ML052430761

Final Ro/Sro Written Examination for the Clinton Initial Examination - July 2005

ML052430761
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: ML052430761 (165)

Text

FINAL ROlSRO WRITTEN EXAMINATION FOR THE CLINTON INITIAL EXAMINATION -JULY 2005

I Question# I 1 1 RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I Cog Level Both I 2 I I I 203000K202 I 2.5 I 2.1 I Lower System/Evolution Name: I Category:

RHR/LPCI: Injection Mode I Plant Systems KA Statement:

Knowledge of electrical power supplies to the following: Valves Which ONE of the following describes the impact of a loss of 480V Unit Sub IB?

A. IE12-F042C, LPCI From RH C Shutoff Valve, will NOT open electrically.

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

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

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

Answer: A Explanation:

A is correct - Per LP85205, Attachment D, the power supply to this MOV is AB MCC IB4. Per CPS 3514.01C006.

Section 2 . I . I , the loss of 480V Unit Sub 1B results in the loss of all of the listed MCC's. including AB MCC I B4.

Without 480VAC motor power. the F042C valve will not open electrically.

B is incorrect - Per LP85434. Figure 7. only a u 4 8 0 V bus s u p p l i e s m the normal power to the inverter's rectifier section. the alternate (backup) power through the bypass transformer. Additionally. this RPS Solenoid Bus B uses =-vital power; 480 V Unit Sub IB is power.

C is incorrect - Per LP85204, page 35, a SLC Pump 'A' start signal closes the lC33-FO04 valve (RWCU Suction Outboard Isolation). while a SLC Pump 'B' start closes the IG33-FO01 valve (RWCU Suction Inboard Isolation). Per CPS 3514.01CW6. Appendix A. page 46. a loss of480V Unit Sub I B disables IG33-F001 (Suction Inboard) and IG33-FO40 (Return Inboard). When operators initiate SLC (start both pumps), the 'A' SLC Pump start will still close IG33-FW4. although the 'B' SLC Pump start will NOT close IG33-FOO1. Because FW4 automatically close. the

'suction side' of RWCU &e automatically isolate.

D is incorrect - Per CPS 35 14.01COO6.Appendix A. page 41, a loss of 480V Unit Sub IB m&& 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 (because of the failed-high PRMs. IRIX-PROWB/D).

I Ohjective:

LP85205.1.12 I Question Source:

New I

I Level of Difficulty:

3.1 References provided to examinee: I None

References:

I LP85205. Residual Heat Removal (RHR)

LP85204, Reactor Water Cleanup (RWCU)

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

CPS 3514.01COO6,4160V Bus IB1 Outage Date Written: I 05/16/05 I Author: I Ryder Comments: Nunc

I Question# I 2 1 RO/SRO I Tier: I (;roup: I KA: I ROIR I SROIR I Cog Level Both I 2 I 1 1 209001 m.01 I 3.0 I 3.1 I Lower SysteWEvolution Name: I Category:

Low Pressure Core Spray System I plant Systems KA Statement:

Knowledge of electrical power supplies to the following: Pump power Which ONE of the following describes the impact of a loss of DC MCC lA?

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

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

C. A running SF Pump will automatically trip IF the Suction Outboard Isolation Valve, ISF004, closes.

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

Answer: D Explanation:

D is correct - Per CPS 35 14.01CO40. Appendix A, page 31, loss of this bus results in the loss of LPCS Pump hreaker control power. Since this pump has no DC control power-dependent starting interlocks, operators c w still start the pump by locally closing the pump motor breaker at 4160V Bus I A l A is incorrect - Per CPS 35 14.01C040. Appendix A. page 31, & I control power for this diesel is lost. both local and remote.

B is incorrect- Per CPS 3S14.01C040, Appendix A, page 32, the 3A breaker for RR Pump 'A' loses its breaker logic cnntrol power. Per CPS S003-5F. the SA breaker will trip if the 3A breaker loses a control power source. The SA breaker is closed ~Y

!I when the RR Pump I S running in FAST speed.

C is incorrect - Per CPS 35 14.01C040. Appendix A, page 32, a running SF Pump will NOT trip (trip interlock is disabled) if the suction valve, ISF004. leaves it open seat.

Obiective: I Question Source: I Level of Difficulty:

LP85209.1.12.3 New I 3.2 References provided to examinee: None

References:

LP85209, Low Pressure.Core Spray, LPCS CPS 3514.01C040, 125 VDC MCC I A Div I Outage CPS S003-5F, RPT BKR LOSS OF DC CONT PWR Date Written: I 02/04/05 I Author: I Ryder Comments: None

References:

LP65380, High Pressure Core Spray. HPCS Date Written: I 6/ 16/05 I Author: I GDSetser Comments: None

1 Question# 1 4 ]

RO/SRO:

n^.L DULII I

I I

SystenVEvolution Name:

Tier:

I I

I Group:

I I KA:

o,,-,,*nq

'li-hl.",

I Cat

,I I

ROIR 2.5 I

I SROIR 2.6 1

I Cog Level Lower I Plant Systems

~ ~~

Standby Liquid Control System A Statement:

nowledge of the physical connections andlor cause-effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following: Plant air 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. Bum out of the SLC Storage Tank OPERATING Heater B. HIGHER Boron concentration in the SLC Storage Tank C. SLC Storage Tank overflow through the top vent D. LOWER Boron concentration in the SLC Storage Tank Answer: B Explanation:

B is correct - This question is written directly from CPS LER 2004-002-00 (see attached references). The continuous sparge 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 correct answer. Per LP8521 I. page 20, if tank level lowers to <l.oOO gallons remaining, the Operating Heater could be damaged due to being uncovered. However, since normal SLC tank level is about 4.000 gallons (LP85211, page 6 ) .uncovering the heater would NOT he the 'earliest' potential impact.

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

Objective: I Question Source: [ Level of Difficulty:

LP85211.1.12.2 New I 2.3 References provided to examinee: None

References:

LP8521 I ,Standby Liquid Control (SLC)

CPS LER 2004-002-00, Mispositioned SLC Air Sparge Valve Results in High Boron Concentration

I Question # I 5 1 RWSRO: I Tier: 1 Group: I KA: I ROIR I SROIK. I Cog Level Both I 2 I I I 212000A1.07 I 34 I 34 I Lower KA Statement:

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR PROTECTION 1 SYSTEM controls including: Rod position information I 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. Confirm the ROD FULL IN light is lit, for that rod, at either RACC panel B. With the Full Core Display in DUAL mode, confirm that EITHER channel's Full-In LED is lit & the numerical display is Blank (NO numerical value).

C. With the Full Core Display in 2 CHANNEL mode, confirm that BOTH channels' Full-In LEDs are lit the numerical display indicates '00'.

D. Confirm a value of 0 (zero) volts on Transient Test Channel 291 Answer: B Explanation:

B is correct - Per CPS 4100.01, Section 2. I .2, as described in this choice A is incorrect - Per CPS 4100.01, Section 2.1.2,and CPS 3304.02, Section 8.2.11.2. There is only single 'All Rods Full In' LED at these RACC panels. There is no individual 'Rod Full In' light for each given rod. This choice is very plausihle to the Candidate who vaguely recalls that there & a way to manually address each given r o d s actual position (including 00). using the ID Generator, at these RACC panels.

C is incorrect - This choice suggests a variation offhe correct answer, 'B',but it i s !ml correct (see the corrcct answer's explanation).

D is incorrect - Per CPS 3304.02, Section 8.2.11.3. A zero (0) volts value is associated with 'all rods full in.'

References provided to examinee: None

References:

LP85401.Rod Control & Information System DB410001, Reactor Scram CPS 4100.01, Reactor Scram CPS 3304.02, Rod Control and Information System Date Written: 1 05/03/05 I Author: I Ryder Comments: None

ROISRO: I Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I I I 215004K6.01 I 3.2 I 3.3 I Higher System'Evolution Name: I Category:

Source Range Monitor (SRM) System I plant Systems KA Statement:

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

A. Scram, ONLY B. Rod block, ONLY C. Rod block AND scram D. NEITHER a scram, NOR a rod block Answer: A

~~

Explanation:

A is correct - Per CPS 5005-1K, rod block is bypassed with the Mode Switch in RUN ('plant operating at rated power'). Per CPS 5005-1 K, scram function is dependent on neither the Mode Switch position. nor IRM Range. It solely depends on shorting link status. Per 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 incorrect - 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 considered the specific impact of having ONLY a single shorting link rcmoved (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.

References:

LP85212. Reactor Protection System LP85215 .~SRMs I I

~

CPS 5005-IK. SRM UPSC ALARM OR INOP I Date Written: I Comments: None 02/06/05 I Author: I Ryder I

1 Question # I 7 1 RO/SRO: I Tier: 1 Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I 1 I 215005 K1.07 1 2.6 I 2.9 I Higher SysteWEvolution Name: I Category:

Average Power Range MonitorILocal Power Range 1 plant Systems Monitor I KPL JLalelne,ll:

LI-L----L Kn owledge of the physical connections andor cause-effect relationships between APRMILPRM and the following:

1 Process computer, performance monitoring system I 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 col~ect- Reference LP8541 I and CPS 9431.60. throughout this discussion. AGAF calculation: AGAF = % CTP +

APRM reading.

Conditions prior to the LPRM failure:

o 90% APRM reading resulting from 33 & LPRM flux signals Conditions post-LPRM failure (had LPRM still feeding the APRM):

o (32 +33) x 90%= 87.3%APRM reading AGAF is now reading: AGAF = 90% + 87.3% = 1.031

1 Question # I 7 I RO/SRO I Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I I 1 215005 Kl.07 1 2.6 I 2.9 I Higher SystemlEvolution Name: I Category:

Average Power Range Monitor/Local Power Range I Plant Systems Monitor I KA Statement:

Knowledge of the physical connections andor cause-effect relationships between APRM/LPRM and the following:

Process computer. performance monitoring system -

Objective: I Question Source: I Level of Difficulty:

LP85211.1.13.3; .1.15.1 New I 3.2

References:

LP8541 I , APRM/LPRM System CPS 9431.60. APRM Gain Adjustment Date Written: I 02/07/05 I Author: I Ryder Comments:

Operational Validity basis for this question: I ) operators must understand the meaning of the AGAF value, rather than

~ just recognize whether an 'as found' value is SATor UNSAT (per Tech Spec SR 3.3.1.1.2); 2 ) opera1or.s musl recognize that an 'as found' value >l.oOO ism-conservative, where a value <1.000 is relatively conservative (albeit still a ptential 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 surveillance requirement (3.3.1.I .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 1.020upper limit; this is the reason for the stern's opening 'ASSUME'

1 Question # I 8 1 KA Statement:

Knowledge of the REACTOR CORE ISOLATION COOLING (RCIC) design features andlor interlocks which provide for the following: Manual initiation Operators are ready to MANUALLY initiate RCIC from 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, lE51-FO46, to fully open, enabling the opening circuit for RCIC Steam Supply Valve, IE51-FO45.

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

C. the Ramp Generator to BEGIN its ramping period.

D. the Ramp Generator to FINISH its ramping period.

Answer: B Explanation:

B is correct - Per LP85217, page 44, and drawing E02-IR199, Sheets 6 and 9. The 'TD' device shown on Sheet 9 must be energized (timcd out) before the initiation signal can enable the F045 opening circuit. CPS 9054.03, Section 8.2.2.2.

validates that this TD device i s calibrated for about 6 seconds. Only after holding the pushbutton depressed for about 6 seconds docs F045 begin to open. Direction on how to manually initiate RClC is found in CPS 3310.01. Section 8.1.3.

A i s incorrcct - Refer to LP85217. pages 44-47, for the explanation related to a11 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 incorrect - The Ramp Generator does not even come into the picture until 6 to 9 seconds after the F045 valves to open. Refer to LP852 17. page 45.

Keferences provided to examinee: None

References:

LP85217, Reactor Core Isolation Cooling CPS Drawing E02-IR199. Sheets 6 and 9, RClC Schematic Diagram CPS 9054.03, RClC Simulated Auto Actuation Test CPS 3310.01, RCIC

1 Question # I 9 I RO/SRO: I Tier: 1 Group: 1 KA: 1 ROIR: SROIR: 1 Cog Level Both I 2 I I I 239002A1.06 I 3.7 I 3.8 I Higher S y s t e d v o l u t i o n Name: 1 Category:

ReliefISafety Valves 1 Plant Systems KA Statement:

Ability to predict and/or monitor changes in parameters associated with operating the RELIEFSAFETY VALVES controls including: Reactor power 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, and describes the reason why?

INITIALLY. indicated reactor...

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

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

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

D. power RISES, because of the reduced feedwater inlet temperature Answer: B B is correct - Per USAR, Section 15. I .4.3.3. the SRV opening initially produces a slight depressurization transient. Per Generic Fundamentals knowledge, the drop in reactor pressure produces more voiding, which initially lowers reactor power (see LP85756S. page 26, for an analogous 'off-normal' pressure transient, which validates the relationship between pressure and power).

A and C are incorrect - Per Generic Fundamentals knowledge, as well as the USAR discussion above. the initial depressurization causes more voiding. which results in an initial RISE of reactor water level ('swcll' transient). The open SRV diverts main steam flow away from (is upstream of) the main steam line flow element (see LP85239. Figure I ) . This results in the Feedwater Level Control System immediately seeing a lower steam flow, not a higher steam flow.

D is incorrect - Per CPS 4005.01,Loss of Feedwater Heating, the SRV opening is a loss of feedwater heating event.

The escape of steam to the suppression pool diverts it away from the main turbine and the extraction stcam supply.

Reactor feedwater temperature lowers (i.e.. a greater amount of core inlet sub-cooling is produced) and this positive reactivity addition should help to raise reactor power. However, this is not the INITIAL power response. The depressuriration event immediately lowers reactor power.

Objective: I Question Source: I Level of 1)ifficulty:

None New I 3.6

References provided to examinee: None

References:

USAR, Section 15.1.4, Inadvertent SRV Opening LP85239, Main Steam System LP85756S, Reactor Operational Physics - BWR Reactor Theory. Chapter 8 CPS 4005.01, Loss of Feedwater Heating

1 Question # I 10 I KA Statement:

it Systems 1 Knowledge of the operational implications of the following concepts as they apply lo REACTOR WATER LEVEL 1 CONTROL SYSTEM: Turbine speed control mechanisms: TDRFP 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, and 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.

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

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

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

Answer: C Explanation:

C i s correct - Per CPS 3 103.01.Sections 2.1.7, 8.1.4. and 8.3.2. The Manual Speed Pot nimt be at zero speed (fully CCW) in order to reset the control logic for the LP and HP control valves. This ensures that when the operator RESETS the TDRFP, immediate control of speed will be available. During a normal plant and fccdwater system startup, the last time that operators have reason to manipulate the Manual Speed Pot is in Section 8.1.4.20(l)(e).

It is here that the pot is set to ZERO speed position and should remain there with 'Feedwater and Feedwater Level Control in their normal configurations'.

A is incorrect - For the reasons described ahove.

B is incorrect - Per CPS 3103.01, Section 8. I .8. the TDRFPs are being controlled by the Master Level Controller.

which means that each RFPT Flow Controller is in AUTOMATIC.

D is incorrect - Per LP85570. page 24, and Figure 13, these XFER switches are in AUTO (pushhutton depressed) whenevcr a n y flow controller is controlling the TDRFP

References:

LP85570. Feedwater Level Control System CPS 3103.01. Feedwater

I Question# I 10 ]

RO/SRO: I Tier: 1 Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I 1 1 259002K5.07 I 2.7 I 2.7 I Lower SystemiEvolution Name: I Category:

Reactor Water Level Control System I Plant Systems KA Statement:

Knowledge of the operational implications of the following concepts as they apply lo REACTOR WATER LEVEL CONTROL SYSTEM: Turbine speed control mechanisms: TDRFP Date Written: I Comments: None 02/08/05 1 Author: I Ryder I

1 Question# I 11 I RO/SRO I Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I I I 261000K4.01 I 3.7 I 3.8 I Higher SystemIEvolution Name: I Catecory:

Standby Gas Treatment System I plant Systems KA Statement:

Knowledge of STANDBY GAS TREATMENT SYSTEM design features and/or interlocks which provide for the following: Automatic system initiation 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, lRIX-PROOlC, fails UPSCALE and produces a trip WITHOUT operator action, which ONE of the following identifies the VG Trains that are RUNNING, and explains why?

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

B. NEITHER, because no VG initiation signal is present.

C. ONLY Train A, because no Train B initiation signal is present D. ONLY Train B, because with the Div 1 NSPS Bus outage, VG Train A auto-initiation signals are disabled.

Answer: B B is correct - Refer to CPS 5140.61 for the one-out-of-two-twice radiation monitor combinations that can produce a VC initiation signal. Although both trains are capable of auto-starting. a trip condition on a single radiation monitor channel (IRIX-PR001C. only) does satisfy the one-out-two-twice initiation logic.

A is incorrect - The Div I NSPS Bus outage does produce a power failure ttip of any of the I RIX-PRO01 channels.

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

not from NSPS (inverter) v)wer. Therefore. the upscale trip produced by the PROOlC failure docs not. alone. satisfy the one- ut-of-two-twice VG iniliation logic.

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

D is incorrect - This is quite plausible to the candidate who believes the NSPS Bus outage produce a power failure trip on the PROOlA channel. and recalls that (per CPS 3509.01COOl, Appendix A. page 2 5 ) the Bus outage disahlcs the auto-start of VG Train A on a LOCA signal (High DW Pressure, Low-Low Level. only). The radiation monitor initiation signals are affected hy this Bus outage.

1 Question # I 11 1 RO/SRO. [ Tier: I Group: I KA: I R O I R I S R O I R 1 Cop, Level Both I 2 I I I 261000K4.01 I 3.7 I 3.8 I Higher SysIellllEvolution Name: I Category:

Standby Gas Treatment System I ~~ant~ystems KA Statement:

Knowledge of STANDBY GAS TREATMENT SYSTEM design features andlor interlocks which provide for the following: Automatic system initiation Objective: I Question Source: I Level of Difficulty: I L LP85216.1.7 I New I 3.8 I CPS 5140.61, IRIX~PROOIA.B, C. DCONTAINMENTEXHAUST CPS 3509.01C00l.Division I NSPS Bus Outage Date Written: I 02/09/05 I Author: I Ryder Comments:

I. Although this question, as framed, leads to a correct answer involving NO automatic system initiation, it does meet the KA. This KA only demands that the candidate demonstrate a knowledge of the system design features andor interlocks; therefore, such a knowledge that leads to a cunclusion that no initiation should result can & come from that same body of knowledge. Admittedly, this question would satisfy any of the 'A3' category KAs; they demand that the operator observe something actually hamening in the system.

2. This is a Higher Cognitive Level (HCL) question for the fullowing reasons:

a. The correct answer demands a knowledge of the one-out-two-twice 'ladder' logic arrangement for a radiation monitor initiation signal to he developed.

h. Elimination of the distracters. particularly choice 'D,demands a knowledge of the relationship between SGTS and the NSPS Bus outage. The candidate must also recognize that the NSPS Bus does not power thc radiation monitor channels. in order to eliminate an otherwise very attractive choice 'A'.

I c. The ahove reasons. combined with the fact that the question demands a 'why' from the candidate.

satisfies the HCL requirements discussed in NUREG-1021, Rev. 9, Appendix A, page 7 of 11.

I Question ## I 12 1 RO/SRO I Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I I I 262001 A2.01 1 3.4 I 3.6 I Higher SystemlEvolution Name: I Category:

A.C. Electrical Distribution I plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Turbinelgenerator trip 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 lB?

A. Cycle the UAT 1B feeder breaker to OPEN then back to AUTO to enable the automatic undervoltage transfer to the RAT.

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

C. 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.

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

Answer: C Explanation:

Ciscorrect-refer to LP85571. FigureJ. andpage31.andtoCPS 3501.01,Section8.1.1. Thenormal feed for6.9 KV Bus IB is Srom UAT IB. When the generator trips, UAT I B (and IA) are lost. forcing the auto-transfer of 6.9 KV Bus IB to thc RAT (its reserve feed). Because the bus is 'dead', procedure section 8.1. I is used to complete the trmsfcr manually. Steps 8.1.4,5, and 7 are featured in this answer choice.

A is incorrect - There are no undervoltage transfer relays on the 6.9 Kv busses. ReSr to LP8557 I pdge 29.

B and D arc incorrect - Whether the hus is alive or dead is irrelevant. The Sync Switch must still be used.

Ohjective: I Question Source: I Level of Difficulty:

LP8557 I . 1.4.6 New I 2.5 I References provided to examinee: [ None

References:

LP85571,Auxiliary Power Date Written: I 04/28/05 [ Author: I Ryder Comments:

The way this question addresses the 'predict' component of this KA dcscrves some mention. The 2"d bullet ofthe stem conditions demands that the candidate: I ) determine if this is an expected condition, and 2) if not, where should the loads have transferrcd to?. Having 'predicted' the desired condition for loads on this bus, the candidate proceeds directly to 4 choices that demand that helshe correct the undesired condition. Helshe can choose the correct answer only by knowing where the loads should have auto-transferred to.

Question revised. Original deemed UNSAT by NRC due to distractor A and B implausibility regarding availability of the UATs. Both distractors modified, explanation modified and additional references cited in distractor A.

GDSetser 6/13/05

1 Question# I 13 I ROISRO: I Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I I 1 262001 A4.05 1 3.3 I 3.3 I Higher SystemlEvolution Name: I Category:

A.C. Electrical Distribution I plant systems KA Statement:

Ability to manually operate andlor monitor in the control room: Voltage. current, power, and frequency on A.C. buses After transfemng loads to the ERAT, operators are preparing to shut down the RAT SVC from 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:

4,220 Volts

-4MVARs Which ONE of the following identifies the FINAL voltage value that the SVC Voltmeter is expected to ramp to AFTER the operator places the RAT SVC control switch to OFF?

A. 4,060 Volts B. 4,140 Volts C. 4,300 Volts D. 4.380 Volts Answer: C Explanation:

C is correct - Refer to CPS 3505.03, Section 8.3 and Appendix A. The - MVARs indication means that the SVC i s acting to hold down voltage. When thc SVC is removed from service, the resulting (uncompensated) voltage will ramp up (to ahove 4.220 volts). The rule of thumb is 20 volts per MVAR. In this casc. that amounts to +80 volts. or a FINAL value or 4.300 volts.

A. B , and D are incorrect - For the reasons described. They have face validity and are plausible t o the candidate who either. does not recall the rule of thumb (applies 40 volts per MVAR instead of 20 volts per MVAR). or cannot distinguish hetween a + MVAR reading and a - MVAR reading.

Objective: I Question Source: 1 Level ofDifficulty:

LP85305.1.10.2 New I 2.6

References:

LP85305, Static VAR Compensator CPS 3505 03. RAT & ERAT Static VAR Compensators

I Question # I 14 1 RO/SRO: I Tier: I Group: I KA: I ROIR. I S R O I R 1 Cog Level Both I 2 I I I 262002K4.01 I 3. I I 3.4 I Lower SystemIEvolution Name: I Category:

Unintenuptable Power Supply (A.CJD.C.) I Piant Systems KA Statement:

Knowledge of UNINTERRUPTABLE POWER SUPPLY (AC/DC) design features and/or interlocks which provide for the following: Transfer from preferred power to alternate power supplies 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.

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

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

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

Answer: B Explanation:

B is correct - Refer to LP85434, pages 1 I , 13. 14. and Figures 2b and 3. A Divisional NSPS Cabinet is normally supplied from the associated 125 VDC Bus. With the Cabinet in its normal operating configuration. the Transfer Switch is in the INVERTER position. This allows the Static Switch to auto-transfer bus loads to the alternate 120 VAC supply.

A is inconect - This describes the impact of a m-Divisional NSPS lnvener Cabinet. See LP85434. Figure 6.

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

D is incorrect - For the reasons associated with the correct answer Objective: I Question Source: I Level of Difliculty:

LP85434.1.4.3 I CPS Operations Exam Bank. Question #I0278 (DIRECT, I 2.1

References:

LP85434, Nuclear System Protection System (NSPS)

Date Written: I 02/10/05 I Author: I Ryder Comments: None

I Question # I 15 I ROISRO T Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I I I 263000A2.02 I 2.6 I 2.9 I Higher SystemlEvolution Name: I Category:

D.C. Electrical Distribution I Plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control. or mitigate the consequences of those abnormal conditions or operations: Loss of ventilation during charging The following statement describes a FACT concerning battery hydrogen production:

The RATE at which a battery produces hydrogen during an EQUALIZING charge is DIRECTLY proportional 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 (VX) 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 ...

A. ( I ) 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 a n d 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 Explanatiun:

B is correct - Concerning Part ( I ) of thc question. refer to information extracted from thc Wch source:

www.de~.state.~a.usldepldenutate/minres/~ms/wehsite/training/batt~r~.~df. This is battcry training program presentation associated with the U.S. Bureau of Mines. These slides show that the rate of hydrogen production ('H')is directlv proportional to the capacity of the battery (in ampere-hours). Since the battery capacity rises over the charging

RO/SRO: I Tier: I Group: I KA: I ROIR: I snom Cog Level Both I 2 I I I 263000A2.02 I 2.6 I 2.9 I Higher SystemlEvolution Name: I Category:

D.C. Electrical Distribution I ant Systems KA Statement:

Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions. use procedures to correct. control, or mitigate the consequences of those abnormal conditions or operations: Loss of ventilation during charging period, so too does the rate of gas production ('H').Concerning Parf (2) of the question. refer to CPS 3412.01, Section 8.2.3.

A, C, and D arc incorrect - For the reasons described above.

