Text

Initial Submittal of the Ro/Sro Written Examination for the Clinton Initial Examination - July 2005
	ML052430760
Person / Time
Site:	Clinton
Issue date:	07/18/2005
From:	Setser G; AmerGen Energy Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML052430228	List: ML052430747; ML052430750; ML052430754; ML052430755; ML052430756; ML052430758; ML052430760; ML052430761;
References
	50-461/05-301 50-461/05-301
	Download: ML052430760 (155)
	v • d • e

INITIAL SUBMITTAL OF THE ROlSRO WRITTEN EXAMINATION FOR THE CLINTON INITIAL EXAMINATION -JULY 2005

I Question# I 1 I

\\

Which ONE of the following describes the impact of a loss of 480V Unit Sub 1 B?

A.

B.

lE12-F042C, LPCI From RH C Shutoff Valve, will NOT open electrically.

The ONLY source of AC power to the RPS Solenoid Bus B Inverter will be via the Bypass Transformer.

The SUCTION side of RWCU will NOT automatically isolate if Standby Liquid Control is initiated.

VC Train B will operate ONLY in the High Radiation Isolation Mode.

C.

D.

Answer: A Explanation:

Aiscorrct-PerLP85205.AttachmentD. thepowersupplytothisMOVisABMCC 184. Per CPS 3514.0lC006, Section 2.1.1, the loss of 480V Unit Sub IB results in the loss of all of the listed MCCs, including AB MCC lB4.

Without 480VAC motor ~OWR, the F042C valve will not open elechically.

B is incorrect - PR LP85434, Figure 1, only a rectifier section, &the altmate backup) power through the bypass tnlnsfomer. Additionally, this RPS Solenoid Bus B uses m-vital powq 480 V Unit Sub IB is C is incorrect - Per LP85204, page 35, a SLC Pump A start signal closes the IG33-FO04 valve (RWCU Suction Outboard Isolation), whilea SLC Pump B start closes the IG33-FOOl valve (RWCU Suction Inboard Isolation). Per CPS 3514.01COO6, Appendix A, page 46, a loss of480V Unit Sub IB disables IG33-FWl (Suction Inboard) and IG33-F040 (Return Inboard). When operators initiate S U : (start both pumps), the A SLC Pump start will still close IG33-FO04, although the B SLC Pump start will NOT close IG33-FO01. Because F W suction side of RWCU D is incorrect - Per CPS 3514.01COO6, Appendix A, page 41, a loss of 480V Unit Sub IB &!s!y disables VC (Control Rom HVAC) Train B (a Div 2 subsystem). Only VC Train A (not B) remains available, and only in the High Rad Isolation Mode @ecause of the failed-high PRMs, IRK-PROWB/D).

480V bus supplies k&!.l the normal POWR to the inverters power.

automatically close, the automatically isolate.

References provided to examinee:

References:

I None I LP85205. Residual Heat Removal (RHR)

LP85204; Reactor Water Cleanup (RWCU)

LP85434, Nuclear System Protection System (NSPS (Inverters))

CPS 3514.01CoO6,416OV Bus IBI Outage Date Written: I 05/16/05 I Author: I Ryder Comments: None

I Question # 1 2 1

References provided to examinee:

References:

ROISRO: I Tier:

1 Group:

I KA:

I R O I R I

SROIR I

Cog Level Both I

2 I

I I 209001 K2.01 I 3.0 I

3.1 I

Lower None LP85209, Low Pressure Core Spray, LPCS CPS 3514.01C040,125 VDC MCC 1A Div 1 Outage CPS 5003-5F, RFT BKR LOSS OF Dc CONT PWR SyitemlEvolution Name:

1 category:

Low pressure core spray System I Plant systems i_-

KA Statement:

Knowledge ofclshical power supplies 10 the following: Pump power Which ONE of the following describes the impact of a loss of DC MCC 1 A?

A.

DG 1A CANNOT be started from the main control room, but CAN be started from its LOCAL CONTROL PANEL.

Reactor Recirculation Pump A will automatically trip IF it is running in B.

SLOW speed.

C.

A running SF Pump will automatically trip IF the Suction Outboard Isolation Valve, 1SF004, closes.

LPCS Pump CANNOT be started from the main control room, but CAN be started at 4160V Bus 1Al.

D.

Answer: D Explanation:

D is correct - PR CPS 3514.01C040, Appendix A, page 31, loss of this bus results in the loss of LPCS Pump breaker control power. Since this pump has no Dc control power-dependent starting interlocks, operators can still starl the pump by locally closing the pump motor breaker at 4160V Bus 1Al.

A is incorrect - Per CPS 3514.01CO40, Appendix A, page 31, dl control power for this diesel is lost, both local and ElUOte.

B is incorrect - Per CPS 3514.01CO40, Appendix A, page 32, the 3A breaker for RR Pump A loses its breaker logic control power. Per CPS 5003-5F, the SA breaker will hip if the 3A breaker loses a control power source. The SA breaker is closed & when the RR Pump is w i n g in FAST speed.

Cis incorrect - PR CPS 3514.01C040, Appendix A, page 32, a running SF Pump will NOT trip (trip interlock is disabled) if the suction valve. 1SF004, leaves it open seat.

Datewritten:

1 02/04/05 I Author: I Ryder Comments: None

[ Question# 13 1

References:

L LPS5380, High Pressure Core Spray, HPCS USAR, Section 6.3.3, ECCS Performance Evaluation Among the following ECCS SYSTEM FAILURES, which one would have the MOST SEVERE impact on the ability to cool the core?

A.

With the reactor initially at rated power. a SMALL-break LOCA occurs and the HPCS Pump will NOT run.

With the reactor initially at rated power, a LARGE-break LOCA depressurizes the plant and the LPCS Pump will NOT run.

After a scram from rated power, reactor water level drops to TAF, and NONE of the ADS Valves will open.

After a scram from rated power, reactor water level is declared UNKNOWN, and NONE of the ADS Valves will open.

B.

C.

D.

Answer: A Explanation:

A is -t

- Per LP85380, pages 4,5 and 7, the HPCS system failure (when compared to the other ~ S W R choices) would have the most smre impact on the ability to cool the core, for both small and large-break LOCAs. The lesson plan discussion is very brief and does not attempt to qualify (describes the analyses) why this is the case. Although the USAR (see attached references) does provide this clarification, such information is beyond scope for the application of this KA to the RO Exam. Therefore, the correct answer has been carefully stated to be consistent with both the lesson plan claim, and the USAR discussion. Similarly, each dishacter has been stated in a way that is certain to mcontlict with any of the assumptions belonging to the USAR analyses, while at the same time providing sufficient plausibility.

B is incorrect - For the reasons described above.

C and D are incorrect - For the m o n s described above. Also. the claim here is that the ADS Valves (an ECCS system) fail to open. There are 9 other non-ADS SRVs available to be opened.

Objective:

I Questlon Source:

I Level of Dimculty:

None New I

3.9

I Question ## I 4 I

References:

L LP8521 I, Standby Liquid Conml (SLC)

CPS LER 2004-002-00. Miswsitioned SLC Air Sparge Valve Results in High L

RO/SRO:

I Tier:

I Group:

I KA:

I ROIR:

1 SROIR 1

Cop. Level Both I

2 I

I

~211000K1.03 I

2.5 I

2.6 I

b W R SystenJEvolution Name:

I catmory:

Standby Liquid Control System KA Statement:

Knowledge of the physical connections and/or cause-effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following: Plant air systems I Plant systems With the plant operating at rated power, the service air SPARGE valve for the SLC Storage Tank has been unintentionally left OPEN.

WITHOUT operator action, which ONE of the following describes the EARLIEST potential impact on the SLC system, as a result of this mispositioned sparge valve?

A.

B.

C.

D.

Answer: B Bum out of the SLC Storage Tank OPERATING Heater HIGHER Boron concentration in the SLC Storage Tank SLC Storage Tank overflow through the top vent LOWER Boron concentration in the SLC Storage Tank Explmatiou:

B is correct - This question is written directly from CPS LER 2004-002-00 (see attached references). The continuous spnrge resulted in tank water evaporation and a rise in boron concentration as a consequence. This LER is included in the lesson plan, LP8521 I, Attachment C, (OPEX) discussion.

A is incorrect - But is quite plausible; so plausible, that the EARLIEST component of the question stem is critical to avoiding a second cmrect answer. Per LP85211, page 20, if tank level lowers to <I,000 gallons remaining, the opaating Heater could be damaged due to being uncovered. However, since normal SLC tank level is about 4,000 gallons (LP8521 I, page 6), uncovering the heater would NOT be the earliest potential impact.

C and D me incorrect - For the reasons associated with the correct answer.

Objective:

I Ouestiun Source:

1 Level of Dlfliculty:

~ ~ 8 5 2 1 1

.I. 12.2 New I

2.3 Date Written: I 04/28/05 I Author: I Ryder Comments: None

[ Question # I 5 1 Cug Level RO/SRO I Tier:

I Group:

I KA:

I ROIR:

I SROIR I

Both I

2 I

I

~212000A1.07 1 3.4 I

3.4 I

LOWR SystemlEvolntiou Name:

I Catet?w:

Reactor Protection System I Plant systems KA Stntemeut:

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR PROTECTION SYSTEM wntrols including: Rod position information BEFORE a scram is RESET, which ONE of the following describes an ACCEPTABLE method to determine that a given control rod HAS FULLY inserted?

A.

B.

Confirm the ROD FULL IN light is lit, for that rod, at either RACC panel.

With the Full Core Display in DUAL mode, confirm that EITHER channels Full-In LED is lit the numerical display is Blank (NO numerical value).

With the Full Core Display in 2 CHANNEL mode, confirm that BOTH channels Full-In LEDs are lit Confirm a value of 0 (zero) volts on Transient Test Channel 291.

C.

the numerical display indicates 00.

D.

Answer: B Explanatiou:

Biswmct-P~CPS4100.01,Section2.1.2,asdanibedinthischoice.

A is incorrect - PR CPS 4100.01, Section 2.1.2, and CPS 3304.02, Section 8.2.1 1.2. There is only single All Rods Full In LED at these RACC panels. There is EQ individual Rod Full In light for each given rod. This choice is very plausible to the Candidate who vaguely malls that there is a way to manually address each given mds actual position (including 00), using the ID Generator, at these RACC panels.

C is incomct - This choice suggests a variation of the correct answer, B, but it is explanation).

D is iuwmct - PR CPS 3304.02, Section 8.2.1 1.3. A zao (0) volts value is associated with all rods rn full in.

wrrect (see the correct ~ S W R S Objective:

I Question Source:

1 Level of DifliculW:

DB410001.I.6 New I

2.7 References provided to examinee:

Refereuces:

I None I LP85401, Rod Control & Information System DB410001, Reactor Scram CPS 4100.01, Reactor Scram CPS 3304.02, Rod Control and Information System Date Written: I 05/03/05 I Author:

1 Ryder Commeuta: None

1 Question # 16 I RO/SRO I Tier:

I Group:

I KA:

I R O I R I

S R O I R Cog Level Both I

2 I

1 I 215004K6.01 I 3.2 I

3.3 I

Higher SystemlEvolutlon Name:

I category:

Source Range Monitor (SRM) System (SRM) SYSTEM RPS The plant is operating at rated power when SRM A fails UPSCALE.

Which ONE of the following describes the plant response if ONLY THE DIV 1 shorting link were loose and had become dislodged (circuit interrupted) before the SRM failed?

I Plant System KA Statement:

Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR A.

Scram,ONLY B.

Rod block, ONLY C.

Rod block AND scram D.

Answer: A NEITHER a scram, NOR a rod block Explauatiou:

A is correct - P a CPS 5lWS-lK, rod block is bypassed with the Mode Switch in RUN (plant operating at rated power). PR CPS 5005-1K scram function is dependent on neither the Mode Switch position, nor IRM Range. It solely depends on shorting link status. PR LP85212, Figure 13, all it takes is the removal of a single shorting link (for

&Division) to enable the non-coincident scram function. SRM A belongs to Div I RPS.

B, C, and D are incomt - For the reasons described above. D is quite plausible to candidate who does recognize that the rod block is already bypassed (RUN), but has never eonsidered the specific impact of having ONLY a single shorting link removed (rather than ALL of them being removed, as would be the case for special testing that might require such a non-coincident scram function). It is also plausible to the candidate who believes that, like the rod block, the SRM scram function is also bypassed in RUN.

Oblective:

Questiou Source:

I Level of DIMeulty:

I LP85215.1.4.5 New I

2.7 References provided to examinee:

References:

I None I LP85212. Reactor Protection System Lp85215, SRMs CPS 5005-IK, SRM WSC ALARM OR INOP Date Written: I ou06/05 I Author: I Rydm Commeuts: None

I Question # I 7 I RO/SRO I Tier:

I Group:

I KA:

I ROIR I

SROIR I Cog Level I

n I

L I 1 i C n m v t n7 I 2.6 I

2.9 I

Higher DOUl

, LI.,"">RI.",

I I

L I

SystemlEvolution Name:

I category:

Average Power Range Monitorbcal Power Range I Plant systems Monitor I

KA Statement:

Knowledge of the physical connections andior causeeffect relationships hehvgn APRMLPRM and the following:

Process computer, performance monitoring system ASSUME the following when answering this question:

The neutron flux levels seen by each of the LPRMs inputting to APRM 'A' are IDENTICAL The plant is operating at power, with the following:

3D-Monicore calculated Core Thermal Power (CTP) is 90%

APRM 'A' is reading 90%

The 'As Found' AGAF reading (3D-Monicore) for APRM 'A' is 1.000 THEN, a single LPRM inputting to APRM 'A' fails to a ZERO value signal The failed LPRM has NOT yet been bypassed Which ONE of the following describes the RESULTING AGAF reading for APRM 'A'?

The AGAF is reading...

A.

Lower than 0.980.

B.

0.980 to 0.999.

c.

1.001 to 1.020.

D.

Higher than 1.020.

Answer: D Explanation:

D is corm3 - Reference LP85411 and CPS 9431.60, throughout this discussion. AGAF calculation: AGAF = % CTP +

APRM d i n g.

Conditions prior to the LPRM failure:

90% APRM reading resulting from 33 gqd LPRM flux signals Conditions poat-LPRM failure ('bad' LPRM still feeding the APRM):

(32 i33) x 90% = 87.3% APRM leading AGAF is now reading: AGAF = 90% + 87.3% = 1.031 o

o A, B, and C are incorrect - For the reasons described above.

I Question# 17 I

SystemEvolutiou Name:

I category:

Average Powex Range MonitorRocal Power Range Plant systems Monitor

~

KA Statement:

Knowledge of the physical connections and/or cause-efect relationships between APRM/LPRM and the following:

Process computer, performance monitoring system

References:

Oblectlve:

I Quatiou Source:

I Level of DiMculty:

LP85211.1.13.3;.1.15.1 New I

3.2 LP8541 I, APRh4mRM system CPS 9431.60, APRM Gain Adjustment Date Written: I 02/07/05 I Author:

1 Ryder Comments:

Operational Validity basis for this question: I) Operators must understand the meaning of the AGAF value, rather than just~ognizewhetheran'asfound'valueisSATorUNSAT@erTechSpecSR3.3.1.1.2);2)operatorsmust recognize that an 'as found' value >I,000 is m-conservative, where a value <I,000 is relatively conservative (albeit still a potmtial Tech Spec concern); and 3) operators must recognize that even a single LPRM failure (hard upscale or hard downscale) can inop the APRM, for sake of the ~ ~ r ~ e i l l a n ~ e requirement (3.3. I, 1.2). Depending on current rod pattern, total rod density, and fuel distribution, the failed-to-zero condition of an LPRM that was already producing a much weaker flux signal (as compared to the other 32 that feed the given APRM), may not affect the aggregate signal enough to push the AGAF value above the 1.020 upper limit; this is the reason for the stem's opening 'ASSUME' statement

1 Question# 18 I Referenca provided to examinee:

Referenca:

ic None LP85217, Reactor Core Isolation Cooling CPS Drawing E02-1R199, Sheets 6 and 9, RCIC Schematic Diagmm CPS 9054.03, RCIC Simulated Auto Actuation Test CPS 3310.01, RCIC C w Level RO/SRO I Tier:

I Group:

I KA:

I ROIR:

I SROIR:

I Both I

2 I

1 1 217000K4.06 1 3.5 I

3.5 I

Higher SystedEvolution Name:

I Catqory:

Reactor Core Isolation Cooling (RCIC)

I Plant systems KA Statement:

Knowledge of the REACTOR CORE ISOLATION COOLING (RCIC) design features andor interlocks which provide for the followiug: Manual initiation Operators are ready to MANUALLY initiate RCIC &om its normal standby lineup.

Which ONE of the following explains why the RCIC Manual Initiate pushbutton MUST be HELD DEPRESSED FOR 6 SECONDS?

To allow enough time for...

A.

RCIC Pump Supply to Turbine Lube Oil Cooler Valve, 1 E5 1 -F046, to fully open, enabling the opening circuit for RCIC Steam Supply Valve, lE51-FO45.

a logic time delay device to energize, enabling the opening circuit for RCIC Steam Supply Valve, lE51-FO45.

the Ramp Generator to BEGIN its ramping period.

the Ramp Generator to FINISH its ramping period.

B.

C.

D.

Answer: B Explanation:

B is c o d - Pa Lp85217, page44, and drawing E02-1RI99, Sheets 6 and 9. The TD device shown on Sheet 9 must be energized (timed out) before the initiation signal can enable the F045 opening circuit. CPS 9054.03, Section 8.2.2.2, validates that this TD device is calibrated for about 6 seconds. Only after holding the pushbutton depressed for about 6 seconds does Fad5 begin to open. Direction on how to manually initiate RCIC is found in CPS 3310.01, Section 8.1.3.

A is incorrect - Refer to LP85217, pages 44-47, for the explanation related to all of the distracters. There is NO electrical connection between the open limit switch for F046 and the opening circuit for F045.

C and D are in-t

- The Ramp Generator does not even came into the picture until 6 to 9 seconds &

the F045 v a l v e s w t o o p e n. Referto LP85217,page45.

Oblectlve:

I Question Source:

I Level of DiMculty:

LP85217.1.15.7 New I

3.0 DateWlitten

1 04/28/05 I Author: I Ryder Comment% None

L With the plant operating at 90% power, a Safety Relief Valve (SRV) INADVERTENTLY OPENS.

Which ONE of the following predicts how a plant parameter INITIALLY responds when the SRV opens, ggi describes the reason why?

INITIALLY, indicated reactor...

A.

water level LOWERS, because of the RFV inventory lost through the open SRV.

power LOWERS, because the SRV opening causes a slight drop in reactor pressure.

water level RISES, because Feedwater Level Control immediately sees the additional steam flow.

power RISES, because of the reduced feedwater inlet temperature.

B.

C.

D.

Answer: B Erohatlon:

B is correct - Per USAR, Section 15.1.4.3.3, the SRV w i n g initially pmduces a slight depmswization transient. Per Gcnmc Fundamentals knowledge, the dmp in reactor prcapure produces more voiding, which initially lowm reactor power (see LP85756S. page 26, fa an analogous 'off-normal' prrssurc transient, which validates the mlationship b e e n prcapure and power).

A and C arc incorrect - Per h a

c Funrlamentps knowledge, as well as the USAR discussion ahove, the initial dcpnapurization CBUS*I more voiding. which resnlts in an initial RISE of reactor water level ('swell' bansieot). The open SRV diverts main steam flow away from (is upstream ot) the main steam line Bow element (see LP85239, Figure 1). This m l t s in the Feedwata Level Contml System immediately s&ing a lower steam flow, not a higher steam flow.

D is incorrect - P a CPS 4005.01, Loss of Feedwata H a h g, the SRV opening is a loss of feedwater heating wmi The cscspe ofsteam to the mpprcasion p l

diverts it away firom the main turbine and the extraction steam supply.

Reactor M w t m tanpaaturc lowcrs (i.e., a g r a t a amount of com inlet sub-cooling is produced) and this positive d v i t y addition should help to raise reactur power. Howcvs, this is not the INITIAL power response. The dcpresswization cvent immediately lowm reactor power.

L Obieftlve:

I Qnation Source:

I Level o f ~ m c n ~ t y :

None NCW I

3.6

I Question# I 10 1

References:

i LP85570, Feedwater Level Control System CPS 3103.01, Feedwater RO/SRO I Tier:

I Group:

I KA:

I R O E I

SROIR I

Cog Level Both I

2 I

1 I 259002K5.07 I 2.7 I

2.7 I

Lower SystemlEvolutiou Name:

1 cateory:

Reactor Water Level Control System KA Statement:

Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM: Turbine sped control mechanisms: TDRFP I Plant systems The plant is operating at rated power with Feedwater and Feedwater Level Control in their NORMAL configurations.

Per CPS 3103.01, Feedwater, which ONE of the following describes the expected CURRENT Feedwater Level Control operating configuration, 4 the reason for that configuration?

A.

TDRFP Manual Speed Potentiometers are set at the LOW SPEED STOP position, to expedite taking manual control of a locked up TDRFP.

RFPT Flow Controllers are in MANUAL, to expedite the Emergency Restart of a tripped TDRFP.

TDRFP Manual Speed Potentiometers are set at the ZERO speed position, to expedite the Emergency Restart of a tripped TDRFP.

RFFT AUTOMAN XFER switches are in MANUAL, to expedite taking manual control of a locked up TDRFP.

B.

C.

D.

Answer: c Erplauatlon:

Ciscorrec-PsCPS3103.Ol,Sections2.1.7,8.1.4,and8.3.2. TheManualSpeedPotmust beatmspeed(hlly CCW) io order to reset the control logic for the Lp and HP wnml valves. This ensures that when the operator RESETS the TDRFP, immediate control of speed will be available. During a normal plant and feedwater system startup, the last time that operators have reason to manipulate the Manual Speed Pot is in Section 8.1.4.20(i)(e). It is here that the pot is set to ZERO speed position and should m a i n there with Feedwater and Feedwater Level Control in their normal configurations.

A is i n c o d - For the reasons described above B is i n c o d - P e r CPS 3103.01, Section 8.1.8, the TDRFPs are being controlled by the Master Level Controller, which means that each RFPT Flow Controller is in AUTOMATIC.

D is inwrrect - Per LP85570, page 24, and Figure 13, these XFER switches are in AUTO (pushbutton depressed) whenever any flow controller is wntrolling the TDRFP.

Objective:

I Question Source:

1 Level of Difncuity:

LP85570.1.14 New I

2.2

i-L-

I Question# I 11 I ROISRO 1

Tier:

I Group:

1 KA:

1 ROJR:

1 SROIR:

1 Cog Level Both I

2 I

1 I 261000K4.01 I 3.7 I

3.8 I

Higher SystemlEvolution Name:

I category:

Standby Gas Tmtment System KA Statement:

Knowledge of STANDBY GAS TREATMENT SYSTEM design features andor interlocks which provide for the following: Automatic system initiation I Plant System The plant is in MODE 4, with the following:

BOTH trains of Standby Gas Treatment (VG) are in a STANDBY lineup The ENTIRE Div 1 NSPS Bus is in an OUTAGE THEN, the CNMT Bldg Exhaust Radiation Monitor, 1RIX-PRO01 C, fails UPSCALE and produces a trip WITHOUT operator action, which ONE of the following identifies the VG Trains that are RUNNING, explains why?

A.

BOTH, because with the Div 1 NSPS Bus outage, the failure of 1RIX-PROOlC completes the one-out-of-two-twice initiation logic.

NEITHER, because no VG initiation signal is present.

ONLY Train A, because no Train B initiation signal is present.

ONLY Train B, because with the Div 1 NSPS Bus outage, VG Train A auto-initiation signals are disabled.

B.

C.

D.

Answer: B Explanation:

B is correct - Rder to CPS 5 140.61 for the oneout-of-two-twice radiation monitor combinations that can produce a VG initiation signal. Although both trains are capable of auto-starting, a trip condition on a single radiation monitor channel (IRIX-PROOIC, only) does not satisfy the one-out-two-twk initiation logic.

A is incorrect - The Div 1 NSPS Bus outage does produce a power failure trip of any of the I RIX-PRO0 I channels.

Per LP85273, page 74 (Attachment B), these radiation monitors get their power from Auxiliary Power System MCCs, a

from NSPS (inverter) ~ W R.

Therefore, the upscale trip produced by the PROOlC failure does not, alone, satisfy the one-out-of-two-twice VG initiation logic.

C is incoma-This is plausible to the candidate who believes the NSPS Bus outage &m produce a power failure trip on the PROOIA channel, who confuses the onwut-of-two-twice logic combinations that produce a VG initiation signal, and who mistakenly associates two of the PRO01 channels with VG Train A and the other two channels with VG Train B. This ladder logic is a fairly common weakness among candidates who have not mastered this system knowledge.

D is incomct-This is quite plausible to the candidate who believes the NSPS Bus outage &m produce a power failure hip on the PR00lA channel, and recalls that @R CPS 3509.01 COOI, Appendix A, page 25) the Bus outage disables the auto-start of VG Train A on a W A

signal (High DW Pressure, Low-Low Level, only). The radiation monitor initiation signals are Wt affected by this Bus outage.

I Question# I 12 ]

L predictions, use, pmcedum to mrmt, conml, or mitigate the consequences of those abnormal conditions or Reactor power is 25% when the following occurs:

Main turbine and generator trip (cause unknown)

ALL LOADS on 6900V Bus 1B become DE-ENERGIZED There are NO indications of electrical faults on any buses or breakers Which ONE of the following describes how the operator restores power to 6900V Bus 1B7 L-A.

Place the Sync Switch to ON for the UAT IB feeder breaker, position the UAT 1B feeder breaker control switch to CLOSE, then place the Sync Switch to OFF.

Verify the RAT source is dead, then position the UAT 1B feeder breaker control switch to CLOSE.

Place. the Sync Switch to ON for the RAT feeder breaker, position the RAT feeder breaker control switch to CLOSE, then place the Sync Switch to OFF.

Verify the UAT 1B source is dead, then position the RAT feeder breaker control switch to CLOSE.

B.

C.

D.

Answer: c EXPIIMUIJU:

Cis cnrrct-Refa toLp85571, Figure4, andpagc31, and to CF'S 3501.01, Section 8.1.1. The normal feed for 6.8 KV Bus IB i s h UAT IB. When the geuuator trips, UAT IB (and 1A) are lost, forCng the auto-hansfer of 6.9 K V Bw IB to the RAT (its reserve a).

Because the bus is 'dead', procedure section 8. I. 1 is used to complete the transfer manually. Steps 8. I.4,5, and 7 me hlui-ed in this answer choice.

A and B cannot recall the 'Msin' (UAT IB) versus 'Resave' (RAT) ~ O W R sources for this 6900V bus.

D is ineonrct-Whether the bus is alive or dead is irrelevant The Sync Switch must still be used.

incorrect - l l c UATs me lost on the garemtor bip. These choices are distracting to the Candidate who t

Obldve:

I Question Source:

1 Level of Dirncuiw:

LP85571.1.4.6 New 1

2.5

1 Question # I 12 1 Referrneu:

LP85571, Auxiliary Power CPS 3501.01, High Voltage Auxiliary POWR System

1 Question # I 13 I i-RO/SRO 1

Tier:

I Group:

I KA:

I R O I k 1

SROIR: I C w Level Both I

2 I

1 1 262001 A4.05 1 3.3 I

3.3 I

Higher SystemlEvolutlon Name:

I Cataory:

A.C. Electrical Dishibution KA Statement:

Ability to manually opaat e and^ monitor in the control room: Voltage, current, power, and frequency on A.C. buses 1 Plant systems

References:

After transferring loads to the ERAT, operators are preparing to shut down the RAT SVC fiom the main control room, in accordance with CPS 3505.03, RAT & ERAT Static VAR Compensators.

EXISTING readings for the RAT SVC, at panel P870, are as follows:

4,220Volts

-4MvARs LP85305, Static VAR Compensator CPS 3505.03, RAT & ERAT Static VAR Compensators Which ONE of the following identifies the FINAL voltage value that the SVC Voltmeter should ramp to AFTER the operator places the RAT SVC control switch to OFF?

A.

4,060Volts B.

4,140 Volts

c.

4,300Volts D.

4,380 Volts Answer: c Exphation:

C is correct - Refer to CPS 3505.03, Section 8.3 and Appendix A. The - MVARs indication means that the SVC is acting to hold down voltage. When the SVC is removed from service, the rsulting (uncompensated) voltage will ramp up (to above4220 volts). The rule of thumb is 20 volts per MVAR In this case, that amounts to +SO volts, or a FINAL value of 4,300 volts.

A, B, and I) are incorrect - For the reasons dseribal. They have face validity and are plausible to the candidate who eith~, does not recall the rule of thumb (applies 40 volts per MVAR instead of 20 volts per MVAR), or cannot distinguish between a + MVAR reading and a - MVAR teading.

Obldve:

I Questloo Source:

1 Level nf Difticuity:

LP85305.1.10.2 New I

2.6 Datewritten

I 02/09/05 I Author: I Ryda Commentl: None 1

1 Question# I 14 1 LP85434.1.4.3 L

Which ONE of the following describes the impact of a loss of the NORMAL power supply to a DIVISIONAL NSPS Inverter Cabinet, with the Cabinet in its NORMAL operating configuration?

A.

The bus loads will REMAIN ENERGIZED as a 125 VDC bus automatically begins to feed the Inverter section.

The bus loads will REMAIN ENERGIZED as a Static Switch automatically transfers them to an alternate 120 VAC supply.

The bus loads will BECOME DE-ENERGIZED and remain that way until operators MANUALLY transfer them using the REVERSE TRANSFER pushbutton.

The bus loads will BECOME DE-ENERGIZED and remain that way until operators MANUALLY transfer them using the TRANSFER SWITCH.

B.

C.

D.

CPS Opsations Exam Bank, Question #I0278 (DIRECT, 2.7 editorial changes only)

L Refereoca:

Answer: B LP85434, Nuclear System Protection System (NSPS)

CPS 3509.01, Inshument Power System Exphatioo:

B is correct - Rcfa to LP85434, page 1 I, 13.14, and Figures 2b and 3. A Divisional NSPS Cabinet is normally supplied from the m i a t e d 125 VDC Bus. With the Cabinet in its normal opting configuration, the Transfer Switch is in theINVERTERposition. This allows the Static Switch to autc-transfer bus loads to the alternate 120 VAC SUPPlY.

