR-10 ofT-104 requires a manual scram, T 101 entry, nArlrAC~c:. rl'::r::atlfln per T -101.

Distracters: B Stack effluent is below the Unusual Event level.

C the primary system breach has not been isolated, Main Stack effluent is well below the GE threshold. An emergency blowdown is not warranted for the given conditions.

o Although the primary system breach has not been isolated, Vent Stack effluent is well below the GE threshold. An emergency blowdown is not warranted for the given conditions.

Psychometrics

_ Level of Knowledge__L ____ QifficlJlty______ H_ HTi'!1~Alto'v\'an<2~_Cr!!lnl..lJe~l HH "----- SRO --

HIGH i 10CFR55.43(b)(4)

Source Documentation Source: o New Exam Item r8J Previous NRC Exam: (PB 2009) r8J Modified Bank Item 0 Other Exam Bank: 0

. _ _ _ _ _ _ _ _ _ _ _ _ _H___ ILT Exam Bank _H _____ H _______ H ____ _

C:!;~~n~e(s ):--l ~:~i~~i~2~~~; PE?APS~1~T~ °4ilnd~H~HsesH 1 Objective:

KIA System: i G2.4 - Emergency Procedures I Plan SRO 4.7 KIA Statement:

G2.4.4 - Ability to recognize abnormal indications for system operating parameters that are entry-level

_c~nditi2I1s foremergen~ancL~bnoi.n1il.LoperatlQ9~()c~dures:_____H __._______HH__ H_H____________ . __

REqUIRE!> MA1ERI~LS =- __~--~O~t;----- __H--H--.-------.------------- ..---______ . _____. ___________._ . _______ _

Notes

.~-

and Comments:

...- - - -.. ---~-- ...- - -... -----~~- .. -----~- ..- ~~- .. -~ ._

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

98. Equipment Operators need to enter a locked high radiation area to manually operate Primary Containment Isolation Valves in order to satisfy a Technical Specification required action. The highest dose rate in the area is 16,000 mRlhr (16 RJhr).

Per RP-PB-460-1 00 1 "Radiation Protection Controlled Keys", which one of the following describes the type of Locked High Radiation Area and the highest level of authorization required for issuing the key?

Tvpe ofLHRA Highest Authorization Required A. Levell Radiation Protection Manager B. Levell Plant Manager C. Level 2 Radiation Protection Manager D. Level 2 Plant Manager 2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer

98SRO Choice Basis or Justification Correct: C RP-PB-460-1001, Level 2 LHRA is an area with dose rates> 1SR/hr.

RP Manager must provide authorization for this entry.

Distracters: A i The level is incorrect. The area is a Level 2 (>1SR/hr), which requires from the RP Manager for issuing the key.

Psychometrics

. Level of Know~dge__ ~ t----- Difficulty______l Time Allowance (mir:lutes) ._L___§RO ___... _

MEMORY I 10CFRSS.43(b)(4)

Source Documentation Source: o New Exam Item ~ Previous NRC Exam: (PB 2008) o Modified Bank Item 0 Other Exam Bank ILT Exam Bank Learning PLOT-1770-3 Objective:

KJA Sy~tem:--* -TG2-:;-=-Radiation Control SRO

__.. ____ _____.____________.____._.. _____.. _--'--_..______._______..__3__..8.c ---_..._-

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,

~ccess to Iqcked _high-~~Q~tio!l areas, ~l1jr1flfil!ers, .etc. _ _ ___ _ ___ _ __ _ _ __ __ _ ___ _

t~~::~::;::~L:~ jNONE_ -------- ---- ------- - - - ---~=~--_=-------

20 II NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011

99. A small steam leak inside the Drywell occurred on Unit 2.

The reactor was depressurized in accordance with T -112 "Emergency Blowdown" due to being unable to restore and maintain drywell temperature below 281 degrees F.

The following conditions existed at the start of the blowdown:

Indicated RPV level was -140 inches

All high-pressure feed sources were unavailable The following conditions exist at the completion ofthe blowdown:

RPV pressure is 35 psig

RPV level is -175 inches

Drywell temperature is 295 degrees 1:;' (TI-250l points 126 and 127)

Multiple failures prevented LPCI ami Core Spray systems from injecting What action is required for these conditions?

Portions ofT-102 "Primary Containment Control" AND T-112 "Emergency Blowdown" are PROVIDED ON THE NEXT TWO PAGES.