Objective: I Question Source: I Level of Difficulty:

XIlnl., I ?n References provided to examinee: I None

References:

I CPS 3412.01, Essential Switchgear Heat Removal (VX)

Date Written: I 05/02/05 I Author: I Ryder Comments:

This question is categorized as Higher Cognitive (HCL), because the Candidate must 'associate' the given stem claim regarding the rate of hydrogen production with hisher understanding of how a battery's capacity changes over the period of being re-charged. It is also HCL because. as a closed-reference question, in order for the Candidate to eliminate choiccs 'C' and 'D', helshe must recognize the danger that would exist if a portable blower (i.e.. a potential spark producing device) were to be started whcn hydrogen is already above a combustible concentration (nominally 4-5%).

1 Question# I 16 I 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 from this Governor malfunction Which ONE of the following:

(1) predicts the response of DG 18 to the governor control signal failure, and (2) describes the action necessary to correct the DG's current operating condition?

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

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

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

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

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

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

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

Answer: A Explanation:

A is correct - Refer t o CPS 9080.02. page 28 CAUTION. The monthly surveillance has the DG loaded in parallel with the off-site power. Per LP 85264. pages 15-16, the electrical governor is 'normally' controlling the engine speed, and when the electrical signal is lost (fails low). the mechanical governor assumes control at a 5% HIGHER governor setpoint. Because the DG is paralleled with off-site, engine speed cannot change. but the DG does pick up mload.

Although we cannot necessarily predict that it will pick up 5% additional load (for a new load of 3.990 KW), i t will cenninly pick up an amount of load that causes the DG to operate -of its 'continuous rating of 3.875 KW' (see Section 6.2.I I of CPS 9080.02). This question suggests that it is this 'operating condition' that nceds 10 he corrected.

Once thc CRS decides to 'correct' the condition, the only way to do so is by corndetely unloading the DG. With the railed electric governor, and with NO procedure guidance that would allow operators to manually lower the mechanical govcmor's setpoint (locally). the required action is to Emergency STOP the machine.

I B is incorrect - For the reasons descrihcd above.

I Question# I 16 I C and D are incorrect -These choices suggest the 'predicted response of the DC if it were paralleled with off-site.

The Candidate is expected to know that the machine is varalleled with off-site when running loaded for the Monthly surveillance.

Objective: I Question Source: I Level of Difficulty:

None New I 3.5

References:

LP85264. Diesel GeneratodDiesel Fuel Oil CPS 9080.02, Diesel Generator I B Operability &e.. the Monthly)

)ate Written: I 04/15/05 I Author: 1 Ryder

omments:

This question is an ROISRO one, and is a an SRO-ONLY question. for the following reason:

I. 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.

2. The last stem condition bullet has pre-empted the need for the SRO to make a decision about whether to correct the DG operating condition, or not.

3. Thc suggested choices for the Part (2) 'required action' are a set of choices from which & the SRO would be expected choose. Rather. each choice challenges the Candidate (both ROISRO) to recognizing what is the only possible/permitted way of correcting the operating condition.

4. Therefore. this question is presented in a way that really amounts lo requiring only several pieces of

'systems' type of knowledge: I ) the configuration that the DG is in before the govcmor failure (i.e..

paralleled with off-site). and on the electric governor; 2) how the DG engine speed and load respond as a result of the 5% off-set between the electric and mechanical setpoints; and 3) recognition of the fact that with there being no electric governor control. the only solution to emergency STOP the DG.

5. As such. this is a question that should be on both the RO and SRO Exams

I Question# I 16 I RO/SRO: I Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I I I 264000A2.01 I 3.5 I 3.6 I Higher SystemlEvolution Name: I Category:

Emergency Generators (DieseUJet) I plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Parallel operation of emergency generator 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 from this Governor malfunction Which ONE of the following:

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

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

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

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

(2) RAISE the SETPOINT for the Mechanical Governor, locally C. (1) Engine Speed RISES, but Load REMAINS THE SAME.

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

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

Answer: A Explanation:

A i s correct - Rcfcr to CPS 9080.02. page 28 CAUTION. The monthly surveillance has the DG loaded in parallel with the off-site power. Per LP 85264, pages 15-16, the electrical governor is 'normally' controlling thc engine speed. and when the electrical signal is lost (fails low). the mechanical governor usumes control at a 5% HIGHER govcrnor setpoint. Because the DG is paralleled with off-site, engine speed cannot change, but the DG does pick up more load.

Although we cannot necessarily predict that it will pick up 5% additional load (for a ncw load of 3.990 KW). it will celtainly pick up an amount of load that causes the DG to operate of its 'continuous rating of 3.875 KW' (see Section 6.2.I I of CPS 9080.02). T h i s question suggests that it is this 'operating condition' that needs to be corrected.

Once the CRS dccides to 'correct' the condition. the only way to do so is by com!Jletelv unloading the DC. With the failed electric governor. and with NO procedure guidance that would allow operators to manually lower the mechanical governor's setpoint (locally), the required action is to Emergency STOP the machine.

1 B is incorrect - For the reasons described above

I Question# I 16 I RO/SRO I Tier: I Group: I KA: I ROIR I SROIR I Cog Level L

DULll I

I I)

L. II II 1L W L " m A 1 n, I WVnI.". 1 7.5 I 3.6 I Higher SystemlEvolution Name: I Category:

Emergency Generators (DieselIJet) I plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS; and (b) based on those predictions, use procedures to conect, control, or mitigate the consequences of those abnormal conditions or operations: Parallel operation of emergency generator C and D are incorrect -These choices suggest the 'predicted response of the DG if it were not paralleled with off-site.

The Candidate is expected to know that the machine is varalleled with off-site when running loaded for the Monthly surveillance.

References:

LP85264. Diesel GeneratorIDiesel Fuel Oil CPS 9080.02, Diesel Generator IB Operability (i.e., the Monthly)

)ate Written: I 04/15/0s I Author: I Ryder

omments:

'his question is an ROISRO one, and is a an SRO-ONLY question. for the following reason:

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

2. The last stem condition bullet has pre-empted the need for the SRO to make a decision about whether to comcct the DG operating condition. or not.

3. The suggested choices for the Pan (2) 'required action' are a set ofchoiccs from which wdy the SRO would be expected choose. Rather. each choice challenges the Candidate (both ROISRO) to recognizing what is the only possible/permitted way of correcting the operating cundition.

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

'systems' type of knowledge: I ) the configuration that the DG is in before the governor failure (i.e..

paralleled with off-site), and on the electnc governor; 2) how the DG engine speed and load respond a? a r ~ s u l of t the S% off-set between the electric and mechanical setpoints; and 3) recognition of the fact that with there being nu electnc govcrnur control, the only solution to emergency STOP the DG.

5. As such, this is aquestion that should be on both the RO and SRO Exams.

1 Question# I 17 I ROISRO: Tier: I Group: I KA: I R O I R I S R O I R I Cox Level Both I 2 I I I 400000K6.06 2.9 I 2.9 I Higher SystemlEvolution Name: I Category:

Component Cooling Water System (CCWS) I Plant Systems KA Statement: ~

Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: Heat exchangers and The plant is in MODE 5 for a refueling outage, with the following:

Shutdown Service Water (SX) Pumps A and C are out-of-service and have Clearances installed THEN, a complete loss of Service Water (WS) occurs SX Pump B starts and runs An electrical failure PREVENTS the associated WS to SX Header Isolation Valve from FULLY closing NLO reports that SX Header Pressure reads 90 psig, locally THEN, a tube leak occurs in the CCW Heat Exchanger (HX) that was operating before the loss of WS Which ONE of the following describes the POTENTIAL consequence of the CCW HX tube leak?

A. Reduced heat transfer due to heat transfer surface fouling B. Higher rate of depletion of the CCW Demineralizer resin C. Radioactive discharge to the environment D. Rising level in the CCW Expansion Tank Answer: C C is correct - Per LP85208, page 29. SX system pressure is lower than CCW system pressure. CCW inventory will he lost through the tube leak and find its way into the Lake.

A. B, and D are incorrect - Per LP85208. page 29. These are a11 indicative of a WS-lo-CCW leak (WS at higher pressure than CCW).

Objective: I Question Source: I Level of Difficulty:

LP85208.1.10.1 New I 2.8 References provided to examinee: I None

References:

I ~ ~ 8 5 2 0Component 8, Cooling Water System Date Written: I 05/02/05 I Author: I Ryder Comments: None

1 Question# I 18 I RO/SRO: I Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I 2 I204000K3.06 I 2.6 I 2.1 I Higher SystemlEvolutiun Name: I Category:

Reactor Water Cleanup System I Plant Systems KA Statement:

Knowledge of the effect that a loss or malfunction of the REACTOR WATER CLEANUP SYSTEM will have on the 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 of placing the RWCU Filter System Functions Interlock Switch to NORMAL?

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

B is correct - Per CPS 3303.02. Sections 8.1.6 and 8.6.5, the NEXT operator action is to return the System Function Interlock Switch to the NORM(al) 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. Per LP85204. page 29, this a n 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 incorrect -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 from the RWCU system and system isolations are not possible.

D is incorrect - This 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 perturbation being suggested by the stem conditions that could cause a RWCU Pump problem (e.& a seal leak) that would result in high radiation levels in that pump area.

Objective: I Question Source: 1 Level of Difficulty:

LP85204.1.15 New I 3.5

ROISRO I Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I 2 1 204000K3.06 I 2.6 1 2.7 I Higher Systedvolution Name: I Category:

Reactor Water Cleanup System I Plant Systems KA Statement:

Knowledge of the effect that a loss or malfunction of the REACTOR WATER CLEANUP SYSTEM will have on the following: Area radiation levels References provided to examinee: None

References:

LP85204, Reactor Water Cleanup System CPS 3303 01, Reactor Water Cleanup System

- CPS 3303.02, RWCU Filter Demineraltzer Operating Procedure

1 Question# I 19 1 RO/SRO. 1 Tier: I Group: I KA: I R O I R I SROIR: I Cog Level Both I 2 I 2 I 201005K5.10 I 3.2 I 3.3 I Higher SystemlEvolution Name: I Category:

Rod Control and Information System (RCIS) I Plant Systems KA Statement:

Knowledge of the operational implications of the following concepts as they relate to ROD CONTROL AND INFORMATION SYSTEM (RCIS): Rod withdrawal limiter 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.

3. 10%; the light above the LO POWER SET PT is ON.

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

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

Answer: D Explanation:

D is correct - Refer to LP85401. pages 23-24, and Figure 4, to Tech Spec 3.3.2. I , and to CPS 3005.01, Section 6.2. for all ofthe answer choices. Reactor power is within the range when the RWL must be OPERABLE (>29%) RTP and at or below the High Power Setpoint (HPSP) of70% RTP). The fact that the LO POWER SET PT light is OFF, while the LO POWER ALM PT light is ON, indicates a --functioning Low Power Function of the RWL. permitting g y type of rod movement (subject to the built-in notch restraints ofthe RWL itsel0. There arc NO Tech Spec administrative restrictions with these conditions.

A is incorrect- Reactor power is within the range when the RWL must be OPERABLE P 2 9 % RTP and at or below the High Power Setpoint (HPSP) of 709,).However. the fact that both of these lights are OFF Indicates that the Low Power Function of the RWL (normally enabled by the Rod Pattern Controller, RPC) is in fact bypassed, making the RWL INOPERABLE. Per TS LCO 3.3.2.1.A. all control rod WITHDRAWALS must be immediately suspended.

Insertions are still allowed.

B is incorrect - Reactor power is below the Low Power Setpoint (LPSP) of 16.7% RTP. However. the fact that the light is ON indicates that the RPC is INOPERABLE. Per Tech Spec LCO 3.3.2.1.8. all normal rod movemenls (in '"d out) must be immediately suspended.

C is incorrect - Reactor power is above the High Power Setpoint (HPSP) of 70% RTP. However. this light being OFF indicates that 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. Insertions are still allowed.

Level of Difficulty:

2.1

1 Question # I 19 I ROISRO: I Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I 2 I 20l005 K5.10 [ 3.2 I 3.3 I Higher SystemlEvolution Name: I Category:

Rod Control and Information System (RCIS) I Plant Systems KA Statement:

Knowledge of the operational implications of the following concepts as they relate to ROD CONTROL AND INFORMATION SYSTEM (RCIS): Rod withdrawal limiter References provided t o examinee: None

References:

LP85401. Rod Control and Information System CPS Tech Spec 3.3.2. I, Control Rod Block Instrumentation CPS 9436.05, RPC Low Power Setpoint Channel Calibration CPS 9030.01C02l. RPC Low Power Setpoint Checklist CPS 3005.01, Unit Power Changes Date Written: I 05/04/05 I Author: I Ryder Comments:

This question is on thc RO Exam (as opposed to he classified as SRO-ONLY) because the RO Candidate needs only to understand the meaning of each of these lights (systems knowledge). recognize the impact of a given lights status on the RWURPC operability, and then recall (from memory) the applicable <I-hour Tech Spec Actions &e.. to immediately suspend rod movements).

I Question # I20 I RO/SRO: I Tier: I Group: I KA: I ROIR I SROIK: I Cog Level Both I 2 I 2 I 202001 A4.01 I 3.7 I 3.7 I Lower SystedEvolution Name: I Category:

Recirculation System I Piant Systems References provided to examinee: None

References:

LP85202, Reactor Recirculation System CPS 3302.01, Reactor Recirculation System CPS 5003-4C, Recirc MG A Interlock Bypass

RO/SRO: I Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I 2 I 214000A3.01 I 3.4 I 3.3 I Lower SystemIEvolution Name: I Category:

Rod Position Information System I Plant Systems KA Statement:

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

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. CRD drive water flow reads 5 gpm.

B. ROD OVERTRAVEL annunciator is NOT received.

C. Red full-out light is LIT on the full-core display D. Rod position is BLANK on the full-core display.

Answer: D 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.

A is incorrecl- Per Section 8.1.10 NOTE. Whether CRDM seals are good (1-3 gpm stall flow indicated), or bad (something higher than 1-3 gpm). stall flow is unaffected by the status of the control rod blade (coupled. or uncoupled).

This is the reason why stall flow I S to be used ONLY as an indication of seal condition.

B is incorrect- Per CPS 3304.02. Section 8.2.6 NOTE, the Rod Overlravel annunciator would be received for an uncoupled rod.

C is incorrect - Per CPS 3304.02. Section 8. I . IO. I , Zd bullet.

Objective: I Question Source: I Level of Difficulty:

LP85401.1.4.9 New I 2.8 Date Written: I 05/02/05 I Author: I Ryder Comments: None

1 Question # I 22 I RO/SRO: I Tier: I Group: 1 KA: I R O I R I S R O I R I Cog Level Both I 2 I 2 I 233000A2.07 I 3.0 I 3.2 I Higher SystellllEvolution Name: I Category:

Fuel Pool Cooling and Cleanup I Plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the FUEL POOL COOLING AND CLEANUP and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High fuel pool 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, and (2) describes the operational impact?

A. (1) Exceeds the ORM OPERATING REQUIREMENT for Spent Fuel Storage Pool temperature.

(2) Movement of fuel assemblies in the pool is NOT permitted.

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

(2) If the reactor is operating, a normal plant shutdown is required C. (1) Exceeds the TECHNICAL SPECIFICATION LCO for Spent Fuel Storage Pool temperature.

(2) Movement of fuel assemblies in the pool is NOT permitted.

D. (1) Results in airborne radioactivity in the Fuel Building (2) If the reactor is shutdown, it must remain shutdown.

Answer: D Explanation:

D is correct - Part ( I ) . per LP85233. page 45. Part (2). per CPS 33 17.01, Section 4.6.

A and C are incorrect - There is no Tech Spec LCO, or ORM OR, related to Spent Fuel Storage Pool Temperature.

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

Objective: I Question Source: I Level of Dificulty:

LP85233.1.14 New I 3.4 References provided to examinee: I None

References:

I LP85233. Fuel Pool Cooling and Cleanup CPS 3317.01, Fuel Pool Coiling and Clc-anup CPS 5040-IF, High Temp Spent Fuel Storage Pool

1 Question# I 22 I RO/SRO: I Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I 2 I 233000A2.07 I 3.0 I 3.2 I Higher SystemlEvolution Name: I Category:

Fuel Pool Cwling and Cleanup I plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the FUEL POOL COOLING AND CLEANUP: and (b) based on