A is incorrect - This describes the impact of a m-Divisional NSPS Invma Cabinet. See LP85434, Figure 6.

C is incorrect - Opaators use this pushbutton to msnually transfer bus loads to the alternate 120 VAC supply per CPS 3509.01, Section 8.1.4.

D is in-

- For the reasons associated with the correct answer.

Date Wrltteo: 1 02/10/05 I Author: I Ryda Commeob. None 1

1 Question# I 15 I L,

The following statement describes a FACT concerning battery hydrogen production:

The RATE at which a battery produces hydrogen during an EQUALIZING charge is DIRECTLY propohonal to the battery capacity (in Ampere-Hours).

Consider the above FACT when answering the following question.

An EQUALIZING Charge of the Div 3 Battery is in progress, when normal battery room ventilation is lost.

Which ONE of the following:

(1) predicts the RATE at which Div 3 Battery Room hydrogen concentration will rise, WITHOUT operator action, AFTER normal battery room ventilation is lost, and (2) describes the required action?

The hydrogen concentration RATE OF RISE will be...

L A.

(1) GREATER in the EARLY hours of the Equalizing Charge.

(2) IF room hydrogen concentration reaches 2%, THEN open the battery room door and ventilate with a portable air blower.

B.

(1) GREATER in the LATER hours of the Equalizing Charge.

(2) IF room hydrogen concentration reaches 2%, THEN open the battery room door and ventilate with a portable air blower.

C.

(1) GREATER in the EARLY hours of the Equalizing Charge.

(2) Open the battery room door; WHEN hydrogen concentration reaches 5%,

THEN ventilate the room with a portable air blower.

D.

(1) GREATER in the LATER hours of the Equalizing Charge.

(2) Open the battery room door; WHEN hydrogen concentration reaches 5%,

THEN ventilate the room with a portable air blower.

Answer: B Explmation:

B is comet - Concaning Part (1) of the question, d m to information extracted from the Web source:

wf This is banery training p r o g m I pnsentation associated with the U.S. Bureau of Mines. These slides show that the rate of hydrogen production ('H) is proportional to the aw~&' of the battery (in amperehows). Since the battery capacity rises over the charging I

I Question # 1 15 I Cog Level RO/SRO I Tier:

I Group:

1 KA:

I R O m.

I SRO1R:

I Both I

2 I

1 I 263000A2.02 I 2.6 I

2.9 I

Higher SystemlEvdution Name:

1 Category:

D.C. Elechid Dishibution I Plant systems

~-

KA Statement:

Ability to (a) d c t the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION and (h) based on those predickons, use proceduresto correct, conml, &mitigate the consequences of those abnormal conditions or operations: Lass of ventilation during charging periad,so~doestherateofgaspmduction('H). ConcemingPart(2)ofthequestion,refertoCPS 3412.01, Section 8.2.3.

I A, C, and D me incoma - For the reasons described above.

I Oblectlvc I

Questiou Source:

1 Level of Dimculty:

None New I

3.0 Referenfes provided to examinee:

Referenca:

1 None CPS 3412.01, Essential Switchgear Heat Removal (VX)

Datewritten: I 05/02/05 I Author: I Rydm Commenb:

This question is categorized as Higher Cognitive (HCL), because the Candidate must 'associate' the given stem claim regarding the rate of hydrogen production with hiher understanding of how a battery's capacity changes over the period of being recharged. It is also HCL because, as a closed-reference question, in order for the Candidate to eliminate choices 'C' and W', hdshe must recognize the danger that would exist if a portable blower (Le., a potential spark producing device) were to be started when hydrogen is already above a combustible concentration (nominally 4-

I Question# I 16 1 L

The plant is operating at rated power, with the Monthly surveillance for DG 1B in progress, with the following:

DG 1B is running loaded at 3,800 KW THEN, the NORMAL control signal to the Woodward Governor is lost CRS determines the need to CORRECT the DG 1B operating condition that has resulted fiom this Governor malfunction Which ONE of the following:

(1) predicts the response of DG 1B to the governor control signal failure, (2) describes the action necessary to correct the DGs current operating condition?

and W

A.

(1) Engine speed REMAINS THE SAME, but Load RISES.

(2) Emergency STOP the DG fiorn the main control room.

(1) Engine Speed REMAINS THE SAME, but Load LOWERS.

(2) RAISE the SETPOINT for the Mechanical Governor, locally.

(1) Engine Speed RISES, but Load REMAINS THE SAME.

(2) LOWER the engine speed using the Governor control switch.

(1) Engine Speed LOWERS, but Load REMAINS THE SAME (2) RAISE the engine speed using the Governor control switch.

B.

C.

D.

Answer: A E1pllo8tion:

A is eareet - Refes to CPS 9080.02, page 28 CAUTION. The monthly surveillance has the DG loaded in parallel with the off-site power. PR LP 85261, pagcs 15-16, the elechical governor is normally controlling the engine speed and wha~

the electrical signal is lost (fails low), the mechanical governor assumes control at a 5% HIGHER governor sapoint Because the Do is paralleled with off-site, engine speed cannot change, but the DG &ss pick up nxm load.

Although we cannot necessarily diet that it will pick up 5% additional load (for a new load of 3,990 KW), it will pick up an amount o i ~ d ~ d thet cauxs thebG tdoperatc Section 6.2.1 I of CPS 9080.02). This question suggests that it is this operating condition that needs to hc c o m t d.

Once the CRS decides to correct the condition, the only way to do so is by CQIQ&.&

unloading the DG. With the feilpd electric governor, and with NO pmcedurc guidance that would allow operators tu manually luwcr the mechanical govanors sapoint (locally), the requid action is to Emergency STOP the machine.

B is i n w m - For the reasons described above.

ofirs continuous rating of3.S75 KW(sre

I Question ## I 16 I

References:

C and D are incorrect - These choices suggest the 'predicted' response of the DG if it were a paralleled with off-site.

The Candidate is expected to know that the machine Sumillsnee.

with off-site when mnning loaded for the Monthly LP85264, Diesel Genwtor/Diesel Fuel Oil CPS 9080.02, Diesel Generator 1B Operahility (i.e., the Monthly)

Oblectlve:

I Ouestlon Source:

I Level of Difficulty:

None New I

3.5 DatcWrilten

1 W/I 5/05 1 Author:

I Ryder Comment%

This question is an ROBRO one, and is an SRO-ONLY question, for the following reason:

1.

It may appear, at first, that the question is presented in a way that is consistent with other 'A2' type exam questions that have been categorized as SRO-ONLY, but a closer look shows that it is quite different.

me last stem condition bullet has pre-cmpted the need for the SRO to make a decision about whether to correct the Do operating condition, M not.

me suggested choices for the Part (2) 'required action' are mf a set of choices from which Qdy the SRO would be exwed choose. Rathex, each choice challenges the Candidate (both RO/SRO) to recognizing what is the only possibldpermitted way of correcting the operating condition.

Th&,

this question is presented in a way that really amounts to requiring only several pieces of

'systems' type of howledge: 1) the configuration that the DG is in before the governor failure (i.e.,

paralleled with off-site), and on the electric govfmor, 2) how the DG engine speed and load respond as a result of the 5% off-set behueen the electric and mechanical setpoints; and 3) recognition of the fact that with them being no electric governor contml, the only solution to emergency STOP the DG As such, this is a question that should bc on both the RO and SRO Exams.

2.

3.

4.

5.

1 Question # I 17 1 L

RO/SRO:

1 Tltr:

I croup:

I KA.

I R O I R I

SROIR I

Cog Level Both I

2 I

I I 400000K6.06 I 2.9 I

2.9 I

Higher SystedEvolnHon Name:

I category:

KA Statement:

Knowledge of the e f k t that a loss or malfunction of the following will have on the CCWS: Heat exchangers and Component Cooling Wata System (CCWS) 1 P ~ M t S ~ t R l U L

Explanatloo:

C is correct - Per LP85208, page 29. SX system pressure is lower than CCW system pressure. CCW inventory will be lost through the tube leak and find its way into the Lake.

A, B, and D are incorrect - Per LP85208, page 29. These are all indicative of a WS-to-CCW leak (WS at higher prssure than ccw).

Date Written: I 05/02/05 I Author: I Ryder Comments: None

I Question# I 18 I ROISRO I

Tier:

I Group:

I KA:

I R O N I

SROIR I Cog Level Both I

2 I

2 I 204000K3.06 I 2.6 I

2.7 I

Higher SystemlEvolution Name:

I Category:

Reactor Water Cleanup System KA Statement:

Knowledge of the &ect that a low or malfunction of the REACTOR WATER CLEANUP SYSTEM will have on the following: Area radiation levels I Plant systems Following a maintenance outage on RWCU Filter A, with the plant operating at rated power, the following conditions exist:

RWCU Filter System Functions Interlock Switch is in the SYS A position RWCU Filter A has just been manually re-filled in preparation for Backwash Backwash Receiving Tank (BWRT) level is reading 38% (local panel)

Operators are unaware that BWRT level is reading about 25% LOWER than ACTUAL level in the Tank Which ONE of the following describes the POTENTIAL consequence associated with the NEXT operator action related to the Backwash of RWCU Filter A?

A.

B.

C.

D.

Answer: B RWCU system isolation on High Differential Flow Higher than normal area radiation level on CNMT el. 778 RWCU system isolation on Equipment Room High Temperature Higher than normal area radiation level on Auxiliary Bldg el. 737 Explanation:

B is correct - Pa CPS 3303.02, Sections 8.7.6 and 8.6.5, the NEXT operator action is to return the System Function Interlock Switch to the NORM(a1) position. Candidates need not recall (from memory) such a procedural action as this; rather, they must only recognize that the filter backwash requires the System Functions Interlock Switch be in NORMAL. Pa LP85204, page 29, this can result in overflowing the Backwash Receiving Tank (BWRT) through a continuous vent connected to the CNMT building HVAC exhaust ductwork This will cause significant contamination and elevated area radiation levels throughout the CNMT spaces.

A and C are inumect - The Filter is still in the Shutdown Mode (see CPS 3303.02, Sections 2.2.8 and 8.6). As such, the Filter is still isolated fium the RWCU system and system isolations are not possible.

D is incorrect - Ibis is the location of the RWCU Pumps. There is no physical, or ventilation air-flow, connection between an existing high area radiation level in the CNMT building and the Auxiliary Building. And, there is no system perhubation being suggested by the stem conditions that could cause a RWCU Pump problem (e.& a seal leak) that would m l t in high radiation levels in that pump area.

Oblectlve:

I Queation Source:

I Level of DIMculty:

Lp85204.1. I5 New I

3.5

I Question# I 18 I RO/SRO I Tier:

I Group:

I KA:

I ROIR:

I S R O I R 1

Cog Level Both 2

2 I 204000K3.06 I 2.6 2.7 Higher I

I I

I SystedEvoluUon Name:

I category:

Reactor Water Cleanup System KA Statement:

Knowledge of the &ecl that a loss or malfunction of the REACTOR WATER CLEANUP SYSTEM will have on the following: Area radiation levels I Plant systems

~

Refereuca provided to examinee:

None

References:

LP85204, Reactor Water Cleanup System CPS 3303.01, Reactor Water Cleanup System CPS 3303.02, RWCU Filter Demineralizer Operating PrOceduIe

I Question# I 19 I RO/SRO I Tler:

I Group:

I KA:

I ROIR:

I SROIR I Cog Level Both I

2 I

2 I 201005Ks.IO I 3.2 I

3.3 I

Higher SystemlEvolution Name:

I Category:

Rod Control and Information System (RCIS)

KA Statement:

Knowledge of the operational implications of the following concepts as they relate to ROD CONTROL AND INFORMATION SYSTEM (RCIS): Rod Withdrawal limiter 1 PlantSystems L

Which ONE of the following describes a situation where the Technical Specifications ALLOW (permit, without administrative restrictions) ALL normal control rod movements (In & Out) to be performed?

Reactor Power is...

A.

40%; the light above the LO POWER SET PT is OFF, and the light above the LO POWER ALM PT is OFF.

10%; the light above the LO POWER SET PT is ON.

75%; the light above the HI POWER SET PT is OFF.

45%; the light above the LO POWER SET PT is OFF. and the light above the LO POWER ALM PT is ON B.

C.

D.

Answer: D Explanation:

Disco~t-RefatoLP85401,pagcy23-24.andFigun:4,toTechSps3.3.2.1,andtoCPS 3005.0I.Section6.2.for all of the answer choiccs. Reactor plwer is within h e range when the RWL must he OPERABLE (>2Y% RrP and at or helow the High Power Squint (HPSP) of 70% RTP). The fact that the LO POWER SET PT light is OFF. whdc Ihe LO POWER ALM PT light is ON, indicates n jn&y-functioning Low Power Function of the RWL, permitting agy rype of rod movemcnt (subjec? to the built-in notch restraints of the RWL itself). T h e arc NO Tech Spec administrative repainions with these conditions.

A is i n c o w - Reacror power is within the range when the RWL must be OPERABLE (>29% RTP and at or helow the High Power Sapoint (HPSP) of 70%). However. the fact thnt both of these lights are OFF indicates that the Low Power Function ofthe RWL (normally cnablcd by the Rod Panern Controller, RPC) is in fact bypassed, making the RWL INOPERABLE. Per TS LCO 3.3.2.1.k all control rod WITHDRAWALS must be immediately suspended Insertions an m'll allowed.

B is inmmCt - R m o r power is below Ihc Low Power Setpoint (LPSP) of 16.7% RTP. However, the fact that the light is ON indicam that the RPC is INOPERABLE. P a Tech Spec LCO 3.3.2. I.B. all normal rod movements (in and out) must be immcdintely suspended.

C is incorrect-Reactor power is above the High Power Setpoint (HPSP) of 70% RTP. However, this light k i n g OFF indicates thnt the High Power Function of the RWL is bypassed, making the RWL INOPERABLE. Per Tech Spec LCO 3.3.2.1.A, all control rod WITHDRAWALS must be immediately suspended. lnsmions arc still allowed.

Objective:

I Question Source:

[ Level of Difficulty:

None New I

2.7

I Question # I 19 I References provided to examinee:

References:

Tler:

I Group:

I KA:

I ROIR:

1 SROIR I

Cog Level I

3.2 I

3.3 I

Higher I -n,nnrvc In I RO/SRO I None LP85401, Rod Control and Information System CPS Tech Spec 3.3.2.1, Control Rod Block Inshumentation CPS 9436.05, RPC Low Power Setpoint Channel Calibration CPS 9030.01CO21, RPC Low Power Setpoint Checklist CPS 3005.01, Unit Power Changes Mom L

L I L U L V V. 2 N. L V I I

I SyatemlEvolution Name:

I category:

Rod Control and Information System (RCIS)

KA Statement:

Knowledee of the owrational implications of the following concepts as they relate to ROD CONTROL AND I Plant systems 1 INFORMITION SLSTEM (RCIS): Rod withdrawal limiter I

L Question # [ 20 ]

i _

RO/SRO 1

Tier:

1 Group:

I KA:

I ROIR I

SROIR I Cog Level I

I 3.7 I

3.7 I

Lower Both 2

2 I 202001 A4.01 1 SystedEvolutlon Name:

I category:

L Recirculation System I Plant Systems RO/SRO 1

Tier:

1 Group:

I KA:

I ROIR I

SROIR I Cog Level I

I 3.7 I

3.7 I

Lower Both 2

2 I 202001 A4.01 1 SystedEvolutlon Name:

I category:

References provided to examinee:

Refereuces:

L None LP85202, Reactor Recirculation System CPS 3302.01, Reactor Recirculation System CPS 5003-4C, Rsirc MG A Interlock Bypass KA Stalemeat:

Ability IO manually opw te andlor monitor in the control room: Recirculation pumps The Reactor Recirc Pumps are being manually transferred from FAST to SLOW speed in accordance with CPS 3302.01, Reactor Recirculation.

Which ONE of the following describes ONE of the operator actions involved in performing this transfer?

In the main control room, the operator...

A.

B.

C.

manually closes the LFMG Motor Breakers, CB-IA(B).

positions the FCVs to about 19% open BEFORE the transfer.

verifies the RECIRC MG A(B) INTERLOCK BYPASS annunciators are extinguished.

positions the FCVs to about 76% open AFTER the transfer.

D.

Answer: A Explanation:

A is correct - Per CPS 3302.01, Section 8.1.3. Although the hansfer sequence logic would automatically close CB-I A and B (see LP85202, page 28, and Figure 7). the operating procedure directs the operator to manually close these breakas before initiating the hansfer sequence.

BandDarein~-PerCPS3302.01,Sections8.1.3.4and8.1.3.7.

About IO%opmBEFOREthetransfer,and about 90% open AFTER the transfer.

C is incorrect-PR CPS 3302.01, Section 8.1.3.1, and CPS 5IXJ3-C. These interlocks are intentionally bypassed for this evolution, causing these annunciators to be in alarm (set extinguished).

I Question# I 21 ]

RO/SRO I

Tier:

I Group:

I KA:

I ROIR:

I S R O I R I

Cog Level Both 1

2 I

2 I 214000A3.01 I 3.4 I

3.3 I

Lower SystemlEvolutlon NaG:

I category:

Rod Position Information System KA Statement:

Ability to monitor automatic operations of the ROD POSITION INFORMATION SYSTEM including:

Full Core Display I Plant systems The RO is performing a control rod coupling check per CPS 3304.02, Rod Control and Information System.

WHILE a continuous withdrawal signal is being applied, which ONE of the following indicates that the control rod is UNCOUPLED?

A.

B.

CRD drive water flow reads 5 gpm.

ROD OVERTRAVEL annunciator is NOT received C.

D.

Answer: D Red 'full-out' light is LIT on the full-core display.

Rod position is BLANK on the hll-core display.

Explanation:

D is correct - Per CPS 3304.02, Section 8.2.6 NOTE, rod position would be blank on the RlDM (full-core display) for an uncoupled rod Aisincome3-PaSEction8.1.10NOTE. WhetherCRDM~saregood(1-3gpmstallflowindicated),orbad (something bigha than 1-3 gpm), stall flow is unaffected by the status of the control rod blade (coupled, or uncoupled).

This is the m n

why stall flow is to be used ONLY as an indication of seal condition.

B is income3 - P a CPS 3304.02, Section 8.2.6 NOTE, the Rod Overtravel annunciator be received for an uncoupled rod Cis incorrect-Per CPS 3304.02, Section 8.1.10.1, 2"6 bullet Objdve:

I Ouestlon Source:

I Level of Dimculty:

LP85401.1.4.9 New I

2.8 References provided to exarnlnee:

References:

I None I CPS 3304.02, Rod Control & Infomation System Datewritten: I 05/02/05 I Author: I Ryder Comments: None

I Question# I 22 I Relwrncea provided to examlnee:

References:

KA Statement:

Ability to (a) predict the impacts of the following on the FUEL POOL COOLING AND CLEANUP: and (b) based on thm predictions, use procedures to c o d, control, or mitigate the consequences of those abnormal conditions or operations: High fuel poOr temperature Which ONE of the following:

(1) describes a POTENTIAL or ACTUAL concern associated with a Spent Fuel Storage Pool Temperature that has risen to 152'F and has STABILIZED there, None LP85233, Fuel Pool Cooling and Cleanup CPS 3317.01, Fuel Pool Cooling and Cleanup CPS 5040-1F, High Temp Spent Fuel StoraKe Pool and (2) describes the operational impact?

A.

B.

C.

D.

Answer: D (1) Exceeds the ORM OPERATING REQUIREMENT for Spent Fuel (2) Movement of fuel assemblies in the pool is NOT permitted.

(1) Results in elevated humidity in the Fuel Building.

(2) If the reactor is operating, a normal plant shutdown is required.

(1) Exceeds the TECHNICAL SPECIFICATION LCO for Spent Fuel (2) Movement of fuel assemblies in the pool is NOT permitted.

(1) Results in airborne radioactivity in the Fuel Building.

(2) If the reactor is shutdown, it must remain shutdown.

Storage Pool temperature.

r Explanation:

I D is correct - Part (I), p a Lp85233, page 45. Part (2). per CPS 3317.01, Section 4.6.

A and C are incomct - There is no T& Spec K O, or ORM OR, related to Spent Fuel Storage Pool Temperature.

B is incorrect - Although Part (1) is correct, there is no procedural requirement for shumng down the plant. See the attached CPS 5040-IF and CPS 33 17.01, Senion 8.2.4, to validate this claim.

1 Question# I22 I Date Written: I 04/29/05 I Author: I Ryda Commenm None This question is presented on the RO Exam (and is not considered an SRO-ONLY type) because the operational impact portion requires only the recall of an operating procedure Precautiofiimitation; it does not require any operational decision-making (reserved for the SROs responsibility), nor does it require any form of application (which might or might not be reserved for the SRO) of the information contained in that PrecautiodLimitation.

1 Question# [ 23 I RO/SRO: I Tier:

I Group:

I KA:

I ROIR I

SROIR I

Cog Level Both I

2 I

2

[ 241OOOK6.01 I 2.8 I

2.9 I

Higher

References:

LP85241, Steam Bypass and Pressure Conml System LP85576, Computer UPS SystetemlEvolution Name:

I category:

Reacorfhbine Pressure Regulating System I Plant systems

~

KA Statement:

Knowledge of the effect that a loss or malfunction of the following will have on the REACTONTURBINE PRESSURE REGULATING SYSTEM A.C. electrical power A main turbine roll-up is in progress, with the following:

Control Building MCC C is de-energized and has clearances installed THEN, FWE MINUTES AFTER the RO depresses the 1800 pushbutton for main turbine Speed Set RFM, UPS Bus 1B is lost Which ONE of the following describes the plant response?

The main turbine...

A.

B.

C.

D.

Answer: c STOPS rolling up and STABILIZES at its CURRENT speed.

TRIPS, and the turbine bypass valves fail OPEN.

TRIPS, and the turbine bypass valves fail SHUT.

RETURNS to 100 RFM and STABILIZES there.

Explanation :

C is Mmct - Per LP85576, page 15, and CPS 35OY.OlCOo6, page 28. With main turbine speed at <75% of rated speed

(.75 x I800 rated rpm = 1350 rpm), the main turbine trips due to de-energidon of the 24 VDC Trip Bus and Electrical Trip Solenoids. Even if the FAST Starting Rate has beRl selected (see LP85241, page 30, and CPS 3105.01, Section 8.1.6), the machine will be running at NO HIGHER than a b u t 900 RPM, a! 5 minutes &a depressing the 1800 RPM pushbutton. In fact, per CPS 3105.01, Section 8.1.7.4 NOTE, it can take 3-4 minutes just see my speed increase on the machine. In this case, at the 5-minute mark, opmton shouldnt expect to see the machine speed any higher than about 200400 p.

A g time all power is lost to the TBV mnml circuits, the TBVs will fail shut. NOTE: The stem mdition regarding CB MCC C being de-energized ensure8 the intended failurn response of the TBVs; i.e., there mg be an auctioneering of power &tween this MCC and UPS IB) to the TBV circuits. Taking this MCC away, ensures the TBVs will fail shut.

I A, B, and D are incorrect ~ For the reasons described above.

Objective:

I Quertion sour-:

I Level of Difficulty:

LP85576.1.13.5 New I

3.6 CPS 35OY.O1C&6, UPS IB Bus Outage CPS 3105.01, Turbine

1 Question # I 23 1 Date Written:

1 05/16/05 I Author: I Ryder Comments: None

Question# 1 24 1 Which ONE of the following describes the impact of a loss of 6.9KV Bus lB?

A.

B.

C.

D.

Reactor Recirc Pump 1B can be started ONLY in SLOW speed.

Motor-Driven Reactor Feedwater Pump (MDRFP) will NOT run.

ONLY ONE Circulating Water hunp will run.

NEITHER Isolated Phase Bus Duct Cooler Fan will run.

Answer: B Explanation:

B is c o w - Pa LP85259, page 12. MDRFP is powered from 6.9KV Bus 1B.

A is i n c a w - P a LP85202, pages 9 and 27. Even a SLOW speed start of the RR Pump 1 B requires 6.9KV Bus I B poWR.

C is i n c a m - P a LP85275, pages 14 and 17. 6.9KV Bus IB powers only one of the 3 CW Pumps (CWP B). The other two powered iium 6.9KV Bus 1A. Th~here are no inter-pump starting permissives or trip signals that would inhibit the running of D is i n c o w - PcrLP85572, pages 8 and 13, and CPS drawing E02-lAP03 (and LP85571, Figure 4 for clarity). One of these fans is powered iium 6.9KV Bus LA, via 480V Unit Sub 1 J. Only the B Fan is lost if 6.9KV Bus IB is lost.

pumps, CWP A and e, on the 6.9KV Bus 1 A.

I ObJective:

I Oueation Source:

I Level of Diffhlty:

LP85259.1.4.4 New I

2.0 References provided to examinee:

References:

1 LP85259. Feedwater swtem 1 None LP85202; Reactor Reckulation System LP85275, Circulating Water System LP85572, Isolated Phase Bus Duct Cooling Datewritten:

1 Ou21/05 I Author: I Ryder Comments: None 1

I Question# I25 I ROISRO I

Tfer:

I Group:

I KA:

I RO1R:

I SROIR:

I Cog Level Both I

2 I

2 1 286MK4.07 1 3.3 I

3.3 I

Higher SystedEvolution N8me:

I categury:

Fire Protection System KA Statement:

Knowledge of FIRE PROTECTION SYSTEM design features and/or interlocks which provide for the following:

Diesel engine protection I Plant systems Y

Operators are testing the automatic start feature of the B Fire Pump. The operator places the Mode Selector Switch in TEST, and the following occurs:

At Time = 0 minutes, the engine begins to crank At Time = 4 minutes, the engine starts and runs At Time = 5 minutes, the engine stabilizes at 130% of rated speed At Time = 6 minutes, both a HIGH ENGINE TEMPERATURE alarm, and a LOW LUBE OIL PRESSURE alarm, are received on the XL3 fire alarm panels 30 seconds later, the operator manually stops the engine by placing the Mode Selector Switch to OFF Which ONE of the following identifies the TOTAL NUMBER of AUTOMATIC engine protective features (i.e., should have prevented the engine fiom running) that FAILED during this test of the B Fire Pump?

A.

1 B.

2

c.

3 D.

4 Answer: B Explanatloo:

B is correct - PR LP85286, pages 21 -25. One engine crank and rest cycle takes 30 seconds; the controller should have allowed no more than 6 total cycles (1 80 seconds...At Time = 3 minutes) before stopping the auto-start sequence and generating a Failure to Start alarm. This was the first failure of a protective action. The engine was allowed to reach 130% of rated sped and continue to run. The engine should have tripped (stopped) at 120% of rated speed. This was the s w n d failure of a protective action.

A is incorrect - For the reasons described above, C and D are inumect - PR the same reference cited above. N e i t h ~

the High Engine Temperature alarm, nor the Low Lube Oil Presure alarm, provide an automatic protective action; they are alarms, only.

Objective:

I Question Source:

1 Level of Dimculty:

LP85286.1.10.4 New I

4.0

I Question# 125 I m m / E x u t l o n Name:

I Category:

I Plant syst 4

Fire Protection System KA Statement:

Knowledge of FIRE PROTECTION SYSTEM design features andor interlocks which provide for the following:

Diesel enfie protection RO/SRO I

Tier:

I Group:

I KA.

I ROIR:

I SROIR.

I Cog Level Both 2

2 I 286000K4.07 I 3.3 3.3 Higher I

I 1

I

References:

LF'85286, Fire Protection and Detection CPS 9071.02, Diesel Fire Pump Capacity CheckdSequmtial Starting Datewritten:

1 02/21/05 I Author: I Ryder Commenb: None

1 Question# I 26 1 ROISRO:

1 Tier:

1 Group:

I KA:

I ROIR:

I SROIR I

Cog Level Both I

2 I

2 I 288000Ki.05 I 3.3 I

3.6 I

Lower L.

SyatemlEvoluiioo Name:

1 catexory:

Plant Ventilation Systems 1 Plant systems KA Statement:

Knowledge of the physical connections and/or cause-effect relationships between PLANT VENTILATION SYSTEMS and the following: Pmcess radiation monitoring system 1

(1) can cause an isolation of the normal Continuous Containment Purge (CCP) lineup, (2) are overridden (considering ALL plant ventilation systems) if both Containment HVAC and Isolation Valve Radiation Interlock Bypass Switches are placed in TOTAL BYPASS?

A.

(1)Two (2) Three B.

(1)Two (2) Four C.

(1)Three (2) Four D.

(1)Three (5) Five Answer: c C is correct - Per LP85455, pages 4,6,49,50,51, and 52. Considaing Part (I)... a total of THREE signals (Conditions) will isolate CCP (Group IO valves); they are: Containment Bldg Exhaust Radiation High; Containment Bldg Fuel Transfer Pool Vent Plenum Radiation High; and Containment CCP Exhaust Radiation High. Considering Part (2)...a total of FOUR signals (Conditions) are overridden by placing these switches in TOTAL BYPASS; they n~ the Same THREE that isolate CCP, &the Fuel Building Exhaust Radiation High signal (which does NOT close the CCP valves).

A, B, and D are incorrect - For the reasons dwcribed above, hut are plausible for any candidate who cannot recall the specific radiation signals that inteaface with CCP and the Interlock Bypass Switches.

Obieaive:

I Questlon Source:

I Level of Difficulty:

LP85455.1.4.11 N W I

3.4 References provided to examinee:

References:

I None I LP8545S, Containment Ventilation and Drywell Purge

I Question# 1 26 1 ROISRO:

I Tier:

I Group:

I KA:

1 ROIR:

I SROIR:

1 Cog Level Both I

2 I

2 I 288000K1.05 I 3.3 I

3.6 I

Lower SystedEvolution Name:

I Cateory:

Plant Ventilation Systems KA Statement:

Knowledge of the physical connections a n d / ~

cau%xfect relationships between PLANT VENTILATION SYSTEMS and the following: Process radiation monitoring system I Plant systems Datewritten: I 04/29/05 I Author: I Ryda Commeuts:

This question is categorized as Lower Cognitive (LCL) because, although a two-part question, there is NO cause-effect relationship between the first and second part; there is no required association, one with the other. Each part demands only one mental process from the Candidate: Part (I) - from memory, recall how many different radiation signals will isolate CCP; Part (2) -from memory, recall how many diferent radiation signals are overridden by the Total Bypass switch.