A. Enter and execute T-116 "RPV Flooding".

B. Establish Shutdown Cooling per T-112 "Emergency Blowdown".

C. Restore RPV level above -172 inches per T-l11 "Level Restoration".

D. Restore RPV level to between +5 and +35 inches per T-lOl "RPV Control".

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 T-I02 "Primary Containment Control" TABLE DW/T-l RPV lEVEL INSTRUMENT STATUS AN RPV LEVEL INSTRUMENT MAY BE USEO TO DETERMINE RPV LEVEL ONLY WHEN THE FOLLOWING CONDITIONS ARE SATISFIEO:

NOTE: USE AVAILABLE POINTS (128 1127 OF TI-2(3)SOI) TO DETERMINE RPV lEVEL INSTRUMENT STATUS ALL RPV LEVEL INSTRUMENTS RPV SATURATION CURVE S 900 14_

-'" RPV SATURATION CURVE ...I,..-... 600

~

!iLL ...z-, 400

'"'('0

<::i- co

....... a

.... - ,~

I....

ILl z' ......

co II..

,~ 200 400 BOO 800 1000 SEE OEHIL

~t: RPV PRESSURE (PSIBl CI 1£ OW TEMP AND RPV PRESS ARE ON THE UNSAFE SIDE OF THE RPV SATURATION CURVE

~ AN INSTRUMENT EXHIBITS AN UNEXPLAINED TREND OR OSCILLATION, THEN THAT INSTRUMENT IS UNAVAILABLE WIDE AND NARROW RANGE INSTS ONLY SHUTDOWN RANGE INST LI -2( 81-2-8-88 ONLY FOR EACH OF THE INSTRUMENTS I N THE HBLE. THE INSTRUMENT LI -2( 8)-2-3-86 READS ON THE SAFE SIDE OF THE CURVE READS ~ THE NIN INDICATED LEVEL ~ THE TEMP NEAR THE DW RENCE LEG VERTI CAL RUNS ( -2( 3)501 PT 126 I 127) ARE BELOW THE MAX RUN TEMP. no z"l I

.,. ,N.,e.TE. lE~ MAX RUN TEMP co Z

.... tOO 80 SAFE '/

I NSTRUNENT IS ABOVE I S BELOW ~=

"', 80 ...,.

NARROW RANGE 10 IN. .. 50Of ......

",I 40 "..

20 WIDE RANGE -120 IN. $00"1'

",N

... ,.!, 0 ...- """'"

u_

.... ('0 -20 (R~V L~:E~ At~KNOWN)

"'I -40 Z ...

100 160200 2(;0300 31i0 400 "liD SOD 660 DW TEMP ON TI-2CnSOl PT 128 I 121 (OF) 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 20 II T -112 "Emere;ency Blowdown" t....-------------..---------,

I CONTINUE RPV DEPRESSURIZATION l_ EB-22 *

(NO) L_

J L (YES) :

t ESTABLISH SHUTDOWN COOLING USE ONLY THOSE RHR PUMPS JtllI REQUIRED TO MAINTAIN RPV LEVEL AB1JVE +5 IN.

L EB-Z3 ***

2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 Answer' Key

99SRO Choice Basis or Justification C With RPV level at -175 inches, the only on-scale level indicators are Fuel Zone. PerT-102, Table DWIT-1, fuel zone level plots on the UNSAFE side of the RPV saturation curve. But since there are no unexplained trends or level is NOT unknown. Level restoration per T-111 is required.

Distracters A I Level,'s not unk n0 w n .Piau . 5bl ef appl"c , ant',ncorrectly app r'e5 Table DWIT-1 and/or believes the given conditions show an "unexplained trend".

B Level is known to be below the point at which T -111 "Level Restoration" actions are required. In addition, SDC cannot be established until RPV level is restored. Plausible based on the T -112 excerpt provided with question.

D Level is known to be below the point at which T -111 "Level Restoration" actions are required. Plausible if applicant does not recall the level control strategies of T-1 01 versus T-111; if level cannot be maintained above -172 inches (as is the case here), T-101 must be exited and entry into T-111 is

required. Once level is mstored above -172 inches, T -101 is reentered I from T -111 .

Psychometrics Level of KnowledgeJ . Qifficulty Time AIIQ"",~nce (r:r!ir1.1:!~~L~_ SRO ....... ~--.