. - the consequences of those abnormal conditions or those nredictions. use mocedures to correct. control. or mitigate

~~~~~~

operations: High fuel pool temperature I Date Written: I 04/29/05 I Author: I Ryder Comments: None This question is presented on the RO Exam (and is considered an SRO-ONLY type) because the 'operational impact' portion requires only the recall of an operating procedure PrecautiodLimitation; it does require any operational decision-making (reserved for the S R O s responsibility), nor does it require any form of 'application' (which might or might not he reserved for the SRO) of the information contained in that PrecautiodLimitation.

I Question# 1 23 1 RO/SRO Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I 2 I 241wOK6.01 I 2.8 I 2.9 I Higher SystemIEvolution Name: I Category:

ReactoriTurbine Pressure Regulating System I ant Systems

- KA Statement:

Knowledge of the effect that a loss or malfunction of the following will have on the REACTOWTURBINE PRESSURE REGULATING SYSTEM: A.C. electrical power Explanation:

C is comect - Per LP85576, page 15, and CPS 3509.01C006, 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-energization of the 24 VDC Trip Bus and Electrical Trip Solenoids. Even if the FAST Starting Rate has been selected (see LP85241. page 30. and CPS 3105.01.Section

8. I .6),the machine will be running at NO HIGHER than about 900 RPM. at 5 minutes after depressing the I 8 0 0 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. operators shouldnt expect to see the machine speed any higher than about 200-400 rpm. & time all power is lost to the TBV control circuits, the TBVs will fail shut. NOTE: The stem condition regarding CB MCC C being de-energized ensures the intended failure response ofthe TBVs; i.e., there m y be an auctioneering of power (between this MCC and UPS IB) to the TBV circuits. Taking this MCC away. ensures the TBVs yJ li fail shut.

A. B. and D are incomect - For the reasons described above.

Objective: I Question Source: I Level of Dimculty:

LP85576. I . 13.5 New I 3.6 References provided to examinee: None

References:

LP85241, Steam Bypass and Pressure Control System LP85576. Computer UPS CPS 3509.01C006. UPS IB Bus Outage CPS 3105.01, Turbine

I Question ## I 23 I RO/SRO: I Tier: I Group: I KA: I R O I R I S R O I R I Coy: Level Both I 2 I 2 I 241000K6.01 I 2.8 I 2.9 I Higher SystenVEvolution Name: I Category:

ReactodTurbine Pressure Regulating System I Plant Systems KA Statement:

Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM: A.C. electrical power Date Written: I 05/16/05 I Author: I Ryder Comments: None

I Question # 1 24 I RO/SRO I Tier: I Group: I KA: I ROIR: I S R O I R I Cog Level Both I 2 I 2 I 259001 K2.01 I 3.3 I 3.3 I Lower SystedEvnlution Name: I Category:

Reactor Feedwater System I Plant Systems KA Statement:

Knowledge of electrical power supplies tn the following: Reactor feedwater pump(s): Motor-Driven only Which ONE of the following describes the impact of a loss of 6.9KV Bus lB?

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

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

C. ONLY ONE Circulating Water Pump will run.

D. Isolated Phase Bus Duct Cooler Fan A will NOT run Answer: B Explanation:

B is correct- Per LP85259, page 12. MDRFP is powered from 6.9KV Bus IB A is incorrect - Per LP85202. pages 9 and 27. Even a SLOW speed start of the RR Pump I B requires 6.9KV Bus I B power.

C is incorrect - Per LP85275, pages 14 and 17. 6.9KV Bus IB powers only one of the 3 CW Pumps (CWP B). The other two powered from 6.9KV Bus I A. There are no inter-pump starting permissives o r trip signals that would inhibit the running o f w pumps. CWP A and C, on the 6.9KV Bus I A .

D is incorrect - Per LP85572, pages 8 and 13. and CPS drawing E02-IAP03 (and LP8557 I , Figure 4 for clarity). One of these fans is powercd from 6.9KV Bus I A, via 480V Unit Sub I J . Only the B Fan is lost if 6.9KV Bus I B is lost.

Objective: I Question Source: I Level of Difficulty:

LP85259. I .4.4 New I 2.0 References provided to examinee: None

References:

LP85259, Feedwater System LP85202. Reactor Recirculation System LP85275. Circulating Water System

. LP85572, Isolated Phase Bus Duct Cooling

I Question# 125 1 RO/SRO. 1 Tier: [ Group: I KA: I R O I R I S R O I R I Cog Level Both I 2 I 2 I 286000K4.07 I 3.3 I 3.3 I Higher SystenJEvolution Name: I Category:

- Fire Protection System I plant Systems KA Statement:

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

Diesel engine protection 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 q should have prevented the engine from running) that FAILED during this test of the B Fire Pump?

A. 1 B. 2

c. 3 D. 4 Answer: B Explanation:

B is correct - Per LP85286. pages 2 1-25. One engine crank and rest cycle takes 30 seconds; the controller should have allowed no more than 6 total cycles ( I 80 seconds. ..At Time = 3 minutes) before stopping the au1o-start sequence and generating a Failure to Start alarm. This was the first failure of a protective action. The engine was allowed to reach 1309, of rated speed and continue to run. Tne engine should have tripped (stopped) at I2070 of rated speed. This was the second failure of a protective action.

A is incorrect - For the reasons described above.

C and D are incorrect - Per the same reference cited above. Neither the High Engine Temperature alarm. nor the Low Lube Oil Pressure almn. provide an automatic protective action; they are alarms, only.

Objective: I Question Source: 1 Level of Difficulty:

LP85286. I . 10.4 New I 4.0

/Question I

1. 25 .. I RO/SRO: I Tier: I Group: I KA: 1 R O I R I S R O I R I Cog Level Both I 2 I 2 I 286000K4.07 I 3.3 I 3.3 I Higher SystemlEvolution Name: I Category:

Fire Protection System I Plant Systems KA Statement:

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

Diesel engine protection

References:

LP85286, Fire Protection and Detection CPS 9071.02, Diesel Fire Pump Capacity Checks/Scquential Starting Date Written: I 02/21/05 I Author: I Ryder Comments: None

1 Question # I 26 I RO/SRO: I Tier: I Group: I KA: I R O I R I S R O I R 1 Cog Level Both I 2 I 2 1 288000K1.05 I 3.3 I 3.6 I Lower SystemlEvolution Name: I Category:

Plant Ventilation Systems I Plant Systems KA Statement:

Knowledge of the physical connections andlor cause-effect relationships between PLANT VENTILATION SYSTEMS

[ and the f&wing:Process radiation monitoring system I Which ONE of the following identifies the TOTAL NUMBER of HIGH RADIATION Isolation CONDITIONS that:

(1) can cause an isolation of the normal Continuous Containment Purge (CCP) lineup, and (2) are overridden (considering ALL plant ventilation systems) if both Containment HVAC 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 Explanation:

C is correct - Per LP8S4S5, pages 4, 6 , 49, SO. 51, and 52. Considering Part ( I ) ...a total ofTHREE signals

('Conditions') will isolate CCP (Croup 10 valves); they are: Containment Bldg Exhaust Radialion 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') u e overridden by placing these switches in TOTAL BYPASS: they are: the same THREE that isolate CCP. &p the Fuel Building Exhaust Radiation High signal (which docs NOT close the CCP valves).

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

I I Objective:

LPXS455. I .4. I I I Question Source:

New I Level of Difficulty:

3.4 References provided to examinee: I None

References:

I LP85455, Containment Ventilation and Drywell Purge

I Question# I 26 1 RO/SRO I Tier: I Group: I KA: I ROIR 1 SROIR: I Cog Level Both I 2 I 2 I 288000K1.05 I 3.3 I 3.6 I Lower S y s t e d v o l u t i o n Name: I Category:

Plant Ventilation Systems I Plant Systems KA Statement:

Knowledge ofthe physical connections andlor cause-effect relationships between PLANT VENTILATION SYSTEMS and the following: Process radiation monitoring system ate Written: I 04/29/05 I Author: I Ryder Comments:

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 (1) -from memory. recall how many different radiation signals will isolate CCP; Part (2) - from memory, recall how many different radiation signals arc overridden by the Total Bypass switch.

[ Question# I27 1 ROISRO Tier: I Group: I KA: I ROIR I SROIR. I Cog Level Both I 3 [ Generics I 2.1.3 I 3.0 I 3.4 I Higher SystendEvolution Name: I Category:

I Conduct of Operations KA Statement:

Knowledge of shift turnover practices -

C. DAY shift on July 25 D. MID shift on July 26.

Answer: D Explanation:

D is correct - Per Section 4.8.3 of OP-AA-I 12-101. The on-coming 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 inc~rrcct- For the reasons described above. This choice would be c o m c t if the procedurc read.. .whichever is m.

B and C are incorrect - For the reasons associated with the correct answer. These choices presumc the candidate incorrectly recalls a preceding five days logs requirement. Additionally. theses choices cause the candidate to ponder the meaning [if preceding five days.

Objective: I Question Source: I Level of Difficulty:

PBADOl2, Objective 5 New I 2.6 References provided to examinee: I None

References:

I OP-AA-I 12-101, Shift Relief and Turnover 1 Date Written: [ 02/22/05 I Author: I Ryder I Comments: None I

~ ~~

ModifiedNRC UNSAT. Original auestion felt to have no correct answer depending on how the words preceding 4 days is in;erpreted (ie 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> OR any part of the last 4 calander days). Changed distractor B and D to to line up with shifts vice absolute times.

I Question # I 28 I RO/SRO: I Tier: I Group: I KA: 1 ROIR I SROIR: I Cog Level Both I 3 I Generics I 2.2.28 I 2.6 I 3.5 I Higher SystelnlEvolutiun Name: I Category:

I Equipment Control KA Statement:

Knowledge of new and spent fuel movement procedures A refueling outage is in progress, when THREE of the SRM Channels fall below 3 counts per second.

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.

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

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

D. 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.

Answer: A Explanation:

THE QUESTION IS JUSTIFIED AS AN RO QUESTION DUE TO THE PRESENCE OF A FACILITY SPECIFIC LEARNING OBJECTIVE. REFER TO LP86610 OBJECTIVE 1.19 A is correct - Per CPS 3703.01, Section 6.IO, the inoperable SRMs cited in the question stem (00 matter which 3 SRM Channels), would require a halt to Core Alterations Per CPS 3703.01, Section 2.2.4. however, this answer choice does NOT describe an evolution that is considered a Core Alteration (i.e.. the evolution meets the criteria of the 2" exception ( ~ b in ) the CORE ALTERATION definition). merefore, this work CAN hc performed with 3 inoperable SRMs. NOTE This claim has been verified against the SRM Tech Spec (3.3.1.2). as well (see that reference.

attached).

B. C. and D me incorrect - Each of thesc choices a Core Alteration. as defined in Section 2.2.4. Therefore. these describe work that CANNOT he pcrformed with 3 inoperable SRMs.

Objective: I Question Source: I Level of Difficulty:

LP86610.1.19 New I 2.9

References:

CPS 3703.01, Core Alterations CPS Tech Spec 3.3.1.2. SRM Instrumentation

1 Question # 128 1 RO/SRO: I Tier: 1 Group: 1 KA: I ROIR 1 SROIR 1 Cog Level Both I 3 I Generics I 2.2.28 I 2.6 I 3.5 I Higher SystedEvolution Name: 1 Category:

I Equipment Control

.md s p a 1 iucl mii\cmenl prorrdures

~

I. The candidate must associate the stems three failed SRMs with the requirements of Section 6.10 of the procedure.

2. The candidate must determine whether any given one of the answer choices idis not a type of Core Alteration.

ModifiedNRC judged UNSAT. Felt to NOT be RO required knowledge. Justified via facility specific learning objective. See specifics in Explanation.. GDSetser 6/13/05

ROISRO. 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 SystemIEvolution Name: I Category:

I Radiological Controls KA Statement:

Knowledge of radiation exposure limits and contamination control, including permissihle levels in excess of those A Site Area Emergency is in progress, when it is determined that an Emergency Exposure is required SOLELY for the purpose of PROTECTING an important piece of PLANT EQUIPMENT.

Which ONE of the following:

(1) identifies the exposure LIMIT (TEDE) for this emergency exposure, gl J

(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. 10 Rem TEDE is the limit for solely protecting property. Part (2): Per EP-AA-I 13-F-02. all Emergency Exposures (no matter the specific limit) require Station Emergency Director authorization.

A, C. and Dare incorrccf - For the reasons described above; but all have face validity and are plausihlc to an uncertain candidate.

Objective: I Question Source: I Level of Dificulty:

None New I 2.8

References:

RP-AA-203, Exposure Control and Authorization EP-AA-I 13-F-02. Authorization for Emergency Exposure

I Question # I 29 I RO/SRO: Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 3 1 Generics I 2.3.4 I 2.5 I 3.1 I Lower SystenJEvolutionName: I Category:

I Radiological Controls KA Statement:

Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized

I Question # 1 30 I RO/SRO: I Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 3 I Generics I 2.4.1 I 3.1 I 3.8 I Higher SystemlEvolution Name: I Category:

I Emergency Procedures and Plan KA Statement:

Knowledge of event based EOP mitigation strategies Using the provided references, answer the following.

A transient has occurred requiring EOP entry and execution.

THEN, main condenser vacuum is completely lost and CANNOT be restored Which ONE of the following mitigation strategies IS ALLOWED by the EOPs?

Using the ...

A. RFPTs to stabilize pressure if needed.

B. MSL Drains to depressurize to the Shutdown Cooling pressure interlock once Cold Shutdown Boron has been injected.

C. Condenser as an Alternate Depressurization System if needed D. MSL Drains to stabilize pressure if needed.

Answer: C Explanation:

C is correct - Stem conditions are not specific regarding which EOPs are being executed only that their execution is in progress. Per CPS 441 I .W. Section 2.2 NOTE. Without a condenser vacuum, the EOPs allow the condenser to be used as a vent path (i.e., an alternate depressurization path). but not as a heat sink. Additionally. the stem conditions indicate that MSlVs (Group I ) and Turbine Bypass Valves have closed on the loss of vacuum (see LP852.5.5,page 20).

Therefore. only where an EOP step states that it is 'OK to defeat RPV vent interlocks' can operators defeat the low vacuum MSIV closure. In this case, this is true if Alternate Depressurization Systems are needed.

A. B, and D are incorrect - For the remow described above. These choices are suggest using the condenser as heat sink when there is no vacuum. These choices are all taken from the Pressure leg of EOP- IA.

Objective: I Question Source: I Level of Difficulty:

None New I 3.9 I EOP-2. RPV Flooding I EOP-3, Emergency RPV Depressurization (Blowdown)

CPS 441O.oOCoO4, Defeating MSUOG Interlocks CPS 441 1 .W, RPV Pressure Control Sources LP8.5255. Condenser Air Removal System

ROISRO: I Tier: I Group: I KA: I R O I R I SROIR I Cog Level Both I 3 I Generics I 2.4.7 I 3.1 I 3.8 I Higher SystemflSvolution Name: I Category:

KA Statement:

I Emergency Procedures and Plan -

Knowledge of event based EOP mitigation strategies Date Written: I 02/24/05 I Author: I Ryder Comments: None MODIFIEDlNRC UNSAT - Original question deemed UNSAT. NRC felt that correct answer would require an unsupported inference that EOP-2, or -3 had been entered. Stremlined stem to NOT be specific as to what EOPs are being executed only that they are. Added phrase if needed to A,C,D. Deleted reference to EOP-2 and EOP-3 in C. GDSetser 6/14/05

I Question# I31 ]

RO/SRO. I Tier: 1 Group: I KA: I ROIR I SROIR I Cog Level Both I 3 I Generics I 2.4.24 I 3.3 I 3.7 I Lower SystemlEvolution Name: I Category:

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 systedequipment considered 'most critical for worst case survivability'?

A. RCIC B. DGIC C. HPCS D. FC Answer: A Explanation:

A is correct - Per CPS 4303.01, Appendix A, Section 1. RCIC is the system considered 'mosl critical for worst case survivability' B and C arc incorrect - But either is plausible to the Candidate who recalls that, per CPS 4200.01. Seclion I .4. HPCS i s the prefcrrcd source of RPV makeup, given that DG IC (Div 3 power) is assumed to be availahle during a SBO.

D is incorrect - BUI provides sufficient plausibility for the Candidate who leans towards giving a greafcr priorily to keeping fission prtxlucts in solution in the Spent Fuel Pool Storage. given that the slem conditions indicate there is 00 RPV inventory conlrol problem &e.. no LOCA).

Objective: I Question Source: I Level of Difficulty:

None New I 3.1 Date Written: I 05/02/05 I Author: I Ryder Comments: None

I Question# I 32 KO/SKO I Tier: 1 Group: KA: [ ROIR: 1 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) I plant Systems KA Statement:

Ability to monitor automatic operations of the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) including: Valve operation 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, IE22-FO04, and HPCS Test Valve to Suppression Pool, IE22-FO23 B. HPCS to CNMT Outboard Isolation Valve, IE22-FO04, and HPCS Second Test Valve to Storage Tank, 1E22-FOll C. HPCS Suction from RCIC Storage Tank Valve, lE22-FO01, and HPCS First Test Valve to Storage Tank, lE22-FO10 D. HPCS Suction from RCIC Storage Tank Valve, IE22-FO01, and HPCS Suppression Pool Suction Valve. 1E22-FO15 Answer: B Explanation:

B is correct - Per CPS 9051.01, Section 8.2, CPS 3309.01, Section 8 1.2. and LP85380, Figure 4. The flowpath for this surveillance is Storage Tank-to-Storage Tank (a Suppression Pool-to-Suppression Pool). Upon receipt of the HPCS 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 FOI I ) stroke closed (show intermediate position).

A is incorrect - 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 throughout the initiation (until an automatic suction swap. if any. occurs later on).

D is incorrect - As descrihed for choice 'C', Fool is already open and remains that way. Interlocks prevent FOOl and FO15 from ever stroking at the same Lime.

Objective: I Question Source: I Level of Difficulty:

LP85380.1.8 NCW I 2.6

1 Question# I 32 1 KOISRO Tier: I Group: 1 KA: 1 ROIR: 1 SROIR: I Cog Level Both I 2 I I 1 209002A3.01 1 3.3 I 3.3 I Higher SystenJEvolution Name: I Category:

High Pressure Core Spray System (HPCS) I Plant Systems KA Statement:

Ability to monitor automatic operations of the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) including: Valve operation References provided to examinee: None

References:

LP85380, High Pressure Core Spray CPS 3309.01, High Pressure Core Spray CPS 9051.01. HPCS Pump Operability This question is categorized as Higher Cognitive (HCL) because the Candidate must associate the pre-LOCA lineup of HPCS (what does the 'Preferred' test lineup mean?) with the specific valves that are necessary to realign (out of the test lineup), post-LOCA.

RO/SRO I Tier: I Group: I KA: I ROIR. I S R O I R I Cog Level Both I 2 I I 1 2180WA3.08 I 4.2 I 4.3 I Higher

- SystemlEvolution Name: I Category:

- Automatic Depressurization System I Plant Systems KA Statement:

Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including: Reactor pressure A. With reactor pressure initially at 900 psig, operators are unable to Inhibit ADS, and ALL of the ADS Valves automatically open.

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

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

D. 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.

Answer: A Explanation:

A is correct - Per LP85239, Attachments F and G,and page IO. Each of the 16 SRVs are capable of relieving reactor pressure at the same rate (about 6.5% of total rated steam flow). There are 7 SRVs that open when operators initiate ADS (also found in LP852 18, page 4).

B is incorrect - Per CPS 3105.04. Section 2. I . 1. Tofal relief capacity is 28.890. for the 6 TBVs, 01U S % for each TBV (as compared to 6,590 for each SRV). Additionally, only 6 TBVs open, as compared to 7 SRVs (when ADS is initiatcd).

C is incorrect - Per LP85239, Attachment G. SRV F051D has a pressure-relief setpoint low enough ( I 103 psig +/- 15 psig, per Tech Spec 3.3.6.5) to open the SRV if pressure peaks at only 1089 psig. As soon 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 +/- 15 psig) low enough to open. shortly after FO51 D opened. T h i s choice, therefore, proposes an event where no more rhan a lutd of 2 SRVs (and n ~ n ofe the 6 TBVs) open to effect a pressure reduction.

D is incorrect - Per LP85241, page 6. the Pressure Regulator is normally set at 930 psig (steampressure. 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 re-opens NONE of these LLS-SRVs. and NONE ofthe 6 TBVs (i.e.. Slearn pressure is still well below its 930 psig Pressure Regulator setpoint).

Objective: I Question Source: I Level of Difficulty:

None New I 3.0

I Question # I 33 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR I Cog Level Both I 2 I I I 218000A3.08 1 4.2 I 43 I Higher System!Evolution Name: I Category:

1 Automatic Depressurization System I Plant Systems KA Statement:

Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including: Reactor References provided to examinee: None

References:

LP852 18, Automatic Depressurization System LP85239. Main Steam System LP85241, Steam Bypass and Pressure Control System CPS Tech Spec 3.3.6.5, Relief and LLS Instrumentalion CPS 3105.04, Steam Bypass and Pressure Regulator

1 Question # I34 I ROISRO: I Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I I [ 2180002.4.31 I 3.3 I 3.4 I Lower SystemlEvolution Name: 1 Catesory:

Automatic Depressurization System I Plant Systems K A Statement:

Knowledge of annunciator alarms and indications, and use of response instructions 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. 5066-5A2,ADS LOGIC B 105 SEC TIMER INITIATED C. 5064-7C, ECCS FLOOR DRAIN SUMP HIGH LEAK RATE D. 5064-5C, SUPPR POOL DIVISION I HIGH TEMPERATURE Answer: B Explanation:

B is correct - Per CPS 5066-SA. This annunciator alarms only when a high drywell pressure (1.68 psig) and confirmed low-low-low water level (-145.5 inches) condition (or confirmed low-low-low level fur >6 minutes) exists. These parameters represent EOP-I and EOP-6 entry conditions. The procedures Operator Actions remind the operators of this.

A is incorrect - Per CPS 5065-6F. This annunciator represents area high temperature &m values, the m u normal values associated with EOP-8 entry C is incorrect - Per CPS 5064-7C.This annunciator monitors the status of various ECCS room (in secondary containment) floor drain sump systems (i.e.. high leakage resulting in excessive pump-down time and/or frequency.

Although this leakage problem may progress to where floor water levels in those rooms require EOP-8 entry. water levels are not there yet. As such, no EOP entry yet exists.

D is incorrect - Per CPS 5064-5C. T h i s annunciator represents a 90°F suppression pool temperature. The EOP-6 entry value is 95°F. As such, no EOP entry yet exists.

References provided to examinee: None

References:

LP852 18, Automatic Depressurization System CPS 50645C response procedure CPS 5064-7C response procedure CPS 5065-6F response procedure CPS 5066-SA response procedure

1 Question# 135 I ROISRO: I Tier: I Group: I K A I R O I R I S R O I R I Cog Level Both I 2 I I I 223002K3.22 I 2.5 I 2.6 I Higher SystenJEvolution Name: I Category:

PCIS/Nuclear Steam Supply Shutoff System (NSSSS) 1 ~ l a nSystems t

KA Statement:

Knowledee ofthe effect that a loss or malfunction of the PCIS/NSSSS will have on the following: Containment 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. INBOARD Containment Equipment. Drain Sump Discharge Valve, IRE021 B. OUTBOARD Containment Equipment Drain Sump Discharge Valve, IRE022 C. OUTBOARD Containment Building Supply Isolation Valve, lVROOlA D. INBOARD Containment Building Supply Isolation Bypass Valve, 1VR002B Answer: A Explanation:

A is correct - Per CPS 400I.02COOl. page 7, this is aGroup 8 valve. Per LP85407, page 5 8 . Div 2 NSPS services the Load Driver fnr the Inboard valves. A ZERO input to the Load Driver produces an energized function to close the Inboard valves.

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

D is incorrect - Although an Inboard (Div 2) valve. this valve belongs to Group 9 (see CPS 4001 . O X O O I . page 9).

Group 9 is not among the other Groups (10, 12. 16, 19, and 20) that are serviced by the same Load Drivers a Groups 8 and 15.

Objective: I Question Source: I Level of Dificulty:

LP85407. I .7 New I 3.0

References:

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

RO/SRO I Tier: I Group: I KA: I R O I R 1 S R O I R I Cog Level Both I 2 I I I 223002K3.22 I

~ 2.5 I 2.6 I Higher SystemEvolution Name: I Category:

PCIS/Nuclear Steam Supply Shutoff Systcrn (NSSSS) I plant Systems KA Statement:

Knowledge of the effect that a loss or malfunction ofthe PCISINSSSS will have on the following: Containment all ofthe Group 8 valves during that JPM performance.

3. Therefore, it i s m unlikely that the following scenario will occur. ..the RO Candidate easily discounts the

'D' dislracter because he/she 'remembers' (from the JPM performance) that IVR002B is not a Group 8 valve.

There is 00 overlap with the Operating Exam.

1 Question # 136 I RO/SRO. I Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I 2 I 223001 2.4.6 I 3.1 I 4.0 I Higher SystemlEvnlutinn Name: I Category:

Primary Containment System and Auxiliaries I Plant Systems KA Statement:

Knowledge of symptom based EOP strategies Using the provided references, answer the following.

Operators are implementing EOP-1, RPV 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 after warm-up, and Containment hydrogen reads 9%.

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

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

D. Hydrogen Mixers are still OFF, THEN hydrogen monitors come on-line after warm-up, with Drywell hydrogen reading <O.S% and Containment hydrogen declared to be UNKNOWN Answer: B Explanation:

B is correct - Once EOP-7 is entered (in this case, on detectable hydrogen). only EOP-7 can cause them to he started.

With both drywell and containment hydrogen reading 2%, operators proceed straight down the left-leg of EOP-7.

proceed through the left-most WAIT step and start the Mixers.

A is incorrect - T h i s choice puts the Candidate solidly into EOP-7. specifically at the right-most leg of EOP-7. where with the Igniters still OFF, operators arc directed to stop the Mixers and prevent igniter restart.

C is incorrect - Per EOP-7, Figure R. thc plant is above the Deflagration Limit, requiring the implementation of the top-most overridc step and the execution of the fight-most leg, where opcrafors are directed to prevent igniter restart and stop the Mixers.

D is incorrect - Once EOP-7 is entered (in this case, because of UNKNOWN Containment hydrogen). only EOP-7 can c a u x them to be s t a n d This choice is similar to the correct answer choice, 'B', except that the Candidate is still WAITING for detectable Drywell hydrogen ( 5 % or higher) before the Mixers can be started.

ROISRO: 1 Tier: I Group: I KA: I ROIR I SROIR I Cog Level Both I 2 I 2 I 223001 2.4.6 I 3.1 I 4.0 I Higher SystemlEvolution Name: I Category:

Primary Containment System and Auxiliaries I Plant Systems KA Statement:

Knowledge of symptom based EOP slrategies

References:

CPS EOP-6, Primary Containment Control CPS EOP-7, Hydrogen Control

I Question # I 37 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel n..L DUUl I I

I I ,

I IA7,VvL n < M I fAi RY II . rU nI 3.5 I 3.6 I Higher I Category:

SysteWEvolution Name:

Partial or Complete Loss of Forced Core Flow Circulation I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE I

LOSS OF FORCED CORE FLOW CIRCULATION: Natural circulation 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. FC will have to be lined up for Alternate Shutdown Cooling.

B. RCIC will have to be lined up for Alternate Shutdown Cooling.

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

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

Answer: C Explanation:

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

A is incorrect- Per CPS 4006.01. Table 2. T h i s lineup requires the RPV head be off (plant in Mode 5 ) .

B is incorrect - Per CPS 4006.01. Tahle 2. In Mode 4, temperature is 200°F or less. well below the 60 prig RCIC system isolation point.

D is incorrect - See the explanation for the correct answer, C

References:

LP85422, Reactor Vessel and lnternals CPS 400601. Loss of Shutdown Cooling Date Written: I 04/29/05 I Author: I Ryder Comments:

This closed-reference question is categorized as Higher Cognitive (HCL) because:

I. The correct answer is neither an Immediate Operator Action. nor a Precaution/Limitation. where the Candidate would be expected to recall such from memory.

2. The stem requires the Candidate to the loss of forced core circulation with the need 10 ensure adequate natural circulation as a substitute.

3. The elimination ofthe distracters requires the several associations described in the Explanation for each

1 Question # I37 1 RO/SRO I Tier: I Group: I KA: I R O I R I S R O I R I CngLevel Both I 1 I 1 I 295001 AKI.01 [ 3.5 I 3.6 I Higher SystemIEvolution Name: I Category:

Partial or Complete Loss of Forced Core Flow Circulation I Emergency and Abnormal Plant Evolutions KA StateTed:

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Natural circulation (above).

Additionally, this question is on the RO Exam (and is not reserved as SRO-only). because the question is E&I actually asking the Candidate to direct an action found in the Subsequent Actions section of the off-normal (Loss of SDC).

Rather, it is framed in such as way that it simply makes use of that Subsequent Action (in Section 4.6) to require the Candidate to demonstrate an understanding of the operational concern that results from having no forced reactor COOIant flow.

L Question # I 38 1 RO/SRO: 1 Tier: I Group: I KA: I I ROIR: SROIR: I cogLevei Both I I I I I 295005 AAl.01 I 3.1 I 3.3 I Higher SystemlEvolution Name: I Category:

Main Turbine Generator Trip I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to operate andlor monitor the following as they apply to MAIN TURBlNE 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. the GREEN lights are lit for RR Pump breakers CB5A and CBSB.

B. Total Core Flow indicates about 45 m l b d h r .

C. both RR FCVs indicate about 10% open.

D. both RR FCV LIMITER ERROR meters read upscale (POSITIVE).

Answer: A

~

Explanation:

A is correct - Per LP85202. pages 23.28, 29. and Figure 8. Whenever the turbine trip scram is enabled (abovc 33.3%

power, nominally). thc EOC-RPT trip is also enabled. The turbine trip automatically downshifts hoth RR Pumps to SLOW speed. Operators will see the red (CLOSE) light extinguish and the green (OPEN) light illuminate for each pump's 6.9KV (Fast speed) breaker, CBSA(B1.

B is incorrcct - Per CPS 9041.01. Figures la. Ib, and 2a. this is the expected total core flow that results from a Flow Control Valve Runhack to about 19% open indication with both RR Pumps still running in FAST speed. The resulting Total Core Flow indication with both pumps running at SLOW speed (after the EOC-RPT trip). with the FCV still nearly wide open, is only about 29 m l h d h r (this value obtained from simulator modeling).

C is incorrect - FCVs remain as is during thc downshift (typically. indicating about 76'35 open). Even if the post-scram levelllevel control transient were to result in a FCV Runback signal heing produced, the FCV will indicated about 19%

open. not IO% open. The '20 Seconds after' statement in the stem allows for the following: I ) the 5-second delay betwccn the scram signal and a generator reverse power trip (see LP85461. page 33). and 2 ) enough time thereafter to suggest that thc FCVs need some time to stoke to a 10% open position.

D is incorrect - The only time the Limiter Error can read on the Positive side of zero (mid-scale). is when a position indication failure has occurred (c.g., LVDT or RVDT feedback signal is sending false position information to the controller). Given that the FCVs haven't moved in this scenario. the Error will still read zero (mid-scale).

Ohjective: I Question Source: [ Level of Difficulty:

LP85202.1.10.7 New I 3.1

I Question# 138 I KA Statement:

Ability to operate andlor monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: Recirculation References provided to examinee: None

References:

LP85202. Reactor Recirculation System LP85461.Main Generator System CPS 9041.01, Jet Pump Operability Test

I Question # 1 39 I RO/SRO: I Tier: I Group: I KA: I ROIR. I SROIR: I ~ o g ~ e v e l Both I I I I I 295004 2.1.14 I 2.5 I 3.3 I Lower SystemlEvolution Name: I Category:

Partial or Complete Loss of D.C. Power I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of system status criteria whlch require notification of plant personnel 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 ProtectionDetection System equipment requiring that power source?

A. 1A B. 1B C. IC D. 1D Answer: B Explanation:

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

This impact is also described in LP85286. page 95 (DCMCC IB is Div II power).

A. C, and D are incorrect - For the reasons described above. Attached, here. are the Load Impact Lists for these 3 buses, verifying there is no direct cause-effect relationship with tire systems. when either of these 3 buses are lost.

References provided to examinee: I None

References:

I LP85286. Fire Protection and Detection CPS 4201.01, Loss of DC Power CPS 4201.01C001(2,3,4), Loss of 125VDC MCC IA(6. C, D) Load Impact Date Written: I 03/02/05 I Author: I Ryder Comments:

Although the SRO would direct the subsequent actions ofthis Abnormal Operating Procedure (4201.01 ). including identifying the need to contact Fire Protection for this bus loss, this question is, nonetheless. categorized as BOTH (RO and SRO) for the following reasons:

I. ROs are expected to recognize the maior imuact of all significant bus losses. especially those &!

(Divisional, Class IE) bus losses such as the DC MCCs IA-ID. without the need to consider what may or may not be described in a procedure.

2. The unisueness ofthe relationship between & bus loss (Division 2) and the fire systems. among these 4 buses, is essentially e knowledge. and is stated g&c& in the Fire Protection lesson plan. LP85286.

and istiedtoaLe2imingObjcctive(.1.10.12)forwhich~theROandSROare responsible.

ROISRO: 1 Tier: \ Group: 1 KA: \ ROIR. 1 SROIR. 1 ~ o e , ~ e v d Both I I I I I 295004 2.1.14 I 2.5 I 3.3 I Lower SystenVEvolution Name: I Category:

Partial or Complete Loss of D.C. Power I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of system status criteria which require notification of plant personnel

I Question# I 40 I ROISRO: I Tier: I Group: I KA: I ROIR SROIK: I CogLevel Rnth I I I I I ---- . ..._

795f!4hAK707

. . I An I AI I 1 nwer SystellllEvolution Name: I Category:

SCRAM I Emergency and Abnormal Plant Evolutions KA Statement:

n  :

Reactor pressure control Following a reactor scram from 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 operators should lower reactor pressure in order to minimize the Feedwater-to-RPV Differential Temperature while cooling down?

A. 800psig B. 700psig C. 600psig D. 500psig Answer: C Explanation:

C is correct - Per CPS 4100.01, Section 4.2 NOTE. Operators should lower reactor pressure to about 600 prig lo minimiLe the RPV-to-FW delta-T that exists during a plant cooldown.

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

Objective: I Question Source: I Level of DiRculty:

None New I 2.7 References provided to examinee: I None

References:

I CPS 4100.01. Reactor Scram At first glance. this question appears to have a potential Operating Exam overlap problem. While it is tNC that one or mnre Simulator Scenarios may progress to where a comparable pressure cnntrol band (550-650 psig) may be established for EOP mitigation strategy purposes, this is purely coincidental relative to this written exam question.

This question is framed entirely in the context of EOP-free. post-scram pressure cnntrol strategies. directed solely from the SCRAM abnormal operating procedure (4100.01). This question does NOT overlap with the Operating Exam.

I Question# 141 I RO/SRO I Tier: I Group: I KA: I ROIR. I SKOIR: I CogLevel Both I I I I I 295016 2.1.32 I 3.4 I 3.8 I Lower SystemlEvolution Name: I Category:

Control Room Abandonment I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to explain and apply system limits and precautions 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, 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.

B. BEFORE starting SX Pump lA, CLOSE the PSW To SSW 1A Header Isolation Valve, 1SX014A; this preempts the consequence of a potential hot short condition that might prevent the valve from automatically closing.

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

D. BEFORE attempting to place RHR B in Shutdown Cooling, VERIFY that Shutdown Cooling Outboard Suction Isolation, lE12-FO08, CAN BE OPENED; the valve may be disabled due to a hot short condition Answer: C Explanation:

C is correct - Per CPS 4003.01COI I , Section 3.0. There is no RHR Pump B flow indication or motor amps indication at the Remote Shutdown Panel (RSP). This is the only way to determine pump flow rate. should such information be needed. Suspicion of degraded flow conditions would be a need for such information: hence, the required action.

A is incorrect - Per CPS 4003.01C002. Section 3.5, running the pump on Recirc (min flow) should be limited to <20 seconds.

B is incorrect - Refer to CPS 4003.01C005. Section 3.0. There is no such requirement for manually shutting this valve before the pump start; rather, the operator is directed to verify it auto-closes when the pump starts.

D is incorrect - Refer to CPS 4003.01CO13, Section 1.4. Operators are cautioned lo beware that a hot short could have disabled IE12-F009 (the Inboard SDC isolation). Although valves, IE12-FO09 4F008 are potentially vulnerable to hot short problems. notice Section 4.8 of this procedure. Operators must unlock and close (place to ON) the motor breaker for F008. The way that CPS has addressed the hot short problem common to both valves (F008 and F009) is by keeping the F008 breaker locked open during normal plant operating conditions (see CPS 33 12.03.

Section 6.3). Hence, the Remote Shutdown Div 2 SDC procedure (4301 .OICO13) is silent on the need for operators lo be concerned about the status of F008. The valve is administratively protected from hot short vulnerability.

I Question# 141 I ROISKO: I Tier: I Group: I KA: I ROIR. I S R O I R I CogLevel Both I I I I I 2950162.1.32 I 3.4 I 3.8 I Lower SystemIEvolution Name: I Category:

Control Room Abandonment I Emergency and Abnormal Plant Evolutions KA Statement:

Ahilir) i o exp1.m md appl) syslcm I i m l l ~and precaurlonr References provided to examinee: None

References:

CPS 4003.01, Remote Shutdown CPS 4003.01C002. RSP - RClC Operation CPS 4003.01C005, RSP - Div I SX operation CPS 4003.01C01 I. RSP - Div 2 LPCl Operation CPS 4003.01C013. RSP - Div 2 Shutdown Cooling Operation

I Question# I 42 I RO/SRO: I Tier: I Group: 1 KA: 1 R O I R 1 S R O I R 1 CogLevel Both I I I I I 295019AK2.1I I 2.5 I 2.6 I Higher SystemlEvolution Name: I Category:

Partial or Complete Loss of Instrument Air I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Radwaste 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 planthystem impact of this loss of air?

A. FC Demineralizers isolate.

B. Main condenser vacuum slowly degrades.

C. WO Chillers automatically shut down.

D. ACTUAL D/P's on the CW Traveling Screens RISE.

Answer: A Explanation:

A is correct - Per CPS 5041-4B. the Fuel Pool Cleanup (FC) demins isolate B is incorrect- Per CPS 5041-48. Because ofthe failure mode ofthe IN66-FO60 valve (fails 'as is') Off-gas is unaffccted. and thereforc condenser vacuum is unaffected. This choicc is attractive 10 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 OUI of loop seal. and a loss of vacuum, IF the MVP were in service. With the plant at rated power. the MVPs are nol in service, and they are isolated.

C is incorrect -Per 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 incorrect- Per CPS 5041-5E. 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 a%designed. The choice is distracting in that is describes how ACTUAL screen d/p would trend if air were lost to the bubblers ...i.e., screens would aauto-start (or auto-shift to Fast speed) on high d/p; consequently, ACTUAL screen d/p will rise.

Sec Figure I of LP85301 for a simplified view of how these air headers are arranged

References:

LP85301, Service and Instrument Air CPS 5041-49,4C, and 5E. a l m response procedures Date Written: I 04/29/05 I Author: I Ryder Comments: None

1 Question # I 4 3 1 RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I 1 I I I 295023AK2.02 [ 2.9 I 3.2 I Higher SystemIEvolution Name: I Category:

Refueling Accidents I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of the interrelations between REFUELING ACCIDENTS and the following: Fuel pool cooling and cleanup 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. Area radiation levels on CNMT el. 828' will rise with EACH fuel bundle transferred.

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

C. Spent Fuel Storage Pool level will LOWER with EACH fuel bundle placed in the pool.

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

Answer: A Explanation:

A is correct - Per LP85233, pages 22. 39. and 46. Without forced FC flow, there is no way to replenish the 1.ooO gallons (approximate) removed from the Containment Transfer Pool (on CNMT el. 828') each time the IFTS transfer tube is flooded for carrying a fuel carriage down to the Fuel Building Transfer Pool. As Containment Transfer Pool level Iowcrs. with each fuel bundleh) transferred, area radiation levels will rise, creating a local radiological harard for personnel on CNMT el. 828'.

B is incnrrcct - Per CPS 3702.01. Section 2.1.5. the transfer tube is designed to handle fuel for up to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> without additional cooling.

C is incorrect - Water would be added to the Spent Fuel Pool from the Containmcnt Transfer Pool tending to RAISE vicc LOWER level.

D is incorrect - Per CPS 3702.01, Section 2.1.7. lethal radiation levels exist in these spaces whenever fuel is transiting the transfer tube. regardless of the status of FC.

1 Question # 43 I 1 RO/SRO I Tier: I Group: I KA: I ROIR. I SROIR: I CogLevel Both I I I I I 295023AK2.02 1 2.9 I 3.2 I Higher SystedEvolution Name: I Category:

Refueling Accidents I Emergency and Abnormal Plant Evolutlons KA Statement:

Knowledge of the interrelations between REFUELING ACCIDENTS and the following Fuel pool cooling and cleanup References provided to examinee: None

References:

LP85233. Fuel Pool Cooling andcleanup System CPS 3702.01, Inclined Fuel Transfer System (IFTS)

CPS 3317.01, Fuel Pool Cooling and Cleanup System

RO/SRO: I Tier: I Group: I KA: I R O I R I S R O I R I Cog Level Both I I I I I 295025 EK3.09 I 3.7 I 3.7 I Lower SystemlEvolution Name: I Category:

High Reactor Pressure I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of the reasons for the following responses as 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 Safety Relief Valves?

A. Overpressure Relief mode of SRV operation B. Low-Low Set initiation C. Overpressure Safety mode of SRV operation D. ADS initiation Answer: B Explanation:

B is correct - Per LP85239, page 16. The 5 LLS SRVs act to reduce the number of SRVs that cycle for given plan1 conditions. prolonging SRV life.

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

Objective: I Question Source: I Level of Difficulty:

None New I 2.1 References provided to examinee: I None

References:

I LP85239. Main Steam System Date Written: [ 03/04/05 I Author: [ Ryder Comments: None

1 Question# I 4 5 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 97'F and slowly rising?

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

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

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

D. 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.

Answer: B Explanation:

B is correct - Refer to EOP-IA, Power leg. Any attempt to slow down the rate of suppression pool temperature rise will delay the resuirement to start SLC before reaching the Boron Injection Temperature (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 distracter that points directly at suppression pool temperature. It provides sufficient distraction. in that it readily attracts one to a very familiar EOP Figure. familiar even to the weakest of Candidates; whcrc the other 3 choices demand a greater investment of time and analysis to determine their specific relationship with the given pool temperature.

C is incorrect - Refer to EOP-I, Figure Z. Even ifoperators are able to STOP the rise in pool temperature (by starting a11 available pool cooling, alone), and therefore stay far below 197'F. it does 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 still 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 temperature. It is not. Even if 'starling all available pool cooling' were able to allow pool temperature to rise no higher than, for example, 140°F, a low enough pool level (15. I feet, in this case) would still exceed the HCL, which is defined by this choice (a paraphrase of Section 12-H of the Bases).

OhLective: I Question Source: I Level of Difficulty:

LP87553.1.6.8 New I 3.7

I Question # 145 I RO/SRO: I Tier: I Group: I KA: I ROIR: I SROIR: I ~ o g ~ e v e l Both I I I I I 295026EK3.02 I 3.9 I 4.0 I Higher SystendEvolution Name: 1 Category:

Suppression Pool High Water Temperature 1 Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of the reasons for the following responses as they relate to SUPPRESSION POOL HIGH WATER TEMPERATURE Suppression pool cooling References provided to examinee: I EOP flowcharts

References:

............ ~

I CPS EOP- I , RPV Control CPS EOP-IA, ATWS RPV Control CPS EOP-6, Primary Containment Control CPS EOP Technical Bases Date Written: I 03/04/05 [ Author: I Ryder Comments:

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

1. The correct answer demands that the Candidate associate the given Suppression Pool Temperature (97°F and rising) with the Boron Injection Initiation Temperature (BIT) derived from Figure G of EOP-IA. The association must recognize that with only 97°F pool temperature, no matter what the power level, there is still time tn potentially delay boron injection by slowing down the rate of pool temperature rise.

2. 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 temperature is well below even the most limiting porlions of Figure P.

I Question# I 46 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I ~og~evel Both I 1 I I I 295027EA2.01 I 3.1 I 3.7 I Higher SystemlEvolution Name: I Category:

High Containment Temperature I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT TEMPERATURE Containment Temperature Using the provided references, answer the following.

A Station Blackout (SBO) is in progress.

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 165F and slowly rising on its ATM

2. Containment Pressure reads 1.5 PSIA and slowly rising on its ATM.

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

1Question# I 46 1 ROISRO: I Tier: I Group: I KA: I ROIR I S R O I R I CogLevel Both I I I I 1 295027EA2.01 I 3.7 I 3.7 I Higher SystedEvolution Name: 1 Category:

High Containment Temperature I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT TEMPERATURE Containment Temperature Explanation:

B is correct - Per CPS 4200.01C003, page 9. I&C personnel use a portable bridge connected to the permanent RTDs at designated cabinet terminal points, extrapolate the corresponding containment temperature and report it to control room operators. With 190F and slowly rising, operators are at the bottom-most IF-THEN step ofthe Containment Temperature leg.

A is incorrect - This would be the choice for a Candidate who believes: I ) the prescribed indicator (ATM) reads in units of absolute pressure (DOES NOT), and has power during an SBO (DOES). and 2) the Containment atmosphere is a saturated one, allowing operators to extrapolate containment temperature using the Steam Tables. The alleged corresponding containment temperature would be 116°F. If this were true, operators would continue to monitor containment temperature for a possible EOP-6 entry at 1 2 2 T C is incorrect - This would be the choice for a Candidate who believes that SPDS is available during an SBO (it is not).

If it were. with 122°F and slowly rising, operators would be at the top-most 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 not. If there were. at 165°F 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.

Ohjective: I Question Source: 1 Level ofDifficulty:

None New I 3.3 r References orovided to examinee: I EOP flowcharts I

References:

Steam Tables CPS 4200.01C003. Monitorina CNMT Temperatures During- a SBO I I EOP-6, Primary Containment Control Date Written: I 05/02/05 I Author: 1 Ryder Comments: None MODIFIED/NRC UNSAT. The choices related to sending personnel into Containment were deemed to be implausible. With that in mind, choice 2 was changed to provide an option that relates to a possible Containment pressure indication available during SBO, and choice 3 was changed to a temperature indication that is NOT available during an SBO. Data in the stem related to Containment Pressure and SP level was deleted since it is not needed to answer the question and introduced confusion.

GDSetser 6f 14/05

I Question # I 4 7 I Which ONE of the following events BY ITSELF requires an entry into one or more of the EOP FLOWCHARTS?

A. After a scram, the Shutdown Criteria are NOT met.

B. After a scram, reactor water level shrinks to +15 inches before recovering.

C. The VF Exhaust CAM, 1RIX-PR019,reaches its ALERT alarm point.

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

Answer: D Explanation:

D is correct - Per EOP-6. Drywell temperature entry is required at 150'F.

A is incorrect - Per EOP-I. Unless there is first an EOP-I entry (there may not he on a low-power scram). there is no way into EOP-IA for failure to meet shutdown criteria. Rather, the REACTOR SCRAM abnormal, CPS 4100.01, Section 4.3.1. directs operators to use one of the EOP support procedures, CPS 441 I .OS, to alternatively insert control rods.

B is incorrect- Per EOP- I . Unless level drops to Level 3 (+8.9 inches), no EOP entry is required. A low-power Scram may not shrink level that low.

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

1 I Objective:

LP87558.1.I I Question Source:

New I Level olDiIliculty:

2.6 References provided to examinee: I EOP flowcharts

References:

I EOP-I, RPV Control EOP-IA, ATWS RPV Control EOP-6. Primary Containment Control EOP-8, Secondary Containment Control CPS 4100.01, Reactor Scram Date Written: I 03/07/05 I Author: I Ryder Comments:

Although plant events on the Simulator Scenario portion of the Operating Exam will require operators to enter EOP-6.

at no time will the iniliatine event he one that requires operators to recognize a 150°F drywell temperature as, uniquely.

the reason for entry. As such, this written exam question does not overlap with the Operating Exam.

I I MODIFIED NRC Enhancement added words BY ITSELF to stem. GDSetser 6/13/05

1 Question# 148 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I I I I I 2950131 EK1.03 I 3.7 I 4.1 I Higher SystemIEvolution Name: I Category:

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

Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL Water level effects on reactor power Consider EOP-IA, ATWS RPV 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 power oscillations are NOT EXPECTED.

A. -40inches B. -65 inches C. - 140 inches D. - 162 inches Answer: B Explanation:

B is correct - Per EOP Technical Bases, EOP-IA, pages 5-14,5-15. 5-17, and 5-18, This level (-65)isjust below the level at which the feedwater sparger is fully uncovered, thus reducing the core inlet subcooling by 65-75%, a point where large-scale core instabilities are not expected to occur. However at this same level (-65), Void Fraction inside the core shroud is still relatively low. This is due to there s t i l l being a sufficient head of wntcr that sustajns a good amount of natural circulation (note: RR Pumps tripped at 4 Y ) . 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 continuc to lower level, will the natural circulation contribution be reduced to a point where the core voids outs (Le., a HIGH Void Fraction will exist inside thc core shroud).

A is incorrect - For the reasons described above. At this level (-40) RR Pumps are still running in SLOW speed (see EOP Tcchnical Bases, page 5-16). Regardless of comparing void contents, fuel damaging core instabilities (power oscillations) at high power and low flow conditions is $ aiJmajor concern because there remilins a good amount of core inlet subcooling with level this high above the feedwater spargcr.

C and D are incorrect- For the reasons associated with the correct answer. At these low levels (- 1 4 0 and -162). fuel damaging power oscillations are NOT expected. However, the whole purpose of intentionally controlling level this low (in either Band B or Band C) is to the void Fraction inside the shroud. thus keeping power down. Again, refertopagcs5-17 and5-18.

Objective: I Question Source: 1 Level of DiBculty:

LP87S53. I .3 New I 3.5 I CPS EOP Technical Bases

I Question # I 48 I RO/SRO: I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I 1 I I I 2950131 EK1.03 [ 3.7 I 4.1 I Higher SystedEvolution Name: I Category:

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

Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL Water level effects on reactor power Date Written: I 05/02/05 1 Author: I Ryder Comments: None

I Question# I 4 9 I RO/SRO: I Tier: I Group: 1 KA: I ROIR: I SROIR: I CogLeveI Both I I I 2 I 295022AA1.04 I 2.5 I 2.6 I Higher SystemlEvolution Name: I Category:

Loss of CRD Pumps I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to operate andlor monitor the following as they apply to LOSS OF CRD PUMPS: Reactor water cleanup system The plant is in MODE 4, with the following:

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 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?

A. CLOSE 1G33-FO41, Drain Flow to Condenser Bypass; THEN, CLOSE IG33-FO31, Drain Flow Orifice Bypass, and FULLY CLOSE IG33-FO33, Drain Flow Regulator.

B. CLOSE IG33-FO46, Drain Flow to Condenser; THEN, VERIFY CLOSED IG33-FO31, Drain Flow Orifice Bypass, and FULLY CLOSE IG33-FO33, Drain Flow Regulator.

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

D. AS NECESSARY to maintain the desired reactor water level, THROTTLE closed IG33-F031, Drain Flow Orifice Bypass.

Answer: A Explanation:

A is correct - Stem conditions indicate that reject is lined up per CPS 3303.01, Section 8. I .6.3. I , for reactor pressure 4 0 psig (Mode 4). with the low drain flow valve, FMI, open, the high drain flow valve, F046. slill shut. and the Drain Flow Orifice Bypass, F031, fully open. With a steady reactor water level being maintained. atrip of the CRD pump would require that the reject flow path be fully secured to maintain level. Therefore. operators implement Section 8.1.6.9. In this case. they will close F041, then close FO3I. and fully close F033.

B is incorrect - For the reasons described above. IG33-FM6 was ncl open for this reject flow lineup.

C is incorrect - Candidate is expected to recognize the pre-trip flow balance that existed (45 gpm in. 45 gpm out). As such, any attempt to throttle down on IG33-FO33 in an effort to maintain water level would take reject flow helow 13 gpm. Per CPS 3303.01, Section 8.1.6.3CAUTION (page 37). this may result in a system isolation on delta-flow.

RO/SRO: I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I I I 2 I 295022AA1.04 I 2.5 I 2.6 I Higher SystemlEvolution Name: I Category:

Loss of CRD Pumps I Emergency and Abnormal Plant Evolutions Objective: I Question Source: I Level of Difficulty:

LP85204.1.3.2 New I 3.7 References provided to examinee: I None

References:

I CPS 3303.01, Reactor Water Cleanup System I CPS 3304.01, Control Rod Hydraulic &Control (RD) I I Date Written: I 05/02/05 I Author: I Ryder

1 Question# I 50 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I 1 I 2 I 295029 2.1.33 I 3.4 I 4.0 I Lower SystemiEvolution Name: I Category:

High Suppression Pool Water Level I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications 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.

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

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

D. An electrical short and fire DESTROYS the MOTOR-OPERATOR for RHR B Shutdown Cooling Suction Valve, 1E12-FO06B; the fire is quickly extinguished.

Answer: C

~

Explanation:

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

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

B is incomect - Per Tech Spec Table 3.3. I.1-1. Function #8. SDV level transmitter trip function is required & in Modes I . 2. and 5(a).

D is incorrcct - 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).

Objective: I Question Source: I Level of Difficulty:

LP85223.1.16 New I 3.3 References provided to examinee: None

References:

LP85205, Residual Heat Removal System CPS Tech Spec 3.6.2.2, Suppression Pool Water Level CPS Tech Spec 3.5.3, RCIC CPS Tech Spec 3.3.1.1, RPS Instrumentation CPS Tech Spec 3.4.9, RHR Shutdown Cooling - Hot Shutdown

I Question# 151 1 Tier: 1 Group: 1 KA: I ROIR I SROIR: I cogLevei J Both I I I 2 H

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

Which ONE of the following identifies the EARLIEST time when operators are PERMI'lTED TO USE the monitors to determine Containment hydrogen concentration?

20 minutes after. ..

A. placing the OVEN TEMP SELECT switch in HIGH.

B. observing the OVEN TEMP ABNORMAL light extinguish C. the STARTUP CYCLE clock begins counting up.

D. RE-OPENING the associated containment isolation valves.

Answer: C Explanation:

C is correct - Refer to CPS 441 1 . I I , Section 2.1.5. In order to utilize monitor readings. there must have been 20-minutes run time (i.e.. sampling time) with the containment isolation valves open. Step 2.1. I has operators place the Oven Temp Sclcct switch in HIGH (at which time, the Oven Temp Abnormal light illuminates, signifying that B w a m -

up has begun). Step 2. I .2 has operators reopen the containment isolation valves (closed on the high drywell pressure scram signal). afler about 20-minutes (i.e.. when the Oven Temp Abnormal light extinguishes). At this point, the monitors are still &sampling. Only when operators depress the 'Enter' key in step 2.1.4.8 does the sampling time (run time) begin. coincident with thc f x t that the 270-second 'Startup Cycle' clock begins to count up.

A. B. and D are incorrect - For the reasons described above.

Objective: I Question Source: 1 Level of Difficulty:

LP85406.I .9 New I 3.0 References provided to examinee: CPS 441 1.1 1, Hydrogen Control System Operation, with the at the top of page 3 (Section 2. I ) redacted

I Question# I 5 1 I Date Written: I 05/16/05 I Author: I Ryder Comments:

This open-reference question is a 'Direct Lookup' type, for the following reasons:

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

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

For these same reasons, this question is also categorized as Higher Cognitive.

I Question # I 52 I RO/SRO. I Tier: I Group: I KA:

Both I I I 2 I 295008AKI.OI I 3.0 I 3.2 I Higher SystemlEvolution Name: I Category:

High Reactor Water Level 1 Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of the operational implications ofthe following concepts as they apply to HIGH REACTOR WATER LEVEL Moisture carryover The plant is operating at rated power when a Master Level Controller failure raises reactor water level AS HIGH AS +60 INCHES, and the Level 8 trip function fails.

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

A. NON-conservative core thermal power calculation B. Failure of SRVs to FULLY RE-SEAT after opening C. Main Turbine excessive vibrations D. Failure of MSIVs to FULLY CLOSE when needed Answer: C