I Question# I27 1 RO/SRO: I Tier:

I Group:

I KA:

I RO1R:

I SROIR I

Cog Level Both I

3 1

Generics I

2.1.3 I

3.0 I

3.4 I

Highher SysfendEvolutiou Name:

I Catqory:

1 Conduct of opaati ons i-.

KA Statement:

Knowledge of shift turnover pradics Consider the following:

You are the on-coming RO-A for dayshift, July 30 You last stood an entire (8-hour) RO watch on dayshift, July 24 Per OP-AA-112-101, Shift Turnover and Relief, which ONE of the following identifies how far back, in time, you are required to review the Narrative Log before relieving the watch?

Back though, at leas t,....

A.

B.

C.

D.

the beginning of SWING shift on July 24.

0000 hours0 days 0 hours 0 weeks 0 months on July 25.

the beginning of DAY shift on July 25.

0000 hours0 days 0 hours 0 weeks 0 months on July 26.

1 Answer: D Explanation:

D is correct - Per Section 4.8.3 of OP-AA-I 12-101, The on-wming RO is required to review the logs through the last previous date on shift, or the preceding four days logs,...whichever is less. The preceding four days limit applies in this case. The four-day period that precedes July 30 begins at OOOO hours, July 26.

A is inwrrect - For the reascms described above. ?his choice would be correcl if the procedure read... whichever is

m.

B and C are inwrrect - F a the m o n s associated with the -t answer. These choices presume the candidate inwrrectly recalls a prrcedug five days logs requirement. Additionally, theses choices cause the candidate to ponder the meaning of preceding five days.

t Objective:

I Queatlou Source:

1 Level of ~ifficu~ty:

PBADOI2, Objective5 New I

2.6 References provided to examlnee:

I None

References:

I OP-AA-112-IOI, Shift ReliefandTumover Date Writteu: I 02/22/05 I Author: 1 Ryder Comments: None

1 Question # [ 28 1 Refereoces:

L A refueling outage is in progress, when THREE of the SRM Channels fall below 3 counts per second.

CPS 3703.01, Core Alterations CPS Tech S m 3.3.1.2, SRM Instrumentation L

Per CPS 3703.01, Core Alterations, which ONE of the following work evolutions CAN be performed?

A.

Removal of an IRRADIATED control rod blade from a FULLY DE-FUELED cell in the core, and its transfer to the Spent Fuel Storage Pool.

Removal of an IRRADIATED fuel bundle from the core, and its transfer to the Spent Fuel Storage Pool.

Transfer of a NEW fuel bundle from the Upper Containment Storage Pool, and its installation into the core.

Transfer of a NEW control rod blade from the Upper Containment Storage Pool, and its installation into a core cell containing ONLY ONE fuel bundle.

B.

C.

D.

Answer: A Explanation:

A is correct - PR CPS 3703.01, Section 6. IO, the inoperable SRMs cited io the question stem (no matter which 3 SRM Channels), would require a halt to Core Alterations PR CPS 3703.01, Section 2.2.4, however, this answer choice does NOT describe an evolution that is wnsidered a Core Alteration (Le., the evolution meets the criteria of the 2" exception ('b')

in the CORE ALTERATION definition). Therefore, this work CAN be performed with 3 inoperable SRMs. NOTE: 'I% claim bas been verified against the SRM Tech Spec (3.3.1.2). as well (see that reference, attached).

B, C, and D are incorrect - Each of t h m choices is a Core Alteration, as defined in Section 2.2.4. Therefore, these describe work that CANNOT be performed with 3 inoperable SRMs.

Obldve:

I Qnestion Source:

I Level of Difiicnlty:

LP86610.1.19 New I

2.9

I Question# 128 I proeedUE.

I

2.

"he candidate must determine whether any given one of the answer choices idis not a type of Core I

Question# I29 I Rdertuca:

RO/SRO I Tier:

I Group:

I KA:

I ROIR:

I SROIR I

Cog Level Both I

3 I

Generics I

2.3.4 I

2.5 I

3.1 I

Lower SyatemlEvolntion Name:

1 Category:

I Radiologid Controls KA Statement:

Knowledge of radiation exposure limits and contamination wnhol, including permissible levels in excess of those authorid A Site Area Emergency is in progress, when it is determined that an Emergency Exposure is muired SOLELY for the umose of PROTECTING an imuortant piece of PLANT Rp-AA-203, Exposure Conhol and Authorization EP-AA-113-F-02, Authorization for Emergency Exposure EQUIPMENT.

Which ONE of the following:

(1) identifies the exposure LIMIT (TEDE) for this emergency exposure, and (2) identifies the HIGHEST level of approval needed to authorize this exposure limit?

A.

(1)7Rem (2) Radiation Protection Management B.

(1) 10Rem (2) Station Emergency Director C.

(1) 15Rem (2) Radiation Protection Management D.

(1)25Rem (2) Station Emergency Director Answer: B Explanation:

B is correct - Part (I): Per RP-AA-203, Section 4.5.3, Table 2. IO Rem TEDE is the limit for solely protecting pmlraty. Pal (2): Per EP-AA-I 13-F-02, all Emergency Exposures (no matter the specific limit) require Station Emagency Director authorization.

A, C, and D are incorrect - For the reasons described above; but all have face validity and are plausible to an uncertain candidate.

Objective:

I Questloo Source:

I Level of DIMculty:

None N W I

2.8

~atewrinen: I 02/23/05 I Antbor: I Ryder Commenw. None

I Question# I 30 I Cog, Level RO/SRO I

Tler:

I Group:

I KA:

I ROIR:

I SROIR I

Both I

3 I

Generics I

2.4.1 I

3.1 I

3.8 I

Higher SyatunflEvoluUon Name:

1 category:

I Emergency Procedures and Plan L

KA Statement:

Knowledge of event based EOP mitigation smtegies L

Using the provided references, answer the following.

An ATWS and LOCA are in progress, with the following:

The MSUOG and IA Interlocks (RPV Level 1) have been defeated per CPS 4410.00C004 THEN, main condenser vacuum is lost and CANNOT be restored Which ONE of the following mitigation strategies IS ALLOWED by the EOPs?

Using the...

A.

RFPTs to stabilize pressure B.

MSL Drains to depressurize to the Shutdown Cooling pressure interlock once Cold Shutdown Boron has been injected Condenser as an Alternate Depressurization System in EOP-3, or as a way to reduce pressure below 50 psig in EOP-2 MSL Drains to stabilize pressure C.

D.

Answer: c Explanation:

C is correct - P a CPS 441 1.W, Section 2.2 NOTE. Without a condenser vacuum, the EOPs allow the condenser to be used as a vent path (i.% an alternate depressurization path), but not as a heat sink. Additionally, the stem conditions indicate that MSIVs (Group I ) and Turbine Bypass Valves have closed on the loss of vacuum (see LP85255, page 20).

Therefore, only where an EOP step states that it is OK to defeat RPV vent interlocks can operators defeat the low vacuum closure. Both EOP-2 and EOP-3 allow this. Stem conditions indicate that operators have already defeated the RPV Level 1 isolation for Group I (MSIVs) (see attached reference).

A, B, and D are incorrect - For the reasons described above. These choica are suggest using the condenser as heat sink when thexe is no vacuum. These choice are all taken from the Pressure leg of EOP-I A.

Obleaive:

I Oueatlou Souree:

I Level of Diffculty:

None N W I

3.9

1 Question # I 30 ]

References provided to ernminee:

References:

EOP flowchart set EOP-1.4, ATWS RPV Control EOP-2, RPV Flooding EOP-3, Emergency RPV Dep&mtion (Blowdown)

CPS 4410.00C004, Defeating MSUOG Interlocks CPS 441 1.09, RPV Pressure Control Sources LP85255, Condenser Air Removal System

[ Question# I31 1 RO/SRO I

Tler:

I Group:

I KA:

I ROIR:

I SROIR:

I Cog Level Both I

3 1

Generics I

2.4.24 I

3.3 I

3.7 I

Lower SystedEvolution Name:

I CltWOIY:

I Emergency Procedures and Plan KA Statement:

Knowledge of loss of cooling water procedures The plant is operating at rated power when a COMPLETE LOSS OF ALL suction capability occurs at the Screenhouse, coincident with a Station Blackout (SBO).

Five minutes later, the following conditions exist:

The reactor is shutdown There is NO LOCA condition Reactor pressure and water level are STABLE in their normal bands Which ONE of the following identifies the system/equipment considered most critical for worst case survivability?

A.

RCIC B.

DG1C C.

HPCS D.

FC Answer: A Explanation:

A is correct - Per CPS 4303.01, Appendix A,Section I, RCIC is the system considered most critical for worst case survivability.

B and C are incorrect - But either is plausible to the Candidate who recalls that, per CPS 4200.01,Section I.4, HPCS is the preferred source of RPV makeup, given that DG IC (Div 3 power) is assumed to be available during a SBO.

D is incorrect - But pmvides sufficient plausibility for the Candidate who leans towards giving a greater priority to keeping fission prcducts in solution in the S p t Fuel Pool Storage, given that the stem conditions indicate there is no RPV inventory control pmblem (i.e., no LOCA).

CPS 4303.01, Loss of the Ultimate Heat Sink CPS 4200.01, Loss of AC Power

I Question# I32 I ROISRO: I Tier:

I Group:

I KA:

I ROIR I

SROIR 1

Cog Level Both I

2 I

I I 209002A3.01 I 3.3 I

3.3 I

Higher SystemlEvolution Name:

I category:

High Pressure Core Spray System (HPCS) 1 Plant systems KA Statemeut:

Ability to monitor automatic opemtions of the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) including: Valve The plant is operating at rated power, with the following:

CPS 9051.01, HPCS Pump Operability (QUARTERLY surveillance), is in progress AT FULL RATED FLOW THEN, HPCS automatically initiates on a High Drywell Pressure signal Which ONE of the following identifies HPCS system valves that will show an INTERMEDIATE position indication (on P601) IMMEDIATELY AFTER the initiation signal is received?

A.

HPCS to CNMT Outboard Isolation Valve, lE22-FO04, and HPCS Test Valve to Suppression Pool, lE22-FO23 HPCS to CNMT Outboard Isolation Valve, lE22-FO04, and HPCS Second Test Valve to Storage Tank, 1E22-FO11 HPCS Suction fiom RCIC Storage Tank Valve, lE22-FO01, and HPCS First Test Valve to Storage Tank, lE22-FO10 HPCS Suction fiom RCIC Storage Tank Valve, lE22-F001, and HPCS Suppression Pool Suction Valve, 1 E22-F015 B.

C.

D.

Answer: B Explanation:

B is correct - PR CPS 9051.01, Section 8.2, CPS 3309.01, saction 8.1.2, and LP85380, Figure 4. The flowpath for this surveillance is Storage Tank-to-Storage Tank auto-initiation signal, the following occur simultaneously (from this pre-initiation lineup): F004 (injection valve) begins to stroke open (shows intermediate position), and HPCS First and Second Test Valves to Storage Tank (FOIO and F01 I) stmke closed (show intermediate position).

A is i n c o d - With the surveillance running using this flowpath, F023 is fully closed and remains that way throughout the initiation.

C is incorrect - Although FOIO begins to stroke closed, FOOl is already fully open (for the surveillance) and remains that way throughcut the initiation (until an automatic suction swap, if any, occurs later on).

D is incorrect - As described for choice C, FOOl is already open and remains that way. Interlocks prevent FOOl and F015 from WR strolring at the same time.

Suppression Pool-to-Suppression Pool). Upon receipt of the HPCS Obldve:

I Question Source:

I Level of Difficulty:

LP85380.1.8 New I

2.6

\\

I Question# I 32 I References prnvlded to euminee:

References:

None LP85380, High Prssure Core Spray CPS 3309.01, High Pmsure Core S p y CPS9051.01,HPCSPumpOpera bility

I Question # 1 33 J RO/SRO: I Tler:

I Group:

I KA:

I ROIR:

I SROIR I Cog Level Both I

2 I

I I 218000A3.08 I 4.2 I

4.3 I

Higher SystemlEvolution Name:

I CateQory:

I Plant systems

x.

Automatic Depressurization System KA Statement:

Ability to monitor automatic operations of the AUTOMAlTC DEPRESSUIUZATTON SYSTEM including: Reactor pressure Which ONE of the following results in the FASTEST INITIAL RATE of reactor pressure reduction?

A.

With reactor pressure initially at 900 psig, operators are unable to Inhibit ADS, and ALL of the ADS Valves automatically open.

With reactor pressure initially at 900 psig, operators open ALL of the Turbine Bypass Valves using the Bypass Jack.

Immediately after a turbine trip fiom 50% power, NO Turbine Bypass Valves open (cause unknown), and reactor pressure PEAKS at 1090 psig.

With the Pressure Regulator at its NORMAL setpoint, ALL of the LLS-SRVs have automatically reclosed, THEN reactor pressure rises and PEAKS at 970 psig.

B.

C.

D.

Answer: A i

Explanation:

A is correct - Per Lp85239, Attachments F and G, and page 10. Each of the 16 SRVs are capable of relieving reactor pressure at the same ratc (about 6.5% of total rated steam flow). There are 7 SRVs that open when openltors initiate ADS (also found in LP85218, page 4).

B is incorrect-PerCPS 3105.04, Section 2.1.1. &&lreliefcapacityis ZS.S%, forthe6TBVs, orah!! 5% for each TBV (as compared to 6.5% for each SRV). Additionally, only 6 TBVs op~1, as compared to 7 SRVs (when ADS is initiated).

C is incorrect - P a LP85239, Attachment G. SRV F051D has a pressursreliefsetpoint low enough (I 103 psig +/- I5 psig, per Tech Spec 3.3.6.5) to open the SRV if pressure peaks at only 1089 psig. As swn as it does, the Low-Low Set (LLS) circuits as activated for all 5 LLS-SRVs. However, only F051C has a LLS opening setpoint (1073 +/- I5 psig) low enough to open, shortly after F05 ID opened. This choice, therefore, proposes an event where no more than a total of 2 SRVs (and 9 p e ~

of the 6 TBVs) open to effect a pressun reduction.

D is incorrect - P a LP85241, page 6, the Pressure Regulator is normally set at 930 psig &em pressure, EX reactor pressure). All 5 LLS-SRVs automatically reclosing suggests reactor pressure has fallen below 957 psig. A rise, again, and a peaking at 970 psig r a q ~ n s NONE of these LLS-SRVs, and NONE of the 6 TBVs (Le., Sem pressure is still well below its 930 psig Rcssure Regulator setpoint).

Objeaive:

Question Source:

I Level of Dimculty:

I None New I

3.0

L Question # I 33 I Reference provided to euminee:

References:

None LP85218, Automatic Depressurization System LP85239, Main Steam System LP85241, Steam Bypass and Pressure Conml System CPS Tech Spec 3.3.6.5, Reliefand LLS Inshumentation CPS 3 105.04, Steam Bypass and Pmsure Regulator Date Written: 1 05/16/05 I Author: I Ryder Comments: None

I Question # [ 34 I References provided to uamlnee:

Refertncu:

1 L

Assuming the alarming condition is VALID, which ONE of the following annunciators, BY ITSELF, reminds the operators that an EOP entry condition ALREADY exists?

A.

5065-6F, SECONDARY CNMT AREA HIGH TEMP B.

C.

D.

Answer: B 5066-5A, ADS LOGIC B 105 SEC TIMER INITIATED 5064-7C, ECCS FLOOR DRAIN SUMP HIGH LEAK RATE 5064-5C, SUPPR POOL DIVISION 1 HIGH TEMPERATURE None LP85218, Automatic Depressurization System CPS 5064-5C response procedure CPS 5064-7C lrsponse p d

m CPS 5065-6F rtsponse procedure CPS 5066-5A response procedure i

Explanation:

B is correct - PK CPS 5066-5k This annunciator a l m s only when a high drywell pressure (1.68 psig) d.

confirmed low-low-low water level (-145.5 inches) condition (or confirmed low-low-low level for >6 minutes) exists. These parameters repnsen( EOP-I and EOP-6 entry conditions. The procedures Operator Actions remind the operators of this.

A is incorrect - P a CPS 5065-6F. This 8nIIUnCiatOI npresmts area high temperature alarm values, not the m e normal d U e 3 d a t e d With EOP-8 fflhy.

C is inconed - PR CFS 5064-7C. This annunciator monitors the status of various ECCS room (in secondary mntainmmt) floor drain sump systems (i.e., high leakage resulting in excessive pump-down time andor frequency.

Although this l&ge problem may progress to where flm water levels in those rooms require EOP-8 entry, water levels m not thae yet As such, no EOP entry yet exists.

D is kcorrect - P a CPS 5064-5C. This annunciator qre.senLs 8 90°F Suppression pool temperature. The EOP-6 mtry value is 95F. As such, no EOP entry yet exists.

Datewritten: 1 05/02/05 I Author: 1 Ryda Commenb: None

I Question# 1 35 1 L

The plant is operating at rated power, when a DIV 2 NSPS circuit malfunction causes the INPUT to the Load Driver, that services Group 8 isolation valves, to fail to ZERO.

Which ONE of the following valves CLOSES as a result of this failure?

A.

B.

C.

D.

Answer: A INBOARD Containment Equipment Drain Sump Discharge Valve, 1RE02 1 OUTBOARD Containment Equipment Drain Sump Discharge Valve, 1RE022 OUTBOARD Containment Building Supply Isolation Valve, IVROOlA INBOARD Containment Building Supply Isolation Bypass Valve, 1VR002B Explanation:

1

References:

i.

LP85407, Containment and Reactor Vessel Isolation Control System CPS 4001.02C001, Automatic Isolation Checklist L.

A is correct - Per CPS 4001.OZC001, page 7, this is a Group 8 valve. Per LP85407, page 58, Div 2 NSPS services the Load Driver for the Inboard valves. A ZERO input to the Load Driver produces an energized function to close the Inboard valves.

B and C are incorrect - 'Ihw Outboard valves would close if the malfunction were to occur in Div 1 NSPS.

D is incorrect - Although an Inboard (Div 2) valve, this valve belongs to Group 9 (see CPS 4001.02C001, page 9).

Group 9 is not among the otha Groups (IO, 12,16,19, and 20) that are seniced by the same Load Driven as Groups 8 and 15.

I Obiective:

I Question Source:

I Level of Difticulty:

LP85407.1.7 New I

3.0

[ Question# 1 35 1 ROSRO 1

Tk:

1 Group:

I KA:

I R O I k I

SROIR I

CogLevel Both 2

I I

I 223002K3.22 I 2.5 2.6 I

Higher SystemlEvdntion Name:

I CateEory:

PCIS/Nuclear Steam Sumly Shutoff System (NSSSS)

KASt.tuncot:

Knowledge of the effect that a loas or malhction of the PCISMSSSS will have on the following: Containment 1 Plant systems L

L&aiwe system I

r all of the Group 8 valve8 during that IPM pafamance.

L&aiwe system I

r all of the Group 8 valve8 during that IPM pafamance.

3.

Therefore, it is unl&ely that the following m a r i n will occur...the RO Candidate easily discounts the D

distracts becnuse hdshe ranembed (from the JPM performance) that IVROO2B is not a Group 8 valve.

there is I I ~

overlap with the Operating Exam.

I Question # 1 36 1

\\-

Using the provided references, answer the following.

Operators are implementing EOP-1, RFV Control, and EOP-6, Primary Containment Control.

Which ONE of the following conditions REQUIRES operators to START the Hydrogen Mixers, or PERMITS the operators to keep the Hydrogen Mixers running if already started?

A.

Igniters are still OFF, THEN hydrogen monitors come on-line d e r warm-up, and Containment hydrogen reads 9%.

Hydrogen Mixers are still OFF, THEN hydrogen monitors come on-line after warm-up, and both Drywell and Containment hydrogen read 2%.

Igniters are still OFF, THEN hydrogen monitors come on-line after warm-up, and Containment hydrogen reads 8% with Containment Pressure at 10 psig.

Hydrogen Mixers are still OFF, THEN hydrogen monitors come on-line after warm-UD, with Drywell hydrogen reading <0.5% and Containment hydrogen declared to be B.

C.

D.

Answer: B Explanation:

B is wnu3 - Once EOP-7 is ataed (in this case, on detectable hydrogen), only EOP-7 can cause them to be started.

With both drywell and contaimnent hydrogen reading 2%. operators proceed straight down the I&-leg of EOP-7, proceed througs the lef-most WAIT step and start the Mixers.

A is incorrect - This choice puts the Candidate solidly into EOP-7, specifically at the right-most leg of EOP-7, where with the Igniters still OFF, opemom are directed to stop the Mixers and prevent igniter restart.

C is incoma - P a EOP-7, Figure R, the plant is above the Detlaption Limit, requiring the implemartation of the top-most ovaride step and the execution of the right-most leg, where owratom are directed to prevent igniter restart and mop the Mixm.

D is inwnu3-Once EOP-7 is entered (in this case, because of UNKNOWN Containment hydrogen), only EOP-7 can cnwc them to be staacd This choicc is similar to the correct answer choice, B, except that the Candidate is still WAITING fa detcdable Ihywell hydqen ($?h or higher) before the Mixers can he started.

I Question # 1 36 1

References:

I Ouestlon Source:

I Level of Difiiculty:

I 9 L Obleetive:

CPS EOP-6, Primary Containment Control CPS EOP-7, Hydrogen Control DateWrftten: 1 05/W05 I Author: I Ryder Comment% None

I Qaestion # I 37 ]

Reference:

RO/SRO 1

Tler:

1 Group:

1 KA:

1 ROIR

[ ~ ~ 0 1 %

I

~ o g ~ e v e l

~~ Both I

1 I

1 1

295001 AKI.01 1

3.5 I

3.6 I

Hinher LPS5422, Reactor Vessel and Intemals CPS 4006.01, Loss of Shutdown Cooling