HIGH I 10CFR55.43(b)(5)

Source Documentation Source: I o New Exam Item cgj Previous NRC Exam: (PB 2008)

cgj Modified Bank Item o Other Exam Bank

~ference(5~' :-ti:~~~7T~~J:i11L~ =~ --.-~ ..

Learning Objective:

+

.1 T PLOT-1560-11

___ ... - -.. ~

KIA System: G2.1 - Conduct of Operations Importance: SRO

. --.l~~ ____..____... ._e_******** _ _ _*

4.3 - - - - - - - - - - - ..

KIA Statement:

I G2.1.45 - Ability to identify and interpret diverse indications to validate the response of another indicator.

...... - - - - - - - - -.... -~--- ......------. -------~------~ ~- ..-

REQUIRED MATERIALS: NONE Notes and Comments:

2011 NRC SRO Written Exam Rev. 1

Peach Bottom Initial Senior Reactor Operator NRC Examination January 2011 100. Unit 2 pre-startup preparations are in progress in accordance with GP-2 "Normal Plant Startup".

The RWM is inoperable and bypassed in accordance with AO 62A.1-2 "Rod Worth Minimizer System Manual Bypass ... "

The conditions of Technical Specification 3.3.2.1 "Control Rod Block Instrumentation" are met With the RWM inoperable, Technical Specification 3.3.2.1 requires verification that movement of control rods is in compliance with the analyzed rod position sequence.

To comply with Tech Spec 3.3.2.1 and AO 62A.1-2, this must be performed by a:

1. Licensed Operator

2. Reactor Engineer

3. Shift Manager A. 1 ONLY B. 2 ONLY C. BOTH 1 and 2 D. BOTH 2 and 3 2011 NRC SRO Written Exam Rev. I

Peach Bottom Initial Senior Reactor Operator NRC Examination January 201 1 Answer Question # 100 SRO Choice Basis or Justification Correct: A With the RWM inoperable, Tech Spec 3.3.2.1 requires control rod movement to be verified in "compliance with the analyzed rod position sequence by a second licensed operator or other qualified member of the technical staff." Tech Spec 3.3.2.1 Bases further defines this as either a Reactor Operator or Senior Reactor Operator. Since there is no defined program to qualify members of the technical staff at Peach Bottom, a second licensed operator is the

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Distractors: B Plausible since Tech Spec 3.3.2.1 requires a licensed operator or "other qualified member of the technical staff" and the applicant may believe a Reactor Engineer meets this requirement. However, per AO 62A.1-2, use of a qualified member of the station technical

..--..--.r.--C---- '~!~b~:~~:eW~!~~i!:03~~~(3:~:i~;:I:~i:l~::~~~~!~;:~:t~:~~~~~!~-;;;ember of the technical staff' and the applicant may believe a Reactor Engineer meets this requirement. However, per AO 62A.1-2, use of a qualified member of the station technical

._______... ___ _ _ ___ ~aff is not allowed since nO..J>Iogral!1..is in..J>lace to..9..ua1ify individualsJor !.l}Ls rolE:!"_._______._

J' I 0

___ . _______ . ______..L ____

Plausible since Tech Spec 3.3.2.1 requires a licensed operator or "other qualified member of the technical staff", and since GP-2 "Normal Plant Startup" previously required a 2 nd licensed operator and the Shift Manger to independently verify the control rod pattern matches the

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MEMORY I 10CFR55.43(b )(2)

Source Documentation Source: D New Exam Item L8J Previous NRC Exam: (PB 2009)

L8J Modified Bank Item 0 Other Exam Bank: ()

3.3.2.1 and Bases Learning PLOT-5062A-9 Objective:

KIA System: G2.2 - Equipment Control Importance: SRO 4.4 KIA Statement:

G2.2.1 - Ability to perform pre-startup procedures for the facility, including operating those controls associated with _plantequipment that coulQ_~ffect rea.2!i\/"i!Y... _ . ________ ._._._. __ . . _______

-~:t;~::~~:::~;~s: ii6~~Th.~pro~~dU~.~1

.- . requirem.. ent for Verifieali.o.*.n.-o-.f.-.. he control rod sequence was different in the 2009 NRC Exam time frame-it used to t.

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v t e Letter Sequence Other
TAC:U01769, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML102880050, ML110200243, ML110200247, ML110200254, ML110200260, ML110620439, ML110630047, ML110630064, ML110630120, ML111580037
MONTHYEAR2010-11-01 [Table View]

ML110630064

Text

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