~~~

Explanation:

C is correct - Refer to LP85422, Figure 21. Instrument Zero elevation is 787 feet. 6 inches. The elevation that corresponds to +60 inches &e.. +5 feet) is 792 feet. 6 inches. The main steam line nozzle elevation is 791 feet: this is 4 feet, 6 inches above where RPV level rose before operators inserted the scram. The only potential consequence of level this high is a reduced drying cfficiency of the moisture separator, and therefore, an increased amount of moisture canyover that could result in high vibration of the main turbine due to water-blade impingement.

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

B is incorrect - With water level getting n o where near the main steam lines, there is no chance for affccling 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 D/P's.

I None

~

References provided to examinee:

References:

[ LP85422, Reactor Vessel & lnternals Date Written: I 01/11/05 I Author: I Ryder Comments: None

I Question # I 53 I ROISRO: I Tier: I Group: I KA: I R O I R I S R O I R I CogLevel Both I I I 2 1 295002AK2.02 I 3. I I 3.2 I Higher SystemlEvolution Name: I Category:

Loss of Main Condenser Vacuum I Emergency and Abnormal Plant Evolutions KA Statement:

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

Circulating Water (CW) Pump B is tagged out for repairs 0 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. The impact of a potential loss of 6.9 KV Bus A B. Windage heating of the LP turbine last stage buckets C. Inability of CW Pump C to automatically trip on high condenser pit level D. The prompt exit from the Power/Flow Map CONTROLLED ENTRY REGION Answer: B Explanation:

B is correct - Post-transient conditions show the plant operating at something less than 300 MWe generator load with too low a condenscr vacuum ( < 2 6 Hg). Per CPS 4004.02, Section 4.1 . I , operating with thcre conditions should be avoided. Per PB400402, this light load-low vacuum condition causes overheating (due to windage) and moisture erosion of the LP turbine last slage buckets. Applicable calculation for post-transient generator load is as follows:

25% actual / 92.5%max allowed = x MWe / 1052 MWe nct x MWe = (.25/.925)X (1052) = 284 MWe A is incorrect - This is no more a concern. now. than it was before the manual trip of CW Pump C. The 6.9 KV A Bus powers both the A and C pumps: a single bus loss with only two pumps running would rcquire a scram. This choice suggests that that vulnerability is greater now (Le.. an IMMEDIATE overational concern) since the CPump is the only one remaining. This is not true.

C is incorrect - Per LP85275. page 28. the high pit level trip 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 incorrect - Pre-transient stem conditions (45% power. 50% rod line) are meant cause the Candidate to ponder the effect of having to reduce recirc flow (andlor rod insertions) in order to bring power down to the post-transient 25%.

Even without the P/F Map as open-reference, 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).

I Question # 1 53 I ROISRO 1 Tier: 1 Group: I KA: I ROIR: 1 snom 1 cog~pvei Both I 1 I 2 1 295002AK2.02 I 3.1 I 3.2 I Higher SystenJEvolution Name: I Category:

Loss of Main Condenser Vacuum I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of the interrelations between LOSS OF MAIN CONDENSER VACUUM and the following: Main turbine Objective: I Question Source: I Level of Difliculty:

None New I 3.0 References provided t o examinee: I None Reierenca: I PB400402, Loss of Vacuum I I LP85275. Circulatine Water System I CPS 4004.02, LOSS ofvacuum CPS 3005.01, Unit Power Changes (for the P/F Operaling Map)

Date Written: I 05/16/05 I Author: I Ryder Comments: None

I Question# I 54 I noma: I Tier: I Group: I KA: I nom I snom I CogLevel Both I 3 I Generics I 2.2.2 I 4.0 I 3.5 I Lower SystemIEvolution Name: I Category:

I Equipment Control KA Statement:

Ability to manipulate the console controls as required to operate the facility between shutdown and designated power Per OP-CL-108-101-1001, General Equipment Operating Requirements, which ONE of the following describes how the operator FULLY CLOSES a THROTTLABLE MOV from the main control room?

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.

B. Places the control switch in the CLOSED position for 1 second intervals, but JUST until the RED light extinguishes.

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

D. Places the control switch in the CLOSED position, JUST until the GREEN light illuminates.

Answer: A Explanation:

A is correct - Per OP-CL-108-IOI-1001, Section 3.5.1. This ensures the valve is fully closed.

B. C, are incorrect - Not the methods descfihed either in & procedure in other operating procedure.

D is incorrect, valve could still be open in this case.