~~

~~~~

~~

SyrtemlE volutlon Name:

I category:

Partial or Complete Loss of Faced Core Flow Circulation I Emergency and Abnormal Plant Evolutions i

KA Statemeat:

Knowledge of the opesational implications of the following concepts as they apply to PARTIAL OR COMPLETE I LOSS OFFORCED CORE FL43W CIRCULATION Natural circulation I

With the plant in MODE 4, which ONE of the following describes an operational implication of a COMPLETE loss of Shutdown Cooling, COINCIDENT WITH having NO Reactor Recirc Pumps available?

A.

B.

C.

D.

FC will have to be lined up for Altemate Shutdown Cooling.

RCIC will have to be lined up for Altemate Shutdown Cooling.

RPV water level will have to be maintained ABOVE the steam separators.

RPV water level will have to be maintained BELOW the steam separators.

Answer: c E ~ p l ~ a t l ~ n :

C is c o r n - Per CPS 4006.01, Section 4.6. Maintaining level above 44 Shutdown Range will provide for some core amling via natural circulation. Pcr LP 85422, page 7, this level corresponds to a water level above the steam separators.

A is ineOrrea - Pes CPS 4006.01, Table 2. This lineup zequim the RPV head be off (plant in Mode 5).

Bisineorrst-PerCPS4006.01,TabIe2. InM~de4,tanpesa~1eis200Forless,wellbelowthe60psigRC1C system isolation point.

D is inmmct - See the explanation foe the eorrst ~ ~ S W R,

2.

L Obieaive:

Qnestion Source:

I Level of Dimcuity:

I None New I

2.1 i

DateWrittcn: I 04/29/05 1 Author: I Ryda Commeutr:

This quwtion is categorized as Higba Cognitive (HCL) because:

I.

The eOrrea auswer is neither an Immediate Operator Action, nor a Precautiodimitation, where the Candidate would be Cxpeaar to d l

such imm memory The stem require4 the Candidate to adequate natursl circulation as a substitute.

The elimination of the dishactas requires the several associations described in the Explanation for each

2.

the loss of forced m e circulation with the need to ensure

3.

I Question # 1 37 I (above).

Additionally, this quesricm is on the RO Exam (and is asking the Candidate to pkpd an action found in the Subsequent Actions section of the off-normal (Loss of SDC).

Rather, it is framed in such sl) way that it simply makes use ofthat Subsequent Action (in Section 4.6) to require the Candidate to demonstrate an understanding of the operational concern that results b m having no foxed reactor coolant flow.

Rsaved BS SRO-only), because the question i s m actually

1 Question# I 38 I ROISRO I Tier:

I Group:

I KA:

I ROIR: I SROIR: I CogLevel Both I

I I

I 1

295005AAI.OI 1

3.1 I

3.3 I

Highs SystedEvdutIon N8me:

I C8ttF.OW:

I Emergency and Abnormal Plant Evolutions L

Main Turbine Gatantor Trip KA Statemed:

Ability to opaatc andor monitor the following they apply to MAIN TURBINE GENERATOR TRIP: Recirculation

~~

~

The plant is operating at rated power when the main turbine trips (cause unknown).

WITHOUT operator action, which ONE of the following describes the status of main control room indications related to Reactor Recirculation (RR), 20 SECONDS AFTER the reactor scrams?

At P680...

A.

B.

the GREEN lights are lit for RR Pump breakers CB5A and CB5B.

Total Core Flow indicates about 45 mlbm/hr.

C.

D.

both RR FCVs indicate about 10% open.

both RR FCV LIMITER ERROR meters read upscale (POSITWE).

Answer: A L

EXplul8Uon:

A is amux - P a LP85202, pages 23,28,29, and Figure 8. Wheneva the turbine trip scram is mablcd (above 33.3%

powa, nominally), the EOC-RFT trip is also enabled. The turbine trip automatically downshifts both RR Pumps to SLOW speed. opaators will see the red (CLQSE) light extinguish and the green (OPEN) light illuminate for each pumps 6.9KV (Fast sped) breaker, CBSA(B).

Bisin~~-PerCPS9041.01,Figurc~

la, Ib,and2a,thisistheexpedcdtotalconflowthatrrsults~aFlow Control Valve Runback to about 19% open indication with born RR Pumps still running in FAST speed. The multing Total Core Flow indication with both pumps running at SLOW speed (atla the EOC-RPT trip), with the FCV still nearly wide open, is only about 29 rnlbdhr (this value obtained from simulator modeling).

C is incorrxt - FCVs ranain as is during the downshift (typically, indicating about 76% open). Even if the post-scram levelilevel control transient were to result in a FCV Runback signal being produced, the FCV will indicated about 19%

open, not 11% open The 20 Smnds after' staternat in the stem allows for the following: 1) the S-sccond delay betwcm the saam signal and a geneator revem pow trip (see LP8546 I, pagc 33). and 2) enough time therafk to suggest that the FCVs need some time to stoke to a lo./. opcn position.

D is inconw - Tbe only time the Limiter Enor UUI rcad o the Positive side of zao (mid-scale), is when a position indication &lure has occumd (e.&, LVDT or RVDT fcedbsck signal is sending false position information to the controller). Given that the FCVs havent movcd in this scenario, the Ermr will still read zao (mid-scale).

1 Question# 138 Referencea provided to examinee:

References:

Tier:

I Group:

1 KA:

1 ROIR:

1 SROIR: I cog~evel RO/SRO.

I I

1 I

295005AAI.01 I

3. I 3.3 Higher Both SystcmfEvdution Name:

1 C8tePOQ':

I Emergency and Abnormal Plant Evolutions Main Turbine Generator Trip None LP85202, Reactor Recirculation System LP85461, Main Generator System CPS9041.01,letPumpOperabilityTest KA Statement:

Ability to opmte andlor monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: Recirculation system 1 Datewritten: I 04/29/05 I Author: I Ryder I Commenb: None I

I Question # 1 39 1 Erpl~o.tlw:

B is comet - PR CPS 4201.01, Section 4.2.2. DC MCC IB loss requires notification of Fire Protection Group to assess the impact on fire systems. PR CPS 4202.01C002, Load Impact List, page 5, this bus powers the fire panels.

This impact is also described in LP85286, page 95 (Dc MCC IB is Div 11 power).

A, C, and D fire incorrect - For the m n s described above. Attached, here, are the Load Impact Lists for these 3 buses, vaifying there is no direct causeeffect relationship with fire systems, when either of these 3 buses are lost.

L Per CPS 4201.01, Loss of DC Power, which ONE of the following DC MCC losses requires that the FIRE PROTECTION GROUP be contacted because of the impact on Fire ProtectiodDetection System equipment requiring that power source?

A.

1A B.

1B References provided to examinee:

References:

c.

IC None LP85286, Fire Protection and Detection CPS 4201.01C001(2,3,4), Loss of 125VDC MCC IA(B, C, D) Load Impact Lists CPS 4201.01, LOSS Of DC POWR D.

1D L

I Objective:

I Quedtion Source:

I Level of Dlffculty:

LP85286.1.10.12 NW I

3.7 DateWritten

I 03/02/05 1 Author: I Ryder Commentx Although the SRO would direct the subsequent actions of this Abnormal -tine Procedure (4201.01 ), including identifying the need to contact Fire htection for this bus loss, this question is, nonetheless, categorized as BOTH (RO and SRO) for the following reasms:

1.

ROs are expected to recognize the (Divisional, Class IE) bus losses such as the DC MCCs 1 A-ID, &bcM the need to consider what may or may not be described m a procedure.

The uniaueness of the relationship between &bus loss (Division 2) and the fire systems, among these 4 buses, is essentially andistied to aLcamingObjectiw(.1.10.12)forwhich~theROandSROareresponsible.

of all significant bus losses, especially those

2.

knowledge, and is stated in the Fire Protection lesson plan, LP85286,

1 Question# I 40 I

\\.._

Following a reactor scram h m 40% power, operators are automatically controlling reactor pressure with the turbine bypass valves.

Per CPS 4100.01, Reactor Scram, which ONE of the following identifies the VALUE to where operaton should lower reactor pressure in order to minimize the Feedwater-to-RPV Differential Temperature while cooling down?

A.

800psig B.

700psig

. C.

600psig D.

5OOpsig Answer: c 1 Explmatioa:

Cismrrect-PerCPS41M).OI, Section4.2NOTE. OperatoAshouldlowerreactorpressure to about600psig to minimize the RFV-to-FW delta-T that exists during n plant cooldown.

A, B, and D an incorrect - For the reason described above.

Objective:

I Question Source:

I Level of Dificully:

None N W I

2.1 References provided to examinee:

Referenfa:

1 CPS4100.01,ReactorScnun 1 None Datewritten: I 04/29/05 I Author: I Ryder Comment% None At first glance, this question appears to have a potential opwting Exam overlap problem. While it is hue that one or more Simulntnr Scenarios may pmgresp to where n comparable pressure conhol hand (550-650 psig) may he established for EOP mitigation strntegy purposes, this is purely coincidental relative to this written exam question.

nilhis q d m is h n e d entirely in the context of EOP-free, post-scmn pressure control strategies, directed solely fiom the SCRAM abnormal operating procedure (4100.01). This question does NOT overlap with the Operating Exam.

[ Question# 1 41 I

\\ __

Following an evacuation of the main control room, operators are establishing control of the plant per CPS 4003.01, Remote Shutdown.

Which ONE of the following describes an associated operator action, and the reason for that action?

A.

If placing RCIC in service for level control, RUN the RCIC pump on RECIRC FLOW at about 60 gpm for AT LEAST 2 MINUTES; this ensures the pump is properly warmed up, without damaging pump intemals.

BEFORE starting SX Pump 1 A, CLOSE the PSW To SSW 1A Header Isolation Valve, lSX014A; this preempts the consequence of a potential hot short condition that might prevent the valve from automatically closing.

If using LPCI B for injection, dispatch an operator to DETERMINE PUMP DP FROM LOCAL INDICATIONS; this is the ONLY way to determine pump flow rate, if degraded flow conditions are suspected.

B.

C.

L D.

BEFORE attempting to place RHR B in Shutdown Cooling, VERIFY that Shutdown Cooling Outboard Suction Isolation, IE12-FO08, CAN BE OPENED; the valve may be disabled due to a hot short condition.

Answer: c Erplmitlon:

C is umect - P s CPS 4003.01COI 1, Section 3.0. Thae is no RHR Pump 8 flow indication or motor amps indication at the Remote Shutdown Panel (RSP). This is the only way to determine pump flow rate, should such infonnaton be needed. Suspicion of degraded flow conditions would be a need for such information; hence, the rcquired action.

A is ineared - P a CPS 4003.01C002, Section 3.5, running the pump on R a i n (min flow) should be limited to <20 ssonds.

B is ineared - Ref= to CPS 4003.01C005, Section 3.0. There is no such requirement for manually shutting this valve before the pump starl; ratha, the operator is directed to verify it auto-closes when the pump starts.

D is in-

- Refa to CPS 4003.01CO13, Section 1.4. Operators are cautioned to beware that a hot short could have disabled 1E12-F009 (the Inboard SDC isolation). Although a valves, IE12-F009 vulnauble to hot short problems, notice Section 4.8 of this procature. Operators must unlock and close (place to ON) the motor brcaka for F008. The way that CPS has addressed the hot short problem common to both valves (F008 and FOOS) is by keeping the F008 breaker locked open dwhg normal plant operating conditions (see CPS 33 12.03, Section 6.3). Hence, the Remote Shutdown Div 2 SDC procedure (4301.01CO13) is silent on the need for operators to be mncaned about the stahts of F008. The valve is administratively protected from hot short vulnerability.

F008 are potentially

I Question # I 41 I References provided to examinee:

References:

Tier:

1 Group:

I KA:

1 ROIR:

1 SROIR 1 CogLevei ROISRO: I 1

I I

295016 2.1.32 I

3.4 3.8 Lower I

I I

Both SyctemlEvolution Name:

I category:

Conhol Room AbandonmRlt KA Sutement:

Ability to explain and apply system limits and precaUhonS 1 Emergency and Abnormal Plant Evolutions None CPS 4003.01, Remote Shutdown CPS 4003.01C002, RSP - RCIC Operation CPS 4003.01C005, RSP - Div 1 SX Operation CPS 4003.01COl I, RSP - Div 2 LPCI Operation CPS 4003.01CO13, RSP - Div 2 Shutdown Cooling Operation I

I Level of DiCficulIy:

Question Source:

Obidvc:

~~

1 Date Wrltten: I 05/02/05 I Author: I Ryda I CommentsNonc I

I Question# 1 42 I RO/SRO I

Tler:

I croup:

I KA:

1 ROR I SROR I CogLevel Both I

1 I

1 I 295019AKz.ll I

2.5 I

2.6 I

Higher SystemIEvolutlon Name:

I category:

Partial 01 Complete Loss of Instrument Air KA Statement:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Radwaste I Emergency and Abnormal Plant Evolutions L

The plant is operating at RATED POWER when the following occurs:

An air line rupture occurs in the RADWASTE BUILDING Instrument Air Ring Header BOTH of the Radwaste Bldg IA Header Isolation Valves, AND BOTH of the Radwaste Building SA Header Isolation Valves SHUT NO other building Service Air or Instrument Air supplies are affected ALL other air header isolation valves are STILL OPEN Which ONE of the following describes the plantkystem impact of this loss of air?

A.

FC Demineralizers isolate.

B.

C.

D.

Main condenser vacuum slowly degrades.

WO Chillers automatically shut down.

ACTUAL D/Ps on the CW Traveling Screens RISE.

Answer: A Explanation:

A is comct - PR CPS 5041-4B, the Fuel Pool Cleanup (FC) demins isolate.

B is incomct-Per CPS 50414B. Because of the failure mode of the IN66-FO60 valve (fails as is) Off-gas is unaffected, and t h d o r e condenser vacuum is unaffected. Tlis choice is attractive to the Candidate who does recall that the Mechanical Vacuum Pump (MVP) Separate Tank Vent Valve will SHUT on this loss of air, causing a blow out of loop seal, and a loss of vacuum, IF the MVP were in service. With the plant at rated power, the MVPs are not in semce, and they are isolated.

C is inwmcl -Pa CPS 5041-4C, these chillers are loads on the Control Building IA Ring Header. With stem conditions indicating there was a successful isolation of the RW Bldg air headers, leaving all others unaffected, these chillers should be unaffected.

D is incomct-Per CPS 5041-SE, the traveling screen air bubblers are loads on the Turbine Building SA Ring Header.

For the same reason as discussed in choice C, these bubblers should continue to function as designed. The choice is dishacting in that is describes how ACTUAL screen d p would m d if air were lost to the bubblers...i.e., screens would & auto-start (or auto-shift to Fast speed) on high Up; consequently, ACTUAL screen d p will rise.

See Figure 1 of LP85301 for a simplified view of how these air headers ax arranged.

Objective:

I Queation Source:

I Level of Difficulty:

LP85301.1.11.3 New I

2.6

L

References:

LP85301, Service and Inshument Air Datewritten: I 04/29/05 I Anthor: 1 Ryda Commenb: None

1 Question # 1 43 ]

RO/SRO I

Tier:

I Group:

I KA:

I ROIR:

I SROIR: I

~ o g ~ e v e l Both I

1 I

295023AK2.02 I 2.9 I

3.2 I

Higher SystemlEvolntion Name:

I category:

u Refueling Accidents Emergency and Abnormal Plant Evolutions I

KA Statement:

Knowledge of the interrelations between REFUELING ACCIDENTS and the following: Fuel pool cooling and cleanup L

Operators have JUST BEGUN transferring 20 spent fuel bundles from the Containment Transfer Pool to the Spent Fuel Storage Pool, when the only available FC Pump trips and CANNOT be restarted.

Which ONE of the following describes a reason why these spent fuel transfers must be STOPPED until an FC Pump can be placed back in service?

A.

B.

C.

D.

Area radiation levels on CNMT el. 828 will rise with EACH fuel bundle transferred.

A FULL carriage in the IFTS transfer tube will raise temperature RAPIDLY enough to damage the he1 before it exits.

Spent Fuel Storage Pool temperature will rise RAPIDLY with EACH fuel bundle placed in the pool.

Area radiation levels in accessible areas around the IFTS will rise to potentially LETHAL levels.

Answer: A Explanation:

A is correct - P a LP85233, pages 22, 39, and 46. Without forced FC flow, there is no way to replenish the 1,000 gallons (approximate) removed from the Containment Transfer Pool (on CNMT el. 828) each time the IFTS transfer Nhe is flooded for canying a fuel &age down to the Fuel Building Transfer Pool. As Containment Transfer Pool level lowas, with each fuel bundle(s) transfcned, area radiation levels will rise, creating a local radiological hazard for personnel on CNMT el. 828.

Bisineorrect-PaCPS3702.01,Section2.1.5,thetransfertubeisdesignedtohandlefuelforupto IOhours without additional cooling.

C isincod-PerCPS3317.01, Section4.14, evenat highheatloadsinthespentfuel storagepoo1,degradedFC conditions will r s u l t in pool temperamre rising on the order of several degrees per hour.

D is i n W d - P R CPS 3702.01, Section 2.1.7, lethal radiation levels exist in these spaces whenever fuel is transiting the wnsfer tube, regardless of the status of FC.

I Objective:

I Quation Sonree:

I Level of Difficulty:

LP85233.1.13.2 New I

3.3

1 Question# I 43 I References provided to examinee:

References:

None LF'85233, Fuel Pool Cooling and Cleanup System O S 3702.01, Inclined Fuel Transfer System (IFTS)

CPS 3317.01, Fuel Pool Cooling and Cleanup System

I Question # I 44 I ROISRO I Tier:

I Group:

I KA:

I ROIR: I SROIR: I ~ o g ~ e v e l Both I

1 I

I 1

295025EK3.09 I

3.7 I

Lower SystedEvdotlon Name:

I category:

~

Hi& Reactor Prepsure 1 Emergency and Abnormal Plant Evolutions L

L KA Statement:

Knowledge of the m o n s for the following responses 8s they relate to HIGH REACTOR PRESSURE Low-low set Which ONE of the following identifies a design feature that acts to prolong the life of MOST of the 16 Safety Relief Valves?

A.

B.

Low-Low Set initiation Overpressure Relief mode of SRV operation C.

D.

ADS initiation Overpressure Safety mode of SRV operation Answer: B Explanatloo:

B is u)ITect - P a LP85239, page 16. The 5 LLS SRVs act to reduce the number of SRVs that cycle for given plant conditions, prolonging SRV life.

A, C, and D arc incorrect - See LP85239, pages 11-13. Neither ofthese functions are directly related to prolonging SRV life.

I I

Obleaive:

I Question Source:

I Level of Dimculty:

None New I

2.7 Referenfa provided to examlnee:

Referenfa:

Datewritten: I 03/04/05 I Author: I Ryder Comment% None I None I LP85239, Main Stcam System

L 1 Question # 1 45 I Which ONE of the following identifies an ADVANTAGE of taking the EOP-6 action to Start all available pool cooling when Suppression Pool Temperature is 97F and slowly rising?

A.

B.

Extends the time that it remains acceptable to INITIATE Containment Sprays.

Extends the time before HAVING to inject boron, if shutdown criteria is NOT met, but reactor power is BELOW 5%.

ENSURES that RCIC Pump damage due to inadequate NPSH will NOT occur if the pump is taking a suction from the suppression pool.

ENSURES that the Containment design temperature limit will NOT be exceeded while the rate of blowdown energy transfer is greater than the containment venting capacity.

C.

D.

Answer: B Explanation:

B is cmmt - Refu to EOP-I A, Power leg. Any attempt to slow down the rate of suppression pool temperature rise will delay the ~ecruirement to start SLC b r e a c h i n g the Boron Injection Temperamre (BIT) of Figure G.

A is incorrect - Refer to Figure 0 of EOP-6. This choice provides face validity in light of the unavailability of a 3d distraaa that points directly at suppression pool temperahue. It provides sufficient distraction, in that it readily a

m one to a very familiar EOP Figure, familiar even to the weakest of Candidates; where the other 3 choices demand a greatex investment of time and analysis to determine their specific relationship with the given pool tanpadtun.

C is incomct-Refer to EOP-I, Figure Z. Even if opcratorJ are able to STOP the rise in pool temperature (by starting ail available pool cooling, alone), and therefore stay far below 197F, it does SQt ENSURE that RCIC pump cavitation is avoidable Figure Z clearly shows that either too low a pool level, or too high a pump flow, can SLl! lead to cavitation and pump damage.

D is incorrect - Refer to EOP-6, Figure P, and to EOP Technical Bases, Section 12-H.

This choice suggests that the Heat Capacity Limit is dependent solely on suppression pool tcmperatu~. It is not. Even if starting all available pool cooling were able to allow pool temperatux to rise no higher than, for example, 140°F. a low enough pool level (15.1 fes, in this casc) would still exceed the HCL, which is defined by this choice (a paraphrase of Section 12-H of the BW).

ObJdve:

I Ouestion Source:

1 Level of Difficulty:

LF87553.1.6.8 N W I

3.7

1 Question# 1 45 ]

CPS EOP-ii, ATWS RFV Control CF'S EOP-6, Primary Containment Control CPS EOP Technical B a m

. Date Written:

1 03/04/05 I Author: I Ryder Comment%

This question is categorized as Higher Cognitive (HCL) for the following reasons:

1.

The correct answer demands that the Candidate &

the given Suppression Pool Temperature (97°F and rising) with the Boron Injection Initiation Tempture (BIT) asswintion must recognize that with only 97OF pool tempemme, no matter what the power level, there is still time to potentially delay boron injection by slowing down the rate ofpool temperature rise.

The elimination of, at least, the choice 'D' distracter demands that the Candidate recognize this choice is describing the operational meaning of the Heat Capacity Limit, and then recognize that a 97°F pool t m p t u r e is well below even the most limiting portions of Figure P.

from Figure G of EOP-I A. The

2.

References provlded to examinee:

Rderemca:

1 EOP flowcharts I CPS EOP-I. RF'V Control

I Question# I 46 I

i.

Using the provided references, answer the following.

A Station Blackout (SBO) is in progress, with the following:

Suppression Pool Level reads 18 feet on its ATM Containment Pressure reads 2.5 psig and slowly rising on its ATM Per CPS 4200.01, Loss of AC Power, ONLY ONE of the following identifies information, about the CURRENT Containment Temperature, that operators should expect to have AVAILABLE and be ABLE to use:

1. Containment Temperature reads 165OF and slowly rising on its ATM.

2. Inside Containment, a hand-held barometer reads I.5 psia and slowly rising.

L

3. Inside Containment, a hand-held i n h e d thermometer reads 122'F and slowly rising.

4. Containment Temperature reads 190°F and slowly rising on a portable resistance-temperature bridge.

Which ONE of the following describes where operators CURRENTLY are in EOP-6, with respect to the CONTAINMENT TEMPERATURE leg?

A.

B.

C.

D.

Monitoring for a possible EOP-6 entry on Containment Temperature At the BOTTOM-most IF-THEN step, about to proceed to EOP-3 At the TOP-most IF-THEN step, waiting for some AC power restoration Have just determined it would be OK To Spray, if an RHR Pump were available Answer: B Explanatim:

B is~-PscPS4200.01C003,page9. I&Cpasonneluscaportablebridgeconnected tothepermanent RTDsat designated cabinet taminal points, extrapolate the corresponding containment temperature and report it to control mom opaators. With 190°F and slowly rising, opemtors are at the bottommost IF-THEN step of the Containment Tempemhm leg.

I Question# I 46 I L

A is inwrrect - This would be the choice for a Candidate who believes: 1) the prescribed method employs sending personnel into a hot containment with a barometer, and 2) the atmosphere is a saturated one, allowing operators to exhapolate containment temperature using the Steam Tables. The alleged corresponding containment temperature would be I16OF. If this were true, operators would umtimie to monitor wntainment temperature for a possible EOP-6 entry at 122'F.

C is incorrect - This would be the choice for a Candidate who believes the prescribed method employs sending personnel into a hot containment with an infrared themometa. It i s m the method prescribed by the procedure. If it were, with 122°F and slowly rising, operators would be at the topmost IF-THEN step of the Containment Temperature leg.

D is incorrect - This would be the choice for a Candidate who believes there is an ATM instrument for Containment Temperature; per CPS 4200.01, Appendix B, there is &. If there were, at 1659 and slowly rising, with Containment Pressure at 2.5 psig, Figure 0 would indicate that it is OK to Spray; operators would be waiting for an available RHR Pump.

References provided to examinee:

References:

Steam Tables EOP flowcharts Steam Tablcq pag es 5-6. photo-wpy, enlarged CPS 4200.01C003, Monitoring CNMT Temperat~res During a SBO EOP-6, Primary Containment Control Oblective:

I Question Source:

I Level of Dimcuity:

None New I

3.3

Which ONE of the following events requires an entry into one or more of the EOP FLOWCHARTS?

A.

After a scram, the Shutdown Criteria are NOT met.

Rcrcrcllur provided to crunb~a?:

References:

B.

C.

D.

Answer: D After a scram, reactor water level shrinks to +15 inches before recovering.

The VF Exhaust CAM, 1RIX-PROl9, reaches its ALERT alarm point.

A break in an SRV discharge connection raises Drywell temperature to 155'F.

E~pLnatlon:

D is --

Per EOP-6. Drywell tanpaature enhy is q i r c d at 150OF.

A is inunrrct -Pa EOP-1. Unlcps there is first an EOP-1 entry (there may not be on a low-powet scram), there is no way mto EOP-IA for failum to m a t shutdown criteria Ratha, the REACTOR SCRAM abnormal, CPS 4100.01, seetion 4.3.1, directs operators to use one of the EOP support procedures, CPS 441 1.08, to alternatively insat control rods.

B is ineOrna-Pes EOP-1. Unless level dmps to b e l 3 (+8.9 inches), no EOP entry is required. A low-power scram may not shrink l e d that low.

C is in-

- Per EOP-8. This CAM must reach its HIGH alarm point before an EOP enhy is required.

EOP flowcharts EOP-I, RPV Control EOP-IA, ATWS RPV Control EOP-6, Rimary Containment Control EOP-8, Skondary Containment Control cPs4100.01,Rcadorsaam I

Obleaive:

I

don Sonm:

I L m l of Dlfflcnlly:

LP87558.1. I New I

2.6

1 Question# I 48 I

+

RO/SRO. I Tier:

I Group:

I KA:

I RON I SROIR: I c o g ~ e v e l Both I

1 I

1 I 2950131 EK1.03 I 3.7 I

4. I I

Higher SystedEvolutiou Name:

I category:

Reactor Low Water Level I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge ofthe operational implications of the following concepts as they apply to REACTOR LOW WATER Consider EOP-1 A, ATWS FWV Control, when answering the following.

WITH reactor power still ABOVE 5%, which ONE of the following identifies an INDICATED reactor water level that can be characterized by the following statement?

Core Void Fraction is relatively LOW; nonetheless, FUEL DAMAGING powa oscillations are NOT EXPECTED.

A.

-40inches B.

-65 inches C.

- 140 inches D.

- 162 inches Answer: B Explanatlon:

B is c o r n - Per EOP Technical Bases, EOP-IA, pages 5-14,5-15,5-17, and 5-18. This level (-653 is just helow the level at which the feedwater sparger is fully uncovered, thus reducing the wre inlet subcooling by 65-75%, a point where large-scale core instabilities are a expected to occur. However at this same level (-65), Void Fraction inside the core shroud is still relatively low. This is due to there still being a sufficient head of water that sustains a good amount of natural circulation (note: RR Pumps tripped at 45). At -65. this natural circulation continues to sweep voids up and out, resulting in a still relatively low Void Fraction. Per pages 5-17 and 5-18. only when operators continue to IOWR level, will the natural circulation contribution be reduced to a point where the core voids outs (i.e., a HIGH Void Fraction will exist inside the core shroud).

A is incorrect - For the reasons described above. At this level (-40)

RR Pumps are still lunning in SLOW speed (see EOP Technical Bases, page 5-16). Regardless of comparing void contents, fuel damaging core instabilities bower oscillations) at high power and low flow conditions is core inlet s u b l i n g with level this high above the feedwater ~ p a r g ~.

C and D are incorrect - For the reasons associated with the COlTect answer. At these low levels (-140 and -1 62), fuel damaging power oscillations are NOT expected. HOWWR, the whole purpose of intentionally conhulling level this low (in either Band B or Band C) is to ~IC%% the void fraction in$.&

the shroud, thus keeping power down. Again, refer to pages 5-17 and 5-18 a major w n m because there remains a good amount of Obleetive:

I Quertiou Source:

I Level of DIMculty:

LF87553.1.3 New I

3.5 References provided to eismlnee:

I None required, but m e s s to EOP flowcharts is OK I CPS EOP-IA. ATWS RFV Conml

I Question# I 48 I E

SRO I R C w Level

4. I Higher SystemlEdntion Name:

Reactor Low Water Level KA Statement:

Knowledee of the d o n a l implications of the following concepts as they apply to REACTOR LOW WATER and Abnormal Plant Evolutions I

I LEVEL water i&i effects on reactor W W

~

I Date Written: I 05/02/05 I Author: I Ryda Commenb: None

The plant is in MODE 4, with the following:

a Operators are maintaining a STEADY reactor water level with the following:

o CRD Pump A is feeding the RPV o RWCU is rejecting to the main condenser at 45 gpm, using ONLY the low flow valve a CRD Pump B is UNAVAILABLE THEN, CRD Pump A trips and CANNOT be restarted Per CPS 3303.01, Reactor Water Cleanup, which ONE of the following describes the operator action with respect to RWCU?

L A.

B.

C.

D.

Answer: A CLOSE 1G33-FO41, Drain Flow to Condenser Bypass; THEN, CLOSE lG33-FO31, Drain Flow Orifice Bypass, and FULLY CLOSE lG33-FO33, Drain Flow Regulator.

CLOSE lG33-FO46, Drain Flow to Condenser; THEN, VERIFY CLOSED 1G33-FO3 1, Drain Flow Orifice Bypass, and FULLY CLOSE lG33-F033, Drain Flow Regulator.

AS NECESSARY to maintain the desired reactor water level, THROTTLE closed 1G33-F033, Drain Flow Regulator.

AS NECESSARY to maintain the desired reactor water level, THROTTLE closed 1G33-F031, Drain Flow Orifice Bypass.

Explanation:

Aiscorren-Stanconditionsindicatcthatreject islineduppesCPS3303.01, Section8.1.6.3.1, forreactorpressure 4 0 psig (Mode 4), with the low drain flow valve, F041, open, the high drain flow valve, F046, still shut, and the Drain Flow Orifice Bypas, F03 I, fully open. With a steady reactor water level being maintained, a trip of the CRD pump would require that the reject flow path be fully s d

to maintain level. Therefore, operators implement Section 8.1.6.9. In this case, they will close F041, then closc F031, and fully close F033.

B isincomct-Forthereasonsdescribedabove. 1 G 3 3 - F 0 4 6 w a s ~ o o p e n f o r t h i s ~ ~ j ~ t flowlineup C is incomct - Candidate is e x w e d to recognize the pre-trip flow balance that existed (45 gpm in, 45 gpm out). As such, any attempt to throttle down on lG33-FO33 in an effort to maintain water level would take reject flow below 13 gpm. P a CPS 3303.01, Section 8.1.6.3 CAUTION (page 33, this may result in a system isolation on delta-flow.

1 Question # I 49 I

References:

RO/SRO.

1 Tier:

I Group:

I KA:

I ROIR 1 SROIR I c o g ~ e v e l Both I

1 I

2 I

295022AA1.04 I 2.5 I

2.6 I

Higher CPS 3303.01, Reactor Water Cleanup System CPS 3304.01, ConIml Rod Hydraulic & Control (RD)

~

SystedEvolution Name:

I Category:

Loss of CRD Pumps KA Statement:

Ability to operate and/or monitor the following as they apply to LOSS OF CRD PUMPS: Reactor water cleanup system D is in-

- Only if the plant were position, before the pump trip. Even if it were to be used, attempting to maintain level with this valve would he incomct for the same reason as described for choice C.

I Emergency and Abnormal Plant Evolutions 50 psig, would operatm have the F03 1 valve in a throttled open

1 Question # 1 50 1 Refemnces:

Tier:

I Group:

I KA:

1 ROIR:

I SROIR: I ~ w ~ e v e l RO/SRO: I Both I

1 I

2 1

295029 2.1.33 I

3.4 I

4.0 I

Lower LP85205, Residual Heat Removal System

~~~

~~

SystemEvdndon Name:

1 Category:

High Suppression Pool Water Level KA Statement:

Ability to recognize indications for system operating paramctcrs which are enny-level conditions for technical specificarions I ~ m q e n c y and Abnormal Plant Evolutions The plant is in MODE 3 with reactor pressure at 140 psig.

Which ONE of the following events requires a Technical Specification ENTRY?

A.

From a standby lineup, the RCIC Turbine Trip Throttle Valve, C002E, trips SHUT due to a BROKEN latch-trip hook assembly.

TWO of the level transmitters for the Scram Discharge Volume High Level Trip Function FAIL their Channel Calibration.

B.

C.

While terminating Suppression Pool makeup, the Supp Pool Fill Valve, 1 SM004, sticks open; pool level rises to 20 FEET, 5 INCHES, before operators can stop the rise.

An electrical short and fire DESTROYS the MOTOR-OPERATOR for RHR B Shutdown Cooling Suction Valve, lE12-FO06B; the fire is quickly extinguished.

D.

Answer: c C is correct - Per Tech Spec 3.6.2.2. Mode 3, below 235 psig, uppa level limit is 20 feet, I inch.

A is incorrect - P a Tech Spec 3.5.3. In Mode 3, RCIC operability is not required until reactor pressure is above 150 psig.

Bisin~-P~TechSpecTable3.3.1.1-1,Function#8. SDVlevel~smittertripfunctionisrequired~in Modes 1.2, and 5(a).

D is incomct - Per Tech Spec 3.4.9. In Mode 3, Shutdown Cooling operability is not required unless reactor pressure is below the RHR cut-in permissive pressure. That pressure setpoint is 104 psig (see LP85205, page 20).

ObfecUve:

I Question Source:

1 Level of Dlfticulty:

LP85223.1.16 New I

3.3 CPS Tech Spec 3.5.3, RCIC-I CPS Tech Soec 3.3.1.1. RPS Instrumentation Date Written: I 03/08/05 I Author: I Ryder Commenb: None 1

Question# 1 51 ]

L Using the provided references, answer the following.

Referenfa provided to exnmloee:

Following a scram on high drywell pressure, operators are placing the Containment H2/02 monitors in service.

CPS 441 1.I 1, Hydrogen Control System Operation, with the NOTE at the top of page 3 (Section 2.1) redacted Which ONE of the following identifies the EARLIEST time when operators are PERMITTED TO USE the monitors to determine Containment hydrogen concentration?

20 minutes after...

A.

B.

C.

D.

placing the OVEN TEMP SELECT switch in HIGH.

observing the OVEN TEMP ABNORMAL light extinguish.

the STARTUP CYCLE clock begins counting up.

RE-OPENING the associated containment isolation valves.

Answer: c Explanatloo:

C i s ~ - R d e s t o C P S 4 4 l 1. 1 1, S ~ o n 2. 1. 5.

Inordn:toutilizemonitorRadings, theremusthavebeen20-minute8 m time (ie., sampling time) with the containment isolation valves open. Step 2.1.1 has operators place the Oven Temp Select switch in HIGH (at which time, the Oven Temp Abnormal light illuminates, signifying that a w m -

up has begun). Step 2.1.2 has operators mopen the containment isolation valves (closed on the high drywell pressure scram signal), &

about 20-minutes (ie., when the Oven Temp Abnormal light extinguishes). At this point, the monitors are still spt sampling. Only when operators depress the Enter key in step 2.1.4.8 does the sampling time (m time) begin, coincident with the fact that the 270-second Startup Cycle clock begins to count up.

A, 6, and D are i n c o m t - For the reawns described above.

I I Level of DiMcully:

Objdve:

Question Source:

LP854406.1.9 New I

3.0

[ Question # I 51 I Duewritten: I 05/16/05 I Author: I Ryder Commentr:

This open-refaence question is a Direct Loohp type, for the following reasons:

I.

The Candidate must first which procedure Section to begin with. Hdshe must recognize (*om the stem) that the associated Containment Isolation Valves have gone shut on the high drywell pressure signal; therefore, seaion 2.1.2 is where placing the monitm in service begins. To accommodate this first piece, the NOTE at the top of page 3 of the procedure has been redacted.

The Candidate must recognize (Systems knowledge) the significance of step 2. I.4.8 (as already described above). The fact that this nction begins the sampling (run) time, is @ explicit in the procedure.

2.

For thesc m e reasons, this question is also categorized ns Higher Cognitive.

1 Question# 1 52 1 L

The plant is o p t i n g at rated power when a Master Level Controller failure raises reactor water level AS HIGH AS +60 INCHES, and the Level 8 hip function fails.

Which ONE of the following describes the POTENTIAL consequence that is AVOIDED by inserting a manual scram?

A.

B.

C.

Main Turbine excessive vibrations D.

Answer: c NON-conservative core thermal power calculation Failure of SRVs to FULLY RE-SEAT after opening Failure of MSIVs to FULLY CLOSE when needed L

Explmatlon:

C is comet - Rda to LP85422, Figure 21. Instrument Z m elevation is 787 feet, 6 inches. The elevation that cornsponds to +M) inches @e, +5 feet) is 792 feet, 6 inches. The main steam line nozzle elevation is 797 feet; this is 4 f q 6 inch= above where RPV level rose More operators insertced the scram. The only potential consequence of level this high is a reduced drying efficiency of the moisture separator, and therefore, an increased amount of moisture cat~yova that could m l t in high vibration of the main turbine due to water-blade impingement.

A is incorrect - The notion of there being an dfect on calculated core thermal power comes from the OPEX discussion in LP85422, page 41. However, as shown in that discussion, a non-conservative situation be associated with a amount ofmoisture carryover, llpt an increased amount.

B is inunrcct - With water level getting no where near the main steam lines, there is no chance for affecting the re.

seating ability of an SRV.

D is incorrect-With water level getting no where near the main steam lines, there is no chance for affecting the ability of the MSIVs to fully seat on their shut seat due to either water in the steam lines, or unanalyzed DPs.

Obleaive:

I Question Source:

I Level of Difficnlm:

None NOW I

3.0 Referencu provided to esamlnee:

References:

I None I LP854'22, Reactor Vessel & Intemals DateWritten: I 0311 1/05 I Author: I Ryder Commeatx None

I Question# 153 1 L

RO/SRO: I Tier:

I GrOUD:

1 KA:

1 ROIR. I SROIR I CogLevel Both I

I I

2 I

295002AK2.02 I

3. I I

3.2 I

Higher SysIcmlEvoluTion Name:

I category:

Loss of Main Condenser Vacuum I Emergency and Abnormal Plant Evolutions

~~

KA Statement:

Knowledge of the interrelations bawm LOSS OF MAIN CONDENSER VACUUM and the following: Main Nrhine The plant is operating at 45% power, on the 50% FCL, with the following:

Circulating Water (CW) Pump B is tagged out for repairs THEN, operators perform an emergency shutdown of CW Pump C due to a major oil leak CW Pump A remains running RO-A lowers reactor power until main condenser vacuum stabilizes Main condenser vacuum is now 25 Hg and STEADY Reactor power is now 25% and STEADY Which ONE of the following describes the NEW operational concern that follows this transient?