Objective: I Question Source: [ Level of Difficulty:

None New I 2.0 References provided to examinee: I None

References:

I OP-CL-108-101-1001, General Equipment Operating Requirements Date Written: I 03/14/05 I Author: I Ryder Comments: None ModifiedNRC UNSAT. Original distractors B and D felt to be IMPLAUSIBLE due to I

reference to Re-opening a valve when the desire is to close it. Changed distractors B and D to only reference various iterations of closing but with reference to different indications. Added explanation for D. GDSetser 6/13/05

1 Question # I 55 1

- RO/SRO: I Tier: I Group: I KA: I ROIR I SROIR I Cog Level

- Both I 3 1 Generics I 2.3.9 I 2.5 I 3.4 I Lower SystenJEvolution Name: 1 Category:

I Radiological Controls KA Statement:

Knowledge of the process for performing a containment purge B. Containment Purge Mode C. Containment Vent Mode D. CCP UNFLTERED Mode Answer: D Explanation:

D is correct - Per CPS 3408.01, Sections 2.2.1 and 4.1. This is the normal operating containment purge mode foi Mode I. and should he used when painting is in progress inside Containment.

A is incorrect-Per CPS 3408.01. Sections 2.1 and2.l.l. Although this mode is avajlable in Mode 1. it uses the Drywell Purge Filter Trains (DWPFTs). and should be used because of the concern described in Section 4.1 (i.e..

reduced capacity of the charcoal filters due to these chemicals/paint fumes).

B is incorrect - Per CPS 3408.01, Section 2.2.2. this mode uses the Drywell Purge Filter Trains (DWPFTs), and should not he used because of the concern described in Section 4. I (Le.. reduced capacity of the charcoal filters due to these chemicalslpaint fumes). Additionally, this mode is only available in Modes 4 and 5 (see Sectiun 8.1.1.4).

C is incorrect- Although specified by CPS 3408.01.Section 2.2.1 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 8.1.1.3).

Ohjective: I Question Source: I Level of Difficulty:

LP85455.1.I4 New I 2.5 References provided to examinee: I None

References:

I CPS 3408.01, Containment Building/Drywell HVAC Date Written: I 04/29/05 I Author: I Ryder Comments: None

I Question# I 56 1 RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I ~og~evel Both I 3 I Generics I 2.1.25 I 2.8 I 3.1 I Higher SystemlEvolution Name: I Category:

I Conduct of Operations KA Statement:

Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance data 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 2) 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. 10 minutes D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 13 minutes Answer: A Explanation:

A is correct - Per CPS 3105.05, Section 8.5.2. VARs to the grid translates to a lagging power factor. At 20.020 volts, the generator is operating at 91% of rated (.91 x 22,wO = 20.020). Section 6. I Table shows rated voltage is 22,000. Operators determine that the maximum allowable MVA load is 91% of rated MVA. Section 6. I Table shows Rated MVA is I179 MVA. Therefore, the maximum allowable MVAs is 1073 MVA (-91 x I179 = 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 I applies (given that no operable cooling banks are available). Per the notes of Section 8.3. I , at 95°F air temperature. operators default to the k r temperature value (104°F). and the higher MVA load value ( I 140 MVA). The resulting time limit at this MVA load is 30 minutes.

I Question # 156 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLeveI Both I 3 I Generics I 2.1.25 I 2.8 I 3. I I Higher SystemlEvolution Name: I Category:

I Conduct of Operations KA Statement:

Ability to obtain and interpret station reference materials such as graphs. monographs, and tables which contain performance data B is incorrect - For the reasons described above. This choice is plausible if the Candidate ina~~rooriatelyinterpolates the 95F air temperature for the correct Max MVA load (1 140 MVA). 95'F is half the interval between 86°F and 104°F: 36 minutes is half the interval between 30 minutes and 42 minutes.

C and D are incorrect - For the reasons associated with the correct answer. These choices are analogous to choice ' A '

and 'E', respectively. They are plausible lo the Candidate who incorrectly translates the 'VARs to the grid' stem condition to a power factor. Per CPS 3105.05, Section 8.5.2. at 91 5% of terminal voltage. the MVA load is limited to 83% of rated MVAs. or 978 MVAs (.91 x .91 x 1179 = 978). This Candidate would then apply the '997.5' MVA IimitofCPS 3504.01, Section 8.3.1. Table 1.

References provided to examinee: CPS 3 105.05, Generator (TG), entire procedure CPS 3504.01,Main Power and UAT Cooling, entire procedure References : CPS 3105.05, Generator (TG)

CPS 3504.01. Main Power and UATCooling Date Written: I 04/29/05 I Author: I Ryder rnmmentq? N~~~

I Question # I 57 I RO/SRO: I Tier: I Group: I KA: I ROIR: I SROIR: I CoxLeveI Both I I I I 1 295003 AA2.02 I 4.2 I 4.3 I Higher SystemlEvolution Name: I Category:

Partial or Complete Loss of A.C. Power I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER: Reactor power. pressure. and level The plant has just scrammed from rated power, THEN the following occurs:

ALL power is lost from 4160V Bus I A l 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 AND ONE recorder on P870 Answer: A Explanation:

A is correct - Per CPS 4200.01, Appendix B. page 24. With 4160 V Bus 181 still powercd, there is a singkreactor pressure instrument still functional (lor t h e m reactor pressure band. 800-1065 psig). at P60l. That i n s t ~ m e n ist 8 2 I-R623B (a paperless recorder). The other recorder instrument. 1LR-SM016. shown in this Appendix is a low-range pressure instrument (0.300 psig, per LP85423, page 17). and is therefore available for determining reactor pressure.

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

C and D are incorrect - But are plausible to the Candidate who, in addition to remembering thc paperless recorders on P601, thinks there are also recorders on P870. This especially attractive to the ILT Candidate. where the Simulator uses a DCS Display screen o n P870 to allow access to key plant Emergency Response paramcters. including Reactor Pressure. This is for training. only: it is a the actual main control room configuration. Besides, this is a DCS Display computer point. and is oat a recorder.

I Objective:

None I Question Source:

New 1

I eve^ of Difficulty:

3.0 References provided to examinee: I None

References:

I LP85423, Nuclear Boiler Instrumentation Date Written: I 03/ 15/05 1 Author: I Ryder Comments:

There is 00 overlap between this question and any part of the Operating Exam. Although Candidates will have need to determine reactor pressure on a number of occasions in the Scenarios. especially, there is no particular time where they

I Question# I 5 7 I RO/SRO I Tier: I Group: I KA: I ROIR: I SKOIR: I CogLevel Both I I I I I 295003AA2.02 I 4.2 I 4.3 I Higher SystemlEvolution Name: I Category:

Partial or Complete Loss of A.C. Power 1 Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER: Reactor power. pressure, and level will have to seek out a yajQ! of other instruments, other than those usually-available (i.e.. DCS displays). to do so.

I Question# I 58 I RO/SRO I Tier: I Group: I KA: I R O I R : I SROIR: I ~ o g ~ e v e l Both I I I 1 I 295021 AK3.04 I 3.3 I 3.4 I Lower System/Evolution Name: I Category:

Loss of Shutdown Cooling I Emergency and Abnormal Plant Evolutions KA Statement:

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

Maximizing reactor water cleanup flow After a Loss of Shutdown Cooling, operators determine there is a NEED to MAXIMIZE the Bottom Head Drain Flow to RWCU, and to MAXIMIZE the cooling capability of RWCU.

Which ONE of the following describes the REASON for operating RWCU in this way?

A. To prevent excessive thermal stress of the Feedwater piping.

B. To minimize thermal stratification of the RPV.

C. To prevent erosion of the RPV bottom head drain line.

D. To minimize Feedwater line check valve flutter.

Answer: B 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 result 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 thermal 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) circulation through the vesscl and prevent bottom head region stratification.

A is incorrect - This choice is derived from another one ofthe 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 temperature when operating at low feedwater flow conditions. There is no cause-effect relationship between the way opcrators have decided to operate RWCU in the stem condition. and the limiting of this delta-T.

C is incorrect - This choice is derived from another RWCU operational concern. Specifically. CPS 3303.01, Section 6.10.1.I . When operators do maximizc thc bottom head drain flow (as suggested in the stem), thcy will necessarily have to comply with limiting that drain flow to 200 gpm, for the purpose of preventing drain linc erosion. However.

the REASON for operating RWCU in the way suggested is not directly related to this specific concern.

D is incorrect- Again. this choice is another opcrational concern. Specifically, CPS 3303.01. Section 6.1 I . Low flow conditions in the feedwater line can cause check valve flutter and excessive valve we=. Operators are advised to return a11 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 direct relationship to the REASON why the Operators have decided to operate RWCU. as suggested in the stem.

Objective: I Question Source: I Level of Difficulty:

LP85204.1.2.6 New I 2.9 References provided to examinee: I None

References:

I CPS 3303.01, Reactor Water Cleanup System

1Question # I 58 1 ROISRO: I Tier: I Group: I KA: I ROIR: I S R O I R I CogLevel Both I I I I I 295021 AK3.04 I 3.3 I 3.4 I Lower SystemlEvolution Name: I Category:

Loss of Shutdown Cooling 1 Emergency and Abnormal Plant Evolutions KA Statement:

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

Maximizing reactor water cleanup flow Dale Written: 1 03/15/05 1 Author: 1 Ryder Comments: None

1Question # I 59 1 ROISRO: I Tier: I Group: I KA: I ROIR: I S R O I R : I CogLevel Both I 1 I 1 I 295024EK2.08 I 4.0 I 4.1 I Lower SystemlEvolution Name: I Category:

High Drywell Pressure I Emergency and Abnormal Plant Evolutions KA Statement:

Knouledgc of ihc inlcmelmons hetween 111Gt1 DRYWELL PRESSURE and the t ~ ~ l l o u ~\ nDgS Assuming that all other logic pcrmisslves are satisfied, which ONE of the following identifies a COMBINATION of HIGH DRYWELL PRESSURE signals that w i l l automatically initiate ADS?

Coincident signals from transmitters..

A. A and C B. B and D C. B and F D. C and E Answer: C Explanation:

C i s correct - Per LP852i8, Figures 4 and 5 . A and E will initiate ADS (Figure 4. Division I logic). 6 and F will initiate ADS (Figure 5, Division 2 logic).

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

Objective: I Question Source: I Level of Difficulty:

LP85218.1.4.1 New I 2.8 References provided to examinee: I None

References:

I LP85218, Automatic Depressurization System Date Written: I 03/16/05 I Author: I Ryder Comments: None

I Question # I60 I 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 from operators in the field Which ONE of the following:

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

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

B. (1) TWO valves (2) STOP the running CY pump.

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

D. (1) TWO valves (2) START a second SF pump.

Answer: A Explanation:

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

(i.e., establishing makeup WITHOUT local operator assistance). however, the CAUTION Suggests that first starting a second CY pump may help prevent CY pump run-out.

B is incorrect - For the reasons described above.

C and D are incorrect - Refer to Section 8.3 of the procedure. These choices are plausible to the Candidate who cannot recall the specific concern for establishing makeup flow in this emergency mode. Section 8.3 utilizes Suppression Pool Cleanup and Transfer (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 available.

[ Question # I 60 1 Objective: I Question Source: I Level of Difficulty:

h

.. I 9 1 References provided to examinee: I None

References:

I CPS 3220.01, Suppression Pool Makeup (SM)

Date Written: I 03/17/05 I Author: I Ryder Comments:

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

1. The pump runout concerns presented in this procedure section are not found in the Precuations/Lirmtations section of the procedure. If they were, then arguably this question would only demand the lowest level of mental processing (simple recall of information) and would be categorized as Lower Cognitive (LCL).

2. Candidates are expected to have procedure sections. for evolutions such as this (i.e., important to protecting a lesser pump from a Dotential runout condition, but not a critical evolution vital to protecting the plant or to mitigating a significant transient), committed to memory.

3. As such, it is expected that m Candidates would have to use a higher level of mental processing to answer Part (2) of the question. This mental processing would be used to 'associate' concepts of potential impact on the system when opening the valves (Le.. recognition that this amounts to placing an tlowpath 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 method for protecting the pump from runout.

4. Accordingly. without applying simple recall (and this being a closed-reference question). several

'Fundamentals' concepts have to be applied to answer Part (2).

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

Operators enter EOP-IA 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. The SLC Pumps have been running for 30 minutes.

B. Suppression Pool Temperature approaches 14S"F.

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

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

Answer: B Explanation:

B is correct- Per EOP-6, Figure P (Heat Capacity Limit, HCL), as pool temperature nears 145°F (approximately). the HCL for a normal pool level (about 19 feet) i s threatened. The HCL is considered one of the 'Critical Parameters' mentioned in CPS 441 1.09, Section 4.3 (and 4.2). Even with an ATWS in progress, that procedure section allows operators lo lower the pressure control hand as necessary to stay within the HCL limil.

A is incorrect - CPS 441 1.W. Section 4.2 prohibits intentionally lowering pressure (in an ATWS) until 'specific reactivity shutdown conditions have been established. Refer to the PRESSURE leg of EOP-I A; specifically: the WAIT sign, and Table X. With both SLC pumps running for only '30 minutes' Cold Shutdown Boron has not yet been injected. Operators must remain at the WAIT sign for about IO more minutes before intenlionally lowering reactor pressure.

C and Dare incorrect -These choices suggest that operators wish to lower pressure simply to get below the shutoff head of either CondensateICondensate Booster Pumps (CD/CB) (-725 psig), or one or more of the low pressure ECCS pumps. CDICB is a 'Preferred ATWS injection system than might be used to maintain the prescribed level hand in EOP-IA, LEVEL leg. Law pressure ECCS (LPCULPCS) is an 'Alternate' ATWS injection that might be used, only i f level cannot be held above TAF (in lhis case. operators would he directed to blow down).

SCRAM Condition Present and Reactor Power Above Emergency and Abnormal Plant Evolutions KA Statement:

Ahtlit) tu dctcrmine a d o r interpret the follorring AS the) apply tu SCRAM COI\DI'I'IO'. PRESEUT AND REACTOR YON E R .ABOVE APRM D0WF;SCAI.F. OR UNKNOWN Rcdctur prerri.rc Ohjerthr: I Queslion Source: I LWCIof I)ifliculty:

~ ~ 8 I 77 9 s ~ Kew I 70 References provided to examinee: EOP flowchms

References:

CPS EOP-IA, ATWS RPV Control CPS EOP-6, Primary Containment Control

. CPS 441 1.09, RPV Pressure Control Sources

1 Question# I 62 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLeveI Both I 1 I I I 295038EA2.01 I 3.3 I 4.3 I LOWER SystemiEvolution Name: 1 Category:

High Off-Site Release Rate

~

I Emergency and Abnormal Plan1 Evolutions KA Stalement:

\hilit) to determine and/or interpret thc fdlouing as they apply to III(;H OFt-SITE RF.I.EASt KATE Off-r~le Per the Clinton Radiological Annex, a radiological release is considered to be an OFF SITE release at a MINIMUM of miles radius around the plant.

A. 0.20 B. 0.40 C. 0.50 D. 0.70 Answer: C Explanation:

C is correct - Refer to EP-AA- 1003, page CL-4. The perfect circle drawn around the Plant is a 0.5 mile radius and is defined as the Site Boundary.

A, B, are ~ 0 . miles 5 from the site boundary.

D while > 0.5 miles does not satisfy the stem condition of MINIMUM Objective: I Question Source: 1 Level of Difficulty:

None New I 3.6

References:

CPS-Offsite Dose Calculation Manual (ODCM)

EP-AA-1003, Clinton Radiological Annex Date Written: I 04/29/05 I Author: I Ryder Comments: This aueslion is suitable for a RO since it is asking . for a recognition for the Site Boundary from the

- only Administrative Emergency Planning procedures. I MODIFIED/NRC UNSAT- Deemed to be at the SRO Only level with no site specific learning objective to support RO knowledge. The question was changed to merely have the candidate recall the location of the site boundary. This also changed the question to LOWER cognitive level.

GDSetser 6/ 14/05

1 Question# I 63 I RO/SRO: I Tier: I Group: I KA: I R O I R I SROIR. I CogLevel Both I 1 I I I 600000AK1.02 I 2.9 I 3.1 I Lower SystemlEvolution Name: I Category:

Plant Fire On-Site

~~

I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of the operational implications of the following as they apply to PLANT FIRE ON-SITE Fire fighting 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. Start a SECOND fire pump, so that TWO are running B. OPEN the FP CNMT and FP Drywell Isolation Valves.

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

D. MANUALLY initiate Halon in the main control room Answer: C Explanation:

C is correct - Per CPS 1893.04, Section 8.1.2.1. P661 contains the Div I SRV logic. This action minimizes the risk that a fire in PO61 could result in the auto-opening of the SRVs (plant depressurization) with 'A' solenoid (Div I )

control switches (OFF-AUTO-OPEN) still in AUTO. Refer to LP85239. pages 14-15, A is incorrect -Per CPS 1893.04, Section 8.1.7. Only one fire pump need be running.

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

D is incorrect- There is 00 requirement for initiating Halon unless an actual tire exists.

References:

LP85239, Main Steam System CPS 1893.04, Fire Fighting Date Written: I 03/19/05 I Author: I Ryder Comments: None

1 Question # I 64 1 RO/SRO. I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I 1 I 2 I 295032EK3.03 I 3.8 I 3.9 I Lower SystenJEvolution Name: I Category:

High Secondary Containment Area Temperature I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledye of the reasons for the followmg responses as they apply to HIGH SECONDARY CONTAINMENT AREA I TEMPERATURE Isolating affected systems I Per EOP-8, Secondary Containment Control, which ONE of the following describes a situation where operators ARE PERMITTED to isolate the GIVEN plant system when its discharge IS THE CAUSE of a high area temperature?

A. RCIC, when it is needed for reactor pressure control in EOP-1.

B. RHR 'B', ANYTIME it is NOT needed for injection in EOP-1, EOP-lA, or EOP-2.

C. RWCU, when it is needed for reactor pressure control in EOP-1A.

D. MSL Drains, WHEN the RPV has been flooded to the Main Steam Lines in EOP-2.

Answer: D Cxplanation:

D is correct - Per EOP-2, bottom-most WAIT steps A and C are incorrect - Each of lhesc describes a situation where a system is 'needed for EOP actions. Isolating such a system is not allowed by EOP-8 (step 3 I).

B is incorrect - 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 not true.

t

~

Objective:

LP87559.1.2.2 I Question Source:

New I Level of Difficulty:

3.1 References provided to examinee: None; access to EOP flowcharts IS OK

References:

CPS EOP-I, RPV Control CPS EOP-IA, ATWS RPV Control CPS EOP-2. RPV Flooding CPS EOP-6. Primary Containment Control CPS EOP-8, Secondruy Containment Control Date Written: I 05/16/05 I Author: I Ryder Comments:

Given this KA. wc have taken an 'applications' approach as a means of creating a question that provides sufficient discrimination. To correctly answer this question, the Candidate must correctly apply the requirements of EOP-8. step

31. The fact that the WAIT steps at the bottom of EOP-2 &operators to isolate the MSL Drains does not detract from the fact that doing so is also consistent with the EOP-8. step 3 I , allowance.

RO/SRO: 1 Tier: I Group: I KA: I ROIR: I SROIR: I cog~evei Both I 2 I I I 205000A1.05 I 3.4 I 3.4 I Lower SystemlEvolution Name: I Category:

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

Ability to predict andlor monitor changes in parameters associated with operating the SHUTDOWN COOLING I SYSTEMiMODE controls including: Reactor water level I 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 plantkystem response when operators INITIATE WARMUP FLOW.

A. Reactor water level SUDDENLY UNCONTROLLABLY LOWERS.

B. Reactor pressure SUDDENLY UNCONTROLLABLY RISES.

C. RHR Pump RAPIDLY approaches RUNOUT.

D. RHR Pump RAPIDLY OVERHEATS Answer: A Explanation:

A is correct - Per CPS 3312.03. Section 5.4.At such a low pressure (<30 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 .T h i s could cause the downstream piping to drain to radwaste when operators open the radwaste discharge valves. If the pipe volume empties. and then the F03 I opens (likely. due to the downstream pressure now being relieved), a sudden, uncontrollable drop in reactor water level (level transient) will occur as that piping refills. See LP85205, Figure 5 .

B is incorrect - This choice has face validity, but is not a concern during this evolution.

C and D are incorrect- Candidate is expected to know that the RHR Pump is not yet running when 'warmup flow' is initiated.

Ohjective: I Question Source: [ Level of Difficulty:

LP85205. I.14 New I 3.0

References:

CPS 3312.03, RHR Shutdown Cooling LP85205. RHR

1 Question# I66 ]

RO/SRO I Tier: I Group: 1 KA: I R O I R I S R O I R I CogLevel Both I 2 I I I 215003K1.06 I 3.9 I 4.0 I Higher SystemlF.volution Name: I Category:

- Intermediate Range Monitor (IRM) System I Plant Systems KA Statement:

Knowledge of the physical connections andlor cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: APRM SCRAM signals Objective: I Question Source: I Level of Difliculty:

LP8441 1. I . 11. I New I 3.0

References:

LP8541 I , APRMILPRM System CPS 3306.01. Sourcellntermediate Range Monitors

I Question # I 66 I ROISRO: I Tier: I Group: I KA I ROIR: I SROIR. I c o g ~ e v e ~

Both I 2 I 1 I 215003K1.06 I 3.9 I 4.0 I Higher SystenVEvolutionName: I Category:

Intermediate Range Monitor (IRM) System I Plant Systems KA Statement:

Knowledge of the physical connections andlor cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: APRM SCRAM signals MODIFIEDNRC UNSAT. Nrc felt that if one assumes a 2% allowance for instrument tolerance, that the value originally provided was <13% therefore there was no correct answer.

Revised correct answer to clearly be >15%. Removed all reference to instrument Drift.

Revised other distractors to account for changed psychometrics. GDSetser 6/15/05

1 Question # I 6 7 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I I I 2 1 295017AA2.03 I 3. I I 3.9 I Lower SystemlEvolution Name:

High Off-Site Release Rate KA Statement:

Ability to determine andlor 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 and/or radiological significance of this alarm?

A. A Technical Specification entry is required.

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

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

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

Answer: C Explanation:

References provided to examinee: None

References:

CPS EOP-9. Radioactivity Release Control CPS 5140.41, alarm response for ORIX-PRO01 CPS 4979.01, Abnormal Release of Airborne Radioactivity Date Written: I 04/29/05 I Author: 1 Ryder Comments: None

I Question # 168 I The plant is operating at power when the following occurs:

0 CCW Effluent Monitor, IRK-PR037, alarms (a valid alarm) 0 IRIX-PRO37 is nearing its HIGH alarm setpoint 0 Source of the alarm is a tube leak on the ONLY available NRHX 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. Stop the RWCU Pumps and isolate the RWCU system.

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

Answer: C Explanation:

C is correct - Per CPS 5 140.49. Operator Action #I, and per CPS 3303.01. Section 8.3.3. Operation of RWCU with the plant above 120°F and CCW isolated from the NRHX cannot continue. operators are directed to remove RWCU from service.

A is incorrect - Per CPS 4001.02COOl. page 3. RWCU delta-flow isolation setpint is 59 gpm after 45 seconds. Per Computer Point E31DA001. the normal. at power, sensed RWCU differential flow (due to instmmcnt calibration impact by system flow dynamics) is about 25 gpm. 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 (i.e.. about 54 gpm). Tnere is no reason for operators to enter the Automatic Isolation procedure.

B is incorrect - For the reason associated with the correct answer, C D is incorrect - 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 aspect of this event. This NRHX tube leak does not constitute a RCS Pressure Boundary leak (defined in Tech Spec 1.1); therefore, Tech Spec 3.4.5.C does not apply here.

Although this Tech Spec is not a 3.0.3entry. this choice is worded in a way that suggests the need for a Candidate to recall shon-term (< I-hour) Tech Specs from memory.

NOTE: Each of these choices suggests that operators have determined the need to isolate the source of the release. The stem avoids explicitly suggesting this in order lo make choices A and D plausible. Tech Spec 3.4.5.A,alone, is sufficient to argue that isolation of the leak is in fact required.

I Question # I 68 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I 1 I I I 295018AAl.03 I 33 I 3.4 I Higher SystemlEvolution Name: I Category:

Partial or Complete Loss of Component Cooling Water I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to operate 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 Objective: I Question Source: [ Level of Difficulty:

LP85204 1.12 New I 2.6 References provided to examinee: None

References:

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, IRK-PR037. a l m response procedure Date Written: I 03/21/05 I Author: I Ryder Comments: None

1 Question# I 69 I RO/SRO: I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I 2 I 2 I 234000Al.03 I 3.4 I 3.9 I Lower SystemlEvolution Name: I Category:

Fuel Handling Equipment I plant Systems KA Statement:

Ability to predict andor monitor changes in parameters associated with operating the FUEL HANDLING EQUIPMENT controls including: Core reactivity level A fuel bundle is loaded on the REFUEL BRIDGE main hoist, operating in Automatic mode, with the following:

The Refuel Bridge is moving THEN. the ROD BLOCK #1 INTERLOCK light EXTINGUISHES on the flat screen alarm panel Which ONE of the following LOCATIONS has experienced the MOST RECENT change in reactivity?

A. Upper Containment Storage Pool Racks Zone B. lFTS Transfer Tube C. Containment Building Upender Zone D. Reactor core Answer: D Explanation:

D is correct - Refer to LP86614, pages 46 and 51. This status of this light is directly related to the bridge position (as sensed by a limit switch on the bridge motion track). The light extinguishes when the bridge motion is A W A Y FROM the reactor core.

A. B, and C are incorrecl- For the reason described above.

I I Objective:

LP86614.1.8.12 I Question Source:

New I Level of Difficulty:

3.0 References provided to examinee: I None

References:

I LP86614, Fuel Handling Equipment Date Written: I 05/20/05 I Author: I Ryder Comments: None

I Question# 1 70 I Radiation Monitoring System I Plant Systems KA Statement:

Knowledge of the effect that a loss or malfunction of the RADIATION MONITORING SYSTEM will have on the following: Control room ventilation Which ONE of the following describes a combination of MCR AIR INTAKE radiation monitor (lRI&PR009A,B,C,D) FAILURES that will automatically shift Control Room HVAC to the High Radiation Isolation Mode?

A. LO FAIL on the A monitor, with COMM FAIL on the B monitor.

B. HI FAIL on the D monitor, with AC POWER FAIL on the C monitor.

C. LO FAIL on the B monitor, with HI FAIL on the C monitor.

D. COMM FAIL on the A monitor, with AC POWER FAIL on the D monitor.

Answer: B Explanation:

B is correct - Per CPS 5 140.64, pages 2 and 3. The ladder logic for this system is: A or D, B or C. Of the several failure types presented in these 4 choices, the only one that does not input lo the isolation logic is the Comm Fail type.

A and D are incorrect - Comm Fail does not trip the isolation logic.

C is incorrect - These two monitors are on the same side of the ladder logic.

Objective: I Question Source: I Level of Dificulty:

LP85447.1.7.1 New I 2.6 References provided to examinee: I None

References:

I CPS 5140.64. IRIX-PR009A-D MCR Air Intake, alarm procedure Date Written: I 03/21/05 1 Author: I Ryder Comments: None

1 Question # I71 1 RO/SRO: I Tier: I Group: I KA: I ROIR. I SROIR: I CogLeveI Both I 2 I I I 223002 2.4.31 I 3.3 I 3.4 I Lower SystemlEvolution Name: 1 Category:

PCIS/Nuclear Steam Supply Shutoff System I Plant Systems KA Statement:

Knowledge of annunciator a l m s 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-lH, 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 Explanation:

A is correct - Per CPS 5067-6D. Operator Action 2.a. Although the Candidate is not expected to know this action from memory, helshe is expected to recognix that this annunciator alarms whenever any one of 9 separate analog trip modules (ATMs) becomes unseated (dislodged) in its rack, and that this produces a Div I trip for the associated parameter (e.&, a Group I isolation half-trip is generated if lB21-N681A (RPV Level I ) is the offending ATM).

B is incorrect - Per CPS 5067-1 H. 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.

C is incorrect - Per CPS 5067-78, This belongs to the Div I 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-SA. This belongs to ATMs associated solely with RClC is not affiliated with the Divl NSPS logic cabinct (P661) or its componcnts. including the Sensor Bypass Switch.

Objective: I Question Source: I Level of Difficulty:

LP85434.1.4.10 New I 4.0 Date Written: I 05/03/05 I Author: 1 Ryder Comments: None

1 Question # I 72 1 ury:

, L.vlrud~tof Operations KA Statement:

Ability to locate and operate components, including local controls The plant is in MODE 3, cooling down, when the following occurs:

Low feedwater flow conditions cause annunciator 5000-2F, RWCU HI D I E FLOW TIMER INITIATED, to alarm This alarm is -(I)-, and the RWCU isolation is procedurally prevented -(2)-:

A. (1) Expected (2) At P632 P642, by manually dialing the associated Differential Flow Timer fully COUNTER-CLOCKWISE.

B. ( I ) NOT expected.

(2) At P8.55, by manually dialing BOTH Differential Flow Timers fully COUNTER-CLOCKWISE.

C. (1) Expected (2) At P632 and P642, by placing the associated RWCU Isolation Bypass switch in BYPASS.

D. (1) NOT expected.

(2) At P85.5, by placing BOTH RWCU Isolation Bypass switches in BYPASS.

Answer: C

~~

Explanation:

C is correct - Refer to CPS 5000-2F. LP85404, page 25, and LP85204, page 16. P a l ( I ) -This condition of low feedwater flow can be expected to cause the listed condition. Pan (2) - A given Isolation Bypass switch defeats the isolation only for the isolation valves controlled by that division. The Outboard isolations are Div I,the lnboards are Div 2 . Therefore both divisions must be bypassed to prevent an isolation..

A. B. and D arc incorrect - For the reasons described above

RO/SRO: I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I 3 1 Generics I 2.1.30 I 3.9 I 3.4 I Lower SystemlEvolution Name: I Category:

I Conduct of Operations KA Statement:

Ability to locate and operate components, including local controls Objective: I Question Source: I Level of Difficulty:

LP85404.l.11.26 New I 2.2 References provided to examinee: None

References:

LP85204, Reactor Water Cleanup System LP85404, Leak Detection System CPS 5000-2F. RWCU Hi Diff Flow Timer Initialed, alarm response Date Written: 1 03/22/05 I Author: I Ryder Comments:

These panels, P632 and PM2. are within the control room complex, hut within the conlrol hoard area of where the operators are normally stationed. That is, they aile hack panels. These hack panels satisfy the intent of this Generic KA. Arguably. this KA does not necessarily demand that the question specifically address local controls. but rather that it address local controls MCR controls.