A.

B.

C.

D.

The impact of a potential loss of 6.9 KV Bus A Windage heating of the LP turbine last stage buckets Inability of CW Pump C to automatically trip on high condenser pit level The prompt exit from the PoweriFlow Map CONTROLLED ENTRY REGION Answer: B Explanation:

B is correct - Post-transient wnditioos show the plant operating at something less than 300 MWe generator load with too low a eondmsa vacuum (46 Hg). Per CPS 4004.02, Section 4. I. 1, operating with these conditions should be avoided. PR PB400402, this light load-low vacuum condition causes overheating (due to windage) and mOiSNIt erosion of the LP turbine last stage buckets. Applicable calculationfor post-transient generator load is as follows:

25% actual 192.5% max allowed = x MWe / 1052 MWe net xMWe=(.25/.925)X(1052)=284 MWe A is incotred - This is no more a concan, now, than it was before the manual hip of CW Pump C. The 6.9 KV A Bus powas both the A and C pump; a single bus loss with only two pumps running would require a scram. This choice sugg*lts thaI that vulnerability is greater now (i.e., an IMMEDIATE cpational concern) since the C Pump is the only one remaining. This is C is incorrect-Per LP85275, page 28, the high pit level hip circuitry for all 3 CW Pumps is electrically powered by the CW Pump A DC Control power. Because CW Pump is still running, this high pit level capability still exists.

D is incorrca - btransient stem wnditions (45% power, 50% rod line) are meant cause the Candidate to ponder the effect ofhaving to reduce mirc flow (andor rod insertions) in order to bring power down to the post-transient 25%.

Even Without the P/F Map as open-refaence, the Candidate is expected to realize that the plant operating well clear of the Controlled Entry Region (the prompt exiting of which would be the IMMEDIATE operational concern).

true.

L References provided to examinee:

References:

1 Question# 153 ]

None PB4OO402, Loss of Vacuum LF'85275, Circulating Water System CPS 4004.02, Loss of Vacuum CPS 3005.01, Unit Pow% Changes (for the P/F Operating Map)

ROISRO I Tier:

1 Group:

I KA.

I ROIR: I SROIR: I c o g ~ e v e l Both I

1 I

2 1

295oMAKZ.02 I 3.1 I

3.2 I

Higher S y r t ~ v o l u l i o n Namt:

I Category:

Loss of Main Condmser Vacuum KA Statement:

Knowledge of the intarelations betwem LOSS OF MAIN CONDENSER VACUUM and the following: Main turbine 1 Emergency and Abnormal Plant Evolutions Obidve:

I Ounnon source:

1 Level of Dimcu)ty:

None New I

3.0 Datewritten

I 05/16/05 I Author: I Ryde Commenta: None

1 Question# 154 ]

~

L

- ROBRO 1

Tier:

I Group:

1 KA:

I ROIR:

I SROIR: I c o g ~ e v e l Both 3

I eenaics I

2.2.2 I

4.0 1

3.5 J

Lower SyitemlEvdution Nnme:

I catmoly:

1 Equipment Control KA Statement:

Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels L

A.

Holds the control switch in the CLOSED position for 1 to 2 seconds after seeing the GREEN light ON with the RED light OFF.

After seeing a closed indication, holds the control switch in the OPEN position until an INTERMEDIATE indication is observed, then re-closes the valve.

Holds the control switch in the CLOSED position and RELEASES the control switch AS SOON AS the RED light extinguishes.

After seeing a closed indication, holds the control switch in the OPEN position for 2 TO 3 SECONDS, then re-closes the valve.

B.

C.

D.

Answer: A Expl.oltlOD:

Aiseorrect-PaOP-CL108-101-1001, Sedon3.5.1. Thisensuresthevalveisfullyclosed.

B, C, and Dare incorrect - Not the methods described either in t&

procedure QI in &u other operating procedure.

Objective:

I Questlon Source:

I Level of Dimculty:

None NeW 2.0 Referenca provided to uamlnee:

References:

I None I OP-CL108-101-1001, General Equipment Operating Requirements Datewritten: I 03/14/05 1 Author: I Ryder Comments: None

I Question# I 55 I L

The plant is operating at RATED POWER when some PAINTING is to be performed inside the Containment Building.

Which ONE of the following identifies the PREFERRED mode of containment purge operations to support this work?

A.

CCP FILTERED Mode B.

Containment Purge Mode C.

Containment Vent Mode D.

CCP UNFILTERED Mode Answer: D Explanation:

Diswmct-PerCPS 3408.01, Sections2.2.1 and4.l. ?%isisthenormal operating containment purgemode for Mode I, and should be used when painting is in progress inside containment.

Aisincorrect-PerCPS3408.01,Sections2.1 and2.1.1. Althoughthismodeisa~ilableinMode 1,itusesthe Drywell Purge Filter Trains (DWPFTs), and should reduced capacity of the charcoal filters due to these chanicaldpaint fumes).

B is incomt - Per CPS 3408.01, Section 2.2.2, this mode uses the Drywell Purge Filter Trains (DWPFTs), and should not be used because of the mcmn described in Section 4. I (i.e,, reduced capacity of the charcoal filters due to these chemicaldpaint fumes). Additionally, this mode is only available in Modes 4 and 5 (see Section 8. I. 1.4).

C is incorrect - Although specified by CPS 3408.01, Section 2.2. I as the preferred mode to be used when painting is being done inside Containment, this mode is only available in Modes 4 and 5 (see Section 8. I. I.3).

be used because of the concern described in Section 4.

I (i.e.,

Objective:

I Ouesiion source:

1 Level of Difficulty:

LP85455.1. I4 New I

2.5 Referearm provided to ex.mlnee:

References:

~ntewritten: I 04/29/05 I Author: I Rydez Cornmenix None 1 None I CPS 3408.01, Containment BuildinglDryw ell HVAC

I Question # I 56 1 L

Using the provided references, answer the following.

ASSUME the following when answering this question:

Main Power Transformer MVA is EQUAL TO Main Generator MVA The plant is operating at power, with degraded grid conditions, and the following:

Main Generator terminal voltage is 20,020 Volts Main Generator is operating at a 0.9 power factor with POSITIVE (+) VARs C Phase Main Power Transformer (MPT C) is operating with NO operable cooling banks Outdoor air temperature is 95°F MPT C oil and winding temperatures are slowly rising Operators have just reduced Main Generator real load ( W e ) to lower the MVA load to the MAXIMUM PERMISSIBLE MVAs for these degraded conditions Which ONE of the following identifies the MAXIMUM amount of time that the Main Generator is permitted to operate with this MVA load?

A.

30minutes B.

36minutes C.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 10 minutes D.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . 13 minutes Answer: A Explanation:

A is correct - P a CPS 3105.05, Section 8.5.2. VARs to the grid translates to a lagging power factor. At 20,020 volts, the perator is operating at 91% of rnted (.91 x 22,000 = 20,020). Section 6.1 Table shows rated voltage is 22,ooO. Opaators determine that the maximum allowable MVA load is 91% of rated MVA. Section 6.1 Table shows Rated MVA is 1179 MVA. Therefore, the maximum allowable MVAs is 1073 MVA (.91 x 11 79 = 1073). Per CPS 3504.01, Section 8.3. I, operators recognize that the degraded cooling condition on MPT C is the limiting concern.

Operators determine that Table 1 applies (given that no operable Moling banks are available). Per the notes of Section 8.3.1, at 95OF air temperature, operators default to t h e m temperature value (1049), and the higher MVA load value (I 140 MVA). The resulting time limit at this MVA load is 30 minutes.

1 Question# I56 I

References:

RO/SRO I

Tier:

I Group:

I KA:

1 ROIR 1 SROIR 1 CogLevel Both I

3 I

Generics I

2.1.25 I

2.8 I

3.1 I

Higher CPS 3105.05, Generator (TG)

CPS 3504.01, Main Power and UAT Cooling SyatemlEvolutlon Nnme:

1 c n t ~ o r y :

I Conduct of operations L

KA Statement:

Ability to obtain and in-station reiaence materials such as graphs. monographs. and tables which contain pdomance data B is in-t

- For the reasons described above. This choice is plausible if the Candidate inamro!xiately interpolates the 95F air temperalure for the comet Max MVA load (1 140 MVA). 95F is half the interval between 86°F and 1049; 36 minutes is half the interval between 30 minutes and 42 minutes.

C and D are inwrrect - For the reasons associated with the correct answer. These choices are analogous to choice A and B, respectively. ll~q are plausible to the Candidate who incomtlytmnslates the VARs to thegrid stem condition to a-pow factor. Per CPS 3105.05, Section 8.5.2, at 91% of terminal voltage, the MVA load is limited to 83% of rated WAS, or 978 MVAs (.91 x.91 x 1179 = 978). This Candidate would then apply the 997.5 MVA limit of CPS 3504.01, Section 8.3.1, Table 1.

Objectlve:

Ouestioa Source:

I Level of Difiiculty:

I None New I

3.5 I References arovided to exuninee:

1 CPS 3105.05. Generator (TG). entire mcedure I

Date Written: I 04/29/05 I Author: I Ryder Comments: None

1 Question # I 57 I w

The plant has just scrammed from rated power, THEN the following occurs:

ALL power is lost &om 4160V Bus 1Al Operators are controlling reactor pressure between 800 and 1065 psig Which ONE of the following identifies main control room REACTOR PRESSURE recorders that are available to determine the CURRENT reactor pressure?

A.

ONE recorder on P601 B.

TWO recorders on P601 C.

ONE recorder on P601 ONE recorder on P870 D.

TWO recorders on P601 ONE recorder on P870 Reference provided to examinee:

References:

LP85423, Nuclear Boiler Instrumentation 1 None CPS 4200.01, Loss of AC Power Datewritten: I 03/15/05 I Author: I Ryder Comments:

There is overlap between this question and any part of the Operating Exam. Although Candidates will have need to determine reactor preamre on a number of oculsions in the Scennrios, especially, there is no particular time where they Answer: A Explanation:

A is correct - P a CPS 4200.01, Appendix B, page 24. With 4160 V Bus IBI still powered, there is a -reactor pressure instrument still functional (for the cUrrent reactor pressure band, 800-1065 psig), at P601. That instrument is B21-R623B (a paperless mrder). The other recorder instrument, ILR-SM016, shown in this Appendix is a low-range pressure instrument (0-300 psig, PR LP85423, page I7), and is therefore mi available for determining reactor pressure.

B is incorrect - But is plausible to the Candidate who forgets that the other recorder is shictly a low-range pressure one.

C and D are income3 - But are plausible to the Candidate who, in addition to remembering the paperless recorders on P601, thinks then are also recorders on P870. This especially atbdCtive to the ILT Candidate, where the Simulator uses a DCS Display senen on P870 to allow access to key plant Emergency Response parameters, including Reactor Pressure. This is for haining, only; it is computer point, and is the actual main conhol room configuration. Besides, this is a DCS Display a recorder.

LQuestion # I 57 I u

will have to seek out a y&&y of other instruments, other than those usually-available (Le., DCS displays), to do so.

I Question# 158 I RO/SRO: I Tier:

I Group:

I KA:

I ROIR:

I SROIR: I ~ o g ~ e v e l Both I

1 I

295021 AK3.04 I 3.3 I

3.4 I

Lower SyctemlEvolution Name:

I category:

Loss of Shutdown Cooling KA Statement:

Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING:

Maximizing reactor water cleanup flow I Emergency and Abnormal Plant Evolutions After a Loss of Shutdown Cooling, operators determine there is a NEED to MAXIMIZE the Bottom Head Drain Flow to RWCU, Which ONE of the following describes the REASON for operating RWCU in this way?

to MAXIMIZE the cooling capability of RWCU.

A.

B.

C.

D.

Answer: B To prevent excessive thermal stress of the Feedwater piping.

To minimize thermal stratification of the RPV.

To prevent erosion of the RPV bottom head drain line.

To minimize Feedwater line check valve flutter.

Explanation:

B is correct - Per CPS 3303.01, Sections 8.2.5 and 8.2.6. With the loss of shutdown cooling, the absence of forced flow through the vessel can m l t in thermal stratification. The Section 8.2.6 NOTE (center of page 58) advises operators to maximize RWCU cooling in order to promote sufficient natural circulation to minimize the concern for thanal stratification. The Section 8.2.5 NOTE (top of page 57) also advises operators to increase bottom head drain line flow to RWCU, to increase the forced (RWCU) cimlation through the vessel and prevent bottom head region stratification.

A is incorrect - This choice is derived from another one of the several operational concerns related to RWCU.

Specifically, CPS 3303.01, Section 6.4.2. Excessive feedwater line stress is prevented by limiting the delta-T between RWCU return temperature and Feedwater temperam when operating at low feedwater flow conditions. mere is no Faus+effect relationship between the way operaton have decided to operate RWCU in the stem condition, and the limiting of this delta-T.

C is incorrect - This choice is derived from motha RWCU operational concern. Specifically, CPS 3303.01, Section 6.10.1.1. When operators do maximize the bottom head drain flow (as suggested in the stem), they will necessarily have to comply with limiting that drain flaw to 200 gpm, for the purpose of preventing drain line erosion. However, the REASON for operating RWCU in the way suggested is not directly related to this specific concem.

D is incorrect - Again, this choice is another opwtional concern. Specifically, CPS 3303.01, Section 6.11. Low flow conditions in the feedwata line can cause check valve flutta and excessive valve wear. Operaton are advised to retum all flows (RWCU and/or Shutdown Cooling) through a single line until there is sufficient flow (- 300 gpm per line) to prevent the valve flutter. This, again, has no d i d relationship to the REASON why the operators have decided to operate RWCU, as suggested in the stem.

v I

I Level of Difficulty:

Objective:

Onntion Source:

LP85204.1.2.6 New I

2.9 References provided to examinee:

References:

I None I CPS 3303.01, Reactor Water Cleanup System

1 Question# 1 58 I KA Statement:

Knowledge of the Teasons for the following response4 as they apply to LOSS OF SHUTDOWN COOLING:

1 Maximi&

reacfor wates cleanup flow I

D8teWritten: I 03/15/05 I Author: I Ryda Comments None

L 1 Question # 1 59 I EXPIUUHO~:

C is cared - Per LP85218, Figurrs 4 and 5. A and E will initiate ADS (Figure 4, Division I logic). B and F will initiate A D S (Figure 5, Division 2 logic).

A, B, and D are incorrect - For the reasons d&bed above.

I Question# I60 I L

Operators are implementing EOP-6, Primary Containment Control, with the following:

Suppression Pool Water Level is rapidly lowering CRS directs the RO to makeup to the Suppression Pool via Cycled Condensate (CY),

WITHOUT any assistance fiom operators in the field Which ONE of the following:

(1) identifies how many valves the RO MUST OPEN to establish makeup flow, (2) describes an additional action SUGGESTED by the procedure BEFORE opening the and L

valve(s)?

A.

(1) ONE valve (2) START a second CY pump.

B.

(1)TWOvdves (2) STOP the running CY pump.

C.

(1) ONE valve (2) STOP the running SF pump.

D.

(1)TWOvalves (2) START a second SF pump.

Answer: A Explmation:

A is correct - Refer to CPS 3220.01, Seaion 8.2. The RO needs only to open ISM004 to establish makeup flow to the pool; one CY pump is normally running to pruvide the flow. The CAUTION advises that, in an emergency, 1 SM004 may be opened without throttling ICYO56. This valve (local) is normally wide open, and would first be fully closed (locally) More the RO opens ISM004. The intent, here, is to provide CY pump run-out protection. For this question,

( i q establishing makeup WTHOUT local o p t m assistance), however, the CAUTION suggests that first starting a second CY pump may help w e n t CY pump nm-out.

B is incorrst - For the reasons described above.

C and D me mcorrect - Refer to Sdon 8.3 of the pmzdum. These choices are plausible to the Candidate who cannot d

l the specific eoncan for emblishing makeup flow in this emergency mode. Section 8.3 utilizes Suppression Pool Cleanup and Tmfer (SF) as an alterative (to using CY) pool makeup source. One SF pump (controlled at control room panel P870) is normally running; the Candidate may choose to stop it. A second SF pump is also availablc

Objective:

I Question Source:

I Level of Diflieulty:

Lp65406.1.2.2 New I

2.1 L

Refereour provided to examlnee:

Recennur:

DateWrnien: 1 03/17/05 I Author: 1 Ryder COraaKntr:

This question is categmized as Higher Cognitive (HCL) for the following reasons:

I None 1 CPS 3220.01, Suppression Pool Makeup (SM)

1.

The pump runout concerns presented in this procedure section are a found in the PrecuationdLimitations section of the procedure. If they wae, then arguably this question would only demand the lowest level of mmtal pmassing (simple recall of information) and would be categorized as Lower Cognitive (LCL).

Candidates are llpt expsted to have procedure sections, for evolutions such as this (i.e., important to pmtecthg a lesser pump from a ps!J&d nmout condition, hut plant or to mitigating a significant msimt). committed to memory.

As such, it is expwted that &

Candidates would have to use a higher level of mental processing to answer Pan (2) of the question. Ibis mental processing wwld be used to associate concepts of potential impact on the system whm o p i n g the valves (i.e,, recognition that this amounts to placing an flowpath burden on the running CY pump), with how this cause lead to pump runout, and then associating these pieces of information to derive a logical mahod for proteaing the pump fmm runout.

Accordingly, without applying simple recall (and this being a closed-reference question), several Fundamentals concepts have to be applied to answe~ Pan (2).

2.

a critical evolution vital to pmtecting the

3.

4.

1 Question # I 61 I SCRAM Condition Present and Reactor POWR Above APRM Dormscale or Unknown Emergency and Abnormal Plant Evolutions The plant is operating at rated power, when the main turbine trips, with the following:

0 Operators enter EOP-1A for an ATWS and soon thereafter start BOTH SLC Pumps Operators are controlling reactor pressure between 800 and 1065 psig with the Turbine Bypass Valves and SRVs Which ONE of the following describes a situation where it would be acceptable for operators to INTENTIONALLY LOWER the pressure control band?

A.

B.

C.

D.

Answer: B The SLC Pumps have been running for 30 minutes.

Suppression Pool Temperalure approaches 145F.

Operators decide to use CD/CB to maintain the prescribed level band.

Operators decide to use low pressure ECCS to maintain the prescribed level band.

Explanation:

B is c o d - Per EOP-6, Figure P (Heat Capacity Limit, HCL), as pool temperaNre nexs 145F (approximately), the HCL for a n o d pool level (about 19 feet) is threatened. The HCL is considered one of the Critical Parameted mentioned in CPS 441 1.09, Section 4.3 (and 4.2). Even with an ATWS in progress, that procedure section allows opesators to lower the pressure control band as necessary to stay within the HCL limit.

A is i n m - CPS 441 1.W, Section 4.2 prohibits intentionally lowering pressure (in an ATWS) until specific reactivity shutdown conditions have ken established. Refer to the PRESSURE leg of EOP-IA, specifically: the WAIT sign, and Table X. With both SIC pumps running for only 30 minutes Cold Shutdown Boron has hj&d Opaatorj must remain at the WAIT sign for about IO more minutes before intentionally lowering reactor pl*rsure.

C and D arc inmrrect - These choices suggest that operators wish to lower pressure simply to get below the shutoff bead of either CondensatdCondensate Booster Pumps (CD/CB) (-725 psig)), or one or more of the low pressure ECCS pumps. CD/CB is a Prefared ATWS injection system than might be used to maintain the prescribed level band in EOP-IA, LEVEL leg. Low pressure ECCS (LPCVLPCS) is an Alternate ATWS injection that might be used, only if level cannot be held above TAF (in this case, operators would be directed to blow down).

yet k e n

L-I I

I 1

1 295037EA2.06 I 4.0 4.1 Higher ROBRO 1

Both 1

SyctcmlEvdution Name:

I category:

SCRAM Condition Present and Reactor Power Above Emergency and Abnormal Plant Evolutions

_APRh4 Downscale 01 Unknown

References:

KA Statement:

Ability to detamine andor interpret the fOlkOWing as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor pressure CPS EOP-IA, ATWS RPV Control CPS EOP-6. Primw Containment Control I

Ouestlou Source:

I Level of Difficulty:

Objective:

LP87553.1.7.9 New 3.0 I

I CPS 441 1.09, RPV &sure Control Sources I

Date Written: 1 03/17/05 I Author: I Ryder Commenw None

1 Question # 1 62 L

Per the Clinton Radiological Annex, which ONE of the following events WOULD BE considered an OFF-SITE radiological release?

A.

A drum containing low-level radwaste falls off a truck and spills its contents on the ground JUST OUTSIDE the Protected Area fence; the spill is immediately contained, and does NOT get into the storm drainage system.

A temporary modification has MC cross-connected with CY when a rupture occurs on the MC Storage Tank, the spill from the tank reaches ONLY AS FAR AS the Nuclear Training Center, it does NOT get into the lake, and does NOT get into the storm drainage system.

The main stack effluent monitor alarms; field teams discover HIGHER than normal radiation levels ONLY AS FAR AS the railroad tracks that run along side of Highway 54, and ONLY in the direction of the AmerGen sign.

Radiologically contaminated water is spilled into the lake on the intake structure side; grab samples show HIGHER than normal amounts of radionuclides, but due to dilution, the problem extends NO FARTHER THAN 500 YARDS from the shoreline, there are NO radionuclides detected on the discharge side of the lake.

B.

C.

D.

Answer: c Explanation:

Ciscomct-RefcrtoEP-AA-1003.pageCG4, andtotheCPS-ODCM,Figure3.1-1.

Theperfect circledrawn around the Plant is a 0.5 mile radius and is defined as the Site Boundary. Highway 54 is tangential to the edge of that Site Boundary. The sites Amergen sign is at the end of the main access road, exactly at the point where the highway touches the Site Boundary edge. The railroad tracks (not shown on the Figure) run along side of the highway on the side cgpg&

the plant This choice describes a gaseous effluent release that has reached beyond the Site Boundary, and is an Off-Site release.

A is incomet - This contained spill goes no further than perhaps 100 yards from the plant, and well within the Site Boundary.

B is i n

m

- The Nuclear Learning Center is on company property well within the Site Boundary, perhaps 400 yards from the MC Storage Tank.

D is incomet - Ihe Site Boundary radius is 0.5 miles from the plant. 500 yards is about 0.28 miles. This liquid release is still on-site.

Objective:

Question Source:

1 Level of DiMculty:

I None New I

3.6

Rderenca:

I Question# I 62 I CPS-Offsite Dose Calculation Manual (ODCM)

EP-AA-1003, Clinton Radiological Annex Tier:

I Group:

I KA:

I ROIR: I SROIR: 1

~ o g ~ e v e l 1

1 1

295038EA2.01 I

3.3 4.3 Higher RO/SRO I I

I I

I Both SyatedEvolutlon Name:

1 catqory:

I Emergency and Abnormal Plant Evolutions High Off-Site Release Rate KA Statement:

Ability to dnamine mdor inter@

the following as they apply to HIGH OFF-SITE RELEASE RATE: Off-site

I Question # I 63 I L-The plant is operating at rated power when the following occurs:

Fire alarm (red flashing strobe light) occurs at MCR panel P661 Operators CANNOT immediately CONFIRM whether or not an ACTUAL fire exists Which ONE of the following describes the REQUIRED operator action?

A.

B.

C.

Start a SECOND fire pump, so that TWO are running.

OPEN the FP CNMT and FP Drywell Isolation Valves.

Place ALL of the SRV control switches in OFF at P601.

Rtftrtnces:

D.

MANUALLY initiate Halon in the main control room.

LP85239, Main Steam System CPS 1893.04, Fire Fighting Answer: c L

Exphation:

C is correct - PR CPS 1893.04, Section 8.1.2.1. P661 contains the Div I SRV logic. This action minimizes the risk that a iirc in P661 could result in the auto-opening of the SRVs (plant depressurization) with 'A' solenoid (Div I) control sm.tch*l (OFF-AUTO-OPEN) still in AUTO. Refer to LP85239, pages 14-15.

A is incorrect - Per CPS 1893.04, Seztion 8.1.7. Only one fire pump need be running.

B is incorrect-Ptr CPS 1893.04, Section 8.1.8. Operators are directed to open these valves only when the fire is in CNMT or the drywell.

D is incorrect - There is rn requirement for initiating Halon unless an actual fire exists.

1 Question# 1 64 I i

RO/SRO:

I Tler:

I Group:

I KA:

I ROIR:

I SROIR: I c o ~ ~ e v e l Both I

I I

2 I

295032EK3.03 I

3.8 I

3.9 I

LOWR.

System/Evolution Name:

I category:

1 Emergency and Abnormal Plant Evolutions

_High Secondary Containment Area Temperature KA Statement:

Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Isolating affected systems Answer: D L

Explanatlou:

D is correct - P a EOP-2, bottom-most WAIT steps.

A and C are inwnect - Each of t h e describes a situation where a system is needed for EOP actions. Isolating such a system is a allowed ky EOP-8 (step 3 I).

B is ineorrat - This choice is incorrect because the word anytime suggests that since the system is not needed for injection (adequate core cooling), operators &

isolate it if it were needed for Containment Spray in EOP-6. This is

& hue.

References provided to emmloec:

Refercuccr:

I None; access to EOP flowcharts is OK I CPS EOP-I. RPV Conml CPS EOP-IA, ATWS RPV Control CPS EOP-2, RPV Flooding CPS EOP-6, Primary Containment Cootrol CPS EOP-8, Secondary Containment Control DateWriltea: I 05/16/05 I Author: I Ryda Commenh Given this KA, we have taken an applications approach as a means of creating a question that provides sufficient discrimination. To correctly answer this qudon, the Candidate must correctly apply the requirements of EOP-8, step 3 I. Ihe faa that the WAIT steps at the bottom of EOP-2 &

operators to isolate the MSL Drains does ml dehact h m the Fact that doing so is also consistent with the EOP-8, step 31, allowance.

I Question # I 65 I

References:

1 RO/SRO:

1 Tier:

I Group:

I KA:

1 ROIR:

1 SROIR: I

~ o g ~ e v e l 1

CPS 3312.03, RHR Shutdown Cooling LP85205. RHR Both I

2 I

1 I

205000A1.05 I

3.4 I

Lower SystemlEvolutiou Name:

I category:

Shutdown Cooling System (RHR Shutdown Cooling I Plant Systems KA Statement:

Ability to predia and/or monitor changes in parametas associated with operating the SHUTDOWN COOLING SYSTEMMODE controls including: Reactor water level With REACTOR PRESSURE LESS THAN 30 PSIG, operators are warming RHR loop B in preparation for placing it in Shutdown Cooling.

Per CPS 3312.03, Shutdown Cooling, which ONE of the following describes a POTENTIAL planthystem response when operators INITIATE WARMUP FLOW.

A.

B.

C.

D.

RHR Pump RAPIDLY OVERHEATS.

Reactor water level SUDDENLY UNCONTROLLABLY LOWERS.

Reactor pressure SUDDENLY UNCONTROLLABLY RISES.

RHR Pump RAPIDLY approaches RUNOUT.

Answer: A Explanation:

A is correct - P a CPS 33 12.03, Section 5.4. At such a low pressure (GO psig), there is a chance that warmup flow (i.e., RHR Pump is not yet running) may not be sufficient to open the RHR Pump Discharge Check Valve, F03 I. This could cause the downstream piping to drain to radwaste when operators open the radwaste discharge valves. If the pipe volume emptiep, and them the F03 I opens (likely, due to the downstream pressure now being relieved), a sudden, uncontrollable drop in reactor water level (level hansient) will occur as that piping refills. See LP85205, Figure 5.

B is income3 - lhis choice has face validity, but is not a concern during this evolution C and D me munrect - Candidate is expected to know that the RHR Pump is not yet running when warmup flow is initiated.

Obiectlve:

I Questlou Source:

I Level of DIMculty:

LP85205. I. I4 New I

3.0

I Question# I 66 I L

RO/SRO.

1 Tier:

1 Group:

I KA:

I ROIR:

1 SROIR: I CogLevel Both 2

I I

1 215003K1.06 1

3.9 I

4.0 I

Higher Syrtem/Evoiutlon Name:

I cat*ory:

Intmnediate Range Monitor ORM) System 1 Plant systems KA Statement:

Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: APRM SCRAM signals Rtlemncn:

L LP8541 I, APRM/LPRM System CPS 3306.01, Sodntermediate Range Monitors E~pl~ation:

D is correct - Per LP85411, pages 24.25, and Figure 14, and per CPS 3306.01, Section 6.4. In MODE 2, the Mode Switch is not in RUN. The APRM Upscale Nextrun Trip setpoint is 15%. An IRM Range IO reading of 40 is within instrument loop tolerance fa this hip signal, at about 13% APRM power, A arid C me incorrect - These equate to Range IO readings of abwt 23 and 30, respectively, and are about 7% and 10%

APRM power, respectively.

B is incorrect - Ofthe disbamrs, this equates to the 2d highest APRM power level (about 10%). Even if it were shown that the 15% AFRM upscale trip inshument loop could reach down as far as IO%, this choice is incorrect bscause, per CPS 3306.01, Section 6.4, the RO should have mged up e reaching this 90 value on Range 9.

Stan conditions are explicit about the fact that opsators ARE ranging up as required by procedures.

Objective:

I Ouestlon Source:

I Level of Difflcuity:

LP8441 I. I.

11.1 New I

3.0

Question # I 67 I ROBRO. I mer:

1 Group:

I KA:

I ROIR:

SROIR I CogLevel.

Both 1

I 2

I 295017AAZ.03 I

3.1 I

3.9 I

Lower Syrtem/Evdutioa Name:

I category:

i High off-Site Release Rate 1 Emergency and Abnormal Plant Evolutions References provided to examinee:

None Referenfa:

CPS EOP-9, Radioactivity Release Control cPS5140.41, almmponseforORIX-PRO01 CPS 4979.01, Abnormal Release of Airborne Radioactivity KKStlteDIent:

Ability to determine andor interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Radiation The plant is operating at power when a VALID, ALERT level alarm occurs on the in-service HVAC Stack Effluent Monitor.