I MOT)IFIED/NRC UNSAT - Distractors B and D were felt to be non-plausible, with the

~~~ ~ ~~~ ~

original part 2 being throwaways. Major re-write of the question. NOTE that this change results in questions being lower cognitive GDSetser 6/13/05

1 Question # [ 73 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I ~ug~evel Both I 3 I Generics I 2.4.43 I 2.8 I 3.5 I Lower SystemIEvulution Name: I Category:

I Emergency Procedures and Plan KA Statement:

Knowledge of emergency communications 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.

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

C. Can MANUALLY operate TWO of the ADS-SRVs.

D. Can inject to the RPV with LPCI Loop C.

Answer: A Explanation:

A is correct - Per CPS 1021.01. Section 8. I .4. ALL CALL feature is available, and so is any plant alarm than can he manually initiated. Per CPS 3842.01. Section 8.1.2, this feature known by how the associated pushhutton is actually labeled: All Page.

B is incorrect - Per CPS 1021.01. Section 8.1.4. The Fuel Building Evacuation alarm can only he automatically initiated (by radiation monitokg); it has no manual feature, either at the RSP or in the MCR.

C is incorrect - Per LP85433, page 16. Only one ADS-SRV (F05 IG) can he manually operated at the RSP.

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

Objective: I Question Source: I Level of Difficulty:

None New I 2.2 References provided to examinee: None

References:

CPS 1021.01, Site Communications CPS 3842.01, Plant Communications Alarm Test LP85433, Remote Shutdown Date Written: I 04/29/05 I Author: 1 Ryder Comments: None

1 Question # I 74 I RO/SRO. I Tier: I Group: I KA: 1 ROIR I SROIR I Cog Level Both I 2 I I I 3OCO00A4.01 I 2.6 I 2.7 I Higher SystemEvolution Name: I Category:

Instrument Air System (IAS) I Plant Systems KA Statement:

Ability to manually operate andlor monitor in the control room: Pressure 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 hecause of a significant air leak?

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

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

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

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

Answer: A Explanation:

A is cnrrect -The RED band covers pressures at or below 70 psig (observed in both the MCR and Simulator configurations). Per CPS 5041.68, the standby air compressor should have auto-started at 80 psig (lowering);

therefore. two are now running. Per CPS 5041-5C. the ring header isolations should close at 70 pSig (lowering). The stem uses one minute as a way to ensure that all portions of the Plant Air system have had sufficient time to sense the lowered pressurc. and to ensure that all affected ring header auto isolations have closed as a result.

B and C are incorrect - For the reasons descrihed above.

D is incorrect - The 3d air compressor is normally in Pull-To-Lock (PTL) and will not. therefure, auto-stan Objective: I Question Source: I Level of DifTiculty:

LP85301.1.7 New I 3.0 References provided to examinee: I None

References:

I CPS 5041-5C. a n d 4 6 alarm response procedures Date Written: I 04/29/05 I Author: I Ryder Comments: None

I Question# I 7 5 I RO/SRO I Tier: I Group: I KA: I ROIR: I SROIR: I CogLevel Both I 2 I I I 300000K5.01 I 2.5 I 2.5 I Higher SystemIEvolution Name: I Category:

Instrument Air System (IAS) I Plant Systems KA Statement:

Knowledge of the operational implications ofthe following concepts as they apply to the INSTRUMENT AIR SYSTEM: Air compressors 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 PLANT AIR system, why?

A. 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.

B. 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.

C. 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.

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 Explanation:

A is correct - Per CPS 3214.01. Section 2. I .a, and CPS 4200.01. Appendix A. No further explanation required.

B is incorrect - Per CPS 3214.01, Section 6.5. The&o way to protect the compressors on power restorillion is by placing them in PTL. Even if the in-field switches are placed in UNLOAD, the compressors can still auto-stal (unless in PTL) and run unloaded C is incorrect- Refer to LP85301, page 24. These are 3-position (CLOSE-AUTO-OPEN) control switches. normally in AUTO. They will in fact auto-reopen when air pressure is restored, hut that's OK. There are no requirements to prevent these valves from reopening in an effort to 'preserve' air (especially Instrument Air) for more important plant equipment.

D is incorrect - For the same reason associated with choice 'C'. Also. these valves CANNOT auto-reopen. They trip shut on low air header pressure (70 psig) and must he manually re-latched and re-opened in the field. There is no need tu 'gag' them shut. Refer to LP85301. pages 22-23.

KA Statement:

Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM Air Compressors References provided to examinee: None

References:

LP85301, Service Air and Instrument Air CPS 3214.01. Plant Air CPS 4200.01, Loss of AC Date Written: [ 04/29/05 I Author: I Ryder

1 Question # I 76 1 RO/SRO: I Tier: 1 Group: 1 KA: I CFR 1 S R O I R I CogLevel SRO I 1 I I 1 295004 2.2.25 I 55.43(b)(2) I 3.7 I Higher SysIemlEvolution Name: 1 Category:

Partial or Complete Loss of DC Power I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of bases in technical specifications for limiting conditions for operations and safety limits Using the provided references, answer the following, The plant is operating at rated power when the following occurs:

At Time = 0000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, a fault in the RAT supply breaker causes 4160V Bus IC1 to transfer to the ERAT At Time = 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />, a fault in the RAT supply breaker causes 4160V Bus 1B1 to transfer to the ERAT At Time = 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />, the supply breaker to Div 1 125 VDC Distribution Panel, lA, 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. 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> after the DC supply breaker fails B. 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after the DC supply breaker fails C. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after the RAT breaker for Bus IB1 fails D. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> after the RAT breaker for Bus IC1 fails Answer: B Explanation:

B is correct- Per TS 3.8.9. Conditions C and D. For Condition C. only the '2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />' completion time applies. Per Condition D. thc plant must be in Mode 3 within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> (2-hr allowed outage time + I2 hours completion time = 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />) after the DC supply breaker failure. To avoid applying the '16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from discovery ...' completion time of Condition C, the SRO Candidate must rccognizelrecall the following portions of TS 3.8.9 Bases: I ) on page B 3.8-78.

the RAT breaker failures for the AC buses do not constitute 'Distribution System' inoperabilities (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 'contiguous' failures which would otherwise require the application ofthe '16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />' constraint.

A is incorrect - See the explanation above. The Candidate will choose this as the answer if helshe inappropriately applies the '16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> ...' completion time olTS 3.8.9,Condition C.

C is incorrect - This is not correct 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 (E offsite circuit) t o m divisional buses (IBI and 1C1) is synonymous with the loss of 'two offsite circuits'

1 Question ._ # I76 I RO/SRO I Tier: I Group: I KA: I CFR I SROIR: I CogLevel SRO I 1 I 1 I 295004 2.2.25 I 55.43@)(2) I 3.1 I Higher System!Evolution Name: I Category:

Partial or Complete Loss of DC Power I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of bases in technical specifications for limiting conditions for operations and safety limits D is incorrect - Per TS 3.8.I , Condition A, then Condition F. Were it not for the more limiting DC distribution problem. this would be the correct answer.

Objective: I Question Source: I Level of Dificulty:

LP85263.1 .I6 New I 2.3

References:

CPS TS 3.8.1,AC Sources - Operating CPS TS 3.8.9, Distribution Systems - Operating (and Bases)

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

1 Question# I77 I RO/SRO I Tier: I Group: I KA: I CFR I SROIR: I CogLevel SRO I I I I I 295016 2.4.6 I 55.43(b)(5) I 4.0 I Higher SystemIEvolutionName: I Category:

Control Room Abandonment I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of symptom based EOP mitigation strategies Using the provided references, answer the following The plant is operating at rated power when the following occurs:

Shift Manager determines that a control room evacuation is required Before leaving, operators place the Mode Switch in SHUTDOWN An ATWS results 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 Which ONE of the following describes the NEXT appropriate operator action?

A. Locally open the SLC Storage Tank Outlets and start the SLC Pumps B. Scram all control rods using the HCU Scram Test Switches C. Terminate and prevent Feedwater, HPCS, and RCIC.

D. Defeat the MSUOG and IA Interlocks.

Answer: C Explanation:

C is correct - Per CPS EOP-IA. Level l e g and CPS 4003.01, Sections 4.3 and 4.4. Operators do have the facilities to prevent HPCS injection (at the Div 3 switchgear, per Section 4.4.4). Feedwater injection (by closing the Feedwater Shutoffs per Section 4.4.3). and RCIC (controllable at the Remote Shutdown Panel). The objective here would be to lower RPV water 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 not fire when the pumps are stmed locally (see LP8521 I , page 21).

B is incorrect - 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 10 solve the ATWS problem, given that the stem conditions indicate a 'hydraulic' type of ATWS exists (Le., scram air headcr has already de-pressurized). This is the NEXT appropriate action.

D is incorrect - Per CPS EOP-IA. Level Icg. and CPS 441O.oOCOO4, this requires accessibility to several main control room panels. Stem conditions indicate the control room is accessible.

KO/SKO: I Tier: I (;roup: I KA: I CFK I SKO IK: I ~ t i g ~ e v r l SRO I I I I I 295016 2 4 6 I 5543th,r51 I J I) I Higher SystemlEvolution Name: I Category:

Control Room Abandonment I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge of symptom based EOP mitigation strategies Objective: I Question Source: I Level of Difficulty:

None New I 2.5 References provided to examinee: I EOP flowcharts

References:

I CPS 4003.01. Remote Shutdown CPS EOP-LA, ATWS RPV Control CPS 441 I .OS.Alternate Control Rod Insertion CPS 441O.OOCoO4, Defeating MSUOG Interlocks LP8521 I , Standby Liquid Control Date Written: I 03/29/05 I Author: I Ryder Comments: None

1 Question# 178 I RO/SRO: I Tier: I Group: I KA: I CFR I SRO IR: I CogLevel SRO I 1 I I [ 295025 EA2.01 I 55.43(h)(2) I 4.3 I Higher SystendEvolution Name: I Category:

High Reactor Pressure 1 Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE Reactor pressure Multiple system failures 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,

-and (2) whether or not that RCS portion's MAXIMUM ALLOWED TRANSIENT PRESSURE value has been EXCEEDED?

A. (1) Recirc pump DISCHARGE piping (2) Has NOT been exceeded.

B. (1) Recirc pump SUCTION piping (2) HAS been exceeded.

C. (1) RPV BOTTOM Head (2) Has NOT been exceeded.

D. (1) RPV TOP Head (2) HAS been exceeded Answer: B Explanation:

B is correct - Refer to Tech Spec SL 2. I .2 Basis for all of the answer choices. The RCS suction piping is at the lowest elevation of any RCS portion. The 1325 psig (steam dome) SL value equates to 1375 psig at the lowest elevation portion of the RCS (i.e.. Recirc suction piping); i.e. a +50 psig difference. Therefore. an overpressure transient peak pressure of 1340 psig (on the control room recorders, which look at steam dome pressure) equates to 1390 psig in the Recirc suction piping. The 'maximum allowed transient pressure value' for any portion ofthe RCS is: I 10% of the pressure value for that portion. The Design pressure value for the Recirc suction piping is 1250 psi&

Therefore. the 'max allowed transient pressure value' is 1375 psig (1.1 x 1250 = 1375). Thercfnre. the overpressure of 1390 psig in the Recirc suction piping portion of RCS exceed this 'max allowed.. .' value. and this RCS portion is clearly the 'most' impacted. relative to the other given RCS choices.

A is incorrect -The Design pressure for the Recirc discharge piping is 1550 psig or 1650 psig. depending on the location relative to the discharge valve. As such, these portions are the 'most' impacted. Additionally. the I I O %

values are 1705 psig and 1815 psig, respectively. Since these portions are a1 elevations higher than the Recirc piping.

the delta-pressure between these portions and the RPV steam dome is something less than +SO p i g . Therefore. the 1340 psig steam dome overpressure transient equates Recirc discharge piping pressures that far helow the respective

' m a allowed ...' values.

C and D are incorrect - A 1340 psig steam dome value equates to that same value throughout the RPV. In Fzct. the

'steam dome' is essentially synonymous with 'top head'. The RPV Design pressure i s 1250 psig. Its 'max allowed'

[ Question# I 78 ]

RO/SHO I Tier: I Group: I KA: I CFR I SROIR: I ~ o g ~ e v e l SRO I I I 1 I 295025EA2.01 I 55.43@)(2) I 4.3 I Higher SystemlEvolutionName: I Category:

High Reactor Pressure I Emergency and Abnormal Plant Evolutions K A Statement:

Ahilil) IO dclcrminc and/or intcrprel the idlouing a [he) appl) to HIGH REACTOR PRESSl'Kh: Kcactor prerrure I ( I 10%) value is 1375 psig. The I340 psig overpressure transient has not exceeded that value.

Objective:

LP87621.1.6 1 Question

.. Source:

New I -L.3" Level of Dificulty:

I I I I References provided to examinee: I None

References:

I CPS Tech Spec SL 2.1 2, Reactor Coolant System Pressure SL (and Basis)

Date Written: I 03/30/05 I Author: I Ryder Comments: None

[ Question# I79 1 I 1 Group: I I I S R O I R I CogLevel

~

RO/SRO: Tier: KA: CFR SRO I I I I 1 295021 AA2.01 I 55.43(b)(6) I 3.6 I Higher SystetnEvolution Name: I Category:

Loss of Shutdown Cooling I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor water heatuplcooldown rate Using the provided references, answer the following.

The plant is in MODE 4, making preparations for refueling, with the following:

The reactor was shut down 6 days ago Reactor water temperature is 122'F THEN, a complete Loss of Shutdown Cooling occurs NO Reactor Recirc 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. 67 minutes B. 90 minutes C. 102 minutes D. 120 minutes Answer: C Explanation:

C is correct - Per CPS 4006.01, Section 4.6.6, operators are allowed to raise RPV water level as high as 1hr main steam lines to gain an initial cooling effect and delay a Mode 3 entry. Section 4.6.4 directs operators to refer to the Heatup Rate and Boil-off Time Curves to assess the heatup rate that can he expected. The SRO Candidate should review the curve labeled 'Reactor Vessel Heatup Rate - Before Refueling' and plot a "'F/Hr" point where '6 days after shutdown' intersects the 'Main Steam Lines' water level curve. The result of this plot yields an approximate 46"FIhr heatup rate.

With current reactor water temperature at 122°F. a Mode 3 entry (200°F) is 78°F away. From this point a simple calculation shows that it will take about 1.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. or 102 minutes. to reach 200°F. Calculation: 78 + 46 = I .7 = I hour, 42 minutes = 102 minutes.

A is incorrect -This choice is plausible to the Candidate who carelessly translates I .7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> (as defined above) lo I hour. 7 minutes (67 minutes). Its plausibility is based on a demonstrated propensity for people lo make exactly this kind of careless mistake.

B is incorrect - Refer to the same explanation as for the correct answcr, 'C'. 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 fact that Candidates readily associate +44 inches with the minimum level necessary to promote adequate natural circulation in the absence of forced cooling flow. SRO

I Question # 179 1 RO/SRO: I Tier: I Group: I KA: 1 CFR I SROIK: I c o g ~ e v e l SRO I 1 I I I 295021 AA2.01 I 55.43@)(6) I 3.6 I Higher SystemlEvolution Name: I Category:

Lass of Shutdown Cooling I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine andlor intelpret the following as they apply to LOSS O F 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 <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> = I hour, 30 minutes = 90 minutes.

D is incorrect - Similarly, this is the calculated result if the Candidate inappropriately applies the Vessel Flange water level cuwe. This level is higher than allowed by CPS 4006.01, Section 4.6.6. Calculation: 78 i39 = 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> = 120 minutes.

Objective: I Question Source: I Level of Dimculty:

None New I 3.0

References:

CPS 4006.01, Loss of Shutdown Cooling Reactor Vessel Heatup Rate -Before Refueling curve, discussed above

I Question # 1 80 1 RO/SRO I Tier: I Group: I KA: [ CFR I SRO IK: I CogLevel SRO 1 I I I 1 600000AA2.09 I 55.43(b)(5) I 2.8 I Higher SystenJEvolution Name: I Category:

Plant Fire On Site 1 Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to PLANT FIRE ON S I T E That a failed tire alarm detector exists Using the provided references, answer the following 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?

A. Restore BOTH of these detectors to an OPERABLE status within 14 days; otherwise, establish a fire watch to inspect the zone hourly.

B. Restore AT LEAST ONE of these detectors to an OPERABLE status within 14 days; otherwise, have a fire watch inspect the zone hourly, thereafter.

C. AFTER declaring these detectors inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a fire watch to inspect the zone hourly.

D. AFTER declaring these detectors inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inspect the zone, and inspect it every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, thereafter.

Answer: C Explanation:

C is correct - Per CPS 1893.01, Appendix D. page 37, TWO inoperable detectors in this zone constitutes having more than halr of the 3 total detectors, that are in this zone. inoperable. This being a HPCS equipment cnne. these detectors are required to be OPERABLE, because HPCS is required to OPERABLE at rated power (Mode 1, per Tech Spec 3.5.1). Per CPS 1893.01. Appendix A. page 21. fire protection impairment Compensatory Measure 9.b applies for this case. Within I hour of declaring the detectors inoperable, a fire watch must be established, then hourly inspections of zone F-lb must commence.

A is incorrect -This would be the required action if there were at least 4 total detectors in fire zone F- I b. This is the action directed by Compensatory Measure 9.a, on page 20.

B is incorrect - This choice has face validity and is plausible based on two premises: I ) The very difficult-to-read Compensatory Measure 9.a making the Candidate vulnerable to misreading the requirements. and 2) an operability restoration technique very often employed within Tech Specs: Le., the idea that as soon as at least one of the 2 detectors can be restored to operability. the ComDensatori Measure can be exited. and no further action would be required. This is a the case.

D i s incorrect - This choice is essentially how a Candidate could easily misread the very difficult-to-read Compensatory Measure 9.b.

1 Question # I SO I RO/SRO I Tier: I Group: I KA: I CFR I SROIR: I CogLevel sno I 1 I I I 600000AA2.09 I 55.43@)(5) I 2.8 I Higher SystemlEvolution Name: I Catezory:

Plant Fire On Site I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine a d o r interpret the following as they apply to PLANT FIRE ON SITE: That a failed fire alarm detector exists Objective: I Question Source: I Level of Diffculty:

None New I 2.2 References provided to examinee: I CPS 1893.01, in its entirety

References:

[ CPS 1893.01, Fire Protection Impairment Reporting Date Written: I 03/31/05 I Author: I Ryder Comments: None MODIFIED/NRC UNSAT. Distractor D originally deemed to also be correct. Added the phrase every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to D to make it definitely incorrect but plasuible since this number is used in the written comp action.

I Question # [ 81 1 The plant is operating at rated power, when the following occurs:

A PARTIAL loss of Drywell Cooling (VP) occurs As aresult:

o Drywell Average Air Temperature rises and STABILIZES at 145.6'F o Drywell-to-Primary Containment d/p rises and STABILIZES at +1.1 psid Which ONE of the following describes:

(1) the required action, and (2) Z P O T E N T I A L consequence of NOT taking that action?

A. (1) Restore the Drywell-to-Primary Containment d/p to within its Tech Spec limits.

(2) Weir wall overflow, should an inadvertent upper pool dump occur.

B. (1) Restore the Drywell-to-Primary Containment d/p to within its Tech Spec limits.

(2) DIRECT communication of the blowdown energy contained in the drywell airspace, to the suppression pool inventory, should a LOCA occur.

C. (1) Restore the Drywell Average Air Temperature to within its Tech Spec limits.

(2) Drywell temperatures in excess of the drywell STRUCTURAL design temperature, should a LOCA occur.

D. (1) Restore the Drywell Average Air Temperature to within its Tech Spec limits.

(2) Drywell temperatures in excess of the drywell EQUIPMENT QUALIFICATION temperatures, should a COMPLETE loss of VP occur.

Answer: B B is correct - Per Tech Spec 3.6.5.4, 1.1 psid is beyond the upper limit of 1.0 psid. Condition A requires that d/p be restored to within the limits within I hour. Refer to Basis for this LCO, B 3.6.5.4, page B 3.6-122. the 'Background 1 discussion prtion that reads ..."The limitation on positive...". This discussion means that too high adrywell-to-CNMT can cause the vents 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 hlowdown energy would 1 communicate directly into the suppression pool inventory. See LP85223. Figure 2 for an illustration of this physical

I Question# 1 81 I RO/SRO I Tier: 1 Group: I KA: 1 CFR 1 SROIR: I CogLevel SRO I I I 2 I 295010AA2.02 I 55.43(b)(2) 3.9 I Higher SystemlEvolutiou Name: I Category:

High Drywell Pressure I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE Drywell pressure A is incorrect - Part ( I ) is correct, hut Part (2) describes the consequence of too low a d/p (i.e., below the lower LCO limit 0 1 4 . 2 psid). Refer to the same page B 3.6 - 122 discussion.

C and Dare incorrect - The stabilized dryweli average air temperature of 145.6F is lower than the entry point for Tech Spec 3.6.5.5 (i.e., 146.53R.

Objective: I Question Source: 1 Level of Difficulty:

LP85223.1.16 New I 3.3 References provided to examinee: None

References:

CPS Tech Spec 3.6.5.4. Drywell Pressure (and its Bases)

CPS Tech Spec 3.6.5.5, Drywell Average Air Temperature (and its Bases)

LP85223. Primary Containment Date Written: I 03/3 1/05 I Author: I Ryder Comments:

Although Part ( I ) is arguably a requirement for both RO/SRO Candidates, Part (2) is not. Pan (2) asks for the potential consequence of not restoring the LCO limits. which is only found i n 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.

I MODLFIED/NRC Enhancement. Deleted all reference to times in all distractors (ie ...within X hours). Changed stem from Drywell-to-Primary Containment d/p rises and STABILIZES at

+1.2 psid to Drywell-to-Primary Containment d / p rises and STABILIZES at +1.1 p s i d .

GDSetser 6/14/05

I Question # 182 I RO1SRO: I Tier: I Group: 1 KA: 1 CFR SROIR: 1 CogLevel SRO I I I 2 I 295011 2.1.14 I 55.43(b)(5) I 3.3 I Lower SystemlEvnlution Name: I Category:

High Containment Temperature I Emergency and Abnormal Plant Evolutions KA Statement:

Knowledge 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, requires NOTIFICATION of the Emergency Response Organization (ERO)?

A. Water level in the LPCS Pump Room rises to 3 inches.

B. Suppression Pool Temperature rises to 112'F.

C. Containment Temperature rises to 188°F.

D. Radiation level in RHR Pump Room 'A' rises to 100 times normal Answer: C Explanation:

C is correct - Per Clinton Radiological Annex EAL's. page CL 3-8, Fission Product Bmier Matrix #I (Containment).

and the FUI action level. Containment temperature at or above 18YF is a 'Potential Loss Containment'. requiring declaration of Unusual Event. Per EP-AA-I 12-100-F-01, Section 1.I.D. this EAL requires NOTIFICATION of ERO personnel (station Management, only).

A is incorrect - 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 room (i.e., 4 inches). Until it is, no E-Plan entry is required. The threshold for the parameter is at the ALERT action level, rather than at the UE level.

B is incorrect - Per EAL page CL 3-8. Fission Product Barrier Matfix #3 (RCS). A Suppression Pool Temperature above I IO'Fdoesm, hy itself, require an E-Plan entry It would, if it were coincident with a stuck-open SRV.

D is incorrect - Per EAL page CL 3-6. action level RU3. The threshold for this parameter is 1.000 times normal. An E-Plan entry is not yet required. Even if we were to consider the 'RA3' (Max Safe = 25 Whr) threshold of Table R4.

What the Candidate is expected to recognize is that the only way that a ' 10 times normal' level could be synonymous with having reached 25 Rlhr, would be for the 'normal' radiation level to 2.5 Rlhr. This conclusion would be implausible for any RHR Pump room.

Objective: I Question Source: I Level of Difficulty:

None New I 2.5 1 EP-AA-I 12-100-F-01, Site Emergency Director Checklist Date Written: 1 04/01/05 I Author: I Ryder Comments: None MODIFIEDiNRC Enhancement changed value in distractor D to 100 vice 10. GDSetser 6/14/05

I Question# 183 I RO/SRO: I Tier: I Group: I KA: I CFR I SROIR I CogLevel SRO I 2 I 1 I 209002 2.4.30 I 55.43(b)(l) I 3.6 I Higher SystemlEvolution Name: I Category:

High Pressure Core Spray (HPCS) I Plant Systems KA Statement:

Knowledge of which events related to system operationdstatus should be reported to outside agencies Which ONE of the following requires a NOTIFEATION (phone call) to the NRC (OTHER than to the on-site Resident)?

A. With the plant at rated power, the MCPR value is determined to be 1.10.

B. Shift Manager discovers that one of the off-going control room operators exceeded the Technical Specification overtime guidelines WITHOUT a deviation being authorized.

C. A power reduction occurs due to HPCS having been INOPERABLE for >14 days.

D. 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 3.5.1. Conditions B and D. Initiation of a plant shutdown (to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) is required. Per Enelon procedure, LS-AA-1020, page 4, item F-aa, a 4-hour report is required for the initiation of a plant shutdown required by Tech Specs (10CFR50.72(b)(Z)(i)).

A is incorrect - Per Tech Spec SL 2.1.1.2. MCPR must be at or above I .09 (2-loop). Stem conditions indicate plant is at rated power (i.e.. can only be in 2.100~). 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. IOCFR50.72(a)(Z)(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. Opcrators have 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to complete this surveillance and resolve this problem before having to enter any LCO. Meanwhile, no NRC notification is required, given this discovery alone.

Objective: I Question Source: 1 Level of Difficulty:

None New I 2.0 References provided t o examinee: None

References:

CPS Tech Spec SL2.1.1, Reactorcore Safety Limits CPS Tech Spec 3.5.1, ECCS -Operating CPS Tech Spec 5.2.2. Unit Staff LS-AA-1020, Reponability Reference Manual 10CFR50.72. Immediate Notification Requirements for Operating Nuclear Power Reactors

I Question # I 8 3 I RO/SRO: I Tier: I Group: I KA: I CFR I SROIR: I CogLevel SRO I 2 I I I 209002 2.4.30 I 55.43(b)(l) [ 3.6 I Higher SystemlEvolution Name: I Category:

High Pressure Core Spray (HPCS) I Plant Systems KA Statement:

Kno\rIrrlge of trhich e\rntr reldterl IO ~)\ICIII operlinnJslalur should hr reponed 10 nulside Jgencim MODIFIEDRVRC UNSAT. Deemed to have no correct answer. Choice C modifed to indicate that a plant shutdown (power reduction) occurs DUE to HFCS inoperability (ie required by ITS). Changed cognitive level to LOWER.

GDSetser 6/14/05

[ Question# 184 I ROISRO: I Tier: I Group: I KA: I CFR I S R O I R : I CogLevel SRO I 2 I 1 I 211OOOA2.04 1 55.43(a) I 3.4 I Higher SystemlEvolution Name: I Category:

Standby Liquid Control System I Plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow Using the provided references, answer the following.

The plant is in MODE 1, with the following:

0 The Standby Liquid Control System Operability surveillance, CPS 9015.01, has just been completed 0 SLC Pump A flow rate ACTUAL VALUE is 41.3 gpm 0 SLC Pump A Dp ACTUAL VALUE is 1260 psid Which ONE of the following:

( 1 ) CORRECTLY INTERPRETS these surveillance results, and (2) describes the required action?

A. (1) SLC Pump A Discharge Check Valve, IC41-FO33A. is NOT opening FULLY.

(2) Take action to establish a 6-week test frequency for IC41-FO33A.

B. (1) A blockage exists somewhere in the SLC Pump A piping.

(2) Take action to establish a 6-week test frequency for SLC Pump A.

C. (1) SLC Pump A Discharge Check Valve, 1C41-F033A, is NOT opening FULLY.

(2) Enter Tech Spec 3.1.7 for SLC subsystem A.

D. ( I ) A blockage exists somewhere in the SLC Pump A piping.

(2) Enter Tech Spec 3.1.7 for SLC subsystem A.

Answer: B Explanation:

B is correct - Per CPS 9015.01D001, 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 hlockage somewhere in the pump piping can yield this combination of low flow-high discharge pressure (and therefore. high &p). Per CPS 9015.01, Section 9. I .4. personnel are directed to double the lest frequency (from quarterly. to 6 weeks) when the pump goes into the ALERT Range.

A and C are incorrect - Per CPS 9015.01Do01. page 3. so long as the pump flow rate is at least 41.2 gpm. the discharge

I Question# 184 I ROISRO: I Tier: I Group: I KA: 1 CFR [ SROIR I CogLevel SRO I SystemlEvolution Name:

2 I I I 211000A2.04 I rntnnnr.,.

I 55.43(a) 1 3.4 I Higher

- .. . . . .- .^

those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow 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 incorrect - Although a blockage somewhere in the pump piping may in fact yield this low flow-high pressure combination, there is no reason to declare the SLC subsystem inoperable. That is, per CPS 9015.01WOI, 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.

Objective: I Question Source: I Level of Difficulty:

None New I 2.5 References provided to examinee: CPS 9015.01, in its entirety CPS 9015.OlWOI, in its entirety

References:

CPS 9015.01, Standby Liquid Control System Operability CPS 9015.01WO1.SLC Pump & Valve Operability Data Sheet Date Written: I 04/02/05 I Author: I Ryder Comments: None MODFIEDNRC Enhancement, Changed wording in disctractors B and D to SLC piping vice specific location (suction/discharge). GDSetser 6/14/05

I Question# I 85 I RO/SRO: I Tier: I Group: I KA: I CFR I SROIR. I ~op~evel SRO I 2 I I I 215005 2.2.25 I 55.43@)(2) I 3.1 I Lower SystemlEvolution Name: I Category:

Average Power Range Monitodhcal Power Range 1 Plant Systems 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. INOP Rod Block C. Neutron Flux -High, Setdown D. Fixed Neutron Flux -High Answer: D 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 INOP 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 review of these shows NO connection to any analysis that tdes credit for the APRM INOP - Rod Block Function.

C is incorrect - Per CPS Tech Spec Bases. pages B 3.3-6 and 7. Although this Function indirectly protects Safety Limit (SL) 2.1.1.1, there is no analysis that takes directlspecific credit for this Function.

References provided to examinee: None

References:

CPS Tech Spec Bases B 3.3.1.1, RPS Instrumentation CPS Tech Spec Bases B 2.0, Safety Limits CPS Tech Spec 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 Drop Accident (CRDA)

I Question# I 86 I ROISRO: 1 Tier: I Group: I KA: I CFR I SROIR: 1 CogLevel SRO I 2 I 1 I 400000A2.03 I 55.41(b)(lO) I 3.0 I Higher SystemIEvolution Name: I Category:

Component Cooling Water System (CCWS) I Plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the CCWS; and (b) based on those predictions. use procedures to correct, control. or mitigate the consequences of those abnormal conditions or operations: HigMlow CCW temperature 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 arc in service CCW HX Shell Side Outlet Temperature is now 106'F and STABLE Which ONE of the following describes:

(I) a consequence of allowing CCW HX Shell Side Outlet Temperature to remain at this temperature,

-and (2) an appropriate action?

A. (1) Operating the CCW HX is excess of its DESIGN limit for Shell Side Outlet Temperature.

(2) Line up an FC Heat Exchanger with cooling supplied by SX.

B. (1) Operating with a CCW Temperature TOO NEAR the temperature that will cause a High CCW Temperature trip of the Service Air Compressors..

(2) Secure all non-essential CCW heat loads C. (1) RWCU pump seal cavity temperature limits have been EXCEEDED.

(2) Secure all non-essential CCW heat loads.

D. (1) RISING radiation levels in the Fuel Building.

(2) Line up an FC Heat Exchanger with cooling supplied by SX.

Answer: A Explanation:

A is correct - Per CPS 3203.01. Sections 6. I and 8.3.1. The given Outlet Temperature is ahove the 105°F Design limit for the HXs. Section 8.3.1 directs operators to place a second FC HX in service, cooled by SX (Shutdown Service Water). Section 8.3. I .4(2)c directs operators to 'consider' shifting a11 FC cooling over to SX (taking about 30% of the total CCW heat load off the CCW system, per Appendix B on page 70). Close scrutiny by the CPS Facility Author and an SRO Validator (incumbent) has determined that the 'intent' of the wording in CPS 3203.01, Section 6. I .2 is as phrased here in this answer choice.

B is incorrect - Per LP85301 Service and Instrument Air, the SACS do not have a HIGH CCW temperature tnp.

C is incorrect - There is CCW supply tcmperature that is defined as 'unacceptably high' for the RWCU Pump Per CPS 3303.01, Section 8.1.1.1 I . operators must only ensure that seal cavity temperature is maintained below 200°F.

I Question# I 86 I RO/SRO: I Tier: I Group: I KA: I CFR I SROIR: 1 ~ o g ~ e v e l SRO I 2 I I I 400000A2.03 I 55.41(h)(10) I 3.0 I Higher SysiemIEvoluiion Name: 1 Category:

Component Cooling Water System (CCWS) I Plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the CCWS; and (h) based on those predictions. use procedures to correct. control, or mitigate the consequences of those abnormal conditions or operations: HigMow CCW temperature D 15 incorrect - There I> no reference. either i n CPS 3203 01 or (PS 3317 01. that a.whiates CCW Outlet Tempemturs uith degrddcd Speni Fuel Storage Pinil cooling ,F(I thdt rc5uII in ele\ateJ radialitin Ik~eIsi n 1lu1area I

I Objective:

~ ~ 8 5 2 0I.814 I Question Source:

New I

I Level of Difficulty:

3.0 CPS 3317.01, Fuel Pool Cooling and Cleanup CPS 5040-IC. High Temp CCW HX Outlet Temperature (alarm procedure)

LP85301 Service and Instrument Air Date Written: I 05/16/05 I Author: 1 Ryder Comments:

Tne following justify this being an SRO only question:

I. Although Part (I)of the question would apply to an RO exam, as well, in that knowledge of the CCW HX design temperature limit has the procedures PrecautionsILimitation Section as its source, Part (2) goes beyond this source, into the procedures Abnormal section.

2. Abnormals section 8.3.1, specifically, step (2)c suggests consideration.... Such direction i s 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 to the job of the on-shift SRO.

3. Although the alarm response procedure (CPS 5040-IC), 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 does not address all of the actions found in Section 8.3.I of the operating procedure MODIFIED NRC Enhancement. Distractor C was deemed implausible, specifically relatec to Demin damage at this low temperature. Replaced B(l) with statement regarding high CCW temperature trip of SACs. This has validity in that there is a high oil temperature trip and a low CCW pressure trip of the SACs. Also changed B(2). Addedreference to S N I A Lesson Plan to disprove this incorrect answer. Changed wording of C(l) from UNACCEPTABLY high CCW supply temperature at the RWCU Pump seals to RWCU pump seal cavity temperature limits have been EXCEEDED.

GDSetser 6/14/05

1 Question # 187 1 A plant power ascension is in progress, per CPS 3004.01, Turbine Startup and Generator Synchronization, with the following:

e Reactor Recirc Pump (RRP) A is about to be transferred to FAST speed e IMMEDIATELY BEFORE the operator positions FCV A for the transfer, the following occurs:

o RRP B trips from 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 m l b d h r on the 65% FCL (2) Immediately scram the reactor, even if NO power oscillations are observed.

B. (1) About 20 m l b d h r on the 60% FCL (2) Verify MFLCPR is at or below 0.970.

C. (1) About 25 m l b d h r on the 55% FCL (2) Isolate RR Loop B using CPS 3302.01, Reactor Recirculation D. ( 1 ) About 20 mlbndhr on the 50% FCL (2) Direct IMD to change the APRM setpoints to those for Single-Loop Operations.

Answer: D

1 Question # I 87 I RO/SRO: I Tier: I Group: I KA: 1 CFR I S R O I R I CogLevel SRO I 2 I 2 I 202001 A2.04 I 55.43(b)(5) I 3.7 I Higher SystenJEvolution Name: I Category:

Recirculation System I Plant Systems KA Statement:

Ability to (a) predict the impacts ofthe following on the RECIRCULATION SYSTEM; and (b) based on those oredictions. use Drocedures to correct. control. or mitieate

" the conseauences ofthose abnormal conditions or I operations: single recirculation pump trip Explanation:

D is correct -

Part (1): Refer to CPS 3004.01, Sections 8.4.5 and 8.4.6, and to CPS 3302.01, Sections 8.1 .I and 8.1.2. The following are the initial conditions BEFORE the trip of RRP 'B': 1) Both RRPs are in SLOW speed, with their FCVs about 90%

open: 2) Reactor power is about 30% with the rod line (FCL) at or near 50% FCL: 3) Thus, a review ofthe P/F Map indicates that Total Core Flow before the RRP 'B' trip is about 38.0 mlbm/hr. The following are the stable flow conditions AFTER the trip of RRP 'B' and the immediate shutting of its discharge valve (Immediate Operator Action.

per CPS 4008.01. Section 3.2: I) RRP 'A' remains unaffected, running in SLOW speed, with its FCV still at about 90% open; 2) Total Core Flow simply migrates down the Same 50% FCL (rod line) and stabilizes at about a little more than half of the initial 38.0 mlbm/hr value (i.e., about 20 mlbmhr). -This is the Generic Fundamental (Pumps and Fluid Flow) of losing one of two. identical, pardkkonfigured pumps. NOTE, ALSO: A review of the P/F Map in this area shows that there is no need to consider the precise slope of the 50% FCL &e.. a so-called 'flatter' sloped FCL, due to fuel design). Whatever slope perturbation there might be in this area of the Map, we are still well helow any point where we would expect to drift into the CONTROLLED ENTRY Region.

Part (2): A review ofthe CPS 4008.01 Subsequent Actions shows there is no specific action that is required 'early' in the case of this specific scenario. From Section 4.4, operators are directed to Section 4.9 for SLO. Stem conditions indicate that Sections 4.9.1.4.9.2.and 4.9.3 are non-issues for this scenario. In Section 4.9.4. operators proceed to CPS 3005.01, to implement SLO. There, in Section 8.4.3,operators are directed to change the APRM setpoints. This is the

'next' action that is required given these stem conditions.

A is incorrect - 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 incorrectly concludes that the transfer takes place at a power level closer to 45.50%. with a 65% rod line (FCL), then the pre-trip total core flow is about 51 mlbmlhr, and the post-trip flow is about 25 mlbmlhr (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 clme to, the Restricted Zone (scram required). The Candidate is expected to recognize that. for these reasons, a pump up-shift would not take place on the 65% FCL.

B is incorrect - This choice is distracting to lhe SRO Candidate who chooses the 63% FCL. with a power closer lo 35%. and a pre-trip flow of about 38.0 mlbmlhr. This choice is very attractive when the SRO Candidate considers Part (2). There is clear indication, in CPS 4008.01, Section 4.2, that some priority should be given the verifying MFLCPR is at or hclow 0.970 for SLO. However. the suggestion, in Part (I).that the RRP up-shift will have been started from the 60% FCL makes this choice absolutely wrong.

C is incorrect - Although a 55% FCL is feasible for pump up-shift conditions. the resulting total core flow is higher than predicted (see explanation for correct answer. 'D).In Part (2). there is NO requirement to isolate the tripped pump loop. In fact. to do so would prohibit the restoration ofthat loop until reactor water temperature is < 2 v F (see CPS 3302.01.Section 6.11.1).

0b.iective: I Question Source: I Level of ~ i f f i c u ~ t y :

None New I 3.0

RO/SRO I Tier: I Group: I KA: 1 CFR I SROIR. I CogLevel SRO I 2 I 2 I 202001 A2.04 I 55.43@)(5) 1 3.1 I Higher SystemlEvolution Name: I Category:

Recirculation System I plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM; and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Single recirculation pump trip References provided to examinee: I None

References:

[ CPS 3004.01, Turbine Startup and Generator Synchronization CPS 3005.01, Unit Power Changes CPS 4008.01, Abnormal Reactor Coolant Flow CPS 3302.01, Reactor Recirculation (RR)

Date Written: I 05/16/05 I Author: I Ryder Comments: None

I Question# 1 88 1 RO/SRO: I Tier: 1 Group: I KA: I CFR I SROIR I CogLevel SRO I 2 I 2 I 219000 2.4.6 I 55.43@)(5) I 4.0 I Higher System/Evolutiun Name: I Category:

RHWLPCI: T o ~ s I P o o Cooling l Mode I Plant Systems KA Statement:

Knowledge of symptom based EOP mitigation strategies From rated power, operators have just manually scrammed the reactor, THEN the following occurs:

Reactor Power is now 20%

SLC Pumps are running Reactor water level is +35 inches Reactor Pressure is 950 psig and STABLE Suppression Pool Level is 19 feet and RISING SLOWLY One SRV has stuck open Suppression Pool Temperature is 93°F and RISING From among the following actions, which ONE is required NEXT?

A. Terminate and prevent injection to establish water LEVEL BAND C.

B. Drain Suppression Pool inventory to stay below the SRV Tail Pipe Limit.

C. Open a Turbine Bypass Valve to stay below the Heat Capacity Limit D. Place all available Suppression Pool Cooling in service Answer: D Explanation:

D is correct - Per CPS EOP-6, Pool Temperature leg. top-most IF-THEN step. Pool temperature is very near point we the CRS would proceed to step 19 ofthis leg. where operators are directed to place a11 available suppression pool cooling in service. Of all the current plant conditions in the stem, this is the highest priority, currently.

A is incorrect - Per EOP-I A, Level 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 override (;.e., I IOF pool temperature).

B is incorrect - Per EOP-6. Pool Level leg, step 20, and Figure Q. Even though the stuck-open SRV is slowly adding to pool inventory, with level currently at 19 feet. and reactor pressure at 950 psig. we are still far below the Figure Q limit of -23 feet. This is not the NEXT required action.

C is incorrect - Per EOP-6. P w l Temperature leg, bottom-most step. Although intentionally lowering pressure lo stay below the Heat Capacity Limit (Figure P) is allowed, even during an ATWS. the current stern conditions of 950 psig and 93°F pool temperature. are still far below the HCL limit of -145F pool temperature. This is not the NEXT required action.

1 Question # 188 I ROISRO: I Tier: I Group: I KA: I CFR I SROIR: I Cog Level SRO I 2 I 2 I 219MM 2.4.6 I 55.43(b)(5) I 4.0 I Higher SystedEvolution Name: I Category:

RHWLPCI: Torus/Pool Cooling Mode I Plant Systems KA Statement:

Knowledge of symptom based EOP mitigation strategies

References:

CPS EOP-I, ATWS RPV Control CPS EOP-6, Primary Containment Control Date Written: I 04/07/05 I Author: I Ryder Comments: None

L Question # I 89 ]

RO/SRO: I Tier: 1 Group: I KA: I CFR I SROIR. I CogLevel SRO I 2 I 2 I 234000 2.1.32 I 55.43(b)(7) I 3.8 I Lower SystenJEvolution Name: I Category:

Fuel Handling Equipment I Plant Systems KA Statement:

Ability to explain and apply system limits and precautions 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 fuel bundle is NOT in its correct core location Which ONE of the following describes the NEXT required action?

A. RELEASE the grapple and contact the Reactor Engineer for further guidance B. Remove the bundle from the core and return it to its CORRECT core location.

C. Remove the bundle from the core and return i t to its UCP rack location.

D. Do NOT release the grapple and contact the Reactor Engineer for further guidance.

Answer: B Explanation:

B is correct - Refer to CPS 3703.01. Section 6.24.I , This question proposes that exact scenario.

A is incorrect - This choice suggests a scenario as described in Section 6.24.2. It is oat.

C and D are incorrect - They hoth have strong face validity plausibility for this &&refercnce question.

Objective: I Question Source: I 1,evel of Difficulty:

None New I 2.0 ~.~

I Question# I 90 I RO/SRO: I Tier: I Group: I KA: I CFR I SROIR: I CogLevel SRO I 3 I Generics I 2.1.4 1 55.43(a) I 3.4 I Higher SystenJEvolution Name: I Category:

I Conduct of Operations KA Statement:

Knowledge of shift staffing requirements 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 Supervisor AND designated STA.

B. While in MODE 3, you have a total of FOUR Fire Brigade Members, one of whom is also the designated Safe Shutdown Qualified Operator.

C. While in MODE 5 , you have a total of THREE Fire Brigade Members, ALL of whom are C Area Qualified.

D. While in MODE 4,you have total of TWO SROs, NEITHER of whom is qualified as STA.

Answer: D Explanation:

D is correct - Per OP-CL-101-102- 1001, all. In Modes 4 and 5 , only 2 SROs are required (for ERO functions of CRS and SM), and w STA is required.

A is incorrect - Per OP-CL-101-102-1001. all. In Modes I . 2. and 3. if either the CRS or SM is also the STA. then a 3d SRO is required and is designated as the Incident Assessor.

B is incorrect - Per OP-CL-101-102-1001, all. In ALL Modes, at least 4 Fire Brigade Members (plus the Leader) are required, n ~ n of c whom can he the designated SSQO.

C is incorrect - Per OP-CL-101-102-1001, all. In ALL Modes, at least 4 Fire Brigade Memhcrs (plus the Leader) are required.

References provided to examinee: I None

References:

I OP-CL-101-102-1001, CPS Minimum On-Shift Staffing Functions Date Written: I Comments: None 04/08/05 1 Author: I Ryder I

I Question # I 91 I RO/SRO. I Tier: I Group: I KA: I CFR I SROIR: I CugLevel SRO I 3 I Generics 1 2.2.23 I 55.43(a) I 3.8 I Lower System/Evolution Name: I Category:

I Equipment Control KA Statement:

Ability to track 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 initiate a Degraded Equipment Log entry for the instrument.

B. Direct IMD to hang an Equipment Status Tag (EST) on the instrument, and the RO to hang a Miniature EST in the control room.

C. Identify the Technical Specification required action in the event the instrument is still INOPERABLE when the Short Duration Time Clock (SDTC) expires.

D. 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.

Answer: C C is correct - Per OP-AA- 108-104, Sections 3.5 and 4.7.1.2, and Attachment 1. The CRS identifies the Tech Spec action that I&C will he 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 plausihlc. especially considering this is a closed-reference question. The DEL is used for equipment that is OPERABLE hut degraded. In this case, the equipment i s INOPERABLE.

B is incorrect - The EST (and Miniature EST) is a process governed by OP-AA- 108-101, and is used to track the status of equipment out of 'normal' (position). This inoperable instrument 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 Spec entry. There is no Tech Spec entry associated with this surveillance test.

Objective: I Question Source: I Level of Difficulty:

None New I 2.3

ROISRO: I Tier: I Group: I KA: I CFR 1 SROIR: I CogLevel SRO I 3 I Generics I 2.2.23 I 55.43(a) I 3.8 I Lower SystemiEvolutionName: I Category:

I Equipment Control KA Statement:

Ability to track limiting conditions for operations References provided to examinee: None

References:

OP-AA-108-104, Technical Specification Compliance OP-AA-108-101. Control of Equipment and System Status OP-CL-101-302-1001, ITS LCO/ORM OWODCM OR Evaluations

[ Comments: None I ModifiedNRC Enhancement. Changed to refer to Degraded Equipment Log. GDSetser 6/14/05

I Question # I 92 I 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(b)(7) 1 3.8 I Lower SystemlEvolution Name: 1 Category:

I Equipment Control KA Statement:

Knowledge of SRO fuel handling responsibilities With all fuel handling equipment operating normally, which ONE of the following REQUIRES the DIRECT SUPERVISION of a Refuel SRO?

Transfer of...

A. NEW fuel from the Fuel Building to the Containment Refuel Floor.

B. IRRADIATED fuel from the Fuel Building to the Containment Refuel Floor.

C. NEW fuel from the New Fuel Storage Vault to the Fuel Building Transfer Pool D. IRRADIATED fuel from the Spent Fuel Storage Pool to the Fuel Building Transfer Pool.

Answer: B Explanation:

B is correct - Per CPS 3703.02. Section 3.5. 1 bullet. Refuel SRO must supervise any irradiated fuel movement between the Fuel Building and Containment.

A is incorrect - Per CPS 3703.02, Section 3.5, ZP*bullet. A Reactor Engineer is allowed to supervise this transfer.

C is incorrect - Per CPS 3703.02, Section 3.5. 3d bullet. This transfcr requires no supervision by a Refuel SRO. only authorization by Control Room Supervision (SM/CRS) or Work Control (WCS).

D is incorrect - Othcr than the movements considered by CPS 3703.02, Section 3.5 (above). only CORE ALTERATIONS requires the supervision of a Refuel SRO. This choice does not describe a Core Alteration Ce., there is no movement of fuel that involves the reactor vessel).

Objective: I Question Source: I Level of L)ifficulty:

None New I 2.0 Date Written: I 04/29/05 I Author: I Ryder Comments: None

I Question# I 93 I RO/SRO I Tier: I Group: I KA: I CFR I SROIR I CogLevel SRO I 3 I Generics I 2.3.2 I 55.43(b)(S) I 2.9 I Higher SystemIEvolution Name: I Category:

I Radiological Control KA Statement:

Knowledge of the facility ALARA program Using the provided references, answer the following The plant is operating at rated power, with the following:

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 m r e d h r 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 operator's entry into this LHRA?

Can enter...

A. ONLY IF accompanied by an RP Tech; an approved RWP is NOT required.

B. ALONE; however, an approved RWP IS REQUIRED, a current survey map is NOT required.

C. ALONE; however, BOTH an approved RWP &JQ a current survey map ARE REQUIRED.

D. ONLY IF accompanied by an RP Tech; an approved RWP IS REQUIRED.

Answer: C 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 & require an approved RWP for HRA and LHRA entries (other than for emergent entries).

A is incorrect - Per 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 to enter without the RWP would amount to NOT operating in accordance with 'plant radiation protection procedures.. .'

B is incorrect - Per RP-AA-460. Section 4.7.1, accompaniment by an RP Tech i s required if a current survey map is not available.

D is incorrect - For the reasons associated with the correct answer. 'C'.

I Question# I 93 I RO/SRO I Tier: I Group: I KA: I CFR I SROIR. I ~og~evel SRO I 3 1 Generics I 2.3.2 I 55.43(b)(5) 1 2.9 I Higher SystemlEvolution Name: I Category:

I Radiological Control KA Statement:

Knowledge of the facility ALARA program Objective: I Question Source: I Level of Difficulty:

None New I 3.0 I References orovided to examinee: I CPS Technical Specification 5.7. in its entirety (2 pages) I

References:

CPS Technical Specification 5.7. High Radiation Area RP-AA-460. Controls for High and Very High Radiation Areas Date Written: I 05/16/05 1 Author: I Ryder Comments: None

1 Question# I 94 I RO/SRO I Tier: I Group: I KA: I CFR I SROIR. I CogLevel SRO I 3 I Generics 1 2.3.1 I 55.43(b)(4) I 3.0 I Lower SysteMvolution Name: I Category:

1 Radiological Control K A Statement: ...

Knonlcdgc 01 10 CFR 20 and rcl3tcJ f x i l i r ) radiation control rr.quirr.mcnlr 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, and (2) how many ADDITIONAL mrem (DDE) she (including the embryo/fetus) is allowed to receive during the REMAINDER of her pregnancy?

A. (1) When she is no longer pregnant (2) 400 mrem B. (1) 12 months from todays date (2) 400 mrem C. (1) 12 months from todays date (2) 500 mrem D. (1) When she is no longer pregnant (2) 350 mrem Answer: B Explanation:

B is correct - Per RP-AA-270, Attachment 3. page I of 1. Given that she has received a total DDE of only 100 mrem since hecoming pregnant, she is limited to a total of 500 mrem DDE for the entire pregnancy. or an additional 400 mrem from the remainder of her pregnancy. Unless she withdraws her declaration hefore-hand. this declaration and its work restrictions automatically expire 12 months from today.

A. C , and D are incorrect- Each is a plausible mis-understanding, or mis-application. of the Attachment 3 requirements cited above.

Objective: I Question Source: I Level of Difficulty:

None New I 4.0

ROISRO: I Tier: I Group: I KA: I CFR I SROIR: I CogLevel SRO I 3 I Generics I 2.3.1 I 55.43(b)(4) I 3.0 I Lower SystedEvolution Name: I Category:

I Radiological Control K A Statement:

Knowledge of 10 CFR 20 and related facility radiation control requirements References provided to examinee: I None

References:

I RP-AA-270, Prenatal Radiation Exposure Date Written: 04/1 I /05 I Author: 1 Ryder Comments:

This question is categorized as Lower Cognitive (LCL) because it only requires the recall of two, independent pieces of information: I ) 500 mrem for the entire pregnancy, and 2) 12 months for the automatic expiration of the declaration's work restrictions. Tnere is no IF-THEN relationship that necessarily exists between the two parts of the question.

This is an SRO-only question because the 'Declaration' i s one that a 'Work Supervisor' must review and approve. In fact. the Work Supervisor (CRSISM in her case), is critical in stipulating the work restrictions for the Dose Equivalent Reduction Action Plan.

I Question# 195 1 RO/SRO. I Tier: I Group: I KA: I CFR I SROIR: I CogLevel SRO I 3 I Generics I 2.1.34 1 55.43(b)(5) I 2.9 I Higher SystemlEvolution Name: I Category:

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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ago, with the following:

Reactor water temperature is 215°F 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 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

B. Direct Chemistry to obtain and analyze a confirmation sample of FEEDWATER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. Suspend control rod withdrawals in preparation for returning to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. Notify the Nuclear Operations Duty Officer of an Action Level 2 condition Answer: D Explanation:

D is correct - Per CY-AB-120-100, Section 4.3.2. For these plant conditions (POAH. Mode 2). Reactor Coolant Chlorides are higher than the Action Level 2 limit (100 ppb). Per Attachment 2, the first requirement is to notify the Nuclear Operations Duty Officer of the Action Level 2 condition.

A is incorrect -This choice has face validity (psychometrically balanced with choice 'B'), and is intended for the Candidate who incorrectly applies Note 'a' of CY-AB-120-100, Section 4.3.2. I Table.

I B is incorrect - This choice is plausible to the Candidate who recognizes Feedwater Conductivity is higher than the

I Question # I 95 I RO/SRO I Tier: I Group: I KA: I CFR I SROIK: I CogLevel SRO I 3 I Generics I 2.1.34 I 55.43(b)(5) I 2.9 I Higher SystedEvolution Name: I Category:

I Conduct of Operations KA Statement:

Ability to maintain primary and secondary plant chemistry within allowable limits Action Level 1 limit of CY-AB-120-1 IO, Table la. and then applies the Action Level I decision tree of Attachment 1.

However, this would be a mistake in the light of Note a associated with the Table l a limits. With the given stem conditions, the steam jet air ejectors cannot he in service (placed in service at or above 150 psig): reactor power is only at the point 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 Chemistry Table 3.3.1-1 shows that the given 150 ppb value for reactor cwlant chlorides is above the 0.1 ppm (i.e., 100 ppb) limit for Modes 2 and 3. The Candidate may refer to OR Action 3.3.1.2. hut neglect to notice that the action to return to Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is not required until after this chloride value has been exceeded for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ago is there to ensure any consideration of applying Note IC of CY-AB-120-100, Section 4.3.2.1 Table, is avoided.

References provided to examinee: CPS Operating Requirements (ORM) 2.3.1. entire section CY-AB-120-100, in its entirety CY-AB-120.1 IO. in its entirety

References:

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

I Question # I96 1 RO/SKO: I Tier: I Group: I KA: I CFR I SROIR: I Cog1,evel SRO I 3 I Generics I 2.4.41 I 55.43(b)(5) I 4. I I Higher SystemlEvolution Name: I Category:

I Emergency Procedures and Plan KA Statement:

Knowledge of the emergency action level thresholds and classifications This question was dropped from the examination based on licensee post-examination comments that there was no correct answer to the question.

P T PI-

"1 -1 Explanation:

C is correct - Part ( I ) : The that an EAL threshold is reached is at Time = +I5 minutes. For EAL ' M U 6 (see CPS Annex page CL 3-1 I). Per EP-AA-I 12-100. Section 2.1. the Shift Manager (SM) would have until Time = +30 minutes to classifyldeclare the event as a UE. and until Time = +45 minutes to complete the required StalelLocal

1Question # I 96 1 ROISRO I Tier: I Group: I KA: I CFR I snom I Cog~eve~

SRO I 3 I Generics I 2.4.41 I 55.43(b)(5) I 4. I I Higher System/Evolution Name: I Category:

I Emergency Procedures and Plan KA Statement:

- Knowledge of the emergency action level thresholds and classifications notifications. However, at Time = +20 minutes, the 'FAI' EAL threshold is reached due to a 'Potential Loss of RCS' (see Annex page CL 3-8). Again, the SM would have until Time = +35 minutes (20+ 15 = 35) to classifyldeclare the event as an ALERT. Per EP-AA- I 11, Section 4.1. the 2"6 NOTE, once this higher classification level is declared, if the UE notification has not yet been made, the UE event is essentially dismissed (without further consideration), in favor 01 the more 'severe' ALERT event declaration. In other words, given these stem conditions, the UE event (loss of annunciators) does not result in a 'First required' StateJLocal agency notification. Rather. the SM has until Time = +50 minutes to complete the ALERT notifications. And since the next plant transient that requires a re-classification (escalation) to an SAE (Le., the RPV Blowdown) doesn't even occur until Time = +55 minutes, the S M does in fact get a chance to complete the ALERT notifications at Time =+SO minutes. This, therefore, amounts to the 'First required' StatelLocal agency notification for these given plant conditions. Part (2): An SAE is the highest classification required for these plant conditions (i.e.. the 'FSI' EAL is reached due to Loss of Containment: see Annex page CL 3-8). Again, per EP-AA- 112- 100. Section 2.1, the SM must declare this escalation (from an ALERT) nu bdICr than Time = +70 minutes (+55 + 15 minutes = +70 minutes)

A is incorrect - For the reasons already described above. Part (I)is plausible to the Candidate who disregards the EP-AA-I I I , Section 4.1, requirements, and mistakenly applies a +30 minute requirement (+I5 + 15 I +30 minute) of EP-AA-I 12-I00, Section 2.1, to the start of the '-clock' for 'MU6'. Part (2)is plausible to the Candidate who recognizes the need to escalate to an ALERT by no later than Time = +35 minutes (FA1 threshold at Time = +20. + I S minutes to classify, per EP-AA-I 12-100, Section 2.1). This Candidate d o e s a r e c o g n i z e that the RPV Blowdown at Time = +55 minutes results in a further escalation to an SAE ('FS I' EAL).

B is incorrect- For the reasons already described above. Part ( I )is plausible to the Candidate who. although correctly waits for the M U 6 threshold clock to become 'active' (i.e., the threshold is met) before applying the +30 minute allowance of EP-AA-I 12-100, Section 2.1. fails to apply the EP-AA-I I I, Section 4.1 requirement that essentially dismisses the M U 6 event. Part (2) is designed to provide psychometric balance with Part (2)of choice 'D'&e.. a time value that is &than its associated Part (I)value). It has sufficient face validity for the thoroughly confused Candidate, as well.

D is incorrect - For the reasons already described above. This choice (both Parts) is plausible to the Candidate who cannot effectively translate the earlier of the EOP-8 actions identified in the stem conditions, and instead simply applies the final state of the plant (RPV Blowdown is progress) and concludes that EAL 'FS I' applies. This Candidate will necessarily recognize that the S M has 15 minutes to classify the SAE (i.e., Time = +55 minutes C 15 minutes = +70 minutes), yielding Part (2)of the answer choice. Similarly, the S M has an additional 15 minutes. from Time = +70 minutes, IO complcte the StateILocal notifications (Time = +70 + 15 minutes = +85 minutes). yielding Part (I)of the answer choice.

References provided to examinee: EP-AA-1003, Clinton Radiological Annex. pages CL 3-6 thru 3- 13 EOP flowcharts

References:

EP-AA-1003, Clinton Radiological Annex EP-AA-112-100, Control Room Operations EP-AA- I 11. Emergency Classification and PARS CPS EOP-8, Secondary Containment Control Date Written: I 05/16/05 I Author: I Ryder Comments: None

I Question # 197 I RO/SRO I Tier: I Group: 1 KA: I CFR I SROIR: 1 ~og~evel SRO I I I I I 295031 2.1.20 I 55.43(b)(S) I 4.2 I Higher SystellllEvolutionName: I Category:

Reactor Low Water Level 1 Plant Systems KA Statement:

Ability to execute procedure steps Using the provided references, answer the following.

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. Leave Level and Pressure; enter EOP-2.

C. Leave Level and Pressure; enter EOP-3.

D. Implement the actions of CPS 441 1.05 for rising pool level Answer: C Explanation:

C is correct - Refer to EOP-IA. Level leg, and Detail C. As soon RPV level drops to - 1 5 0 Wide Range, thls instrument becomes unusable, with a containment temperature above 100°F (175°F is indicated in the stem conditions).

Operators must immediately transition to the Fuel Zone Range instruments. Because reactor pressure is about 6W psig (still far ahovc the depressurized (0 psig) calibration conditions for the Fuel Zone instmments). and Wide Range instruments are reading essentially the same 3s actual level before they become unusable, the Fuel Zone will indicate well below TAF (-162) when the operators operationally transition to them. As such the CRS has no choice but to irnplemcnl 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 Down.

A is incorrect - Per EOP-6, Containment Temperature leg, and Figure 0. The existing Containment Temperature (175F)lContainment Pressure (2.0 p i g ) combination has us on the bad side of the Containment Spray Initiation Limit curve, Figure 0. Until things change (likely to be an additional rise in Containment Pressure). it is OOt OK to Spray. This choice is the NEXT required action.

B is incorrect - This would be the NEXT action if there were usable RPV water level instruments (i.e.. no available Fuel Zone instruments) when the Wide Range instruments dropped below the minimum usahle level of Derail C . The CRS would invoke the top-most override step of the EOP-IA Level leg. by declaing RPV water level unknown and transitioning 111 EOP-2 for RPV flooding. Because there are indications in the given stem conditions to cause the Candidate to conclude that the Fuel Zones are not avdable, this choice is a the NEXT required action.

RO/SRO: I Tier: I Group: I KA: 1 CFR I SROIR I CogLevel SRO I I I 1 I 295031 2.1.20 I 55.43@)(5) I 4.2 I Higher I Category:

SystemlEvolution Name:

Reactor Low Water Level 1Ability to execute procedure steps I Plant Systems 1 I

D is incorrect - 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 1.05)is associated not only with a pool level high enough to threaten the SRV Tail Pipe Limit of Figure 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 Question Source:

M.

II Level of Difficulty:

I"

References:

CPS EOP-IA, ATWS RPV Control CPS EOP-6, Primary Containment Control

I Question # I 98 I ROISRO. I Tier: 1 Group: I KA: I CFR I SROIR: I CogLevel SRO I 1 I I I 295003 AA2.04 I 55.43(b)(5) I 3.7 I Higher SystedEvolution Name: I Category:

Partial or Complete Loss of A.C. Power [ Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

Using the provided references, answer the following.

With the plant operating at rated power, a COMPLETE LOSS of AC Power (including Div 3) occurred 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. Div 1 B. Div2 C. Div3 D. BOP Answer: B Explanation:

B is correct - Refer to CPS 42W.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 ofRClC to maintain RPV inventory, and the impact of the SRVs adding energy (heat) to the Suppression Pool. All Divisional batteries (including the Div I battery for RClC support) are 4-hour batteries (see USAR. Section 8.3.2.1.2.1). With RClC adequately controlling level, with there being plenty of decay heat and, therefore. steam pressure to support RCIC. and with RClC having been on its battery for only 15 minutes. now, there is no 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 temperature significantly below its Tech Spec limit of95"F. per Tech Spec 3.6.2.1 Refer to Figure P, Heat Capacity Limit. of EOP-6. With post-scram reactor pressure between 800 and 1065 psig, the pool's Heat Capacity Limit is no where near being threatened (happens as pool temperature nears -145°F). Therefore.

there is no urgent need to restore any of the Divisional (1.2, 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 00 urgent need to restore Div 3 power to enable HPCS as an alternate injection source. With the RCIC/SRV feedlbleed combination providing adequate core cooling and there being no urgent need to restore the main condenser, or a vacuum. there is no immediate need lor restoring Reactor Recirc. CCW. Plant Chilled Water. Plant Service Water, Plant Air. Condensate, or a CRD Pump. Therefore. there is no urgent need to restore non-divisional (BOP) bus power. In the end, the SRO Candidate should recognize that restoring Div 2 bus power is t h e m priority. because it re-energizes the plant security systems, allowing card-reader access throughout the plant to suppon 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).

I Question # I 98 I RO/SRO I Tier: I Group: I KA: I CFR I SROIR: CogLevel I SRO I 1 I I [ 295003AA2.04 I 55.43@)(5) I 3.7 I Higher SystemlEvolution Name: I Category:

Partial or Complete Loss of A.C. Power I Emergency and Abnormal Plant Evolutions KA Statement:

Ability to determine andlor interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER: System lineups A, C. and D are incorrect - For the reasons described above.

I Objective: I Question Source: I Level of Difficulty:

None New I 2.3 I References provided to examinee: I CPS 4200.01, excluding: pages 3.4, 14-20, and &Iof the Appendices (A, B. I I and C); double jeopardy concerns

References:

I CPS 4200.01, Loss of AC Power USAR. 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 Date Written: I Comments: None 04/13/05 I Author: I Ryder 1

I Question # I 99 I KA Statement:

Ability to determine and/or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATION: Combustible limits for drywell 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 RPV Pressure is 215 psig and slowly lowering 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 PRIOR TO ERO ACTIVATION?

A. Prevent Igniter restart.

B. Attempt to re-start the Igniters.

C. Sample containment/drywell for hydrogen.

D. Slowly inject with Preferred Systems to re-establish Level Band 'C'.

Answer: A Explanation:

A is correct - 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 heen directed by EOP-IA means that RPV level dropped helow TAF (see the bottom-most 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 conditions' to determine what actual hydrogen levels must be. Here, it suggests that with RPV level helow 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 4412.00COOl. page 2 012. Section

3. I . Because of the removal of the PASS panel H2/02 monitors at CPS, this section instmcts operators to consider hydrogen and oxygen levels to he 'unknown'. if both MCR Monitors stop working. In either case. the NEXT action, therefore, is to 'prevent igniter restart' per the top-most override of EOP-7. Stem conditions indicate that the

1 Question# 1 99 1 ROISRO: I Tier: I Group: I KA: I CFR I S R O I K . I CogLevel SRO I 1 I 2 I 500000EA2.03 I 55.43@)(5) I 3.8 I Higher Sytedl;!ulutiim Name: I Category High Ciintainment Hydrogen Concentration I Fmergen.'y and Ahnorinal P l m F:\oIdtim, KA Statement:

Ability to determine and/or interpret the following as they apply IOHIGH PRIMARY CONTAINMENT HYDROGEN igniters are currently off.

B is incorrect - For the reasons described above.

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 lo recall the explicit requiremenis of CPS 4412.00CoOlI Section 3.1. hdshe will likely apply the lop-most IF-THEN action of the EOP-7 override. It is important to understand why this choice does represent a 2d correct answer. ..An understandinglapplicationof the EOP-7 BGes 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 he 'unknown' For this choice to he correct, would mean that & by & going to the 4412.M)CWI procedure could the CRS declare the hydrogen levels to be unknown. ..this is not true. What's more, the choice's wording suggests that the NEXT action would be to wait for Chemistry's actual sample (!ml readily available with the absence of the PASS monitors). The SRO candidate must recognize that ERO support is needed to I ) determine and develop an hydrogen sampling method. and 2) to actually get the sample. Step conditions indicate that the ERO is not yet activated. 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 represent the NEXT required action.

D is incorrect - Once 7 ADS Valves are open in EOP-3, that EOP directs operators to return to the Level leg of EOP-IA. However. upon returning to EOP-IA, at step 7, operators must wait until the RPV has denressurized helow 138 psig. per Table J. before attempting to re-establish Level Band 'C Objective: I Question Source: I Level of Difficulty:

None New I 3.0 References provided to examinee: 1 EOP flowcharts

References:

~~ I CPS EOP-IA. ATWS

~~ ~ RPV Control CPS EOP-3, RPV Blowdown CPS EOP-6, Primary Containment Control CPS EOP-7, Hydrogen Control CPS 4412.OOCoO1, Sampling Containment and Drywell for Hydrogen CPS EOP Technical Basis document Date Written: I 05/16/05 I Author: I Ryder Comments: None MODIFIED/NRC UNSAT. Added bullet to stem regarding RPV pressure to avoid distractor D possibly being correct. NRC also wanted further justification as to why distractor C could not be correct. The wording was changed for clarity but the justification provided above provides ample justification as to why this cannot be the NEXT required action.

[ Question # 1 100 I RO/SRO: I Tier: I Group: I KA: I CFR I SROIR: I CogLevel SRO I 2 I I I 239002 A2.03 I 55.43(b)(5) I 4.2 I Higher SystenVEvolutionName: I Category:

ReliefBafety Valves I Plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the RELIEFEAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open SRV A plant startup is in progress, with REACTOR POWER AT lo%, when the following occurs:

ONE SRV inadvertently opens and is STUCK OPEN The Immediate Operator Actions for CPS 4009.01, Inadvertent Opening Safety/Relief Valve, HAVE been performed The SRV is still STUCK OPEN 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) 270°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 Explanation:

C is correct -Per CPS 4009.01, Section 1. I , tailpipe temperature is expected to be >37S'F for a stuck open SRV (Note:

this value is derived from empirical CPS test data, and is recognized as being inconsistent with the predicted temperature denved from the Steam Tables/Mollier Diagram.). Per Section 4.3. operators will attempt to shut the SRV by pulling its solenoid fuses. Per Section 4.7. if it remains open, operators are directed to shifi CNMT Bldg HVAC.

then perform a Rapid Plant Shutdown IAW CPS 3005.01. Per CPS 3005.01. Section 8.5.1, a Rapid Plant Shutdown

[ Question # I 100 I RO/SRO: I Tier: I Group: I KA: I CFR 1 SROIR: I CogLevel SRO I 2 I I I 239002 A2.03 I 55.43(b)(5) I 4.2 I Higher SystemlEvolutioo Name: I Category:

Relief/Safety Valves I Plant Systems KA Statement:

Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct. control, or mitigate the consequences of those abnormal conditions or operations: Stuck open SRV looks like the following: I ) Evacuate Containment (already done as part of the Immediate Operator Action 3. I . of CPS 4009.01); 2) lower core flow to 43 mlbm/hr with Recirc (because the stem conditions indicate the plant is only in MODE 2 (pressurization in progress, at 750 psig), both RR Pumps are still running in SLOW speed and total core flow is already <43 mlbm/hr: therefore. this 'action' is already done; and 3) place the Mode Switch in SHUTDOWN: this is the&o action remaining to complete a Rapid Plant Shutdown, given these stem conditions.

A is incorrect - Per CPS 4009.01,Section I . I , this is an expected temperature for a leaking SRV, a an open SRV.

B is incorrect - Per CPS 4009.01, Section 1.1, this is the Hi-Hi Temperature alarm setpoint, but is still far below the range (>375"F) expected for an open SRV. Additionally, even though a 'stuck open SRV' is an ently condition (Symptom) for CPS 4005.01, Loss of Feedwater Heating. there is NO entry required given these stem conditions (i.e.,

at 750 p i g . in MODE 2, the plant is well below the 21.6%reactor power applicability threshold for the Loss of Feedwater Heating off-normal (see CPS 4005.01, NOTE at the top of page 2 of 8).

D is incorrect - Even though 40WF may he an expected tailpipe temperature for a stuck open SRV @e.. >375"F). the second part of this choice is incorrect for the same reason attributed to choice 'B'.

Objective: I Question Source: I Level of Difficulty:

DB400901.1.3.1 New I 2.5 References provided to examinee: I None

References:

I CPS 4005.01, Loss of Feedwater Heating CPS 4009.01, Inadvertent Opening SafetyIRelief Valve CPS 3005.01, Unit Power Changes CPS 3002.01, Hearup and Pressurization MODIFIED/NRC Enhancement. NRC felt that 375F called out in 4009.01 could possibly only apply if at rated pressure, therefore stem condition changed to indicate 10%. which implies rated pressure. GDSetser 6/14/05