Which ONE of the following describes the operational andlor radiological significance of this alarm?

A.

B.

C.

D.

A Technical Specification entry is required.

An EOP-9, Radioactivity Release Control, entry is required.

NEITHER the ON-SITE, NOR the OFF-SITE exposure limits are being exceeded.

The ON-SITE exposure limits MIGHT have been exceeded; the OFF-SITE exposure limits are NOT being exceeded.

Answer: c Datewritten: I 04/29/05 1 Author: I Ryder Commenia: None

1 Question# 68 3 ROBRO 1

Tler:

1 Group:

I KA:

I RO1R:

1 SROIR. 1 CqLevel Both I

1 I

295018AA1.03 I 3.3 I

3.4 I

Higher SyltemlEvdotioa Name:

I category:

L Partial or Complete Loss of Component Cooling Wata 1 Emergency and Abnormal Plant Evolutions KA Statement:

Ability to operate andor monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT 1 COOiJNGWATER: AEected systems so 88 isolate damaged pations I

The plant is operating at power when the following occurs:

0 CCW Effluent Monitor, IRM-PR037, alarms (a valid alarm) 0 1RM-PRO37 is nearing its HIGH alarm setpoint 0

Source of the alarm is a tube leak on the ONLY available NRHX 0 Leak rate is about 29 gpm Which ONE of the following describes the required operator action?

A.

Enter CPS 4001.02, Automatic Isolation.

B.

Isolate CCW from the NRHX and open the RWCU Heat Exchanger Bypass, 1G33-F104.

C.

D.

Stop the RWCU Pumps and isolate the RWCU system.

Commence a noimal plant shutdown within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Answer: c Erplnnation:

Ciseomct-PaCPS5140.49,OpaatorAction#l,andperCPS3303.01,Section8.3.3. OperationofRWCUwith the plant above 120°F and CCW isolated from the NRHX cannot continue. Operators are directed to remove RWCU fimn saviw.

A is incorrect - PR CPS 4M)I.O2C001, page 3. RWCLJ delta-flow isolation setpoint is 59 gprn afier 45 seconds. Per Computer Point E31DA001, the normal, at power, sensed RWCU differential flow (due to instrument calihration impct by system flow dynamics) is about 25 ~pm. The system should still be un-isolated with the tube leaking at only 29 gpm; total system delta-flow is well below the 59 gpm isolation setpoint (is., about 54 gprn). Them is no reason for operators to enter the Automatic Isolation procedure.

B is incoma - For the reason associated with the eomct answer, c.

D ia incomct - This choice suggests that Tech Spec 3.0.3 applies. It does not. Its plausibility is rooted in its attraction to the Candidate who wants to consider the RCS Leakage aspen of this event. This NRHX tube leak does constitute a RCS PreMUre Boundary leak (deiined in Tech Spec 1.1); therefore, Tech Spec 3.4.5.C does not apply here.

Although this Tech Spzc is not a 3.0.3 entry, this choice is worded in a way that suggests the need for a Candidate to d

l short-term (< I-hour) Tech Specs from memory.

NOTE: Each of these choices suggests that +tors have determined the need to isolate the source of the release. The stem avoids explicitly Suggesting this in orda to make choices A and D plausible. Tech Spec 3.4.5.A, alone, is sufficient to argue that isolation ofthe I&

is in fact required.

I Question# I 68 1 References provided to examinee:

References:

RO/SRO I

Tier:

I Group:

I KA:

I ROIR: I SROIR. I ~ o g ~ e v e l Both I

1 I

295018AA1.03 I

3.3 I

3.4 I

Higher SystemIEvolution N8me:

I Category:

Partial or Complete Loss of Component Cooling Water I Emergency and Abnormal Plant Evolutions None CPS 3303.01, Reactor Water Cleanup System CPS 4001.01, Automatic Isolation CPS 4001.02C001, Automatic Isolation Checklist CPS Tech Spec 3.4.5, RCS Operational Leakage CPS Tech Spec 1.1, Definitions CPS 5140.49,l RIX-PRO37, alarm response procedure KA Statement:

Ability to opaate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Affected systems so as to isolate damaged portions I

Objective:

I Oueation Source:

I Level of Difficulty:

LP852W. 1.12 N W I

2.6

I Question# I71 I ROISRO: I Tier:

I Group:

I KA:

I ROIR:

I SROIR: I ~ o g ~ e v e l Both I

2 I

I I

223002 2.4.31 I

3.3 I

3.4 I

Lower SyitemlEvolutlon Name:

I C1tegory:

PCIS/Nuclear Steam Supply Shutoff System I Plant systems L

KA Statement:

Knowledge of annunciator alarms and indications, and use of the response procedures Which ONE of the following identifies an annunciator for which the alarm response procedure directs operators to place the Div 1 Sensor Bypass Switch in BYPASS if the alarming (tripped) condition CANNOT otherwise be removed?

A.

5067-6D, DIV 1 TRIP UNIT OUT OF FILE B.

5067-1H, INBOARD LOSS OF ISOLATOR POWER C.

5067-7B, LDS P632 ISOLATOR CARD POWER LOSS D.

5063-8A, DIV 1 SAFETY ASSOCIATED ATM TROUBLE Answer: A Explmatlon:

A is correct - Per CPS 5067-6D, Operator Action 2.a. Although the Candidate is not expected to know this action fmm memory, hdshe is expected to recognize that this annunciator alarms whenever any one of 9 separate analog trip modules (ATMs) hecomes unseated (dislodged) in its rack, and that this produces a Div 1 trip for the associated parameter (e.g., a Group 1 isolation half-trip is generated if IB21-N681A (RPV Level I) is the offending A m ).

B is incorrect - Per CPS 5067-1H. This annunciator is associated solely with the MSlV Leakage Control System, and is not affiliated with the Divl NSPS logic cabinet (P661) or its components, including the Sensor Bypass Switch, Cis incomt - Per CPS 5067-78. mihis belongs to the Div 1 portion of Leak Detection (LDS) and is not affiliated with the Divl NSPS logic cabinet (p661) or its components, including the Sensor Bypass Switch.

D is incorrect - Per CPS 5063-8k This belongs to ATMs associated solely with RCIC is not affiliated with the Divl NSPS logic cabinet (P661) or its components, including the Sensor Bypass Switch.

L Objective:

I Onestion Source:

I Level of Difficulty:

LP85434.1.4.10 New I

4.0 References provided to examinee:

References:

Datewritten: I 05/03/05 I Author: I Ryda Commenh: None I None 1 CPS 5067-6D. IH, and 7B, and CPS 5063-8& alarm response procedures

I Question # 1 72 1 i-The plant is in MODE 3, cooling down, when the following occurs:

Low feedwater flow conditions cause annunciator 5000-2F, RWCU HI DIFF FLOW TIMER INITIATED, to alarm CRS decides to prevent an unnecessary RWCU isolation Which ONE of the following describes:

(1) how procedures direct operators to PREVENT this RWCU isolation, (2) the automatic response of RWCU, if operators are able to DEFEAT ONLY ONE and division before the Differential Flow Timers time out?

A.

(1) At P632 (2) RWCU isolates.

(1) At P855, manually dial BOTH Differential Flow Timers fully (2) RWCU does NOT isolate.

(1) At P632 in BYPASS.

(2) RWCU isolates.

(1) At P855, place BOTH RWCU Isolation Bypass switches in BYPASS.

(2) RWCU does NOT isolate.

P642, manually dial the associated Differential Flow Timer filly COUNTER-CLOCKWISE.

B.

COUNTER-CLOCKWISE.

C.

P642, place the associated RWCU Isolation Bypass switch D.

Answer: c 1 Explanatton:

C is c o r n - Refa to O S 5000-2F. LP85404, page 25, and LP85204, page 16. Pan (I ) - The Div I RWCU Isolation Bypass switch is lofatcd on P632 (Div I Leak W o

n System panel); similarly, the Div 2 switch is on P642 (Div 2).

At cach of thw panels, P632 and P612, is also that division's Diff-Flow Timer (referred to in CPS 5000-2F as I E3 I-R615A and B). Although it is physically possible to manually dial back a running timer, tt is not the mnhod prescribed by pmcedurrs, including CPS 5000-2F. Pan (2) - A given Isolation Bypass switch defeaL% the isolatiun only fur thc isolation valves conholled by that division. The Outboard isolations are Div I, the Inboards are Div 2. Even if opaatm only manage to place one of the switches in BYPASS M o r e 45 seconds has expired (the Tim-setpoint), the valves conbvlld by thc oppon'fe Division will close. Thus. RWCU will still isolate.

A, B, and D arc in-t

- For the reasons desaibcd above.

i

[ Question# 1 72 I Oblative:

I Ouution Source:

I Level of Difficulty:

LP854M.l.lI.26 NeW I

2.2 Refermcu provided to ex8mlnn:

References:

I None 1 LP85204. Reactor Wata Cleanur, System LP85404, Leak Detection System CPS 5000-2F. RWCU Hi Diff Flow Timer Initiated, alarm response

~htewritten: I 03/22/05 I Author: 1 Ryder Comment%

These panels, P632 and P642, are within the control m m complex, but m.l within the control board area of where the opaators me. normally stationed. That is, they are back panels. These back panels satisfy the intent of this Generic KA. Arguably, this KA does not nffessarilv demand that the question specifically address local conh-ols, but rather that it address local controls PI MCR controls.

c

I Question# I73 I L

RO/SRO I

Tkc 1

Group:

I KA:

1 R O ~ R : I S R O ~ R : 1 c o g ~ e v e l Both I

3 I

Generics I

2.4.43 I

2.8 I

3.5 I

Lower Aiscorrect-PerCPS 1021.01,Section8.1.4. ALLCALLfeatureisavailable,andsoisanyplantalarm thancan be manually initiated. P a CPS 3842.01, Section 8.1.2, this feature known by how the associated pushbutton is actually labeled: All Page.

B is incorrect - P a CPS 1021.01, Section 8.1.4. The Fuel Building Evacuation alarm can only he automatically initiated (by radiation monitoring); it has no manual feature, either at the RSP or in the MCR C is incorrect - Per LP85433, page 16. Only one ADS-SRV (F05iG) can be manually operated at the RSP.

D is incorrect - Per LP85433, page 6. Although LPCl Loop C is a Div 2 subsystem, it cannot he operated to inject from the Remote Shutdown Panel area. The other Div 2 LPCI Loop, B, be used.

SystemlEvolutlon Name:

I catecory:

I Emergency Pmcedures and Plan 1

KA Statement:

Knowledge of emergency mmmunications systems and techniques Which ONE of the following identifies operational features available at the Remote Shutdown Panel area?

A.

Can initiate the ALL PAGE mode for the Gaitronics system, and sound the plant GENERAL PURPOSE alarm.

Can sound the CONTAINMENT EVACUATION alarm, and sound the FUEL BUILDING EVACUATION alarm.

Can MANUALLY operate TWO of the ADS-SRVs.

Can inject to the RPV with LPCI Loop C.

B.

C.

D.

Answer: A Referen& proviaed to eumlnee:

References:

None CPS iO21.01, SiteCommunications CPS 3842.01, Plant Communications Alarm Test LP85433, Remote Shutdown 0bIHth.e:

I Questlon Source:

I Level of Dlffleulty:

None New I

2.2 i

I Question# 74 I

Group:

I KA:

1 ROIR: I SROIR: I

~ o g ~ e v e l RO/SRO I Tier:

Both I

2 I

I 1

300000A4.01 I

2.6 1

2.1 I

Higher SptemlEvdntIon Name:

I Category:

rnsfnlment Air system ( IAS)

I Plant systems

\\-

KA Satemcut:

Ability to manually operat e andlor monitor in the w n m l mom: Prcssurc gauges The SERVICE AIR HEADER PRESSURE gauge on P800 has three colored bands: Green, Yellow, and Red.

WITHOUT operator action, which ONE of the following describes the expected status of Plant Air components ONE MINUTE after an operator sees the reading on the gauge drop into the RED band because of a significant air leak?

A.

TWO service air compressors are running; one or more automatic ring header isolations are CLOSED.

ONLY ONE service air compressor is running; all automatic ring header isolations are OPEN.

TWO service air compressors are running; all automatic ring header isolations are OPEN.

ALL THREE service air compressors are running; one or more automatic ring header isolations are CLOSED.

B.

C.

D.

Answer: A Explanation:

A is wmcI -The RED band wvers prssures at or below 70 psig (observed in both the MCR and Sirnulalor wnfigudtons). Per CPS SWI-6B. the standby air compressor should have auto-stanal at 80 psig (lowering).

thdom. two arc now m i n g. Per CPS S041-5C. the ring header isolations should close at 70 psig (lowering). The stem use8 'one minute' as a way to ensure that all pomons of the Plant Air system have had sufficient time 10 scnse the lowatd pressurr and to a w n that all affected ring header auto isolations have closed ar a rnuit B and C BTC incom -For the reasons described above.

D in in-

- The 3" air compressor is normally in Pull-To-Lock (PTL) and will mt, therefore, 3Ut0-stm.

Oblative:

I Questloo Source:

I Level of Dimculty:

Lp85301. I.7 NCW I

3.0 References provided to examinee: I None Referencea:

1 CPS5041-5C,and-6Balarmres~onseproood UreS DateWritteu: I 04/29/05 I Author: 1 Ryder Commenb None

1 Question# 175 1 A STATION BLACKOUT is in progress.

Per CPS 4200.01, Loss of AC, which ONE of the following describes what operators are directed to do with the P U N T AIR system, why?

In the main control room, place all 3 Service Air Compressor control switches in pull-To-Lock, to prevent their auto-restart (when power is restored) until their support systems are also made available.

In the field, place all 3 Service Air Compressor Mode Selector switches in the UNLOAD position, to prevent auto-restart of the compressors (when power is restored) until their support systems are also made available.

In the main control room, place the Containment SERVICE Air Header Isolation Valves control switches in the CLOSE position, to preserve available air for vital plant equipment systems, when plant air pressure is restored.

A.

B.

C.

D.

In the field, gag SHUT the SERVICE Air Ring Header Isolation Valves for the Radwaste, Turbine, Control, and Auxiliary Buildings, to preserve available air for vital plant equipment, when plant air pressure is restored.

Answer: A Explauatiou:

A is correct-PaCPS 3214.01, Section 2.14 and CPS4200.01, Appendixk No furtherexplanation required.

B is incorrect-Pa CPS 3214.01, Section 6.5. The Q&!Y way to protect the compresson on power restoration is by placing them in FTL.. E m if the in-field switches are placed in UNLOAD, the compressors can still auto-start (unless in PTL) and run unl&

C is incorrect - Refer to Lp85301, page 24. These are 3-position (CLOSE-AUTO-OPEN) control switches, normally in AUTO. ney will in fsct auto-nopen when air pressure is restored, but thats OK. There are no requirements to prevent these valves from reopening in an effort to preserve air (especially Inshument Air) for more important plant equipment.

D is incorrect - For the same reason associated with choice T.

Also, these valves CANNOT auto-reopen. They Rip shut on low air heada prcssurc (70 pig) and must be manually mlatched and mpened in the field. There is no need to gag them shut Refer to LP85301, pages 22-23.

Obiective:

I Question Source:

I Level of DiMculty:

~~4200oi.i.3.3 New I

3.0

Referencn provided to examinee:

Referencu:

None LF'85301, Scnicc Air and Insbument Air CPS 3214.01, Plant Air CPS 4200.01, Loss ofAC

[ Question# I76 1 v

Using the provided references, answer the following.

The plant is operating at rated power when the following occurs:

At Time = 0000 hours0 days 0 hours 0 weeks 0 months , a fault in the RAT supply breaker causes 4160V Bus 1C1 to transfer to the ERAT At Time = 0300 hours0.00347 days 0.0833 hours 4.960317e-4 weeks 1.1415e-4 months , a fault in the RAT supply breaker causes 4160V Bus 1B1 to transfer to the ERAT At Time = 1500 hours0.0174 days 0.417 hours 0.00248 weeks 5.7075e-4 months , the supply breaker to Div 1 125 VDC Distribution Panel, 1 A, trips open and will NOT re-close If NONE of these failures can be corrected, which ONE of the following identifies the LATEST time by when the plant MUST be in MODE 3?

The plant must be in MODE 3 no later than...

A.

B.

C.

13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months after the DC supply breaker fails.

14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months after the DC supply breaker fails.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months after the RAT breaker for Bus 1B1 fails.

D.

Answer: B 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after the RAT breaker for Bus 1 C 1 fails.

Exptaoatloo:

B is correct - Per TS 3.8.9, Conditions C and D. For Condition C, only the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months completion time applies. Per Condition D, the plant must be in Mode 3 within 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months (241 allowed outage time + 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months completion time = 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months ) atter the DC supply breaker failure. To avoid applying the 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery... completion time of Condition C, the SRO Candidate must recognizdrecall the following portions of TS 3.8.9 Bases: 1) on page B 3.8-78, the RAT breaker failures for the AC buses do not constitute Distribution System inoperahilities (because the ERAT source is still available to the buses); 2) once this is recognized, then the Candidate will avoid looking at these failures (both AC buses, followed by the DC panel) as a string of wntiguous failures which would otherwise require the application of the 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months constraint.

A is incomt - See the explanation above. The Candidate will choose this as the answer if hdshe inappropriately applies the 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ... completion time ofTS 3.8.9, Condition C.

C is incorrect - This is not w m t because the DC breaker problem is more limiting (as described above). It is plausible to the Candidate who inappropriately applies TS 3.8.1, Condition C; i.e., thinks the loss of the RAT supply offsite circuit) to Qm divisional buses (IBI and ICI) is synonymous with the loss of two offsite circuits.

1 Question # I76

References:

SROIR: I ~ o g ~ e v e l RO/SRO I Tier:

I Group:

I KA:

I CFR I

SRO 1

1 I

295004 2.2.25 I 55.43@)(2) 1 3.7 Higher I

I I

SyatemlEvolution Name:

I category:

I Emergency and Abnormal Plant Evolutions L

Partial or Complete Loss of DC Power CPSTS3.8.l,ACSources-Opaating CPS TS 3.8.9, Distribution System-Operating (and Bases)

KA Statement:

Knowledge of basa in technical specifications for limiting conditions for operations and safety limits D is incorrect - Per TS 3.8.1, Condition A, then Condition F. Were it not for the more limiting DC distribution problem, this would be the correct B~SWR.

Objective:

Ouation Source:

I Level of DiMculty:

I LP85263.

I. 16 N W I

2.3

I Question# I 77 1 L

RO/SRO I Tler:

I Group:

I KA:

I CFR I SROIR: I c o g ~ e v e l SRO I

1 I

295016 2.4.6 I 55.43@)(5) I 4.0 I

Higher SystemlEvdution Name:

I category:

Control Room Abandonment KA Statement:

Knowledge of symptom based EOP mitigation strategies I Emergency and Abnormal Plant Evolutions The plant is operating at rated power when the following occurs:

An ATWS results Shift Manager determines that a control room evacuation is required Before leaving, operators place the Mode Switch in SHUTDOWN Operators arm and depress the Manual Scram pushbuttons and manually initiate ARI The SLC Pumps will NOT start Personnel abandon the control room with the following:

o Main turbine is on line o Reactor power is 35%

o Scram air header is DE-PRESSURIZED o Main control room is UNINHABITABLE and INACCESSIBLE v

L Which ONE of the following describes the NEXT appropriate operator action?

A.

B.

C.

D.

Locally open the SLC Storage Tank Outlets and start the SLC Pumps.

Scram all control rods using the HCU Scram Test Switches.

Terminate and prevent Feedwater, HPCS, and RCIC.

Defeat the MSUOG and IA Interlocks.

Answer: c Explanation:

C is correct - Per CPS EOP-I& Level leg, and CPS 4003.01, Sections 4.3 and 4.4. opadton d~ have the facilities to prevent HPCS injection (at the Div 3 switchgear, per Section 4.4.4), Feedwata injection (by closing the Feedwater Shutoffs pa Section 4.4.3), and RCIC (conhullable at the Remote Shutdown Panel). The objective here would be to IOWR RPV watu level to -60 inches and establish Level Band B.

A is incorrect - Besides there being no procedure guidance for this, the SLC squib valves will arestartedlocally(seeLP85211,page21).

B is incotrec-This is allowed by CPS EOP-I A, Power leg, and CPS 441 1.08, Section 2.6. However, this is time intensive and would be a vain attempt to solve the ATWS problem, given that the stem conditions indicate a hydraulic type of ATWS exists (Le., scram air header has already deqessurized). This is a the NEXT appropriate action D is incorrect - Per CPS EOP-LA, Level leg, and CPS 441O.OOCOO4, this requires accessibility to several main control room panels. Stem conditions indicate the control room is nnt accessible.

fire when the pumps

I Question # I 77 1 References provided to examinee:

References:

Oblealve:

I Ouestlon Source:

I Level of Difficulty:

None New I

2.5 EOP flowcharts CPS 4003.01, Rmote Shutdown CPS EOP-IA, ATWS RF'V Control CPS 441 1.08, Alternate Control Rod Insertion CPS 441O.OOCOO4, M a t i n g MSUOG Interlocks WE521 1, Standby Liquid Control

1 Question # 1 78 ]

L Multiple system fhilures have resulted in Reactor Pressure rising to a PEAK of 1340 psig, as indicated on control room recorders.

Which ONE of the following identifies:

(1) the Reactor Coolant System (RCS) portion MOST impacted by this overpressure transient,

. I..

the d d t a - p s & k thw portions and &e RPV steam iome is something less thank0 psig. Therefore, the 1340 psig steam dome overprcw~re transient equatw Recirc discharge piping pressures that far below the respective m a allow ed... values.

C and D are incomd - A 1340 psig steam dome value equates to that same value throughout the RPV. In fact, the

-steam dome is essentially synonymous with top h d. The RFV Design pressu re is 1250 psig. Its max allowed and (2) whether or not that RCS uortions MAXIMUM ALLOWED TRANSIENT PRESSURE value has been EXCEEDED?

L A.

(1) Recirc pump DISCHARGE piping (2) Has NOT been exceeded.

(1) Recirc pump SUCTION piping (2) HAS been exceeded.

B.

C.

(1) RPV BOTTOM Head (2) Has NOT been exceeded.

D.

(1) RPV TOP Head (2) HAS been exceeded.

Answer: B v

Esplamtfon:

B is comet - Refer to Tech Spa SL 2.1.2 Basis for all of the answer choices. The RCS suction piping is at the lowest elevation ofmy RCS portion. The 1325 psig (stcam &me) SL value equates to 1375 psig at the lowest elevation portion ofthe RCS @e., Rccirc suction piping); i.c. a +50 psig difference. Therefore, an overpressure transient peak pr**surr of 1340 psig (on the mnhul room recorde~~,

which look at steam dome pressure) equates to 1390 psig in the Rsirc suction piping. The maximum allowed transient pressure value for any portion of the RCS is: 110% of the Q&II pr**surr value for Thueforc, the max allowed hansient pressure value is 1375 psig (1.1 x 1250 = 1375). Therefore, the overpressure of 1390 psig in the bArc Suction pipiing poltion of RCS dprs exceed this max allowed... value, and this RCS portion is clearly the most impacted, relative to the other given RCS choices.

A is incorrrct - The Design pressme for the Recirc discharge piping is 1550 psig or I650 psig, depending on the loeation relative to the dischsrge valve. As such, these portions are & the most impacted. Additionally, the I 10%

values arc 1705 win and 1815 asie remectivclv. Sine the% Dortions are at elevations higher than the Recirc piping,

@on.

The Dcsign prcssurr value for the Recirc suction piping is I250 psig.

L 1 Question# I 78 1 I

CFR I SROIR: I c o g ~ e v e i RO/SRO I

Tier:

I Group:

I KA:

SRO 1

I 1

I 29502SEAZ.OI I 55.43@)(2) I 4.3 I

Higher SystedEvdution Name:

I Catmow:

High Reactor Pressun

[ Emagency and Abnormal Plant Evolutions KA sbtcmcot:

Ability IO duennine and/or intspFn the following as they apply IO HIGH REACTOR PRESSURE: Reactor pressure (1 10%) value is 1375 psig. The 1340 psig overpressure transient has exceeded that value.

Obldve:

I Question Source:

1 Level of Diffculty:

LP87621.1.6 New I

2.3 References provided to uamlnee:

References:

I None I CPS Tech S m SL 2.1.2, Reactor Coolant System Pressure SL (and Basis)

Lhltewrfttm: I 03/30/05 1 Author: I Ryda Commentr: None

I Question# 1 79 1 c

Using the provided references, answer the following.

L L

The plant is in MODE 4, making preparations for refueling, with the following:

The reactor was shut down 6 days ago 0

Reactor water temperature is 122°F THEN, a complete Loss of Shutdown Cooling occurs NO Reactor Re& Pump is available There is NO readily available means of restoring shutdown cooling If operators were to MAXIMIZE the reactor water level, as allowed by procedures, how long will it take before a Mode change is required?

A.

61minutes B.

9Ominutes C.

102 minutes D.

120minutes Answer: c Explmitlon:

C is correct - Per CPS 4006.01, Section 4.6.6, opRators arc allowed to raise RPV water level as high as the main steam lines to gain an initial cooling &ea and delay a Mode 3 miry. Section 4.6.4 directs operators to refer to the Heatup Rate and Boil-off Time Curves to assess the heatup rate that can be expected. The SRO Candidate should review the w

e labeled 'Reactor V e l Heatup Rate - Before Refueling' and plot a ""F/Hr" pint where '6 days after shutdown' inters& the 'Main Steam Lines' wata level curve. The RsUlt of this plot yields an approximate 46"Fihr heatup rate.

With current reactor water t e m p a N ~

at 122"F, a Mode 3 a i r y (200°F) is 78'F away. From this pint a simple calculation shows that it will take about 1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , or 102 minutes, to reach 200°F. Calculation: 78 + 46 = I.7 = I hour, 42 minutes = 102 minutes.

A is i n c o d - This choice is plausible to the Candidate who carelessly translates 1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months (as defined above) to I how, 7 minutes (67 minutes). Its plausibility is based on a demonshated propensity for people to make exactly this kind of careless mistake.

B is incorrect - Refer to the same explanation as for the un'rect answer, "2, This is the calculated time if the Candidate were to believe that the maximum allowed RPV water level is +44 inches Shutdown Range. The idea that the Candidate might believe this is soundly based on the fsa that Candidates readily associate +44 inches with the minimum level nsessary to promote adequate natural circulation in the absence of forced cooling flow. SRO

L 1 Question# I 79 1 ROBRO 1

Tier:

1 Group:

1 KA.

1 CFR I SROIR: I CqLevel.

SRO I

I 1

I 295021 AAz.01 1 55.43(bX6) I 3.6 I

Higher Syrtem/Evolotion Name:

I cstegory:

Loss 0fShutdovm Cooling I Emergency and Abnormal Plant Evolutions uASt.tcwat:

Ability to &tennine andor interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor water heatup/cooldown rate Candidate is expected to recall that CPS 4006.01, Section 4.6.6 allows operators to raise level as high as the main steam lines. Calculation: 78 + 52 = 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months = I hour, 30 minute8 = 90 minutes.

D is incarcct - Similarly, this is the calculated result if the Candidate inappropriately applies the Vessel Flange water level m e. This level is higha than allowed by CPS 4006.01, Section 4.6.6. Calculation: 78 i 39 = 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months = 120 minutes.

Objective:

I Quation Swrce:

I Level of D I ~ ~ ~ C I I I ~ ~ :

None New I

3.0 Meno-provlded to emmian:

RdLlL?.CCl:

I The Reactor Vase1 Healup Rare - Before Refueling curve, discussod nbo\\e I CPS 4006.0 I. Loss of Shutdown Cooline.

I Reactor Veskl Heatup Rate-Before &fueling curve, discussed above

~atewfitten: I 03/31/05 I Author: I Ryda Commeutc: None 1

L The plant is operating at rated power when TWO of the smoke detectors in Fire Zone F-lb are determined to be INOPERABLE.

Which ONE of the following describes the required action?

RO/SRO 1

Tier:

I Group:

I KA:

I CFR I SROIR: I CogLevel suo I

1 i

1 1

600000AAZ.09 1 55.43@)(5) 2.8 I

Higher SystenIEvolution Name:

I category:

Plant Fire On Site KA Statement:

Ability to determine andor intaprrt the following as they apply to PLANT FIRE ON SITE: That a failed fire alarm detectorexists I Emergency and Abnormal Plant Evolutions A.

B.

C.

D.

Restore BOTH of these detectors to an OPERABLE status within 14 days; otherwise, establish a tire watch to inspect the zone hourly.

Restore AT LEAST ONE of these detectors to an OPERABLE status within 14 days; otherwise, have a fire watch inspect the zone hourly, thereafter.

AFTER declaring these detectors inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch to inspect the zone hourly.

AFTER declaring these detectors inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inspect the zone, and inspect it hourly,-thereafier.

Answer: c 1 Explanation:

C is COtTEcl-Per CPS 1893.01, Appendix D, page 37, TWO inoperable detectors in this zone constitutes having more than half ofthe 3 total detectors, that are in this zone, inopaable. This being a HPCS equipment zone, these detectors an required to be OPERABLE, because HPCS is required to OPERABLE at rated power (Mode 1, per Tech Spec 3.5.1). PaCPS 1893.01, Appstdix A, page21, finpPotccrion impainnent Compensatoty Measure 9.b applies for this case Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of declsring the detectors inoperable, a fire watch must be eptablished, then hourly inspections of zone F-1 b must commence.

A is incomcl - This would be the required action if t h e were at least 4 total detectors in fire zone F-l b. This is the action directed by Canpensamry Measure 9.4 on page 20.

B is incomcl - This choice has face validity and is plausible based on two premises: I) The very difficult-to-read Compemsatory Measure 9.a making the Candidate vulnwble to misreading the requirements, and 2) an operability restoration technique very o b

employed within Tech Specs; i.e., the idea that as soon as at least one of the 2 detectors can be m t o d to operability, the Compensatory Measure can be exited, and no further action would be required. This ismthecase.

D is incaren - This choice is essentially how a Candidate could easily n&~& the very difficult-to-read Compensatory Measure 9.b.

L-I Question # 1 80 1 ROISRO. I Tier:

I Group:

I KA:

1 CFR I SROIR: I ~ogLeve1

~

SRO 1

I 1

1 6ooooOAAZ.09 I 55.43@)(5) I 2.8 I

Higher

~~

SymmlEvoluUon Name:

I Catcpory:

Plant Fire On Site KA Statement:

Ability to determine and/or interpret the following as they apply to PLANT FIRE ON SITE: mat a failed fire alarm detector exists 1 Emagency and Abnormal Plant Evolutions I

Queatlou Source:

1 Level of Dilficuity:

Objeclive:

None New I

2.2 References provided to esamiaee:

References:

I CPS 1893.01, Firehtccfjon lmpairmmt Reporting I

I CPS 1893.01, in its entirery DateWritten: I 03/31/05 1 Author: I Ryda Comments: None

I Question# I 81 I

~

Explanation B is coned - Pa Tah Spec 3.6.5.4, 1.2 psid is beyond the uppa limit of 1.O psid. Condition A requires that d/p he restond to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Refa to Basis for this LCO, B 3.6.5.4, page B 3.6 - 122, the Background discussion portion that reads... %e limitation on positive.. _. This discussion means that too high a drywell-to-CNMT can awe the mt8 to be already uncovered (cleared) at the onset of a DBA LOCA (as a result of the downward force on the annulus water level). If a LOCA, then, were to occur, the RPV blowdown energy would communicate directly into the suppression pool inventoty. See LP85223, Figure 2 for an illustration of this physical anangemcnt.

The plant is operating at rated power, when the following occurs:

A PARTIAL loss of Drywell Cooling (VP) occurs Asaresult:

o Drywell Average Air Temperature rises and STABILIZES at 145.6F o Drywell-to-Primary Containment d/p rises and STABILIZES at +I.2 psid Which ONE of the following describes:

(1) the required action, (2) the POTENTIAL consequence of NOT taking that action?

and L

A.

B.

C.

D.

Answer: B (1) Restore the Drywell-to-Primary Containment d p to within its Tech Spec (2) Weir wall overflow, should an inadvertent upper pool dump occur.

(1) Restore the Drywell-to-Primary Containment d/p to within its Tech Spec (2) DIRECT communication of the blowdown energy contained in the drywell limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

airspace, to the suppression pool inventory, should a LOCA occur.

(1) Restore the Drywell Average Air Temperature to within its Tech Spec (2) Drywell temperatures in excess of the drywell STRUCTURAL design limits within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

temperature, should a LOCA occur.

(1) Restore the Drywell Average Air Temperature to within its Tech Spec (2) Drywell temperatures in excess of the drywell EQUIPMENT limits within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

QUALIFICATION temperatures, should a COMPLETE loss of VP occur.

I Question# 181 I

, References provided to examinee:

None CPS Tech Spec 3.6.5.4, Drywell Pressure (and its Bases)

CPS Tech Spec 3.6.5.5, l3ywell Average Air Temperature (and its Bases)

LP85223, Primary Containment

References:

RO/SRO 1

Tier:

I Group:

I KA:

I CFR I SROIR: I

~ o g ~ e v e l SRO I

1 I

2 I

295010AAZ.02 I 55.43(bH2) I 3.9 I

Higher System/Evolulion Name:

I c.txory:

High Dry well Pressure I E m ~ g ~ l ~ y and Abnormal Plant Evolutions KA Statement:

Ability to determine andlor intapret the following as they apply to HIGH DRYWELL PRESSURE: Drywell pressure A is incorrect - Part (I) is corm%, but Part (2) describes the consequence of too low a d/p (i.e,, below the lower LCO limit of 4. 2 psid). Refer to the same page B 3.6 - 122 discussion.

C and Dare incorrect - The stabilized drywell average air temp-ature of 1456°F is lower than the entry point for Tech Spec 3.6.5.5 (Le., 146.53F).

Objective:

I Question Source:

I Level of..

DiMculty:

DateWrltteu: I 03/31/05 I Author: I Ryder Comments:

Although Part (I) is arguably a requirement for both RO/SRO Candidates, Part (2) is not. Part (2) asks for the potential consequence of not restoring the LCO limits, which is only found in the Tech Spec Bases (as well as in the USAR).

Whats more, it is the Part (2) requirement that applies the KA statements ability to interpret portion. This question is in fact presented at an SRO-only level.

1 Question# [ 82 1 Refereom:

IT:

t Evolutions EP-AA-1003. Clinton Radioloeical Annex r:

KA statemcot:

Knowlcdgc of system status criteria which require the notification of plant personnel Using the provided references, answer the following.

With the plant in MODE 3, which ONE of the following, BY ITSELF, rquires NOTIFICATION of the Emergency Response Organization (ERO)?

A.

B.

C.

D.

Answer: c Water level in the LPCS Pump Room rises to 3 inches.

Suppression Pool Temperature rises to 112°F.

Containment Temperature rises to 188°F.

Radiation level in RHR Pump Room 'A' rises to 10 times normal.

ExpL.nitlon:

C is cmred - P a Clinton Radiological Annex EAL's, page CL 3-8, Fission Product Barrier Matrix #I (Containment),

and the FUI action level. Containment temperahue at 01 above 185'F is a 'Potential Loss Containment', requiring declaration of Unusual E m t. Per EP-AA-I 12-100-F-01,Section I.l.D, this EAL requires NOTIFICATION of ERO pcrsonncl (statim Management, only).

A is incomct - Per EAL page CL 3-13, EOP-8 Table W, and the HA4 action level. The given water level is not above the max safe value for that mom (i.e. 4 inches). Until it is, no E-Plan entry is required. The threshold for the p ~ r a m e t ~

is at the ALERT action level, rather than at the UE level.

B is incorroct-P s EAL page CL 3-8, Fission Product Barrier Matrix #3 (RCS). A Suppression Pool Temperature above I I O O F does D is incorrixt - PR EAL page CL 3-6, action level RU3. ?he threshold for this parameter is 1,000 times normal. An E-Plan entry is not yet requid Even if we wcre to considex the 'RA3' (Max Safe = 25 Rihr) threshold of Table R4.

What the Candidate is expected to resosize is that the only way that a '10 times normal' level could be synonymous with having rcsched 25 whr, would be for the 'narmsl' radiation level to 2.5 RAW. This conclusion would be implausible for any RHR Pump room.

by itselc require an E-Plan mhy. It would, if it were coincident with a stuck-open SRV.

objective:

I Quation Source:

1 Level of Difficulty:

None New I

2.5

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

~~

P-AA-I 12-100-F-01, Site E~&B~cY Director Checklist I

I Question # I83 I References provided to examinee:

References:

Which ONE of the following requires a NOTIFICATION (phone call) to the NRC (OTHER than to the on-site Resident)?

A.

B.

With the plant at rated power, the MCPR value is determined to be 1.lo.

Shift Manager discovers that one of the off-going control room operators exceeded the Technical Specification overtime guidelines WITHOUT a deviation being authorized.

None CPS Tech Spec SL 2.1.1, Reactor Core Safety Limits CPS Tech Spec 3.5.1, ECCS - Operating CPS Tech Spec 5.2.2, Unit Staff LS-AA-1020, Reportability Reference Manual 10CFRS0.72, ImmediateNotification Requirements for Operating Nuclear Power Reactors C.

D.

With the plant at rated power, HPCS has been INOPERABLE for 14 days.

Shift Manager discovers that the LPCI A quarterly surveillance, performed 30 days ago, was reviewed and signed off, but was INCOMPLETE.

Answer: c Explanation:

C is correct - Per Tech Spec 3.5. I, Conditions B and D. Initiation of a plant shutdown (to he in Mode 3 within I2 hours) is required. Per Exelon procedure, LS-AA-I 020, page 4, item F-aa, a 4-hour report is required for the initiation of a plant shutdown required by Tech Specs (IOCFRSO.72@)(2)(i)).

A is incorrect - PR Tech Spec SL 2. I. 1.2. MCPR must be at or above 1.09 (2-loop). Stem conditions indicate plant is at rated power (i.e., can only be in 2-loop). No SL has been violated.

B is incorrect - This choice refers to Tech Spec (Administrative Controls) 5.2.2.e. Per LS-AA-1020, page 9, item T-03, because this choice describes a Tech Spec violation that solely administrative in nature, no NRC reporting is required. 10CFRSO.72(a)(2)(i)(B) agrees with this.

D is incorrect - This choice suggests an application of Tech Spec SR 3.0.3. is in order for a now overdue (beyond 25% grace period) surveillance. Operators have 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to complete this surveillance and resolve this problem before having to enter any LCO. Meanwhile, no NRC notification is required, given this discovery alone.

I Objective:

I Question Source:

I Level of Difficulty:

None New I

2.0

I Question# I84 1 v

Using the provided references, answer the following.

The plant is in MODE 1, with the following:

The Standby Liquid Control System Operability surveillance, CPS 9015.01, has just been completed SLC Pump A flow rate ACTUAL VALUE is 41.3 gpm SLC Pump A Dp ACTUAL VALUE is 1260 psid Which ONE of the following:

(1) CORRECTLY INTERPRETS these surveillance results, (2) describes the required action?

and A.

(1) SLC Pump A Discharge Check Valve, lC41-F033A, is NOT opening FULLY.

(2) Take action to establish a 6-week test frequency for lC41-FO33A.

(1) A blockage exists somewhere in the SLC Pump A DISCHARGE.

(2) Take action to establish a 6-week test frequency for SLC Pump A.

(1) SLC Pump A Discharge Check Valve, lC41-F033A, is NOT opening (2) Enter Tech Spec 3.1.7 for SLC subsystem A.

(1) A blockage exists somewhere in the SLC Pump A SUCTION.

(2) Enter Tech Spec 3.1.7 for SLC subsystem A.

B.

C.

FULLY.

D.

Answer: B Explanation:

B is correct - Per CPS 9015.01DOO1, page 3, the SLC Pump flow rate (Qr) is in the ALERT Range, while the pump D/P (Dp) is much higher than normal (in fact, high outside of the Acceptable Range). Only a blockage somewhere in the pump discharge piping can yield this combination of low flow-high discharge pressure (and therefore, high Up).

Per CPS 901 5.01, Section 9.1.4, personnel a~ directed to double the test fresuency (from q~arterly, to 6 weeks) when the pump goes into the ALERT Range.

AandCareincol~ect-PerCPS9015.01DOOI,page3, solongasthepumpflowrateisatleast4l.2gpm, thedischarge L.

I Question# I 84 I Refereuees provided to examinee:

References:

check valve is expected to exercise (open), fully. There is no reason to interpret that a failing check valve is in any way responsible for the low flow-high pressure combination.

D is i n m m t - Although a blockage somewhere in the pump SUCTION piping may in fact yield this low flow-high pressure combination, there is no Teason to declare the SLC subsystem inoperable. That is, per CPS 9015.01 DOOl, page 3, although the 1260 psid actual value is high outside the Acceptable Range for Dp, there is no requirement to enter the Tech Spec.

CPS 9015.01, in its entirety CPS9015.01WOI,initsentirety CPS 9015.01, Standby Liquid Control System Operability CPS 9015.01DO01, SLC Pump & Valve Operability Data Sheet I

1 1 Level of Difnculty:

^.-

Ouestion Source:

Objective:

I Question # I 85 1 Average Power Range Monitorha1 Power Range Monitor Plant systems KA Statement:

Knowledge of the bases in technical specifications for limiting conditions for operations and safety limits

References:

Per Technical Specification (or ORM) Bases, which ONE of the following identifies an APRM related Instrumentation Function that IS SPECIFICALLY relied upon by an accident analysis?

Average Power Range Monitor...

A.

INOPTrip B.

MOP Rod Block C.

D.

Answer: D Neutron Flux - High, Setdown Fixed Neutron Flux - High CPS Tech Spec Bases B 3.3.1.1, RPS Instrumentation CPS Tech Spec Bascs B 2.0, Safety Limits Explanation:

D is correct - Per CPS Tech Spec Bases, pages B 3.3-9 and B 3.3-30a This Function is relied upon by the Control Rod Drop Accident analysis of USAR Section 15.4.9.

A and B are incorrect - Per CPS Tech Spec Bases, page B 3.3-6. This RPS Trip Function is not assumed in any safetylaccident analysis; rather, it is retained in Tech Specs be virtue of being part of the NRC-approved licensing basis. The MOP Rod Block is found in the Operating Requirements Manual (ORM), where Bases Section 5.2.1 refers back to the Control Rod Block Instrumentation Bases (of Tech Spec 3.3.2.1) and Power Distribution Limits Bases (of Tech Spec 3.2). A renew of these shows NO connection to any analysis that takes credit for the APRM MOP - Rod Block Function.

C is inco-t-Per CPS Tech Spec Bases, pages B 3.3-6 and 7. Although this Function indirectly protects Safety Limit(SL)2.1.1.1, thereisnoanalysisthattakesdirect/spcificcreditforthisFunctioo.

Objective:

I Question Source:

[ Level of DlMculty:

None New I

2.5 CPS Tech S&

Bases B 3.3.2.1, Control Rod Block Instrumentation CPS ORM, Section 2.2.1, APRM Control Rod Block Instrumentation CPS USAR, Section 15.4.9, Control Rod Drq, Accident (CRDA)

Date Written: I 04/04/05 I Author: I Ryda Commento: None I

I Question# I 86 I The plant is operating at rated power, during prolonged hot summer conditions, with the following:

Abnormally high CCW heat load conditions exist ALL available CCW Pumps and HXs are in service CCW HX Shell Side Outlet Temperature is now 106°F and STABLE Which ONE of the following describes:

(1) a consequence of allowing CCW HX Shell Side Outlet Temperature to remain at this temperature, and (2) an appropriate action?

A.

B.

C.

D.

Answer: A (1) Operating the CCW HX is excess of its DESIGN limit for Shell Side (2) Line up an FC Heat Exchanger with cooling supplied by SX.

(1) Operating with a CCW HX Shell Side Outlet Temperature TOO NEAR the temperature that will cause CCW Demineralizer resin damage.

(2) Bypass the CCW Demineralizer.

(I) UNACCEPTABLY high CCW supply temperature at the RWCU Pump (2) Secure all non-essential CCW heat loads.

(1) RISING radiation levels in the Fuel Building.

(2) Line up an FC Heat Exchanger with cooling supplied by SX.

Outlet Temperature.

seals.

Explmatioo:

A is correct - Per CPS 3203.01, Sections 6.1 and 8.3.1. The given Outlet Temperature is above the 105°F Design limit for the HXs. Section 8.3. I directs operators to place a second FC HX in senice, cooled by SX (Shutdown Service Water). Section 8.3. I.4(2)c direns operaton tn consider shifting all FC cwling over to SX (taking about 30% of the total CCW heat load off the CCW system, per Appendix B on page 70). Close smtiny by the CPS Facility Author and an SRO Validator (incumbent) has determined that the intent ofthe wording in CPS 3203.01, Section 6.1.2 is as phrased here in this answer choice.

B is incorrect - Per CPS 3203.01, Section 4.6, CCW demin resin damage isnt a concern until CCW temperature nears

1 Question # I 86 1 Rdereucea provided to examinee:

None Referencn:

CPS 3203.01, Component Cooling Water CPS 3303.01, RWCU CPS 3317.01, Fuel Pool Cooling and Cleanup CPS 5040-IC, High Temp CCW HX Outlet Temperature (alarm procedure)

C is income3 - There is pp CCW supply temperature that is defmed as unacceptably high for the RWCU Pump Per CPS 3303.01, Section 8.1.1.1 I, operators must only ensure that seal Disin--Thereisnoretaence,eitherin CPS 3203.01 orCPS 3317.01,thatassaciates~CCW Outlet Temperature with degraded Spent Fuel Storage Pool cwling.(FC) that result in elevated radiation levels in that area temperature is maintained helow 200°F.

I I

Objectfve:

I Quntion Source:

1 Level of Difficulty:

LP85208.1.14 New I

3.0 DmteWritten

1 05/16/05 I Author:

1 Ryder Comments:

The following justify this being an SRO only question:

1.

Although Part (1) of the question would apply to an RO exam, as well, in that knowledge of the CCW HX design tcmpaature limit has the procedures PmautiondLimitation section as its source, Pati (2) goes beyond this source, into the procedures Abnormal section.

Abnormals section 8.3.1, specifically, step (2)c suggests consideration.... Such direction is always reserved for the SRO, only. The fact that such procedural guidance exists only in an Operating Procedure (i.e., CPS has no Loss of CCW off-normal operating procedure) does not automatically disqualify this guidance; it is still Although tbe alarm response procedure (CPS 5040-1 C), usually considered to be the ROs first procedural line of defense, includes the Operator Actions to verify a properly operating temperature control valve, and to vent the CCW HXs, it docs the operating pmcedure

2.

to the job of the on-shifi SRO.

3.

address all of the actions found in Section 8.3.1 of I

] Question ## I 87 I c

A plant power ascension is in progress, per CPS 3004.01, Turbine Startup and Generator Synchronization, with the following:

Reactor Recirc Pump m)

A is about to be transferred to FAST speed MEDIATELY BEFORE the operator positions FCV A for the transfer, the following occurs:

o RRP B trips fiom SLOW speed to OFF o Operators immediately shut the RRP B Discharge Valve Which ONE of the following:

(1) PREDICTS the resulting TOTAL CORE FLOW, after flow stabilizes and BEFORE any additional operator action is taken, and (2) describes the NEXT required action?

A.

(1) About 25 mlbm/hr on the 65% FCL (2) Immediately scram the reactor, even if NO power oscillations are observed.

B.

(1) About 20 d b &

on the 60% FCL (2) Verify MFLCPR is at or below 0.970.

(1) About 25 mlbm/hr on the 55% FCL (2) Isolate RR Loop B using CPS 3302.01, Reactor Recirculation.

(1) About 20 mlb&

on the 50% FCL (2) Direct IMD to change the APRM setpoints to those for Single-Loop C.

D.

Operations.

Answer: D

[ Question# 187 1

~~

Erplana%on:

Dbcor~&-

Part (1): Refa to CPS 3004.01, Sections 8.4.5 and 8.4.6, and to CPS 3302.01, Sections 8.1.1 and 8.1.2. The following are the initial conditions BEFORE the hip of RRP B:

1) Both RRPs are in SLOW speed, with their FCVs about 90%

open; 2) Reactor power is about 3Wh with the rod line (FCL) at or near 50% FCL; 3) Thus, a review of the P/F Map indicates that Total Core Flow More the RRP B hip is about 38.0 mlbmhr. The following are the stable flow conditions AFTER the trip of RRP B and the immediate shutting of its discharge valve (Immediate Operator Action, p a CPS 4008.01, Section 3.2 1) RRP A remains unaffected, running in SLOW speed, with its FCV still at about 90% opn;

2) Total Core Flow simply migrates down the same 50% FCL (rod line) and stabilizes at ahuut a little more than half of the initial 38.0 mlbm/hr value (Le., about 20 mlbm/hr).

and Fluid Flow) of losing one of two, identical, parallelsonfigured pumps. NOTE.: A review of the PIF Map in this area shows that there is 11p need to consider the precise slope of the 50% FCL (i,e., a so-called flatter sloped FCL, due to fuel design). Whatever slope m a t i o n there might be in this area of the Map, we are still well below any point when we would expect to drift into the CONTROLLED ENTRY Region.

Put (2): A review of the CPS 4008.01 Subsequent Actions shows there is no specific action that is required early in the case of this specific swnario. From Section 4.4, operators are directed to Section 4.9 for SLO. Stem wnditions indicate that Sections 4.9. I, 4.9.2, and 4.9.3 are non-issues for this scenario. In Section 4.9.4, operators proceed to CPS 3005.01, to implement S M. Then. in Section 8.4.3, operators are directed to change the APRM Wpoints. This is the next action that is required given these stem conditions.

A Ir hcorrect - This choice is distracting to the SRO candidate who cannot recall (from memory, no reference is provided here) where in the plant power ascension (at what reactor power) we transfer the RRPs from Slow to Fast. If that Candidate incomctly mncluda that the transfer takes place at a power level CIOSR to 4550%. with a 65% rod line (FCL), then the p h i p total m flow is about 51 mlbm/hr, and the post-trip flow is about 25 m l b h r (i.e., a little more than half of the initial flow). Once this is determined, a migration down the 65% FCL, places the plant either firmly in, or too close to, the Restricted Zone (scram required). The Candidate is expected to recognize that, for these reasons, a pump upshift would a take place on the 65% FCL.

B is incorrect - This choice is distracting to the SRO Candidate who chooses the 60% FCL, with a power closer to 35% and a pwhip flow of about 38.0 mlbmihr. Ihis choice is very attractive when the SRO Candidate considers Part (2). Rers is clear indication, in CFS 4008.01, Section 4.2, that some priority should be given the verifying MFLCPR is at or Mow 0.970 for SLO. HOWR, the suggestion, in Part (I), that the RRP upshift will have been started from the 6Wh FCL makes this choice absolutely wnmg.

C h incorrect - Although a 55% FCL is feasible for pump upshift conditions, the resulting total core flow is higher than predicted (see explanation for correct answer, D).

In Part (Z), there is NO requirement to isolate the tripped pump loop. In fact, to do so would prohibit the restoration of that loop until reactor water temperature is < 200°F (see CPS 3302.01, Section 6.11.1).

This is the Generic Fundamental (Pumps

1 Question# 187 I Rtfertnca prodded to eraminet:

Rtftrtnca:

None CPS 3004.01, Turbine Startup and Generator Synchronization CPS 3005.01, Unit POWR Changa CPS 4008.01, Abnormal Reactor Coolant Flow CPS 3302.01, ~eaetor Recirculation (RR)

L.

Using the provided references, answer the following.

From rated power, operators have just manually scrammed the reactor, THEN the following occurs:

. Erphuatfon:

D is cor&- Per CPS EOP-6, Pool TempaarUre leg, topmost IF-THEN step. Pool temperature is very near point we the CRS would pmceod to stcp 19 of this leg, where qmators are directed to place all available suppression pool cooling in Service. Ofall the current plant conditions in the stem, this is the highest priority, currently.

A is income3 - Per EOP-IA, Lcvel leg, and Figure G. Only through implementing the IF-AND-AND-AND-THEN override is Level Band C required Stem conditions do not support the Suppression Pool temperature above Figure G portion of this oveSride (k, 110°F pool tem-).

B is incorrect - Per EOP-6, Pool Level leg, step 20, and Figun Q. Even though the stuck-open SRV is slowly adding to pool inventory, with level nmently 81 19 feet, and reactor pressure at 950 pig, we are still far below the Figure Q limit of -23 feet. This is mt the NEXT required action.

C is mcorrect - Per EOP-6, Pool Tnnperahm leg, bottom-most step. Although intentionally lowering pressure to stay below the Heat Cap8city Limit (Figure P) is a l l ~ ~ e d,

even during an ATWS, the current stem conditions of 950 psig and 93F pool tanpcrature, an still far below the HCL limit of -145F pool temperature. This is not the NEXT required action.

a Reactor Power is now 20%

a SLC Pumps are running a Reactor water level is +35 inches a Reactor Pressure is 950 psig and STABLE a

Suppression Pool Level is 19 feet and RISING SLOWLY a One SRV has stuck open a

Suppression Pool Temperature is 93°F and RISING From among the following actions, which ONE is required NEXT?

A.

B.

C.

D.

Terminate and prevent injection to establish water LEVEL BAND C.

Drain Suppression Pool inventory to stay below the SRV Tail Pipe Limit.

Open a Turbine Bypass Valve to stay below the Heat Capacity Limit.

Place all available Suppression Pool Cooling in service.

Answer: D

1 Question# I 88 1

References:

1:

t systems CPS EOP-I, ATWS RPV Contml CPS EOP-6, Primary Containment Control I Knowledge of symptom based EOP mitigation strategies I

Objective:

Qneation Source:

1 Level of Difficulty:

I None New I

2.5

I Question# 189 I Core Alterations are in progress, with the following:

A fuel bundle has been removed from the Upper Containment Fuel Storage Pool (UCP)

That same fuel bundle has just been placed in the core, BUT the grapple has NOT been released THEN, the Refuel SRO recognizes the fie1 bundle is NOT in its correct core location Which ONE of the following describes the NEXT required action?

A.

B.

C.

D.

Answer: B RELEASE the grapple and contact the Reactor Engineer for further guidance.

Remove the bundle ftom the core and return it to its CORRECT core location.

Remove the bundle from the core and return it to its UCP rack location.

Do NOT release the grapple and contact the Reactor Engineer for further guidance.

Exphnatlou:

B is correct - Refer to CPS 3703.01, Section 6.24.1. This question proposes that exact scenario.

A is i n c o w - This choice suggests a scenario as described in Section 6.24.2. It is mi.

C and D are incorrect - They both have strong face validity plausibility for this &&reference question.

t Objective:

I Question Source:

I Level of DiMculty:

None New I

2.0 References provided to examluee:

Reference%

1 None 1 CPS 3703.01, Core Alterations

~itewritten: I 05/03/05 I Author: I Ryder Comment% None

I Question# I 90 I L'

Considering the MINIMUM staffing requirement for either the SROs, or Fire Brigade Members, which ONE of the following describes a situation that SATISFIES the respective requirement?

A.

While in MODE 1, you have a total of TWO SROs, one of whom is the Shift Manager, and the other is BOTH the Control Room Supenisor AND designated STA.

While in MODE 3, you have a total of FOUR Fire Brigade Members, one of whom is also the designated Safe Shutdown Qualified Operator.

While in MODE 5, you have a total of THREE Fire Brigade Members, ALL of whom are 'C Area Qualified'.

While in MODE 4, you have total of TWO SROs, NEITHER of whom is qualified as B.

C.

D.

L Objective:

I Question Source:

I Level of Difficulty:

None New I

2.0 Answer

D Explanation:

Discorrect-PaOP-CL-101-102-1001, all. InMcdes4and5, only2 SROsarerequired(forER0functionsofCRS and SM), and no STA is required.

Aisincorrsn-PaOP-CLLOI-102-IO01,all.

InModes 1,2,and3,ifeithatheCRSorSMisalsotheSTA, thena 3" SRO is required and is designated as the Incident Assessor.

B is in~-PerOP-CLIOI-102-1001, all. In ALLModes, at least 4 FireBrigade&!&zs (plus thekader) are

required, C isinconrct-PerOP-CL101-102-1001,all.

InALLModes,at least4FireBrigadeMembers(plustheLeada) are required.

of whom can be the designated SSQO.

Dnte Written: I 04/08/05 I Author: I Ryder Commenh: None

1 Question# I91 I RO/SRO 1

Tler:

I Group:

I KA:

1 CFR I SROIR: I ~ o g ~ e v e l SRO I

3 I

Generics I

2.2.23 I 55.43(a) 1 3.8 I

Lower SystemlEvolutinn Name:

1 C.*ory:

I Equipment control L

KA Statement:

Ability to hack limiting conditions for operations IMD is about to commence a surveillance test, with the following:

The surveillance test will cause a TECH SPEC-REQUIRED plant instrument to be INOPERABLE for the duration of the test Performance of the surveillance test does NOT require an LCO ACTION entry Which ONE of the following describes a CRS required action, PRIOR to IMD beginning the surveillance test?

A.

Direct the RO to hang an Adverse Condition Monitoring Tag on the annunciator window associated with the instrument.

Direct IMD to hang an Equipment Status Tag (EST) on the instrument, and the RO to hang a Miniature EST in the control room.

Identify the Technical Specification required action in the event the instrument is still INOPERABLE when the Short Duration Time Clock (SDTC) expires.

Identify the Maximum Out of Service Time (MOST) for the instrument and direct IMD to notify the control room if the test is still in progress within 30 minutes of the MOST.

B.

C.

D.

Answer: c Explanation:

C isco~ect-PaOP-AA-108-104, Sections3.5 and4.7.1.2.andAttachment I. TheCRS identifiestheTech Spec action that I&C will be directed to take should the instrument not be OPERABLE at the end of the Short Duration Time Clock.

A is incorrect - For an ILT Candidate, this choice has enough face validity to make it plausible, especially considering this is a closed-reference question. There is Q such tag, although there is an Adverse Condition Monitoring and Contingency Planning program defined by OP-AA-IO8-Il1.

That pmgnun has ng connection to the situation described in the stem.

B is incorrect-IheEST(and Miniature EST) is aprncess governed by OP-AA-108-101, and isused totrack the status of equipment out of normal (position). This inoperable insmment does not constitute such a condition.

D is incorrect - The MOST concept is governed by OP-CL-101-302-1001 and is bounded by equipment conditions that result in a Tech Spee enhy. There is no Tech Spec enhy associated with this surveillance test.

Objective:

I Question Source:

I Level of Difficulty:

None New I

2.3

i Referenca provided to examinee:

References:

1 Question# I 91 I None OP-AA-108-104, Technical Specification Compliance OP-AA-108-101, Conhol of Equipment and System Status OP-CL-101-302-1001, ITS LCO/ORM OWODCM OR Evaluations I

Group:

I KA:

1 CFR I SROIR: I c o g ~ e v e l RO/SRO 1

Tier:

SRO 3

I Generics I

2.2.23 I 55.43(a)

I 3.8 Lower I

I

~~

SystemlEvdutIon Name:

I category:

I Equipment Conhol KA Statement:

Ability to track limiting conditions for options

I Question # I 92 1 RO/SRO: I Tier:

I Group:

I KA:

I CFR I SROIR: I Cog~evel SRO I

3 I

Generics I

2.2.29 I 55.43@)(7) I 3.8 I

L O W S SyateovEvolution Name:

I category:

L.

I Equipment Control KA Statement:

Knowledge of SRO fuel handling responsibilities L

With all fuel handling equipment operating normally, which ONE of the following REQUIRES the DIRECT SUPERVISION of a Refuel SRO?

Explanation:

B is comt - Per CPS 3703.02, Section 3.5, I" bullet Refuel SRO must supervise any irradiated fuel movement between the Fuel Building and Containment.

A is incorrect - PS CPS 3703.02, Section 3.5,2" bullet. A Reactor Engineer is allowed to supervise this transfer C is incorrect - PsCPS 3703.02, Section 3.5,3" bullet. This hansfa requires no supervision by a Refuel SRO, only authorization by Control Rmm Supervision (SMICRS) or Work Control (WCS).

D is incorrect - Other than the movements considered by CPS 3703.02, Section 3.5 (above), only CORE ALTERATIONS nquk the supervision of a Refuel SRO. This choice does mt describe a Core Alteration &e., there is no movement of fuel that involves the reactor vessel).

Transfer of...

A.

B.

C.

D.

NEW fuel from the Fuel Building to the Containment Refuel Floor.

IRRADIATED fuel from the Fuel Building to the Containment Refuel Floor.

NEW fuel from the New Fuel Storage Vault to the Fuel Building Transfer Pool.

IRRADIATED fuel from the Spent Fuel Storage Pool to the Fuel Building Transfer Pool.

Answer: B

I Question# I93 I I_

RO/SRO: I Tier:

I Group:

I KA:

I CFR I SROIR: I c o g ~ e v e l SRO I

3 I

Generics I

2.3.2 I 55.43@M5) I 2.9 I

Higher SystedEvolution Name:

I category:

I Radiologiml Control KA Statement:

Knowledge of the facility ALARA program Using the provided references, answer the following.

L The plant is operating at rated power, with the following:

0 An operator needs to enter a Locked High Radiation Area (LHRA) to verify a valve position The LAST KNOWN Dose Rate (DDE at 30 cm), AT RATED POWER, was 1,200 mrem/hr for this LHRA The need to enter this LHRA does NOT involve any emergency situation Which ONE of the following describes the MINIMUM radiological control REQUIREMENTS applicable to the operators entry into this LHRA?

Can enter...

A.

B.

ONLY IF accompanied by an RF Tech; an approved RWP is NOT required.

ALONE; however, an approved RWP IS REQUIRED, a current survey map is NOT required.

ALONE; however, BOTH an approved RWP REQUIRED.

C.

a current survey map ARE D.

ONLY IF accompanied by an Rp Tech; an approved RWP IS REQUIRED.

Answer: c Explanation:

C is correct - Per RP-AA-460, Section 4.7. I, an RP Tech can substitute for a current survey map. Per Section 4.4, RP procedures do require an approved R W for HRA and LHRA entries (other than for emergent entries).

A is incorrect - P a RP-AA-460, Section 4.4 (as described above), RP procedures require the RWP. However, this choice is very plausible to the Candidate who opts for applying the exemption of Tech Spec 5.7.4, without regard for the fact that tn enta without the R W would amount to NOT o p t i n g in accordance with plant radiation protection pmcedUreS...

B is incorrect - P a RP-AA-460, Section 4.7.1, accompaniment by an RP Tech is required if a current survey map is not available.

D is incorrect - For the m o n s associated with the correct answer, C.

References provlded to esrmiuee:

References:

CPS Technical Specification 5.7, in its entirety (2 pages)

RP-AA-460, in its entirety CPS Technical Specification 5.7, High Radiation Area RP-AA-460, Controls for High and Very High Radiation Areas

I Question# I 94 1 RO/SRO I Tier:

1 Group:

I KA:

1 CFR I SROIR: 1 c o g ~ e v e i SRO I

3 I

Generics 1

2.3.1 I 55.43@)(4) I 3.0 I

LOW-SyrtemlEvolutlou Naw:

I category:

I Radiological Control L

KA Statement:

Knowledge of IO CFR 20 and related facility radiation control requirements Consider the following related to one of your crews NLOs:

After discovering she is ALREADY 3 MONTHS PREGNANT, she formally submits a Declaration of Pregnancy, TODAY Exposure records reveal that she has received 100 mrem (DDE) in the LAST 3 MONTHS She IS choosing to abide by the work restrictions prescribed in a Dose Equivalent Reduction Action Plan Which ONE of the following identifies:

(1) when her work restrictions AUTOMATICALLY expire, (2) how many ADDITIONAL mrem (DDE) she (including the embryo/fetus) is allowed to and receive during the REMAINDER of her pregnancfl A.

(1) When she is no longer pregnant (2) 400 mrem (1) 12 months h m todays date B.

(2) 400 mrem C.

(1) 12 months h m todays date (2) 500 mrem (1) When she is no longer pregnant (2) 350 mrem D.

Answer: B Explanation:

B is eomct-PR RP-AA-270. Attachment 3, page I of I. Given that she hss d v e d a total DDE of only 100 mrem since becoming prcgnnut, she is limited to a total of 500 nuem DDE for the entire pregnancy, or an additional 400 nnem from the mnainds of her pregnancy. Unless she withdraws her declaration befomhand, this declaration and its work reshictions automatically expire 12 months from today.

1 A, C, and D arc incorrect - Each is a plausible mis-understanding, or mis-application, of the Attachment 3 requirements cited above.

I I

Obiective:

I Question Source:

[ Level of Difficulty:

None New I

4.0

L

-Date Written: I 04/11 /OS I Author: I Ryder Comments:

This question is categorized as Lower Cognitive (LCL) because it only requires the recall of two, independent pieces of information: 1) SO0 nuem for the entire pregnancy, and 2) 12 months for the automatic expiration of the declarations work restrictions. There is no IF-THEN relationship that necessarily exists W e e n the two parts of the question.

This is an SRO-only question because the Declaration is one that a Work Supervisor must review and approve. In fact, the Work Supervisor (CRS/SM in her case), is critical in stipulating the work restrictions for the Dose Equivalent

~eductioo Action pian.

I Question# I94 I RO/SRO:

I Tier:

I Group:

I KA.

I CFR I SROIR: I c o g ~ e v e l SRO 3

1 Generics I

2.3.1 I 55.43@)(4) I 3.0 Lower I

I System/Evolution Name:

I Category:

I Radiologicd Control KA Ststemenl:

Knowledge of 10 CFR 20 and related facility radiation control requirements

1 Question # 1 95 I RO/SRO I Tier:

I Group:

I KA:

I CFR 1 SROIR I CogLevel SRO I

3 I

Generics I

2.1.34 I 55.43@)(5) I 2.9 I

Higher SyatedEvolutiou Name:

I catepory:

I Conduct of operations KA Statement:

Ability to maintain primary and secondary plant chemistry within allowable limits Using the provided references, answer the following.

FOLLOWING a refueling outage, the plant entered MODE 2 in preparation for a plant startup 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ago, with the following:

Reactor water temperature is 2WF Reactor Power is at the Point of Adding Heat (POAH)

Chemistry makes the following sample data available to the CRS:

o Feedwater Conductivity is 0.30 pS/cm o Feedwater Oxygen is 280 ppb o Reactor Coolant Conductivity is 0.50 pS/cm o Reactor Coolant Chlorides is 150 ppb Continuous Conductivity and Oxygen monitors AGREE with the above sample data L

i-Which ONE of the following describes the NEXT required action?

A.

Direct Chemistry to obtain and analyze a confirmation sample of REACTOR COOLANT within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

B.

Direct Chemistry to obtain and analyze a confirmation sample of FEEDWATER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Suspend control rod withdrawals in preparation for returning to MODE 3.

Notify the Nuclear Operations Duty Officer of an Action Level 2 condition.

C.

D.

Answer: D Explanation:

D is correct - Per CY-AB-120-100, Section 4.3.2. For these plant conditions (POAH, Mode Z), Reactor Cwlant Chlorides are higher than the Action Level 2 limit (100 ppb). Per Attachment 2, the & requirement is to notify the Nuclear Operations Duty officer of the Action Level 2 condition.

A is incorrect - This choice has face validity @sychometrically balanced with choice B), and is intended for the Candidate who incorrectly applies Note a of CY-AB-120-100, Section 4.3.2.1 Table.

B is incorrect - 7his choice is plausible to the Candidate who recognizes Feedwater Conductivity is higher than the Action Level 1 limit ofCY-AB-I2O-l IO, Table la, and then applies the Action b e l I decision tree of Attachment I.

I Question# I95 1

..~...

RO/SRO I Tier:

I Group:

I KA:

I CFR I S R O I R I CogLevel SRO I

3 I

Generics I

2.1.34 I 55.43(bH5) I 2.9 I

Higher SystemiEvolution Name:

I category:

I Conduct of opaah ons KA Statement:

Ability to maintain primary and secondary plant chemistry within allowable limits However, this would be a mistake in the light of Note a associated With the Table I a limits. With the given stem conditions, the steam jet air ejectors cannot be in service @laced in service at or above 150 psig); reactor power is only at the pint of adding heat (about IRM Range 6 or 7).

C is incorrect - One of the open-references provided to the Candidate for this question is the ORM section 2.3. I for Reactor Coolant Chemishv. Table 3.3.1-1 shows that the given 150 ppb value for reactor coolant chlorides is above the References provided to examinee:

References:

0.1 ppm (Le.. 100 pph) limit for Modes 2 and 3. The Candidate may&

to OR Action 3.3.1 2, hut neglect IO nntice that the action to return 10 Mode 3 within I2 hours is not required until after this chlondc value has k n

excccded for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months CPS Operating Requirements ( O W ) 2.3.1, entire section CY-AB-120-100, in its entirety CY-AB-I20-110, in its entirety CPS OR 2.3.1, Reactor Coolant System Chemistry CY-AB-120-100, Reactor Water Chemistry CY-AB-120-1 IO, Condensate and Feedwater Chemistry L

NOTE - The Feedwater Oxygen sample value given in the stem is there for psychometric balance between Feedwater and Reactor Coolant. The stem statement regarding...8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ago is there to ensure any consideration of applying Note c ofCY-AB-120-100, Section 4.3.2.1 Table, is avoided.

Date Written:

04/29/05 I Author: I Ryder Comments: None I

i RO/SRO I Tier:

I Group:

I KA:

I CFR I SROIR: I ~ o ~ ~ e v e l SRO I

3 I

Generics 1

2.4.41 1 55.43(bX5) I 4.1 I

Higher SystemlEvolutioo Name:

I category:

1 Emagency Procedures and Plan

-KA Statement:

Knowledge of the emergency action level thresholds and classifications i

L The plant is operating at rated power, when the following occurs:

At Time = 0 minutes, ALL annunciators on P877 are lost due to a blown power supply At Time = +20 minutes, an UNISOLABLE primary system discharge causes operators to enter EOP-8 because an Area Temperature has JUST REACHED its EOP-8 entry value At Time = +55 minutes, as directed by EOP-8, operators perform an RFV Blowdown Which ONE of the following identifies the LATEST time:

(1) by when the FIRST required StatdLocal agency NOTIFICATION must be completed, (2) by when the event MUST be ESCALATED to the HIGHEST Classification Level and necessary for these plant conditions?

A.

(1) Time = +30 minutes (2) Time = +35 minutes (1) Time = +45 minutes (2) Time = +40 minutes (1) Time = +50 minutes (2) Time = +70 minutes (1) Time = +85 minutes (2) Time = +70 minutes B.

C.

D.

Answer: c C is correct-Pan (I): T h e w that an EAL threshold io CPS AnnexpageCL3-11). P~EP-AA-I12-IW),Scction2.l,thcShifl Managa(SM) would haveuntilTime.-30 minutes to classify/dalarc the event as a UE, and until Time = -45 minutes to complete the required Statw Local notifications. However, at Time = t2O minuts, the FAI EAL threshold is reached due to a Potmtial Loss uf RCS (see Annex page CL 3-8). Again, the SM would have until Time = -35 minutcs (20 - 15 = 35) to classify/declare the evmtaqanALERT. PaEP-AA-Ill.

Section4.1.thc2dNOTE,oncethishighaclassification level i s d e c l d. ifthe UE notification has not yd b madc. the UE event is essmtially dismissed (without funha consideration). in fhor of is at Time = -15 minutcc. for EAL MU6 (soe

1 Question # I 96 1 the more sevae ALERT event declaration. In other words, given these stem conditions, the UE event (loss of annunciators) does m l t in a First required StsrJLOcal agency notification. Rather, the SM has until Time = +50 minutea to complete the ALERT notifications. And since the next plant transient that requires a re-classification (escalation) tn an SAE (i.s, the RPV Blowdown) docsnt even occur until Time = t55 minutes, the SM does in fact get a chance to campleh the ALERT notifications at Tune = +50 minutes. nis, therefore, amounts to the First required StatJLOcal agmcynotificatim forthesegiven plant conditions. Part (2): An SAE is the highest classification required for the plant conditions (is, the FSI EAL is reached due to Loss of Containment; see Annex page CL 3-8). Again, per EP-AA-I 12-100, W o n 2.1, the SM must &lare this escalation (from an ALERT) no later than Time = +70 minutes (+55 + 15 minutes = t70 minutes).

A is incomd - F a the reasons already d d b e d above. Pan (I) is plausible to the Candidate who disregards the EP-M-111, Sectian4.1,rcquimncnts, andmi&enly applies a+30minuterequirement (+I5 + 15 = t30 minute) of EP-AA-I 12-100, Section 2.1, to the atad ofthe th@&j clock for MU6. Part (2) is plausible to the Candidate who recognizes the need to d a t e to an ALERT by no later than Time = +35 minutes (FA1 threshold at Time = +20, +I5 minutes to classify, pcr EP-AA-I 12-100, Scction 2.1). ?his Candidate does recognize that the RPV Blowdown at Time=+55minutesmltsinafurtherescalatimtoanSAE(FSlEAL).

B is in-

- For the reasons ekrady described above. Part (I) is plausible tn the Candidate who, although correctly waits for the Mu6 tluesbold clock to become aaive (Le., the threshold is met) bdore applying the +30 minute allowanceofEP-AA-I12-100, Section2.1, failstoapplythehe-AA-Ill, Section4.1 requimentthutessentially dismisses the MU6 evmt Part (2) is deigned to pruvide psychnmettic balance with Part (2) of choice D (i.e., a time d n e that is &than its associated Part (1) value). It has sufficient face validity for the thoroughly confused candidate, as well.

D is inc4n?ect - For the reasons already described above. This choice (both Parts) is plausible to the Candidate who m o t efktively translate the &

of the EOP-8 actions identified in the stem conditions, and instead simply applies the final state of the plant (RPV Blowdown is props) and concludes that EAL FSI applies. This Candidate will nexssarily rsognize that the SM has 15 minutes to classify the SAE (i.e., Time = +55 minutes + 15 minutes = t70 minutes), yielding Part (2) ofthe answer choice, Similarly, the SM has an additional 15 minutes, from Time = +n?

minutes, to complete the Statubcal notifications (Time = +70 + 15 minutes = +85 minutes), yielding Part (I) of the m e r choice.

Reference8 p d d d to exmioee:

Reierenca:

EP-AA-1003, Clinton Radiological Annex, pages CL 3-6 thru 3-1 3 EOP flowcharts EP-AA-1003, Clinton Radiological Annex EP-AA-I 12-100, Conbol Room opaations EP-AA-I 11, Emagency Classification and PARs CPS EOP-8, Seumdary Containment Contml

I Question# I 97 I RO/SRO 1

Tier:

I Group:

I KA:

I CFR I SROIR: I CogLevel SRO I

I I

I 1

295031 2.120 1 55.43@)(5) 1 4.2 I

Higher Sy8tsmiEvolution Name:

I catqory:

1 Plant systems L

Reactor Low Water Level KA Statement:

Ability to execute procedure stew Using the provided references, answer the following.

L An ATWS and LOCA are in progress, with the following:

Reactor pressure is 600 psig and slowly lowering Operators are injecting with ALL available PREFERRED ATWS Systems Reactor water level is -149 inches Wide Range and slowly lowering Containment Temperature is 175°F and slowly rising Containment Pressure is 2.0 psig and slowly rising Suppression Pool Level is 19 feet, 5 inches, and slowly rising Which ONE of the following describes the NEXT required action?

A.

Start Containment Sprays.

B.

C.

Leave Level and Pressure; enter EOP-2.

Leave Level and Pressure: enter EOP-3.

D.

Implement the actions of CPS 441 1.05 for rising pool level.

Answer: c ExDhation:

C is Mmct - Refer to EOP-IA, Level leg, and Detail C. As soon RPV level drops to -150 Wide Range, this instrument becomes unusable, with a containment temperahre above 100°F (175°F is indicated in the stem conditions).

Opwtors must immediately transition to the Fuel Zone Range insmments. Because reactor pressure is about 600 psig (still far above the depmmmed (0 psig) calibration conditions for the Fuel Zone instruments), and Wide Range instruments are reading essentially the same as mal level before they become unusable, the Fuel Zone will indicate well below TAF (-162) when the operators opmtionally transition to them. As such the CRS has no choice but to implement the bottom-most step of the EOP-IA Level leg. The NEXT action is to Leave Level and Pressure, and enter EOP-3 to Blow Dorm.

A is incorrect - P a EOP-6, Containment Tempaahue leg, and Figure 0. The existing Containment Temperature (175F)Kontainment Prcssurc (2.0 psig) combination has us on the bad side of the Containment Spray Initiation Limit C U N ~,

Figure 0. Until things change (likely to be an additional rise in Containment Pressure), it is mX OK to Spray. This choice is B is incorrect - This would be me NEXT action if there wae 9p usable RPV water level instruments (i.e., no available Fuel Zone instruments) when the Wide Range instruments dropped below the minimum usable level of Detail C. The CRS would invoke the *most override step of the EOP-IA Level leg, by declaring RPV water level unknown and transitioning to EOP-2 for RPV flooding. Because there are 9p indications in the given stem conditions to cause the Candidate to conclude that the Fuel Zones are not available, this choice is nc!t the NEXT required action.

the NFXT required action.

1 Question# I97 I RO/SRO I

Tler:

I Group:

1 KA:

I CFR I SROIR: I CogLevel SRO I

1 I

I I

295031 2.1.20 I 55.43@)(5) I 4.2 I

Higher SystemlEvolutioo Name:

1 category:

Reactor Low Water Level KA Statement:

Ability to execute procedure steps I Plant systems D is incomt - This choice suggests the need to give a priority to the slowly rising suppression pool level (per the Pool Level leg of EOP-6). This procedure (CPS 441 1.OS) is associated not only with a pool level high enough to threaten the SRV Tail Pipe Limit ofFigure Q, but in fact provides the actions necessary to protect in-Containment equipment in the event that such equipment becomes submerged. Stem conditions suggest a pool level that is no where near this high level.

Objective:

I Queation Source:

I Level of Difficulty:

New I

2.0 Date Written: I 05/03/05 1 Author: I Ryder Comments: None

I Question # I 98 I RO/SRO 1

Tler:

I Group:

I KA:

I CFR I SROIR: I c o g ~ e v e l SRO I

I I

1 I

295003AA2.04 I 55.43@)(5) 1 3.7 I

Higher systemlEvdunon Name:

I category:

L Pnrtial OT Complete Loss of kC. Power KA Statemeoh Ability to determine a n d h interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER System lineups I Emergency and Abnormal Plant Evolutions Using the provided references, answer the following.

L L

With the plant operating at rated power, a COMPLETE LOSS of AC Power (including Div 3) occured 15 MINUTES AGO, and is still in progress, with the following:

RCIC is being used to control reactor water level at about +35 inches SRVs are being used to control reactor pressure between 800 and 1065 psig THEN, power is returned to the station via the RAT Which ONE of the following identifies the AC buses that should be re-energized FIRST?

A.

Divl B.

Div2 C.

Div3 D.

BOP Answer: B Explanation:

B is correct - Refer to CPS 4200.01, Section 4.2.3 NOTE. This question requires the SRO Candidate to consider the overall existing plant status in light of the sustained ability of RCIC to maintain RFV inventory, and the impact of the SRVs adding energy (heat) to the Suppression Pool. All Divisional batteries (including the Div 1 battery for RClC support) are 4-hour batteries (see USAR, Section 8.3.2.1 2.1). With RCIC ndqately controlling level, with there being plenty of decay hcat and, therefore, steam prepsure to support RCIC, and with RClC having been on its battery for only 15 minutes, now, there is &Q urgent need to restore power to the Div I battery charger. Before the loss of AC occurred, suppression pool level is understood to have been within Tech Spec limits (at least 19 feet, per Tech Spec 3.6.2.2). and suppression pool tempemture significantly below its Tech Spec limit of 95F, per Tech Spec 3.6.2.1.

ref^ to Figure P, Heat Capaeity Limit, of EOP-6. With post-scram reactor pressure between 800 and I065 psig, the pools Heat Capacity Limit is no where near being thmatened (happens as pool tempemture nears -14YF).

Therefore, there is urgent need to restore any of the Divisional (I, 2,3,4) power necessary to re-establish the main condenser as the preferred heat sink With RClC capably controlling level (as already described), there is no urgent need to restore Div 3 power to enable HPCS ns an alternate injection source. With the RCIC/SRV feedibleed combination providing adequate core cooling and there being no urgent need to restore the main condenser, or a vacuum, there is no immediate need for restoring Reactor Recirc, CCW, Plant Chilled Water, Plant Service Water, Plant Air, Condensate, or a CRD Pump. Therefore, there is n~ urgent need to restore non-divisional (BOP) bus power. In the end, the SRO Candidate should recognize that restoring Div 2 bus p o

~

is the ix&e&t priority, because it reenergizes the plant security systems, allowing card-reader access throughout the plant to support system restorations, and because it re energizes 4160V Bus IBI, which supports the main turbine turning gear (which otherwise has to be manually jacked to prevent post-shutdown rotor bowing; see CPS 4200.01, Section 4.2.1).

1 Question# I98 I References provided to examinee:

References:

RO/SRO I Tier:

I Group:

I KA.

1 CFR I SROIR I Cog~evel SRO I

I I

1 I

295003AA2.04 I 55.43@)(5) I 3.7 I

Higher SystemlEvnlution Name:

I Category:

Partial or Complete Loss of AC. Power KA Statement:

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER: System lineups I Emergency and Abnormal Plant Evolutions CPS 4200.01, m:

pages 3,4,14-20, and &I of the Appendices (A, B, and C); double jeopardy wncerns CPS 4200.01, Loss of AC Power US= Section 8.3.2.1.2.1, Batteries CPS Tech Spec 3.6.2.1, Suppression Pool Average Temperature CPS Tech Spec 3.6.2.2, Suppression Pool Water Level A, C, and D are incorrect - For the reasons described above.

I Question# I99 I L

Using the provided references, answer the following An ATWS is in progress, with the following:

Operators are controlling RPV level using Level Band C THEN, a LOCA occurs Operators enter EOP-3 and open 7 ADS Valves Hydrogen Igniters have tripped OFF (cause unknown) and are still OFF NOW, the MCR Hydrogen Monitors have been warmed up and are JUST beginning to come on-scale THEN, BOTH MCR Hydrogen Monitors STOP working (fail downscale, cause unknown)

Which ONE of the following describes the NEXT required action?

A.

Prevent Igniter restart.

B.

C.

D.

Answer: A Attempt to re-start the Igniters.

Direct Chemistry to sample for hydrogen.

Slowly inject with Preferred Systems to re-establish Level Band C.

Explanation:

A is wrrect - Refer to EOP-7, topmost override step, or to the EOP-6, top-most override step. The SRO Candidate may believe that when the hydrogen monitors stop working, the NEXT action is to sample...for hydrogen per 4412.OOCOOl. However, the Candidate is expected to recognize that a Blowdown having been directed by EOP-IA means that RPV level dropped below TAF (see the bottommost step of EOP-IA Level leg). How should the SRO Candidate interpret this with respect to actual hydrogen levels, despite the lack of functional monitors? Refer to the EOP Technical Bases for EOP-7, page 9-6. This discussion proposes that the CRS should use judgment based on plant wnditions to determine what actual hydrogen levels must be. Here, it suggests that with RPV level below TAF, hydrogen production should be suspected, and so, with the absence of monitors, the CRS should consider that actual levels do exist, and therefore require mitigation by EOP-7 actions. Refer to CPS 4412.OOCOO1, page 2 of 2, Section 3.1. Because of the removal of the PASS panel HZ102 monitom at CPS, this section instructs operators to wnsider hydrogen and oxygen levels to be unknown, if both MCR Monitors stop working. In either case, the NEXT action, therefore, is to prevent igniter restart per the topmost override of EOP-7. Stem wnditions & indicate that the igniters are currently off B is incorrect - For the reasons described above.

1 Question# I99 1 References provided to examinee:

References:

RO/SRO I Tier:

I Group:

I KA:

I CFR I SROIR: I

~ o g ~ e v e l SRO I

1 I

2 I

50oooOEA2.03 I 55.43@)(5) I 3.8 I

Higher EOP flowchmts CPS EOP-IA, ATWS RPV Control CPS EOP-3, RPV Blowdown CPS EOP-6, Primary Containment Control CPS EOP-7, Hydrogen Control CPS 4412.00COO1, Sampling Containment and Drywell for Hydrogen CPS EOP Technical Basis document SystemlEvolution Name:

I Category:

High Conlainmat Hydrogen Concentration I Emergency and Abnormal Plant Evolutions KA Stntemeut:

Ability to determine and/or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN

~~

~

I CONCENTRATION: Combustible limits for &ell C is incorrect - This is the most likely choice for the uncertain Candidate. If the Candidate either, fails to consider the below TAF...hydrogen production... levels unknown concept presented in the EOP Bases, or fails to recall the explicit requirements of CPS 4432.OOCOO1, Section 3.1, hdshe will likely apply the topmost IF-THEN action of the EOP-7 ovaride. It is important to understand why this choice docs understandinglapplication of the EOP-7 Bases on page 9-6 is && that is required for the CRS to determine that the NEXT required action is to prevent igniter restart, having declared the hydrogen levels to be unknown For this choice to be correct, would mean that Q&

by first going to the 4412.OOCOOl procedure could the CRS declare the hydrogen levels to be unknown...this ism me. Whats more, the choices wording suggests that the NEXT action would he to wait for Chemistrys actual sample readily available with the absence of the PASS monitors). In the meantime, if the igniters were to be started with a high hydrogen concentration present, a combustible situation could result.

==

Conclusion:==

this choice does not represent the NEXT required action.

D is incorrect - Once 7 A D S Valves are open in EOP-3, that EOP directs operators to rehun to the Level leg of EOP-IA. However, upon rehlrning to EOP-I A, at step 7, operators must wait until the RPV has depressurized below I38 psig, per Table J, represent a 2 cornst answ er...An attempting to re-establish Level Band C.

Objective:

I Oueation Source:

I Level of DiMculty:

None New I

3.0

I Question # 1 100 1 L

A plant pressurization is in progress, with REACTOR PRESSURE AT 750 PSIG, when the following occurs:

ONE SRV inadvertently opens and is STUCK OPEN The Immediate Operator Actions for CPS 4009.01, Inadvertent Opening SafetyhZelief Valve, HAVE been performed The SRV is still STUCK OPEN L

Which ONE of the following:

(1) identifies an expected tailpipe temperature, for the STUCK OPEN SRV, on the temperature recorder at P614, and (2) describes the required operator action?

A.

(1)27O"F (2) Attempt to shut the SRV by pulling its solenoid fuses; if the SRV remains open, shift CNMT HVAC to CCP Filtered Mode, then place the Mode Switch in SHUTDOWN.

B.

(1)310"F (2) Attempt to shut the SRV by pulling its solenoid fuses, AND enter CPS 4005.01, Loss of Feedwater Heating.

C.

(1)380°F (2) Attempt to shut the SRV by pulling its solenoid fuses; if the SRV remains open, shift CNMT HVAC to CCP Filtered Mode, then place the Mode Switch in SHUTDOWN.

D.

(1)400°F (2) Attempt to shut the SRV by pulling its solenoid fuses, AND enter CPS 4005.01, Loss of Feedwater Heating.

Answer: c L

Explnnntlon:

C is correct - Per CPS 4009.01,Section I. I, tailpipe temperature is expected to be >375"F for a stuck open SRV (Note:

this value is derived from empirical CPS test data, and is mgnized as being inconsistent with the predicted temperature derived from the Steam TablesiMollier Diagram.). P a Section 4.3, operators will attempt to shut the SRV

i-

. References provided to examinee:

None CPS 4005.01, Loss of Feedwater Heating CPS 4009.01, Inadvertent Opening SafetylRelief Valve CPS 3005.01, Unit Powerchanges CPS 3002.01, Heatup and Pressurization References I Question# I 100 I ROISRO:

I Tier:

I Group:

I KA:

I CFR I SROIR: I CogLevel SRO I

2 I

I 1

239002A2.03 1 55.43(bH5) I 4.2 I

Higher SystemlEvduHou Name:

I category:

RelieflSafety Valves I Plant systems KA Statement:

Ability to (a) predict the impacts of the following on the REUEF/SAFETY VALVES; and (b) based on those predictions, use procedures to umcct, control, or mitigate the wnequences of those abnormal conditions or operations: Stuck open SRV hy pulling its solenoid fuses. Per Section 4.7, if it mains open, operators are directed to shift CNMT Bldg HVAC, then perform aRapidPlant Shutdown IAW CPS 3005.01. PRCPS 3005.01, Senion 8.5.1, aRapidPlant Shutdown looks like the following: 1) Evacuate Containment ( a l d y done ns part of the Immediate Operator Action 3.1, of CPS 4009.01 1; 2) Iowa core flow to 43 mlbm/hr with Recim (because the stem conditions indicate the plant is only in MODE 2 (pressurization in ptugms, at 750 psig), both RR Pumps are still running in SLOW speed and total core flow is already -43 mlbmmr; therefore, this action is allready done; and 3) place the Mode Switch in SHUTDOWN; this is the d y action remaining to complete a Rapid Plant Shutdown, given these stem conditions.

A is i n c o w - Per CPS 4009.01, Section 1.I, this is an expected temperature for a leaking SRV, an open SRV.

B is incorrect - PR CPS 4009.01. Section 1.1, this is the Hi-Hi Temperahue alarm setpoint, but is still far below the range (>375T) expected f a an open SRV. Additionally, even though a stuck open SRV is an entry condition (Symptom) for CPS 4005.01, Loss of Feedwater Heating, there is NO entry required given these stem conditions (i,e.,

at 750 psig, m MODE 2, the plant is well below the 21.6% reactor power applicability threshold for the Loss of Feedwater Heating off-normal (sa CPS 4005.01, NOTE at the top of page 2 of 8).

D is incorrect - Even though 400°F may be an expected tailpipe temperature for a stuck open SRV @e., >375F), the second pan of this choice is mwrrect for the same reason attributed to choice B.

t Objective:

I Quertlou Source:

I Level of DiMeulty:

DB400901.I.3.1 New I

2.5 L-

ML052430760

Text

References:

References:

References:

References:

References:

References:

3.0 DateWlitten

References:

References:

2.6 Datewritten

References:

3.5 DatcWrilten

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

Reference:

References:

References:

3.7 DateWritten

References:

References:

References:

References:

References:

References:

References:

References:

3.0 Datewritten

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

3.0 DmteWritten

References:

2.0 Answer

References:

References:

References:

References:

References:

Navigation menu

Search